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Handbook of nuclear chemistry Handbook of nuclear chemistry Document Transcript

  • 63 Nuclear Safeguards Verification Measurement Techniques M. Zendel1 . D. L. Donohue1 . E. Kuhn2 . S. Deron2 . T. Bı´ro´3 1 International Atomic Energy Agency, Vienna, Austria 2 Retired from the International Atomic Energy Agency, Vienna, Austria 3 Institute of Isotopes, HAS, Budapest, Hungary 63.1 Introduction ................................................................. 2896 63.2 Safeguards Verification Measurement Procedures .......................... 2898 63.2.1 Diversion Strategy ............................................................. 2899 63.2.2 Type of Material ............................................................... 2899 63.2.3 Significant Quantity ........................................................... 2899 63.2.4 Type of Facility ................................................................ 2900 63.2.5 Material Balance Area and Measurement Points ............................. 2900 63.2.6 Material Stratification for Sampling .......................................... 2900 63.2.7 Type of Defect ................................................................. 2901 63.2.8 Sampling Plan ................................................................. 2901 63.2.9 Inspection Activities for Safeguards Verification Measurements ............ 2901 63.2.10 Inspection Frequency ......................................................... 2902 63.2.11 Detection Probability ......................................................... 2902 63.2.12 Safeguards Approach .......................................................... 2903 63.2.13 Classification of Methods ..................................................... 2903 Nondestructive Assay ......................................................... 2903 Destructive Analysis ........................................................... 2904 63.2.14 Evaluations of Accountability Verification Measurements ................... 2904 63.3 Non-Destructive Assay (NDA) .............................................. 2905 63.3.1 Introduction ................................................................... 2905 63.3.2 Safeguards Environment and Measurement Conditions ..................... 2907 63.3.3 Gamma Ray Spectrometry .................................................... 2909 Gamma Ray Detectors ........................................................ 2909 Low-Resolution Gamma Spectroscopy (LRGS) .............................. 2911 High-Resolution Gamma Spectroscopy (HRGS) ............................ 2913 63.3.4 Neutron Counting Techniques ................................................ 2915 Neutron Detectors ............................................................ 2915 Gross Neutron Counting ..................................................... 2917 Neutron Coincidence Counting .............................................. 2918 Multiplicity Coincidence Counting ........................................... 2921 Active Neutron Coincidence Counting ....................................... 2922 Attila Ve´rtes, Sa´ndor Nagy, Zolta´n Klencsa´r, Rezso˝ G. Lovas & Frank Ro¨sch (eds.), Handbook of Nuclear Chemistry, DOI 10.1007/978-1-4419-0720-2_63, # Springer Science+Business Media B.V. 2011
  • 63.3.5 Spent Fuel Measurement ...................................................... 2923 Gamma Methods .............................................................. 2924 Neutron Methods ............................................................. 2925 Combined Gamma/Neutron Methods ....................................... 2926 Optical Methods .............................................................. 2927 63.3.6 Unattended NDA Systems .................................................... 2928 Unattended Gamma-Based NDA Systems .................................... 2930 Unattended Neutron-Based NDA Systems ................................... 2931 Other Unattended NDA Systems ............................................. 2934 63.3.7 Other NDA Techniques ....................................................... 2935 Physical Property Measurement .............................................. 2935 Calorimetric Techniques ...................................................... 2937 X-Ray Measurements ......................................................... 2938 Analytical NDA Techniques at Laboratories .................................. 2940 63.3.8 New and Novel Technologies ................................................. 2941 New Technologies ............................................................. 2941 Novel Technologies ............................................................ 2944 63.4 Laboratory Analysis for Nuclear Material Accountability Verifications ... 2950 63.4.1 Introduction ................................................................... 2950 63.4.2 Bulk Measurement, Sampling, Conditioning, and Shipment of Safeguards Inspection Samples ............................................... 2950 Spent Fuel Solutions .......................................................... 2951 Uranium Hexafluoride in Pressurized Cylinders ............................. 2952 Plutonium Oxide Powders .................................................... 2954 Uranium Dirty Scrap Materials ............................................... 2956 63.4.3 Safeguards Analytical Laboratories ........................................... 2957 Off-Site Laboratories .......................................................... 2957 On-Site Laboratories .......................................................... 2958 63.4.4 Isotopic Analysis .............................................................. 2960 Isotopic Analysis by Mass Spectrometry ..................................... 2960 238 Pu Abundance by Alpha Spectrometry ..................................... 2967 Gamma Spectrometry of Nuclear Material Samples ......................... 2970 63.4.5 Elemental Assay ............................................................... 2971 Ignition Gravimetry of U, Pu, Th ............................................ 2971 Uranium Titration ............................................................ 2972 Plutonium Titration .......................................................... 2973 Controlled Potential Coulometry of Plutonium ............................. 2975 Isotope Dilution Assays ....................................................... 2978 Spectrophotometric Determination of Hexavalent Plutonium .............. 2982 X-ray Absorption and Fluorescence Spectrometry ........................... 2983 Assay of Alternative Nuclear Materials ....................................... 2984 2894 63 Nuclear Safeguards Verification Measurement Techniques
  • 63.5 Environmental Sampling and Analysis to Verify the Completeness of State Declarations ........................................................... 2985 63.5.1 Introduction ................................................................... 2985 63.5.2 Sampling, Conditioning, and Shipment of IAEA Safeguards Environmental Inspection Samples ........................................... 2988 Cotton Swipe and Other Swipe Materials .................................... 2988 Air Filters ...................................................................... 2989 Water, Soil, Vegetation, and Biota Samples .................................. 2990 Process and Structural Materials ............................................. 2991 63.5.3 Safeguards Analytical Laboratories ........................................... 2991 Clean Laboratory for Safeguards ............................................. 2991 The Network of Analytical Laboratories (NWAL) of the IAEA ............. 2994 63.5.4 Sample Screening Methods ................................................... 2994 Gamma Spectrometry ......................................................... 2995 X-Ray Fluorescence Spectrometry ............................................ 2995 63.5.5 Bulk Sample Analysis ......................................................... 2995 Tracers ......................................................................... 2996 Sample Preparations and Separations ........................................ 2997 Thermal Ionization Mass Spectrometry ...................................... 2998 Inductively Coupled Plasma Mass Spectrometry ............................ 2998 63.5.6 Particle Analysis ............................................................... 2999 Sample Preparation ........................................................... 3000 Thermal Ionization Mass Spectrometry ...................................... 3000 Secondary Ion Mass Spectrometry ........................................... 3001 Scanning Electron Microscopy with X-Ray Spectrometry ................... 3002 Nuclear Safeguards Verification Measurement Techniques 63 2895
  • Abstract: This chapter deals with the ‘‘nuclear safeguards’’ verification system and describes procedures and measurement methods that allow the safeguards inspectorates/authorities to verify that nuclear materials or facilities are not used to further undeclared military activities. These procedures and methods provide the strong technical basis upon which the safeguards inspectorates/authorities issue their conclusions and receive the broadest international acceptance regarding the compliance of participating states with their obligations. 63.1 Introduction Nuclear safeguards stands for the ‘‘methods developed to safeguard the peaceful activities against diversion of nuclear material by the risk of early detection – controlling nuclear material as a measure of arms reduction’’ (IAEA 1985). More specifically, IAEA safeguards has been described as a comprehensive set of internationally approved technical and legal measures, applied by the IAEA, to verify the political undertakings of states not to use nuclear material to manufacture nuclear weapons and to deter any such use (IAEA 1998b). Nuclear material is defined as any source material (natural uranium, depleted uranium and thorium, excluding uranium ore) or special fissionable material (plutonium-239, uranium-233, ura- nium enriched in the isotopes 235 or 233) (IAEA 2002, No. 4.1) encountered in the applica- tions of nuclear energy. Nuclear safeguards systems have been introduced because of the fears raised by the devastating effects of the nuclear bombs detonated over Hiroshima and Nagasaki at the end of the Second World War in August 1945 (Fisher 1997). This chapter focuses on safeguards verification techniques and does not address other important safeguards measures such as containment and surveillance, near real time accountancy (NRTA), collection and analysis of satellite imagery, and ‘‘external information’’ on national programs and international trade and exchanges. These measures merit separate chapters outside the scope of this handbook. The chapter emphasizes the procedures and methods used to verify the accuracy of declared inventories and flows of nuclear materials but also those involving trace elements in ‘‘environmental samples’’ and novel technologies searching for undeclared nuclear material and activities. All procedures and methods (except new and novel technologies) discussed here are implemented by the IAEA and to a great extent by national or regional inspectorates. In the wake of the arms race, which started between the Atlantic Alliance and the Soviet Union, the United Nations set up the International Atomic Energy Agency (IAEA) in 1957. The UN delegated to IAEA the task of promoting the peaceful uses of atomic energy and establishing an international safeguards system, with the aim to stop the proliferation of nuclear weapons and foster disarmament. Yet up to 1971, IAEA was limited to exercise its safeguards on nuclear materials or facilities that were acquired through its technical assistance or were placed voluntarily in its custody (IAEA 1961; IAEA 1968). The nuclear Non- Proliferation Treaty (NPT) (IAEA 1970), which was open for signature in 1968, earmarked a major step toward a global international safeguards system. As of 2010, more than 190 states have become signatories of the treaty, covering indeed the dominant part of nuclear activities in the world. NPT entered into force in 1970. It requires the signatories to enter a safeguards agreement with the IAEA, whereby they renounce to nuclear weapons and place all their nuclear materials under IAEA control. Based on its Statute, the IAEA obtained approval of its NPTsafeguards system (IAEA 1971) in 1971. NPTassures the international community that all nuclear materials and activities under safeguards outside the so-called Nuclear Weapon States (the five states that possessed nuclear weapons when the NPTentered into force: China, France, 2896 63 Nuclear Safeguards Verification Measurement Techniques
  • Soviet Union (now Russia), United Kingdom, and United States of America) are used exclu- sively for peaceful purposes, contributing to dispel mistrust among states (IAEA 2002). With the expulsion of Iraqi armed forces from the territory of Kuwait in early 1991, the UNSCOM Special Commission and the IAEA (UNSC687 resolution) uncovered a comprehensive nuclear weapon program that Iraq had failed to declare in breach of its commitments as party to NPT. This led the IAEA member states to endorse a strengthening of IAEA safeguards measures providing the framework to detect undeclared nuclear material and activities. In 1997, the IAEA Board approved the model of ‘‘Additional Protocols (AP)’’ (IAEA 1997), and by 2009, the IAEA had concluded APs with 136 states and the European Atomic Energy Community (EURATOM). The AP spells out measures beyond the scope of INFCIRC-26, -66 or -153 type agreements providing broader access for the inspectorate to virtually all locations where nuclear material is handled or suspected and to detailed information on the state’s nuclear activities. The combination of activities under the comprehensive safeguards agreement and AP could determine that a state’s nuclear program is for peaceful purpose only and has been completely and correctly declared. This has drastically changed the safeguards system toward an information driven, state-level approach whereby the IAEA concludes for a state with increased confidence that no nuclear material has been diverted from declared sites and that there is no evidence of clandestine nuclear activities elsewhere. This state-level approach takes into account all available information to the IAEA such as inspection results, design informa- tion verification, open source evaluation including tracking of black-market nuclear networks, satellite imagery and environmental sampling. EURATOM, instituted as early as 1958 by its six founding states (Belgium, France, Germany, Italy, Luxembourg and the Netherlands), established a comprehensive international nuclear safeguards system soon after the signature of a US-EURATOM Cooperation Agree- ment (1959). Argentina and Brazil set up a joint safeguards inspectorate (ABACC) in 1991. IAEA safeguards agreements foresee operating safeguards state systems of accounting and control over nuclear materials (SSAC). Partnerships between the SSACs, ABACC, EURATOM, and the IAEA are a strong factor in optimizing the resources of all parties, while the IAEA retains the capability to reach independent conclusions. With the indefinite extension of the NPT in 2005, the IAEA has been confirmed into its responsibilities in the operation and strengthening of a worldwide nuclear safeguards system in cooperation with the relevant national and regional institutions and the United Nations Organization. International safeguards systems have certainly succeeded in limiting proliferation of nuclear weapons, but the world is still a long way from being free from their menace. Most non-nuclear weapons states are weary of unfair burden upon those subject to NPT, while nuclear weapon states appear reluctant to progress toward nuclear disarmament. A few states continue to use this as an argument for not signing NPTat the cost of mistrust by neighboring countries and international community. Others like Israel consider that existing safeguards systems or other treaties are not sufficient to ensure their national security and refuse to renounce to nuclear weapons. Serious conflicts in the Middle East, in the Korean peninsula, and Asia make much current headlines about further risks of proliferation. International safeguards, evidently, can at best contribute to build and maintain international confidence. However, their effectiveness (but also their limitations) had a positive impact on the negoti- ations of the successive arms limitation treaties and on the declaration of seven Nuclear Weapon Free Zones: the Antartic (1961), Space (1967), South America (Tlatelolco Treaty 1968), Seabeds and Ocean Floors (1972), the Pacific (Rarotonga Treaty 1986), South East Asia (Bangkok Treaty 1995), and Afrika (Pelindaba 1996). The signatories of the four regional Nuclear Safeguards Verification Measurement Techniques 63 2897
  • treaties have the obligation to enter comprehensive safeguards agreements with the IAEA. Synergies between nuclear safeguards measures and future progresses toward a global nuclear disarmament, could be considered for the IAEA and CTBTO – the UN organization created for the control of the compliance with the Comprehensive Nuclear Weapon Test Ban Treaty, yet under ratification – as both are based in Vienna. The IAEA has acquired vast experience in safeguards verification techniques. It may be called to take on new roles in the future, such as verifying fissile material from dismantled weapons or verifying compliance with a potential global ban on the production of fissile material for weapons. It could thus contribute to both nonproliferation and disarmament. The IAEA and a commission of eminent persons (IAEA 2008) have recently reviewed IAEA activities to meet future challenges up to 2020 and beyond. The increasing spread of nuclear material, technology, and know-how may pose increased proliferation risks in a globalized world. Safeguards, which will remain a core mission for the IAEA, must be further strength- ened to cope with expanded growth and spread of nuclear power generation – and particularly, the establishment of new facilities for uranium enrichment, spent-fuel reprocessing, or processing of direct-use nuclear material. This will require new technologies to make safe- guards more effective and efficient. Multilateral fuel-cycle centers and proliferation-resistant nuclear facilities could facilitate future safeguards implementation. Meeting future challenges will require a robust IAEA ‘‘toolbox’’ containing: the necessary legal authority to gather information and carry out inspections, state-of-the-art technology, particularly for the detection of clandestine nuclear activities, a high-caliber workforce, and sufficient resources. 63.2 Safeguards Verification Measurement Procedures The safeguards systems, based on regional and international treaties, consist of activities at the headquarters and on-site inspections to verify that states comply with their obligations derived from the related specific agreements. At Headquarters, state reports and declarations are being received and evaluated in context with all other available information such as inspection data, open source information and satellite imagery to assess the peaceful nature of a state’s nuclear program. Databases on nuclear material inventories and transfers are being maintained. In addition, inspection data acquired from implementing technical devices in the field are being archived and used for preparing and coordinating on-site inspections. On-site inspections are performed to independently verify nuclear material and activities and are considered the most powerful tool for safeguards purposes. The technical objective of NPT is ‘‘the timely detection of diversion of significant quantities of nuclear material from peaceful nuclear activities to the manufacture of nuclear weapons and other nuclear explosive devices or for purposes unknown and deterrence of such diversion by risk of early detection’’ (IAEA 2002). To this aim, safeguards inspectorates regularly physically verify the inventories and transfers of declared nuclear materials subject to safeguards agree- ments. The type and amount of nuclear material are thereby inspected by direct physical and/or analytical measurements, using in the first place the nuclear characteristics of the material (emitted gamma or neutron radiation, isotopic composition, etc.) but also other ‘‘non- nuclear’’-type measurements, such as weighing, calorimetry, Cherenkov radiation detection, etc. The goal is to verify the correctness and completeness of the nuclear material accountancy reports of the state and to confirm that no ‘‘safeguards significant quantities’’ of nuclear 2898 63 Nuclear Safeguards Verification Measurement Techniques
  • materials are missing. In addition, IAEA inspectors are authorized to take ‘‘environmental samples’’ in specified locations within or near nuclear facilities in states having signed an ‘‘additional protocol,’’ which defines means for the IAEA to confirm the absence of signatures of undeclared activities in the facility process or its environment. All these measures are implemented in close cooperation with the facility operator and the state or regional safeguards authorities, but in a way that guarantees that the IAEA can draw its own independent conclusions. International safeguards systems have thus two basic pillars, the first one being the state’s declaration regarding the nuclear activities and nuclear material inventory and accountancy in the nuclear facilities in the state, the second being a complex verification system to confirm that the state’s accountancy system and its declarations are both correct and complete. The state accountancy reports are expected to be based on operator’s own accountability data. The following definitions and descriptions are used in order to quantify and qualify the inspection goals (IAEA 2002). 63.2.1 Diversion Strategy Diversion of nuclear material, a particular case of noncompliance, could include: The undeclared removal of declared nuclear material from a safeguarded facility. The use of a safeguarded facility for the production of undeclared nuclear material (e.g., production of high enriched uranium in an enrichment plant or plutonium in a reactor through irradiation). The use of nuclear material specified and placed under safeguards in such a way as to further any military purpose (applicable to specific safeguards systems). The diversion strategy is a hypothetical scheme, which the diverter could consider to divert nuclear material or to misuse items subject to safeguards. It could include one or more concealment methods, i.e., actions to reduce the probability of detection, such as tampering safeguards equipment or accountancy, falsifying documents and declarations, etc. Verification principles, methods, and procedures are designed to uncover possible diversion strategies according to the type of nuclear material and the characteristics of the facility. 63.2.2 Type of Material Nuclear material is classified according to the element contained and, for uranium the degree of enrichment. Six classes are defined at this time: plutonium (Pu), high enriched uranium (HEU), uranium-233 (233 U), depleted uranium (DU), natural uranium (NU), low enriched uranium (LEU), and thorium (Th). 63.2.3 Significant Quantity To set a target for the quantification of the inspection goal, the IAEA currently adopts the following quantities (> Table 63.1) called ‘‘significant quantities’’ (SQs) as estimates of the amount of material that would be sufficient to manufacture a nuclear explosive device. Nuclear Safeguards Verification Measurement Techniques 63 2899
  • 63.2.4 Type of Facility Safeguards procedures vary also according to the type of facility and the physical form under which nuclear materials appear. Two types of facilities are considered. Item facility: Power reactors (Light Water Reactor (LWR), On-load Reactor (OLR)), Research Reactor (Material Testing Reactor (MTR), Fast Breeder Reactor (FBR), (TRIGA) and Critical Assemblies, Nuclear Material Storage (dry and wet). Bulk handling facility: fuel fabrication, reprocessing, conversion, and enrichment. 63.2.5 Material Balance Area and Measurement Points A material balance area (MBA) defines an area in or outside of a facility such that: ● The quantity of nuclear material in each transfer into or out of each MBA can be determined. ● The physical inventory of nuclear material in each MBA can be determined when necessary, in accordance to specified procedures. A key measurement point (KMP) is a location where nuclear material appears in a form that may be measured to determine material flow or inventory. In addition to the above, whenever the state has accepted the additional measures foreseen in the model protocol INFCIRC/540, IAEA inspectors are granted access to ‘‘location-specific’’ environmental sampling points where they may take samples of air, water, vegetation, soil, smears, etc., to confirm the absence of undeclared nuclear material or activities at the specified location (IAEA 1997). 63.2.6 Material Stratification for Sampling In order to make a meaningful statistical evaluation of the results of accountancy verifications, the inspector’s measurements must be planned in a way, which will provide independent estimates of the overall measurement uncertainties. All factors influencing these uncertainties must be considered, such as the material type and form, the sampling procedure, and the measurement method. According to theory and experience, the inspectorate detection . Table 63.1 IAEA Significant Quantities (SQs) Material type Isotopic composition SQ HEU 235 U > 20% 25 kg 235 U LEU 235 U < 20% 75 kg 235 U Natural U (NU) 235 U = 0.7% 10 t NU Depleted U 235 U < 0.7% 20 t DU 233 U 8 kg 233 U Plutonium <80% 238 Pu 8 kg Pu Thorium 20 t Th 2900 63 Nuclear Safeguards Verification Measurement Techniques
  • capability is optimal if the materials (items, batches, and lots) being verified are grouped into ‘‘material strata’’ having similar features, and physical and chemical properties. In practice, the inspector and the operator agree on a common stratification. 63.2.7 Type of Defect A difference between the declared amount of nuclear material or nonnuclear material and the material actually present for verification purposes is called a defect. Gross defect refers to an item or batch that has been falsified to the maximum extent possible so that all or most of the nuclear material is missing (e.g., a spent fuel (SF) assembly is substituted with a dummy assembly containing no nuclear material). Partial defect refers to an item or batch that has been falsified in such a way that some fraction of the declared amount is yet actually present. Bias defect refers to an item or batch that has been slightly falsified so that only a small fraction of the declared amount is missing. 63.2.8 Sampling Plan The term sample has two meanings: ● In statistical sampling, a sample is a subset of items selected from a defined group (population) of items. ● In material sampling for analysis, a sample is a small quantity of material taken from one item or container for measurement. Statistical sampling plan procedures are applied to determine, in a given stratum, the number of items to be verified by each of the relevant measurement methods (nondestructive assay (NDA), weighing, sampling, and destructive analysis (DA)). The sampling plan is based on inspection by attributes and ensures that – based on the assumption that the operator’s declaration has been falsified by the goal amount – at least one defect will be correctly identified as a defect with probability (1Àb), where b is the nondetection probability. The preselected value of 1Àb is typically 90% for high and 20% for low probability levels. The goal amount is usually 1 SQ (see > Sect. 63.2.3). 63.2.9 Inspection Activities for Safeguards Verification Measurements The physical verification activities consist of the following steps (IAEA 1997, 2004): 1. Advance planning of an inspection is done at the inspectorate headquarter or inspectors’ regional office and includes: (a) A review of the facility design information and the accountancy reports. (b) A stratification of the declared nuclear material available for verification in groups or ‘‘strata’’ of items of similar characteristics at each KMP. (c) A statistical sampling plan for each material stratum to ensure a predetermined detection probability for the specific types of verification measurements and defect testing. Nuclear Safeguards Verification Measurement Techniques 63 2901
  • 2. During inspection at the facility: (a) For each KMP: (i) Selected items or batches are measured using NDA. (ii) Their weight or volume are determined, the operator’s scales or tanks calibration and declared tare weights are confirmed. (iii) Statistically significant defects in the above tests are investigated. (iv) Material samples are taken, as required by the sampling plan, and conditioned, to ensure that the integrity of the analytical information is maintained during transport. (b) Environmental samples are taken at each ‘‘environmental sampling point,’’ as specified in the sampling plan. 3. The samples are analyzed in the laboratory selected by the inspectorate. 4. The inspection data and the results of the analyses are collected and evaluated statistically, usually at the inspectorate headquarter. 63.2.10 Inspection Frequency A minimum time necessary for manufacturing a nuclear explosive device has been defined according to the material category, its irradiation status and suitability for conversion into components of nuclear explosive devices. Direct use material (such as Pu containing less than 80% 238 Pu, HEU and 233 U) can be used without transmutation or further enrichment. Unirradiated material does not contain substantial amounts of fission products thus it requires less time and effort to be converted to components of nuclear explosive devices than irradiated direct use material like spent reactor fuel. Indirect use material (DU, NU, LEU and Th) must be further processed to produce direct use material. These categories set the period of time used as the objective for timely detection of a diversion (timeliness component of the inspection goal) and govern the inspection frequencies (number of inspections per year). Material categories and typical inspection frequencies are shown in > Table 63.2. 63.2.11 Detection Probability Because safeguards systems cannot in practice cover all material during an inspection, the inspectors take only random samples of the total population for verification. The goal is to assure that the detection probability for detecting a defect (a falsified or missing item, a certain . Table 63.2 Material categories and typical inspection frequencies Material category Material type Inspection frequency Direct use material unirradiated* Purified Pu, HEU 1 month Direct use material irradiated Spent fuels 3 months Indirect use material (235 U < 20%, NU, DU, Th) 1 year 2902 63 Nuclear Safeguards Verification Measurement Techniques
  • quantity missing from items or bulk material) applied to each material stratum is not below a predetermined level. 63.2.12 Safeguards Approach The type of the safeguards agreement and facility, as well as all the above factors are taken into consideration when the safeguards approach for the particular facility is prepared. This approach will determine also the verification methods and the instruments that will be installed permanently at facilities and applied during inspections or used in the analytical laboratories on samples taken by the inspectors. 63.2.13 Classification of Methods > Table 63.3 lists the four classes of methods and instruments implemented for safeguards verification measurements. Nondestructive Assay A nondestructive assay is a measurement of the nuclear material content or of the element or isotopic concentration of an item without producing significant physical or chemical changes in the item. It is generally carried out by observing the radiometric emission or response from the item and by comparing that emission or response with a calibration based on essentially similar items whose contents have been determined through destructive analysis. There are two broad categories of NDA: (a) Passive assay, in which the measurement refers to spontaneous emissions of neutrons or gamma rays or to the total decay energy. (b) Active assay, in which the measurement refers to a stimulated emission (e.g., neutron- or photon-induced fission). . Table 63.3 Classification of methods and instruments Class Implementation mode NDA unattended mode Instruments permanently installed, possibly combined with remote monitoring via satellites or other communication means NDA attended mode Instruments kept at the facility or hand-carried/shipped by the inspector Nuclear material sample DA samples taken at the facility and shipped to an analytical laboratory for elemental and isotopic assay of fissile element Environmental sampling Environmental samples, mainly swipes taken from inside or outside surfaces at facilities or any other location, analyzed in specialized laboratories for signatures of potential undeclared activities Nuclear Safeguards Verification Measurement Techniques 63 2903
  • Destructive Analysis Determination of nuclear material content and, if required, of the isotopic composition of chemical elements present in the sample. Destructive analysis normally involves destruction of the physical form of the sample. In the context of IAEA safeguards, determination of the nuclear material content of an item sampled usually involves: (a) Measurement of the mass of the sample. (b) The taking of a representative sample. (c) Sample conditioning (if necessary) prior to shipment to the Safeguards Analytical Labo- ratory for analysis. (d) Processing of the sample to the chemical state required for the analysis (e.g., dissolution in nitric acid). (e) Determination of the concentration of the nuclear material (U, Pu, Th) present in the sample (i.e., elemental analysis) using techniques such as chemical titration, controlled potential coulometry, gravimetrical analysis, isotope dilution mass spectrometry, and K-edge densitometry. (f) Determination of the isotopic abundance ratios of U or Pu isotopes (i.e., isotopic analysis) using, inter alia, techniques such as mass spectrometry, gas mass spectrometry, and thermal ionization mass spectrometry. 63.2.14 Evaluations of Accountability Verification Measurements The majority of the nuclear material samples, collected during safeguards inspections and sent to an analytical laboratory for measurements, serve several purposes. (i) They are taken to verify the correctness of declarations for the amounts of nuclear material (element (U, Pu, Th) and isotope amounts (235 U, 233 U)) in inventories and in transfers into or out of a facility; (ii) The analytical results obtained for the verification samples are also evaluated for the purpose of verifying the quality and functioning of the operators’ measurement systems; these should ‘‘. . .conform to the latest international standards or be equivalent in quality to such standards’’ (IAEA 1972). The operator-declared data and the measurement results, obtained for the inspection samples, are stored in the inspectorate in a central operator–inspector database where they can be accessed for subsequent evaluations (IAEA 2004). This database also contains the results of verification measurements by nondestructive assay (NDA) methods. In the IAEA data analysis, various statistical techniques (IAEA 1998a) are used to derive separate estimates of the operator’s and inspector’s uncertainty parameters based on the collection of historical operator–inspector differences. The results of these evaluations are ‘‘performance values,’’ typical for a specific facility and for each stratum (material type) and measurement method combination. The actually observed ‘‘verification measurement perfor- mance’’ is then used for the planning (sample size calculations), the conduct (establishing reject limits), and the evaluation (material balance) of inspections in a given facility. The nuclear material contained and processed in a facility is stratified into items or batches that have similar physical and chemical characteristics. Grouping the material into strata simplifies verification and makes it possible to formulate the sampling plans needed to verify a material balance and to calculate its uncertainty. In calculating sampling plans, generally 2904 63 Nuclear Safeguards Verification Measurement Techniques
  • three levels of ‘‘defects’’ (differences between the declared amount of nuclear material and the material actually present) are considered. In most situations NDA measurements serve for the purpose of detecting ‘‘partial and gross defects,’’ while sampling for DA copes with the detection of ‘‘bias defects.’’ The sample size for a stratum defines the number of items to be verified in order to be able to draw conclusions about the total population. The total sample size can then be allocated among the accountancy verification methods for gross, partial and bias defects (IAEA 1998a; Jaech and Russell 1991). The uncertainty values generally used are the verification performance estimates for the given facility and strata. The International Target Values could serve as a substitute (IAEA 2001). For every inspection, the analytical results are evaluated by an operator–inspector paired comparison. Differences exceeding the 3-sigma limits, calculated from the respective verifica- tion performance estimates, are classified as discrepant and subject to further investigations. The ‘‘material unaccounted for’’ (MUF) over a ‘‘material balance period,’’ declared by the operator for each ‘‘material balance area’’ and its statistical significance are examined, where: MUF ¼ PB þ X À Y À PE with PB being the beginning inventory, X the sum of all inventory increases, Y the sum of all inventory decreases, PE the ending inventory. For the declared MUF, its uncertainty sMUF is calculated from the verification performance estimates, as derived from the historical operator–inspector differences for this facility. The declared MUF is expected not to exceed 3 Â sMUF; otherwise it is concluded that MUF is statistically significant. The calculated sMUF value is also compared to the International Standard of Accountancy (IAEA 2002, No 6.35) which is based on the operating experience at the various types of bulk handling facilities and defines the uncertainty with which a facility operator is expected to be able to close a material balance. If sMUF is larger, it is concluded that the facility measurement system does not meet this standard. In both cases, a significant finding will trigger follow-up evaluations, and activities with respect to the facility accountancy and measurement system will be initiated. For all samples collected during inspections, the analytical results and their evaluation will be reported to the corresponding state as part of an inspection statement. Any significant finding in the material balance evaluations will also be reflected in the annual Safeguards Implementation Report. 63.3 Non-Destructive Assay (NDA) 63.3.1 Introduction Nondestructive assay (NDA) for safeguards describes analytical techniques to measure, check, and verify the amount of nuclear material or of the elemental or isotopic concentration of an item without producing significant physical or chemical changes in the item. It allows inspectors to determine both the quantity and composition of nuclear material without ever sampling it directly. Ultimately, NDA techniques provide for the independent verification of the total amount of nuclear material held at a nuclear facility. The main nuclear materials of interest are uranium (U) and plutonium (Pu). Usually, no single measurement method can Nuclear Safeguards Verification Measurement Techniques 63 2905
  • directly determine the total amount of either Pu or U and only a combination of appropriate NDA techniques provide the total mass of the respective element. The most widely used NDA instruments rely on the detection of nuclear radiation such as gamma rays and/or neutrons. Physical measurement techniques are also used with available instruments that measure heat, weight, liquid volume, thickness, and light emission/absorption. These physical techniques may be applied by themselves, or they may be used in combination with other nuclear measurements to provide quantitative measurements of the nuclear material. The general reference on the theory and application of passive NDA (PANDA) is given by reference (Reilly et al. 1991) and its addendum (Reilly et al. 2007). The development of NDA instruments for the measurement of nuclear materials has evolved over many years. Various national and international institutions have developed NDA techniques with the goal of increasing the technical competence of safeguards systems. NDA instruments range in size and complexity from small portable units for use by safeguards inspectors during on-site verification of nuclear materials, to large in situ NDA systems designed for routine in-plant use (e.g., plant operator equipment, subject to independent authentication). Most of the NDA equipment has to withstand demanding environments such as high radiation fields, variable humidity, and high and low temperatures. In addition, such equipment must be reliable with consistently reproduced performance when operated by different users. In general, NDA techniques are less expensive, nonintrusive on the operation of the nuclear facility, less time consuming than destructive analysis (DA) techniques, and are amenable to automation. NDA measurements can be performed on large quantities of nuclear material without breaching the container or containment of the material. Principally, NDA significantly reduces the need for DA sampling. However, the accuracy associated with NDA verification measurements is generally less than it would be if the items were verified at the final process stage at which sampling for DA and direct measurement of the nuclear material content become possible. In many instances, NDA is the only technically feasible solution to perform verification, e.g., for valuable finished products (such as fresh fuel assemblies/pins) and also when direct access to nuclear material is impossible or undesirable (e.g., spent fuel). NDA measurements can be made outside of glove boxes, transport containers, on solutions inside processing systems, and on materials packaged for storage and disposition. NDA methods are also well suited to the verification of inhomogeneous bulk materials such as waste, where representative DA sampling cannot be performed. NDA offers another important advantage over traditional DA methods: measurements can be performed in a timely manner, both in situ and during inspection activities. In the context of neutron and gamma-ray measurements, NDA techniques used by the inspectorate can be categorized as passive NDA or active NDA. Passive NDA refers to tech- niques that measure radiation emitted spontaneously from nuclear material. This method is often applied to Pu samples, because of the large spontaneous fission rate of the even-even Pu isotopes. Active NDA, on the other hand, refers to techniques that measure induced radiation responses from a sample, often using an external neutron source. These active methods are usually applied to perform uranium measurements where the spontaneous fission rate is low. In addition to the quantitative measurements performed by an inspector, in some cases a qualitative measurement (attribute testing) is sufficient, e.g., simply confirming the presence of a representative isotope based on a typical gamma ray. 2906 63 Nuclear Safeguards Verification Measurement Techniques
  • 63.3.2 Safeguards Environment and Measurement Conditions Safeguards inspectors routinely use NDA to perform the verification of nuclear materials present in the complete range of nuclear fuel cycle facilities. Today, a broad range of modern NDA instruments and techniques are available for the detection, identification, assay and verification of nuclear material in a wide variety of physical and chemical forms. The IAEA alone has authorized over 100 different types of NDA instruments for inspection use, while several NDA instruments are under evaluation and others under development (IAEA 2003; Zendel 2008). If safeguards verification is to be effective, inspectors have to perform indepen- dent measurements in order to verify declared material quantities. Selection of the correct equipment for a particular measurement task is a very important part of inspection activities and a thorough knowledge of an NDA technique’s capability to meet the goal of the inspection activity is required. A high level of standardization of NDA equipment can reduce implemen- tation costs significantly because less maintenance and training is required. The diversity of nuclear materials present in the large variety of nuclear facilities to be safeguarded necessitates an equally diverse ‘‘toolbox’’ of NDA instrumentation. Nuclear mate- rial can be grouped into two main types: bulk material and item material. The first group comprises powders, pellets, solutions, scrap, and metals contained in various process and storage containers, whereas the second group consists mainly of fuel elements and pins in various storage configurations. Further distinctions are made in respect of the strategic value of the nuclear material as unirradiated direct use (e.g., Pu-oxide and mixed oxide (MOX) powders), irradiated direct use (e.g., spent fuel), and indirect use material (depleted, natural and low enriched uranium). The hardware and software of an NDA system depend on the specific conditions (e.g., environment, sample matrix, etc.) of the measurement task. In most cases, different NDA instruments are required to obtain the total quantity for a given nuclear material sample. Only a combination of the results of several calibrated NDA techniques provides for a credible assessment of the type and quantity of nuclear material. The nature of a particular inspection activity may require the customization of equipment and methods applied. The equipment and method should be simple and user friendly, and should provide results in a short time period, because the available inspection time is usually very limited. Furthermore, emphasis is placed on compact, lightweight, rugged, and reasonably priced equipment, which can be carried by the inspectors and thus remain completely under their control. For large and complex facilities, the inspectorates require facility resident and specific safeguards equipment systems with features for attended, unattended, and remotely controlled operations. Some of the safeguards systems are integrated in the process equipment of the facility. The dimensions of the measurement head (the sensor) for these systems are often strongly conditioned by factors like sample geometry, plant requirements and detection efficiency, all of which factors impose boundary conditions that demand an individual design. Calibration is necessary in all quantitative NDA measurements to relate measured responses (e.g., neutron coincidence rate or specific gamma intensity) to nuclear material characteristics. An accurate measurement depends crucially on the effective calibration of measurement instrumentation. This calibration is based on similar items whose nuclear material content is very accurately known. The resulting calibration functions including all necessary correction factors (such as those relating to neutron multiplication or gamma-ray Nuclear Safeguards Verification Measurement Techniques 63 2907
  • absorption effects) are incorporated into the software or firmware associated with a given measurement technique. In certain specific cases, the procurement and use of standards that are nominally identical to the measured items is possible. The use of calibration standards on a one-to-one basis is, in general, not feasible. Other alternatives, such as predictive modeling methods based on Monte Carlo calculations (mainly with the code MCNP (Monte Carlo Nuclear Particle)) are being increasingly employed for the prediction of responses for given samples under well-defined conditions (Bourva et al. 2007; Lebrun et al. 2007). A recent version of the Monte Carlo code, MCNPX (Monte Carlo Nuclear Particle Extended), can directly simulate the singles, doubles, and triples count rates from a known neutron source (see > Sect. In addition to meeting the urgent practical need for the reduction of the number of calibration standards, a good computational method, once developed, could play an important role in helping to identify false physical standards. Measurement accuracy and quality control (QC) are important issues and require a regular measurement control program to provide standards against which to check the measurement performance of equipment. Normalization standards are applied independently of calibration standards to verify that an instrument is working properly and providing authentic data. Such a standard may be: (a) a well-characterized radiation source (e.g., 252 Cf source for neutron coincidence counters or sources of known energy and intensity for gamma spectrometers); or, (b) a properly authenticated sample of plant-specific material that has been kept under safeguards seal. In some instances, safeguards inspectorates do not own the NDA equipment and need to share the instruments with third parties, e.g., operators. In this case, documented joint-use procedures should define arrangements for data sharing, authentication, recalibration and use of standards and software, maintenance, repair, storage, and transportation. Sensors, together with their electronics and data generators, are security critical compo- nents, as they are the prime sources of independent safeguards data. Therefore, any unauthorized access to potentially vulnerable components must be prevented by containing these components in tamper-indicating housings and by restricting their servicing, repair, and replacement to inspectorate staff. A sensor mounted with its data generator in a single tamper- indicating enclosure constitutes a ‘‘smart sensor.’’ NDA measurements are usually performed on awell-stratified inventory grouping items with similar weights, locations, and properties together. The inspector applies a random sampling plan based on the number of items per stratum, the measurement performance of the selected NDA instrument, the significant quantity and the non-detection probability. For verification purposes, random sampling plans are calculated to detect three types of defects: gross, partial, and bias. In some cases, where it is not possible to take samples or to obtain representative samples (e.g., inhomogeneous material such as dirty scrap), partial defect NDA measurements are performed in lieu of bias defect tests. The rejection limit for a given NDA measurement is set at three-sigma level of the performance-based measurement errors. Performance-based mea- surement errors are derived from the collection of historical operator–inspector differences obtained for each MBA/stratum/measurement method combination. The performance- based measurement error provides the overall uncertainty combining all sources of uncer- tainties associated with the measurement (Aigner et al. 2002; IAEA 2001). Real performance values for NDA techniques applied to safeguards have been reviewed extensively (Guardini et al. 2004). The hypothesis of the random sampling plan is based on the assumption that no defect and hence no ‘‘outlier’’ is present in the stratum; otherwise the verification of the stratum has failed. 2908 63 Nuclear Safeguards Verification Measurement Techniques
  • In most cases, partial and bias defect testing by NDA are accomplished through the application of destructive analysis (DA) methods to random samples derived from the selected items. 63.3.3 Gamma Ray Spectrometry Most nuclear materials of concern to safeguards emit gamma rays that can be used for identification and quantitative measurement purposes. Gamma-ray devices measure the energy and intensity of gamma rays emitted from nuclear material. The transmitted gamma ray intensity is mainly a function of gamma ray energy, absorber composition, absorber thickness, and measurement geometry – each of these parameters must be carefully considered in the final analysis. Spectral analysis of emitted radiation identifies the type of a nuclear material by its characteristic gamma rays. Further analysis of selected gamma ray intensities provides quantitative characteristics such as enrichment and isotopic composition. In some cases, additional parameters must be measured (e.g., active length, weight, etc.) to calculate the total mass of the element. The gamma spectrometry technique, used in nuclear safeguards for measuring enrichment, requires less expensive equipment than the mass spectrometry technique and the equipment required is very easy to operate and maintain. Gamma spectrometry techniques, if properly implemented, can offer very precise results within relatively short measurement times. Both low- and high-resolution gamma-spectrometric measurements are essential verification tools for safeguards purposes. Main attributes of various gamma ray-detector systems are listed in > Table 63.4. Gamma Ray Detectors Three main types of instruments are presently used for gamma spectrometry: inorganic scintillation counters (usually activated sodium iodide – NaI – crystals), semiconductor detectors (usually high-purity germanium – HPGe – crystals), and gas-filled detectors (e.g., high-pressure xenon ionization chambers). In all cases, the interaction of a gamma ray with the detector results in an electrical signal, whose intensity is proportional to the energy of the incoming gamma ray. The signal is amplified, processed in a pulse processing electronic chain, counted, and analyzed by a multichannel analyzer (MCA). The resulting gamma spectrum (which shows the number of events as a function of gamma energy) exhibits isotope- characteristic gamma peaks. Finally, the spectrum is analyzed using specialized software, performing peak fitting, background subtraction, peak intensity calculation, external or intrinsic calibration and calculation of the relative isotopic abundances. Semiconductor detectors change their conductivity upon the impact of radiation by pro- ducing a flow of free electrons (n-type) or positive holes (p-type) in the semiconductor material. This results in a collection of charge at the electrodes, when a voltage is applied to the semiconductor. Germanium detectors presently offer the best resolution but must be cooled by liquid nitrogen – they are capable of resolving complex gamma spectra and determining the isotopic composition of essentially all nuclear materials present in the nuclear fuel cycle. Recently, electrically cooled HPGe became available with sufficient resolution to measure the isotopic composition of Pu-samples. These systems are especially important for future unattended and remote systems as well as for field applications where liquid nitrogen is not available. Nuclear Safeguards Verification Measurement Techniques 63 2909
  • Room temperature semiconductor detectors – cadmium/zinc/telluride (CdZnTe) and CdTe detectors in particular – have a proven record in safeguards verification measurements and related applications and have been in use for more than 15 years (Arlt et al. 1992, 1993; Arlt and Rundquist 1996). Peltier-cooled CdTe detectors operated at just below 0 C can achieve a resolution at relatively low gamma-ray energies (<200 keV), which are sufficient to perform selected isotopic measurements for uranium and plutonium. Silicon-pin photo diodes can serve as detectors for X-ray and gamma ray photons. They are used as gamma monitors in high-radiation fields. CdZnTe and CdTe detectors are ideal for field measurements and for the design of small detection probes that can be operated in close proximity to the items to be verified, even if there are space restrictions. They have become the most versatile room temperature gamma detectors, complementing the classical NaI and liquid nitrogen-cooled germanium detectors, by providing a medium resolution and a reasonable efficiency. In many cases, the use of these detectors has helped to increase both the efficiency and effectiveness of NDA methods applied in the nuclear safeguards field. Scintillation detectors consist of a scintillator (usually inorganic crystals such as NaI) and a photomultiplier tube. The scintillator emits light upon the absorption of radiation. This light . Table 63.4 Gamma-ray systems System Detector Verification task Typical performance values (%)a HRGS HPGe 240 Pu-effective (%) in fresh PuO2, MOX, scrap 0.3–1 240 Pu-effective (%) in fresh fuel 1 Uranium enrichment in UF6 cylinder 2 (LEU)-10 (DU) HM-5 NaI Search and Identify nuclear material Attribute Active length 0.5 cm 235 U enrichment in fresh fuel (NAIGEM) 3 MMCN 235 U enrichment powder/pellets 3 UBVS 235 U enrichment in UO3 3 MMCC CdZnTe 235 U enrichment in fresh fuel assemblies 5 FMAT Fresh MOX assemblies underwater Attribute IRAT Irradiated nonfuel items Attribute SFAT Spent fuel (no movement required) Attribute CBVB Spent fuel bundles (CANDU) Attribute CBVS Spent fuel bundles in stacks (CANDU) Attribute SMOPY CdZnTe + FC LWR spent fuel 5(attribute/consistency) CRPS CdZnTe (+FC) Radiation profiling of dry storage casks Attribute HSGM IC Dose rate of irradiated items Attribute FDET IC + FC Burnup and SF confirmation 10(attribute/consistency) a Average measurement time ~300 s. 2910 63 Nuclear Safeguards Verification Measurement Techniques
  • is collected and converted into an electronic signal by a photomultiplier tube. Scintillation detectors operate at room temperature and are cheaper and more robust than germanium detectors. Scintillation counters have a low energy resolution, but high detection efficiency. However, they experience a distinct shift with temperature changes and need to be stabilized frequently. Improving detection capability, specifically resolution and sensitivity, is a continuous challenge. Recently, LaCl3 and LaBr3 scintillation detectors have been introduced (Synthfeld et al. 2006). This type of detector combines an improved detector sensitivity relative to CdZnTe detectors with a reasonable resolution (2% at 122 keV), in comparison with NaI(Tl) detectors of the same size. LaBr3 detectors are already commercially available and implemented for isotope identification and are employed in uranium enrichment applications that profit from their superior resolution and sensitivity. Gas-filled detectors record ionization of the gas in the chamber caused by the gamma interaction. The ion current is proportional to the amount of energy deposited by the gamma ray. Gas detectors feature long-term stability that cannot be matched by scintillator or solid-state detectors because the charge transport properties of the gas are not significantly affected by changes in temperature and the effects of radiation. This high stability is very important for detectors in unattended monitoring applications, where background tempera- ture and radiation varies significantly. High-pressure xenon ionization chambers have emerged recently as gamma-ray spectrometers. Low-Resolution Gamma Spectroscopy (LRGS) Low-resolution gamma spectroscopy (LRGS) is simple to use and easy to implement under field conditions. LRGS applications in safeguards range from performing the quantitative verification of enrichment levels to the purely qualitative detection of spent fuel attributes and the presence of nuclear material. Mini multichannel analyzers connected to NaI detectors (MMCN) are routinely used to verify the enrichment of uranium in powders and pellets (Arlt et al. 1997). The basic mea- surement procedure involves viewing a uranium sample through a collimator with a NaI detector. The enrichment is deduced from the intensity of gamma rays attributed to 235 U (e.g., gamma ray at 186 keV). Under a well-defined geometry, the measured count rate of the 186 keV photons is proportional to the 235 U abundance. Because of the strong attenuating properties of uranium compounds, infinite thickness for the 186 keV gamma rays is required and achieved with rather thin samples (3 mm for uranium metal, 15 mm for UF6). The standardized procedure controls the geometry and utilizes specially designed support stands with collimators to provide a quantitative assessment of enrichment within minutes. The technique works very well for pure and homogeneous uranium materials. Mini multichannel analyzers connected to CdZnTe detectors (MMCC) are the preferred instruments for fresh fuel verification giving more credible results than NaI-based systems (Arlt et al. 1993, 1997). The probe of the CdZnTe based system is less than 1 cm in diameter and can be inserted into the water tube or control rod guide tube of fuel assemblies and can therefore be implemented entirely in situ without any problems arising from interference resulting from radiation emitted by adjacent fuel assemblies. In many instances, qualitative results are sufficient to characterize the nuclear material (e.g., Spent Fuel Attribute Tester (SFAT) to confirm the presence of the 137 Cs peak at 662 keVas Nuclear Safeguards Verification Measurement Techniques 63 2911
  • an attribute for spent fuel). In some cases, the absence of specific gamma rays confirms the absence of a particular nuclear material or distinguishes between nuclear and nonnuclear material (e.g., irradiation attribute tester (IRAT) used to confirm absence of nuclear material in closed containers stored in spent fuel ponds). The handheld monitor (HM-5, Fieldspec) is a battery-powered, digital, low-resolution, gamma spectrometer (Jung et al. 2007). The HM-5 is a lightweight, easy to operate and very frequently used instrument for safeguards purposes. This device combines various functions such as dose rate measurement, source search, isotope identification, enrichment measure- ment. It uses a small scintillation NaI detector and energy selective electronics with a digital display. The HM-5 is a typical instrument for attribute measurements (if background permits) and capable of detecting different energy ranges of gamma rays, allowing qualitative verifica- tion of the presence of either plutonium or uranium in unirradiated nuclear material. The HM-5 is used to measure the active length of light water reactor (LWR) fuel assemblies. The measured length combined with data on uranium mass per unit of length, obtained from a neutron coincidence collar, enables the inspector to determine the total uranium mass in the fuel assembly. Recently, enhancement of the HM-5 has enabled it to measure the enrichment of unirradiated uranium materials. The instrument is also widely used as a key tool during complementary access (CA) activities to confirm the absence of undeclared nuclear materials (> Fig. 63.1). The Fresh MOX Attribute Tester (FMAT) consists of a stainless steel cylinder housing shielding and collimation, a CdZnTe detector and a preamplifier (Aparo et al. 1999). A multiwire cable connects the waterproof measurement cylinder with a data acquisition/ control unit (operated above water). The FMAT is used to verify fresh mixed oxide (MOX) fuel stored in spent fuel ponds awaiting loading to the reactor core. It clearly distinguishes between . Fig. 63.1 Hand-held low-resolution gamma spectrometer (HM-5, Fieldspec) 2912 63 Nuclear Safeguards Verification Measurement Techniques
  • the gamma rays of 235 U (186 keV) and 241 Pu (208 keV) and measures key plutonium gamma rays to evidence that an interrogated item exhibits the unique characteristics of fresh MOX. The Uranium Bottle Verification System (UBVS) is an integrated system combining a weighing scale and NaI detector for 235 U enrichment measurement of reprocessed UO3 in large storage bottles. An inspector determines the weight of the bottle ($1,300 kg for full bottles) utilizing a dedicated scale. The weighing scale platform is sealed in a tamper-proof enclosure and is connected via a tamper-proof conduit to the measurement cabinet where the weighing terminal is stored. An ultrasonic gauge (ULTG) is used to measure the thickness of the bottle wall to allow for the attenuation correction of the subsequent enrichment measure- ment. The enrichment level is measured using either a calibrated HM-5 or a calibrated NaI detector connected to a multichannel analyzer. The acquired spectra are evaluated using a special computer code (NaIGEM). The code calculates the 235 U enrichment by a peak- fitting technique and allows for wall thickness and matrix corrections. A simple evaluation of the 186 keV peak by regions of interest alone would fail due to the strong interference of 228 Th stemming from the 232 U decay. Therefore, the masses for 235 U and U total composition are derived from a combined evaluation of the weighing results, the enrichment and the uranium concentration in UO3. High-Resolution Gamma Spectroscopy (HRGS) High resolution gamma spectroscopy (HRGS) is widely used in safeguards to verify the isotopic composition of plutonium or uranium in unirradiated nuclear material. The isotopic information for plutonium (238 Pu, 239 Pu, 240 Pu, 241 Pu, 242 Pu and 241 Am) is needed for neutron assays to convert measured neutron responses stemming from a limited number of Pu-isotopes (including Am) to total plutonium mass. Similarly, the heat output measured by calorimetry is correlated with the Pu-isotopics to obtain the total mass of Pu. Plutonium emits a complex spectrum of X- and g-rays that are interpreted using dedicated software (Multi-Group Analysis (MGA)) (Gunnink 1990) either embedded in the multichannel analyzer software or as a stand- alone application. MGA mainly exploits the complex 94–104 keV region. MGA calculates the abundances of 238 Pu, 239 Pu, 240 Pu, and 241 Pu (242 Pu has indistinct gamma rays and is estimated from isotopic correlation). The abundances of 241 Am and, if present in the Pu sample, 235 U and 237 Np, and their respective element ratios are determined simultaneously. The TARGA software contains an upgraded version of MGA as an engine, and provides a comparison tool for operator-declared and inspector-measured values. An alternative method – fixed energy response function analysis with multiple efficiencies (FRAM) – considers the higher energies, requiring specific hardware settings (Sampson et al. 2003). This allows for measurements in containers with steel wall thickness of greater than 10 mm. The measurement of uranium samples usually involves only two main isotopes – 235 U and 238 U – providing the uranium enrichment of the sample. The spectral analysis of the g-spectra is performed using a modified MGA code – MGAU (Gunnink et al. 1994). HRGS is used to determine the 235 U enrichment of uranium hexafluoride (UF6) in shipping cylinders. In this case, low-resolution gamma ray systems would fail to provide accurate assays due to the complexity of the gamma spectra caused by the presence of additional radionuclides (e.g., 228 Th) stemming from natural decay chains plated on the container walls. In addition, for feed (Natural Uranium) and tails (Depleted Uranium) cylinders, the signal-to-noise ratio for the 186 keV peak with LRGS exceeds measurement tolerances and, therefore, HRGS is needed. Nuclear Safeguards Verification Measurement Techniques 63 2913
  • High Purity Germanium (HPGe) detectors cooled by liquid nitrogen are the backbone of HRGS applied by Safeguards. The Pu-sample is placed, in its original packaging, on a planar HPGe detector, acquiring a spectrum in the energy range 0–614 keV within 300–1,000 s. The spectrum is then analyzed using MGA code. Typical precisions and accuracies range between 0.5% and 2% reliability for all isotope abundances except 242 Pu. The content of nuclear material in waste drums is usually very low and not homogeneously distributed. However, bulk-handling facilities such as reprocessing and fuel fabrication plants often accumulate a huge number of waste containers that together contain significant amounts of nuclear material. Representative sampling for DA is not feasible and NDA is the only means of determining the content of nuclear material in the waste. This is also true for the determination of holdup in bulk facilities. Although several NDA methods exist to tackle the problem of holdup accounting in bulk facilities, the related measurement uncertainties often exceed the specified goal. The Segmented Gamma Scanner (SGS-IQ3) is a commercially available gamma assay system and uses transmission-corrected passive assay techniques to determine isotopics and to quantify individual isotopes (usually 239 Pu or 235 U) within items of scrap and waste (Booth et al. 1997; Mayer et al. 2004). The SIQ3 system utilizes a 15 cm, 4p low background, steel shield with three collimated HPGe detectors (coaxial) to perform quantitative gamma assay measurements in combination with three additional HPGe detectors (planar, optimized for low energy and high resolution) to measure the plutonium isotopics. Three transmission sources are mounted oppo- site the coaxial detectors to perform the transmission measurements. The quantities are deter- mined both by summing the transmission-corrected result for each detector, and by summing the spectra and performing a quantitative assay based on the average density of the drum. The in situ object counting system (ISOCS) is an intrinsically, numerically calibrated gamma spectrometry system incorporating a well-characterized HPGe detector (Bronson and Young 1997). The system is commercially available and is used to verify nuclear materials, in particular uranium, contained in holdup and waste. The calculation of the efficiency versus energy function is based on user-defined models that take into account all physical parameters describing geometry and sample matrix. The acquired gamma spectrum is analyzed using calibration information. The ISOCS calibration method is a powerful tool enabling calibration of the detector for a wide variety of source geometries and activity distributions. The cascade header enrichment meter (CHEM) is another application that uses HPGe detection to qualitatively confirm the absence of HEU in centrifuge cascade header pipes of enrichment plants (Close et al. 1998). The technique uses an external radiation source (57 Co) and a special collimated HRGS with specific software to control, perform, and evaluate X-ray fluorescence (XRF) and passive measurements. The XRF measurement provides the amount of total uranium in the UF6 gas. The deposit on the inside surface of a header pipe requires two measurements of the 235 U gamma ray at 186 keV with different geometries to determine the amount of 235 U in the gas alone. The level of enrichment of uranium in the gas of the header pipe can then be determined independently of the gas pressure. Gamma/X-Ray/Weighing (GXW) method can simultaneously determine plutonium ele- ment concentration and isotopic composition both in solid and liquid samples from a single HRGS measurement (Dragnev et al. 1997; Parise et al. 2003). It exploits the full spectroscopic information contained in a gamma spectrum from a plutonium sample using several gamma- spectrometric analysis techniques such as enrichment-meter-type measurements and passive X-ray fluorescence analysis (XRF). In combination with weighing, this method determines the total Pu content in a sample. 2914 63 Nuclear Safeguards Verification Measurement Techniques
  • 63.3.4 Neutron Counting Techniques NDA based on neutron measurements plays an important role in the qualitative and quanti- tative analysis of nuclear material, in particular plutonium in bulk and item form. Plutonium samples have a high rate of spontaneous fission neutrons, while uranium samples are typically interrogated using an induced fission neutron signature. Neutrons are primarily emitted from nuclear material in three ways: 1. Spontaneous fission of uranium, plutonium (in particular involving even isotopes of plutonium), and curium (in spent fuel). 2. Induced fission from fissile isotopes of uranium and plutonium, typically by means of a low-energy neutron source. 3. a-particle induced reactions, involving light elements such as oxygen and fluorine. Contamination of the plutonium-bearing materials with other spontaneous fission nuclides (e.g., 244 Cm) will strongly interfere with the neutron assay. 244 Cm is the strongest neutron emitter and accounts for more than 95% of the neutrons in spent fuel although the share of curium is only about 0.5% of the plutonium content. A few ppm of 244 Cm will lead to a significant overestimation of the result. However, 244 Cm is an important ‘‘tagging’’ nuclide that could be used for neutron measurements in combina- tion with known concentration ratios of curium, plutonium and uranium (Rinard and Menlove 1997). This technique can be applied only if chemical processes do not change the element ratio, i.e., there is no separation. This is the case for determining plutonium and uranium composition in selected process wastes, e.g., the leaching process of spent fuel or vitrification process. Falsification of the neutron count by adding a neutron source (such as 252 Cf) could fool the neutron assay equipment. However, the subsequent presence of 252 Cf in the product would be a serious problem and a very strong indicator of attempted spoofing. Main attributes of various neutron detector systems are listed in > Table 63.5. Neutron Detectors Neutrons can only be detected by indirect methods, e.g., via nuclear reactions producing charged particles. The electrical signal produced by the resulting charged particles can then be processed by the detection system. 3 He gas detectors are the most commonly used neutron detectors in safeguards. The detection principle is based on the 3 He (n, p) 3 H reaction. This reaction produces a proton with recoil energy of 732 keV that ionizes the surrounding gas and generates an electronic signal. Thermal neutrons have a high absorption cross section for the 3 He(n, p)3 H reaction. The neutron absorption cross section decreases with orders of magnitude as the neutron energy increases, hence moderation of neutrons is essential to achieving a reasonable detection efficiency of the counting system. 3 He gas detectors have been proven robust and effective. They are commercially available with various diameters, lengths, and gas pressures. BF3 detectors are occasionally used based on the 10 B(n, a)7 Li reaction. BF3 detectors are less sensitive to gamma radiation fields but are less efficient. Recently, solid-state neutron radiation devices with boron carbide diodes have been developed, which demonstrate very promising potential for future applications such as miniaturized handheld neutron detection devices. Nuclear Safeguards Verification Measurement Techniques 63 2915
  • Fission chambers have a thin layer of 235 U plated on the inner wall of a gas-filled chamber. Neutrons will cause fission of 235 U producing high-energy fission fragments ($90 MeV). The fission fragments cause ionization in the stopping gas, which could then be transformed to an electronic signal. The fission chambers have the highest tolerance versus gamma dose rates (up to roughly 104 Gy/h) of any of the available neutron detectors because the short ranging fission fragments deposit a much larger quantity of energy in the stopping gas in comparison to the gamma rays. They are the only neutron detectors capable of measuring highly active spent fuel. The inherently low efficiency of fission chambers is compensated for by the large number of neutrons available for counting. . Table 63.5 Neutron systems System Detector Measurement task Typical performance values (%)a HLNC 3 He Pu in PuO2 0.5(pure)-3(scrap) Pu in MOX 3–5 Pu in pins/assemblies 1 INVS Pu in small samples (powder, pellets, liquids) 1.5 Pu in MOX scrap 2.5 FPAS Pu in fresh MOX pins 2 GBAS Pu in in-process materials 5–10 BNCN Pu in fast critical assembly fuel plates 5 DRNC Pu in fast critical assembly fuel drawers 3 UFBC Pu in Powders, MOX pins and fresh FBR fuel 1–2 UWCC Pu in fresh MOX fuel assemblies (underwater) 2–3 SFCC Pu in spent fast breeder reactor fuel (underwater) 8 WCAS Pu in large waste container 11 WDAS Pu in waste drums 8 PNCC Pu under various field conditions PSMC Pu in impure samples (MOX) 2 ENMC Pu in pure samples (MOX) 0.5 Pu in impure samples (MOX scrap) 2–3 UNCL 235 U in LWR fresh fuel 2–4 AWCC 235 U in HEU, bulk UO2 and pellets 3–5 235 U in LWR fuel 1–5(burnable poison) AEFC 235 U in spent fuel from research reactors 10 SFNC Fission chamber Presence of spent fuel Attribute a Average measurement time $300 s. 2916 63 Nuclear Safeguards Verification Measurement Techniques
  • 3 He and BF3 detectors are sensitive to high gamma radiation fields, which produce a high pile-up and mask the neutron signal. Under such conditions, fission chambers are used. Plastic and liquid (organic) scintillators are often used for fast-neutron detection because of their fast response and modest cost. Their function is based on the elastic scattering of the neutron on light elements (mostly carbon and hydrogen). The proton absorbs the kinetic energy from the neutron, producing heat and visible light. The visible light is collected in a photomultiplier tube coupled to the scintillator and converted to an electronic pulse. However, gamma radiation also produces visible light while interacting with the scintillator. This sensitivity to gamma radiation severely limits the application of scintillators in the selective neutron detection process. Europium-activated lithium iodide (enriched in 6 Li) as a scintillator can detect neutrons and gammas simultaneously (Mukhopadhyaya and Mchugh 2004; Syntfeld et al. 2005). Neutrons are detected via the reaction 6 Li(n, t)a + 4.78 MeV. The high-energy pulses from the neutron events can be well discriminated from the pulses stemming from the interactions with the gamma radiation. Gross Neutron Counting Gross neutron counting for safeguards purposes is applied in searching for undeclared nuclear material and activities, process monitoring, measuring holdup in glove boxes, and in the assay of spent fuel. The handheld neutron monitor (HHNM) is a portable ($4 kg) neutron detection device with three 3 He proportional neutron counters, a GM counter and integrated electronics, which provide a means of searching for and localizing neutron radiation sources. A measurement sequence consists of background and verification measurements. When a predetermined threshold is exceeded, the detector triggers an alarm and records the relevant information. The DG-5 is a portable device for gamma and neutron detection using a liquid scintillator. The presence of neutrons is a strong indicator of nuclear material and activities. Neutrons are difficult to shield by heavy metals in contrast to gamma radiation. They need lightweight elements to lose their energy and be absorbed while gamma radiation would penetrate this neutron shield to a large extent. Therefore, safeguards inspectors use the DG-5 to scan for undeclared nuclear material and activities. The spent fuel plutonium canister counter (SPCC) is designed to operate underwater and to count neutrons emitted from natural uranium spent fuel rods stored in dry stainless steel canisters (Menlove et al. 2002). The counter uses three 3 He detectors located inside watertight and Pb-shielded pipes spaced 120 apart and surrounding a cylindrical cavity. SPCC deter- mines 240 Pu mass from total neutron counts corrected for background, 238 U spontaneous fission and multiplication. The portable neutron uranium holdup (PNUH) monitor system is a neutron counting system to determine the quantity of uranium holdup within the cascade halls of an enrichment plant (Beddingfield and Menlove 2002). PNUH is custom built and the detector head is a polyethylene-moderated assembly with 25 3 He neutron proportional tubes. PNUH measures the total neutron signals at various prescribed locations. Measurement data are acquired with standard INCC software and evaluated with specialized PNUH software using distributed source term analysis (DSTA). Nuclear Safeguards Verification Measurement Techniques 63 2917
  • Neutron Coincidence Counting During fission, multiple neutrons (multiplicity $2–3 for plutonium) are emitted contempo- raneously from the splitting nucleus. Because they arrive in the detector simultaneously, these neutrons can be distinguished from all other neutrons (e.g., those produced by (a, n) reactions or by neutron generators), which are not correlated in time. The fission neutrons are referred to as coincident neutrons. The neutrons emitted by a plutonium-bearing sample consist of spontaneous fission neutrons from even-even plutonium isotopes (238,240,242 Pu), from the interaction of a-particles with low-Z elements (e.g., O, C, F, Al, etc.) and induced fissions from 239,241 Pu (‘‘multiplication effect’’) caused by thermalized spontaneous fission neutrons and (a, n) neutrons. The spontaneous fission rates are sufficiently high to permit direct measure- ment of the fission neutrons (passive neutron counting). Emitted neutrons are moderated and then detected with 3 He tubes. Amptek circuits amplify the output pulses from the 3 He tubes and convert the pulses above a discriminator threshold to digital pulses. The 3 He neutron counters detect all neutrons arising from both (a, n) and fission reactions. Sophisticated pulse processing electronics, called neutron coincidence counting circuits (shift register), measure the number of neutrons that are detected within a predefined time interval (gate width), and differentiate between time correlated (coincidence) neutrons emitted from the fission events and single neutrons created as a result of a-particle interactions. The measured coincident neutrons (doubles) are proportional to the mass of the even-even Pu isotopes (240 Pueffective = 2.43 238 Pu + 240 Pu + 1.69 242 Pu). The absolute plutonium mass is determined from the mass of 240 Pu effective and the isotopic abundances. Induced fissions exhibit higher neutron multiplicity than the spontane- ous fission events; hence, they contribute to the enhancement of coincidence response and introduce nonlinearity in the response for higher amounts. The plutonium content of samples in this type of measurement can typically range from the gram level up to several kilograms. Standard methods have been developed for multiplication corrections. The main parameters characterizing neutron coincidence are the average neutron lifetime (die-away time t, typically $50 ms) and the neutron detection efficiency (e) in terms of their magnitude and uniformity. High detection efficiency is important for coincidence counting because the probabilities of detecting dual and triple coincidences are proportional to e2 and e3 , respectively. A flat spatial efficiency profile is needed to obtain comparable results for different sample positions, dimensions, and filling heights. A short die-away time is important as it minimizes the accidental coincidence count rate from any high background of random neutrons. Both parameters are influenced by the moderator. Reducing the moderator may decrease detection efficiency and die-away time. Less-moderated counter assemblies detect neutrons before thermalization (epithermal counters) and therefore have a significantly reduced die-away time (t % 22 ms). The detector system is calibrated using known standards that are subject to DA sampling. Two calibrations are normally used: (a) ‘‘Passive’’ calibration curve (using measured doubles versus 240 Pu-effective mass). This calibration curve provides the most accurate results for measurement situations where the singles neutron background has a significant uncertainty. This is especially true for small samples where the neutron multiplication is small. For this mode of calibration, a well-matched set of standards is required and the multiplication is built into the calibration curve. 2918 63 Nuclear Safeguards Verification Measurement Techniques
  • (b) ‘‘Known-Alpha’’ multiplication calibration curve can be used when the material is pure. Then the a-factor (ratio of the uncorrelated to spontaneous fission neutron events) is accurately predicted by the isotopic ratios. Impurities, e.g., fluorine, moisture, signifi- cantly change the calculated neutron response. This mode of measurement always has the best statistical precision and the calibration curve is a straight line. Interactive software (e.g., IAEA neutron coincidence counting (INCC)) functions as a hardware interface, controlling setup for the coincidence counting electronics. The software supports the inspector in setting calibration and measurement parameters and inputting sample data (e.g., Pu isotopics, declared mass of plutonium, etc.) based on the operator’s declaration or inspector’s own analysis. The code updates the input to the date of measurement providing decay-corrected values for the operator–inspector data comparison. The actual assay cycle is preceded by a sequence of normalization and background measurement cycles. Each sequence must pass all built-in quality control criteria for acceptable results. When the measurements are completed, the Pu mass is calculated based on the selected calibration parameters and the measured neutron count rates (singles, doubles, and triples) and an operator–inspector difference for the plutonium mass along with propagated errors is recorded. The computer routine facilitates the re-evaluation of results using different mea- surement parameters, e.g., change of calibration curve or using different data sets of isotopics. The neutron measurements are influenced by a number of physical and environmental factors such as filling heights, changes in density, presence of nearby reflectors and additional neutron sources. These factors could change the number of neutrons available to induce fission and hence lead to false coincident events. The same coincidence measurement technique is used for ‘‘active neutron counting techniques’’ whereby the fissile isotopes are irradiated by neutrons from an external source to induce fission. The resulting multiple induced fission neutrons are then measured using standard coincidence counting methods to separate the signals from induced fission neutrons from the signals caused by single neutrons. The active neutron counting technique measures the fissile isotopes of uranium (235 U) and plutonium (239 Pu, 241 Pu). The high-level neutron coincidence counter (HLNC) has a detector head composed of a polyethylene cylinder with an embedded array of 18 3 He tubes (Krick and Menlove 1979). The moderator slows down the energetic neutrons to thermal velocities. The words ‘‘high-level’’ are included in the name because the counting and sorting electronics can perform at a high rate, such as 100,000 cps of single neutrons. The HLNC is specifically designed to measure high spontaneous fission neutron rates from samples containing up to several kilograms of plutonium. The HLNC is the basic model – a whole ‘‘family’’ of instruments exists with various measurement configurations to fit shape and size of the item being measured (Menlove 1983; Menlove et al. 1994). In most instances, these systems are facility resident or integrated into the facility process and can be operated either in attended or unattended mode. The counter is used to measure plutonium in bulk material (e.g., PuO2, mixed PuO2 – UO2 (MOX)) or plutonium in unirradiated MOX fuel assemblies and pins (> Fig. 63.2). The inventory sample counter (INVS) is a small counter within the HLNC family (Miller et al. 1991; Sprinkle et al. 1993) with a high neutron counting efficiency (35%), designed to minimize perturbation from moisture, bagging, or other matrix materials in the sample. It is used to perform high-precision measurements of small plutonium samples, such as process samples, e.g., Pu pellets, powders, and solutions in vials. The samples are weighed with high accuracy ($ Æ0.1 mg) and the plutonium result is then extrapolated to the total item quantity. Nuclear Safeguards Verification Measurement Techniques 63 2919
  • The Universal Fast Breeder Counter (UFBC) is a thermal neutron coincidence counter designed to assay FBR fuel assemblies and other types of plutonium fuel (Menlove et al. 1984). UFBC has 7.0% efficiency using 12 3 He tubes surrounded by polyethylene and a thin cadmium sleeve. The uniform counting (flat response) region in the detector head (141 cm high and 30.5 cm in diameter) is 105 cm. UFBC can measure assemblies with plutonium loadings of up to 16 kg (24% Pu). The fuel pin assay system (FPAS) is used to determine the Pu-content of fresh MOX fuel pins stored on trays with results that correspond to greater than 2% with the declared data (Cowder and Menlove 1982; Miller et al. 1989). FPAS has a relatively flat response for 1.2 m. The system primarily operates in attended mode by the inspector identifying and checking the positioning of the pin tray in the counter. The system could be transformed to an unattended system using a robotic conveyor for the tray delivery/removal in combination with a radiation- triggered ID camera to record the tray identification. The glove box assay system (GBAS) measures the in-process material of plutonium in a specific process glove box (Miller et al. 1989). This system is a very large coincident neutron counting system (160 cm high, 100 cm long, and 7.6 cm wide) that can be positioned and raised to cover the front and back sides of very tall MOX process glove boxes. Six slabs were originally installed in pairs on either side of a glove box. Each slab contains twenty 152 cm 3 He tubes. Monte Carlo calculations were used to design the detector and study its response before installation. GBAS can be moved into place around randomly selected glove boxes containing large amounts of MOX materials in process. Experience has shown a measurement uncertainty of $5% for neutron assay. The birdcage neutron coincidence counter (BCNC) verifies plutonium content in fast critical assembly fuel plates stored in containers called birdcages (Krick et al. 1985a). . Fig. 63.2 High-level neutron coincidence counter (HLNC) 2920 63 Nuclear Safeguards Verification Measurement Techniques
  • A birdcage can store up to 2 kg of plutonium. Neutron coincidence measurements are performed without removing the fuel plates from the birdcages. The custom-designed detector body consists of 20 3 He proportional counter tubes embedded in polyethylene, providing an efficiency of about 5%. The drawer neutron counter (DRNC) has been designed to perform the assay of plutonium in fast critical assembly fuel drawers (Krick and Menlove 1980). Eight tubes (2.5 cm diameter by 91 cm active length) were used in the system. The principal feature of the neutron coincidence detector is a 7 cm by 7 cm by 97 cm detector channel, which provides a uniform neutron detection efficiency of 16% along the central 40 cm of the channel. The underwater coincidence counter (UWCC) is a transportable system for measuring fresh MOX fuel stored under water (Eccleston et al. 1998). It is a modified version of a standard Fork detector (FDET) whereby the ionization and fission chambers have been replaced with sensitive 3 He tubes embedded in a high-density polyethylene measurement head. The UWCC measures neutrons coming from a segment of the MOX fuel in ‘‘multiplication corrected’’ coincidence mode and provides total Pu, once the isotopics and the active fuel length are known. The waste crate assay system (WCAS) measures plutonium content of large waste con- tainers for high- and low-activity waste (Menlove et al. 2001). WCAS is a passive neutron coincidence counter operating in 4p geometry and can work in high radiation fields up to 110 R/h. The system employs a combination of shielded and unshielded 3 He detectors (98 3 He tubes in total) embedded in a polyethylene matrix and has a detection efficiency of $10%. For the low active waste, all tubes are used to determine 240 Pu effective by the coincidence rates. The high active waste measurement utilizes only the shielded tubes (78 mm steel) to obtain 244 Cm contents by the singles rates. The amount of plutonium and 235 U in the wastes are calculated with Cm/Pu ratio and Cm/235 U ratios, known from the stream average ratios at the waste generating sites. WCAS has a small 252 Cf source of known source strength that can be positioned in an automated sequence at a fixed number of locations adjacent to the waste container wall. The measurement with and without source provides a matrix correction factor for a given configuration. WCAS can measure a wide range of plutonium masses from a few milligrams to tens of kilograms within a matrix of 4,000 kg of mixed metal waste. The waste drum assay system (WDAS) measures the residual small plutonium amounts of in-process wastes in 200 liter drums. The system uses a modified neutron coincidence counter with a counter comprising 60 3 He tubes ($20% efficiency) with low background. WDAS applies the add-a-source correction technique that corrects for the effects of the waste matrix on neutrons (Menlove et al. 1993, Menlove 1995). A small 252 Cf source is placed in various positions near the external surface of the sample drum. The changes in the 252 Cf coincidence counting rate provide a matrix correction for the plutonium inside the drum. The portable neutron coincidence counter (PNCC) consists of four individual slab detectors with four 3 He tubes each that can operate in multiple modes and configurations (Thornton et al. 2006). The detector is lightweight and portable ($15 kg) to address flexibility of measurement requirements for various field environments. PNCC has about 12% detection efficiency. Multiplicity Coincidence Counting Normal coincidence counting techniques rely on the detection of two coincident neutrons (doubles) and making an assumption based either on the multiplication or on the (a, n) neutron rate. These two analysis methods – passive calibration and known-alpha – require Nuclear Safeguards Verification Measurement Techniques 63 2921
  • traceable calibration standards and the results can be subject to bias due to changes in sample multiplication or composition; however, they are useful for identifying these changes in the material. Multiplicity counting uses the additional information from events when three coincident neutrons are emitted per fission (triples) (Ensslin et al. 1998). This additional information is obtained from the measurable multiplicity distribution and allows solving all three unknowns, namely, 240 Pu-effective mass, multiplication, and the (a,n) neutron rate. Therefore, the mass of plutonium in the sample can be calculated directly without making any assumptions about its chemical and physical composition. Multiplicity counting can be applied to all plutonium samples, but is beneficial primarily in measuring impure samples. For some material categories (e.g., small Pu samples, process residues with high (a,n) neutron rate) multiplicity counting may not be helpful because of the limited precision of the triple coincidences. Multiplicity counting requires high efficiency as the detected triples rate is proportional to the efficiency cubed. The counters are designed to minimize die-away time and deadtime. Conventional coincidence counters can be used for multiplicity analysis, but their lower efficiencies and longer die-away times lead to very long counting times. The plutonium scrap multiplicity counter (PSMC) is a high-efficiency neutron multiplicity counter with cadmium lining designed for measuring impure samples such as MOX scrap materials (Nakajima et al. 1997). The PSMC contains 80 3 He-filled tubes (at 4-atmosphere fill pressure) arranged in four rings embedded in polyethylene and has an efficiency of $55%. Epithermal-neutron multiplicity counters (ENMCs) use $120 high-pressure 3 He tubes (10 atm) in closely packed rings with less moderator material (Langner et al. 2006; Asano et al. 2006). The higher pressure and the large number of tubes capture more of the thermalized and epithermal neutrons resulting in an efficiency of about 65%. The reduced thickness of high- density polyethylene shortens the neutron die-away time to only 22 ms. This combination improves the statistical precision by a factor of 5–20 compared with traditional thermal- neutron counters depending on the characteristics of the samples (Stewart et al. 2000). Sufficient counting statistics are important for unfolding the multiplicity distribution to determine the triples. In addition, the singles background rate must be accurately measured and shielding used on the outside of the neutron detector to reduce the singles neutron background. The background for doubles and triples is very small and relatively constant. Although designed for multiplicity counting, ENMC is a very powerful ‘‘normal’’ neutron coincidence counter due to its high efficiency and optimized design parameters. In some cases, statistical errors for small samples have been reduced to less than 0.15% using an ENMC; in impure samples, the statistical error dominates the systematic error and can be as high as 3% (over 100 min measurement time). Active Neutron Coincidence Counting The fissile isotopes of uranium (235 U) and plutonium (239 Pu, 241 Pu) can be measured using ‘‘active neutron counting techniques.’’ This technique uses an external neutron source to induce fission in the fissile plutonium and uranium content of the sample. The multiple induced fission neutrons are then measured using standard coincidence counting methods. The technique is mainly applied to determine the mass of 235 U in uranium-bearing samples (from LEU to HEU) in powder, metal, pellets, fresh fuel elements, and waste drums. It can be operated either with or without a cadmium liner (fast or thermal mode). 2922 63 Nuclear Safeguards Verification Measurement Techniques
  • The uranium neutron coincidence coLlar (UNCL) can be operated in either an active or a passive mode to measure the 235 U and the 238 U mass per unit length of LWR fuel assemblies (Menlove 1981). UNCL has four polyethylene slabs that surround a central cavity where a fuel assembly is placed for measurement. 3 He tubes are embedded in three of the four slabs. The fourth polyethylene slab contains a cylindrical well for the AmLi neutron source. The active mode uses neutrons from a low-intensity (5 Â 104 n/s) AmLi neutron source to cause fission of 235 U in the fuel and the neutrons from induced fission in 235 U are counted using coincidence electronics. When no interrogation source is present, the passive neutron-coincidence count rate from spontaneous fission gives a measure of 238 U linear mass density (g/cm). The UNCL measures only one section of the fuel, approximately 40 cm long. The measured coincidence neutron rate is proportional to the 235 U mass per unit length. Together with a measurement of the active length of fuel (e.g., using an HM–5), the UNCL can verify the operator-declared total 235 U content of a fuel assembly. Measurements on pressurized water reactor (PWR) and boiling water reactor (BWR) assemblies have demonstrated that, for measurements of similar fuel assemblies used for calibration, the overall accuracy for assay of 235 U linear density is 1–2%. For absolute measurements, using calibration parameters from other facilities and fuel assemblies, the accuracy is in the range of 2–4%, depending on how closely the standards match the assemblies assayed and how precisely calibration has been performed. When operated in passive mode, the UNCL can be used to confirm 238 U linear density to within 10%. The 235 U sensitivity enables detection of a removal or substitution of three to four rods in a PWR assembly and a single rod in a BWR assembly. Assemblies containing Gd as a neutron poison need to be analyzed using a poison correction factor calculated by Monte Carlo simulations (Menlove and Pieper 1987). The active well coincidence counter (AWCC) is designed to measure 235 U in HEU metals, bulk UO2 materials and light water reactor (LWR) fuel pellets (Krick et al. 1985b; Menlove et al. 1996). The AWCC has 42 3 He tubes in two rings, achieving an efficiency of about 36%. The system uses two AmLi neutron sources, each mounted in one of the end plugs of the assay chamber of the coincidence counter body. The neutron sources induce fissions in isotopes not having significant spontaneous fission activity, e.g., 235 U. The induced-fission neutrons of the sample are quantified with standard coincidence counting techniques. The use of the AWCC to verify total 235 U content provides an important advantage over gamma-ray techniques for some applications (e.g., for large and/or heterogeneous samples). The high neutron penetra- bility permits the AWCC to measure the entire 235 U content of a sample rather than just the surface layer that would be measured using gamma-ray techniques. The AWCC can be operated either with or without a cadmium liner (thermal or fast mode). Although designed for active neutron assay, it is also a good passive neutron counter if the AmLi sources are removed. 63.3.5 Spent Fuel Measurement NDA methods are presently the only feasible means of verifying spent fuel. Several instruments are available for measuring spent fuel assemblies stored underwater. Direct measurement of plutonium and uranium in spent fuel is normally impossible because the primary neutron and gamma signals come from higher actinides (e.g., spontaneous fission neutrons from the 242 Cm and 244 Cm isotopes) and fission products. The measurement of gamma and neutron radiation from spent fuel assemblies can be used to confirm various attributes and to estimate burnup, which can be correlated to the quantity of plutonium in the assembly. The variability of the fuel Nuclear Safeguards Verification Measurement Techniques 63 2923
  • designs is a challenge for the adaptation of these techniques, particular in the area of the quantitative verification of irradiated fuel from research reactors that might contain HEU. The most powerful verification tools for spent fuel in wet storage are the improved Cherenkov viewing device (ICVD), safeguards MOX python (SMOPY) device and the FORK detector (FDET). The ICVD is a nonintrusive instrument based on the detection of Cherenkov light specific to high irradiation effects in water caused by spent fuel. SMOPY and FDET measure total neutron and gamma radiation simultaneously. A passive gamma emission tomograph is currently being tested to be able to detect defects at a pin level. Gamma Methods The spectrometric performances of CdZnTe detectors, their robustness, and simplicity are key to their wide application for the verification of irradiated materials in spent fuel (Lebrun et al. 2000; Lebrun and Carchon 2003). High-resolution gamma spectrometric measurements from fission products have also been used to estimate burnup. The fission product nuclide must be long-lived in comparison to the fuel irradiation time. The measured gamma-ray intensity must be corrected for cooling time decay; if the cooling time is not known, it can be estimated from the relative activities of various fission products. Errors can be greatly reduced by measuring the ratio of the activities of two nuclides having a known dependence on burnup. Although capable of giving good estimates of burnup, this method is time consuming, because of the need to place detector systems under- water and to move each fuel assembly to the detector position for verification to be performed. The spent fuel attribute tester (SFAT) is a routine inspection instrument for gross defect detection. It provides a qualitative verification of the presence of spent fuel through the detection of particular fission product gamma rays – either from 137 Cs (662 keV) for fuel which has cooled for longer than 4 years or from 144 Pr (2,182 keV) for fuel with shorter cooling time (Carrasco et al. 1997). The equipment comprises a stainless steel watertight housing containing a collimated NaI, CdTe, or CdZnTe detector that can be submerged in a storage pond. SFAT takes measurements from the top of a fuel assembly as it sits on top of the underwater storage rack and does not require movement of the fuel. The SFAT is applied where Cherenkov viewing cannot provide conclusive verification (e.g., for low burnup or long cooling time, where the Cherenkov radiation is too weak or where the water in the storage pond is not clear enough). The irradiated item attribute tester (IRAT) differentiates irradiated nonfuel items from irradiated fuel items that are stored in spent fuel storage ponds (Aparo et al. 1999). It consists of a stainless steel cylinder housing shielding and collimation, a miniature CdZnTe detector, and a preamplifier. A multiwire cable connects the measurement cylinder (watertight) and data acquisition/control unit (operated above water). The IRAT detects gamma radiation characteristic of either fission products contained in spent fuel or activation products contained in irradiated structural materials. The IRATuses the detection of key fission product gamma rays as evidence that an item being measured has the characteristics of spent fuel. A statistical test compares the measured gamma ray spectrum to a background measurement in order to determine if a certain gamma emitter is present. The presence of such fission/ activation products serves as evidence that an item once underwent fission (137 Cs for cooling time > 4 years; 144 Pr, 137 Cs, 134 Cs, or 95 Zr/Nb cooling time < 4 years) or, in the case of a structural item, was once exposed to a significant neutron flux (60 Co). The measurement with 2924 63 Nuclear Safeguards Verification Measurement Techniques
  • IRAT requires movement of the spent fuel as the detector approaches the item from the side. IRAT is very similar to FMAT with differences in the collimator part and the type of CdZnTe detector (> Fig. 63.3). The high-sensitivity gamma monitor (HSGM) is an enhanced underwater survey meter, consisting of a Geiger–Mueller probe and a microprocessor unit. It measures gamma and X-rays in the 60 keV to 3 MeV range over a total dose rate range of 1 mSv/h (0.1 mR/h) to 300 Sv/h (30,000 R/h). The HSGM is powered by an internal chargeable battery (or by facility power if the battery is discharged beyond a certain level). The technique provides a qualitative confirmation of irradiated items. The method cannot readily distinguish between irradiated nonfuel items and genuine spent fuel items, as it only detects the radioactivity of each item. It does not provide any information on burnup, cooling time, plutonium content, or any other quantitative property. In contrast to ICVD or SFAT, each item to be verified must be isolated from neighboring items. Thus, movement of items is a requisite unless a sufficient separation already exists. The CANDU (CANada Deuterium Uranium) spent fuel bundle verifier for baskets (CBVB) employs a highly collimated and shielded CdTe, suspended on an automatic winch whose speed can be set for scanning either storage baskets or stacks. It verifies the presence of irradiated CANDU fuel bundles stacked in baskets under water. The CANDU bundle verifier for stacks (CBVS) moves vertically along the space between columns of trays (10 cm gap) of spent fuel and uses a CdZnTe detector for bundle identification (Ahmed et al. 2001). It employs thick lead shielding to protect the electronics and detector. The CBVS is unable to verify the spent fuel at the bottom layer of a stack due to limited accessibility of a large-size detector part through the funnel structure. During inspection, the tray must be moved. In addition, the large size of the scanning part is both heavy and difficult to handle. Neutron Methods The advanced experimental fuel counter (AEFC) is used for the characterization of spent fuel from research reactors stored underwater (Menlove et al. 2007). The AEFC can be operated in . Fig. 63.3 The irradiated item attribute tester (IRAT) Nuclear Safeguards Verification Measurement Techniques 63 2925
  • either passive neutron mode or active neutron mode, using an AmLi neutron source to generate fissions in the fuel item. The counter is transportable ($70 kg) and consists of a polyethylene moderator containing six boron-lined 3 He tubes with a lead shield surrounding the measurement cavity (117 mm in diameter) to reduce the gamma dose to acceptable levels. The radiation tolerant 3 He tubes are embedded in the moderator forming two different measurement sets. One set (inner row) measures neutron coincidences to distinguish fission neutrons from background. The second set (outer row) is placed further back within the polyethylene moderator and is therefore much less sensitive to AmLi neutrons than to fission neutrons, so that the signal is approximately proportional to the fission rate in the fuel item. The detector also contains a collimated ionization chamber to record the gamma emission profile correlated with the relative burnup of the measured item. The spent fuel coincident counter (SFCC) is an underwater neutron coincident counter for the verification of operator-declared plutonium content in canned fast breeder reactor spent fuel (Bytchkov et al. 2001; Lestone et al. 2002). The SFCC is hermetically sealed; it operates $5 m below water level in a fixed position in a spent fuel storage pond. SFCC has a ring of 20 3 He tubes embedded in polyethylene ($15% detection efficiency) and shielded from fission product gamma rays by a 7 cm thick lead ring. A single ionization chamber measures the gamma-ray dose from the spent fuel to determine the appropriate operational parameters to avoid gamma-ray pile effects in the 3 He tubes. The plutonium isotopics are calculated based upon validated burnup chains code and specially developed iterative software in combination with MCNP modeling converts the measured single and double neutron count rates to plutonium mass. An inspector is guided by the software through all measurement, evaluation and decision making (accepted/rejected) processes. The SFCC easily distinguishes irradiated fuel from nonfuel items, which are loaded to the reactor to replace discharged assemblies. The spent fuel neutron counter (SFNC) is a prototype neutron-detector system that verifies closely packed spent fuel assemblies stored in a spent fuel pond (Ham et al. 2002). The system contains a fission chamber moderated by a polyethylene cylinder housed in a watertight stainless steel enclosure. The SFNC measures total neutron signals from long-cooled spent fuel assemblies while in their storage position, without requiring them to be moved. The technique can detect a missing fuel assembly. These measurements are performed underwater in a gap between four assemblies. Combined Gamma/Neutron Methods The ratio of neutron to gamma ray data, when combined with other complementary infor- mation, is used to characterize a particular type of fuel. To simplify the verification of spent fuel, a technique involving gross gamma ray and neutron measurements can be used. The Fork DETector (FDET) measures gross gamma and neutrons from a spent fuel assembly, which can be correlated with the operator’s declared data on burnup and cooling time (Halbig et al. 1985). Separate detector heads are used to measure BWR- fuel and PWR- type fuel. The system is operated in the spent fuel pond and consists of a microprocessor- controlled, battery-operated electronics unit, a preamplifier, 5–6 meter long pipes to support the detector underwater, and a special watertight measuring head. This head has two pairs of fission chambers for neutron counting and a single pair of ionization chambers for the total gamma flux, which form a fork-type measurement cavity. The pairing of detectors on two sides of an assembly minimizes the sensitivity to assembly-detector geometry and burnup 2926 63 Nuclear Safeguards Verification Measurement Techniques
  • asymmetries. The operator has to move the fuel assembly to the measurement cavity. Interac- tive software prompts the user through the measurement procedure and simultaneously collects neutron and gamma data. The software can also support unattended measurements. The safeguards MOX Python (SMOPY) device combines gross neutron counting with low- resolution gamma spectroscopy to characterize any kind of spent fuel (Lebrun et al. 2001). The SMOPY uses online interpretation tools for the evaluation of measurement data. The system contains a well-shielded and collimated CdZnTe gamma detector and a fission chamber. It is placed over the storage hole of the spent fuel assembly. The assembly is lifted through the open measurement cavity and it can either be scanned or selective parts measured. The SMOPY can verify and distinguish irradiated MOX fuel from LEU fuel and can confirm the burnup of a spent fuel assembly. The SMOPY device can also be operated in active mode using an AmLi source. This has been successfully demonstrated for the underwater verification of canisters containing residues of irradiated HEU. This application is based on total neutron counting and detects the difference between active background and active background plus induced fissions. The cask radiation profiling system (CRPS) records radiation profiles (‘‘fingerprints’’) from spent fuel storage containers at the time of positioning casks in the dry silo and compares them with a fingerprint taken at the time of subsequent verification (Thevenon et al. 2008). A collimated and calibrated CdZnTe detects gamma spectra in scanning mode while ascending down the verification tube at constant speed. The verification tube is adjacent to the silo and ranges from the top to the bottom. The position information of the detector is provided by a pulser, which emits a signal for every constant interval rotation of the motor’s winch wheel that drives the detector within the verification tube. All raw data are acquired and processed with dedicated software on the inspector’s laptop. The CRPS provides a tool to reverify a dry storage cask in its storage position to demon- strate that the cask content has not changed (in the event of loss of continuity of knowledge) or as a part of periodic routine requirements for reverification. The maintenance and/or restora- tion of continuity of knowledge of a spent fuel dry storage container by a reproducible fingerprint require a systematic management of fingerprints over a long period. A database has been developed for the storage and evaluation of fingerprints to secure and effectively compare fingerprints while taking into account decay and changes in the measurement hardware configuration. Automatic unfolding of the radiation profiles to count the number of baskets loaded in the storages is also possible. The CRPS could be run with a pair of detectors to perform neutron (fission chamber) and gamma profiling. Optical Methods The improved Cherenkov viewing device (ICVD) is the instrument most commonly used by safeguards inspectors to verify spent fuel (Trepte et al. 1996). Observation of the Cherenkov radiation from irradiated reactor fuel assemblies is used to obtain qualitative confirmation of the presence of spent fuel in storage pools. Cherenkov radiation, which results from the interaction of the intense beta radiation from spent fuel with the water in the storage pool, varies in intensity and – for highly active fuel – can be seen in darkness with the naked eye. For fuel with low burnup and/or long cooling times, the Cherenkov intensity is very low, but can be seen using a light-amplifying capability in the night vision device. The ICVD employs light filtering techniques (e.g., ultraviolet filters). Special photo cathodes have been developed with Nuclear Safeguards Verification Measurement Techniques 63 2927
  • appropriate spectral sensitivity for Cherenkov light to minimize background light interference and even permit use of the ICVD under normal facility lighting. Observation of the Cherenkov radiation from irradiated reactor fuel assemblies is used to obtain qualitative confirmation (attribute testing) of the presence of spent fuel in storage by scanning rows of assemblies from the pool bridge. Characteristic patterns formed by the arrangements of rods and holes in fuel assemblies have to be observed to ensure a valid verification. A well-trained inspector can easily detect the presence of an inactive ‘‘dummy assembly’’ surrounded by highly active neighbors. ICVD cannot be used when fuel is stored in canisters. Furthermore, absorption and reflection of the Cherenkov light caused by additives or particles in the water and its turbulence in spent fuel ponds may result in failure to verify the spent fuel. It is also difficult to see fuel assemblies that are double-stacked in the storage pool. The digital Cherenkov viewing device (DCVD) is used to verify assemblies with long cooling times and/or low burnups, which have weak Cherenkov signals that cannot be seen with a standard ICVD (Chen et al. 2006a). Apart from its higher sensitivity, the DCVD can record and document individual scans for subsequent reanalysis. It has the potential to quantify the Cherenkov glow from spent fuel assemblies as a function of irradiation history and cooling time. Further development efforts to enable partial defect detection are promising (Chen et al. 2006b). The optical fiber radiation probe system (OFPS) is used to verify irradiated bundles in the spent fuel bay of CANDUstations (Kim et al. 2006). The OFPS consists of a scanning actuator, an optical fiber scintillator coupled to a flexible optical fiber, data acquisition electronics, and a PC. The use of an optical fiber scintillator for CANDU spent fuel verification has the benefit of detecting gross gamma rays in storage ponds without being hindered by the funnel structure. Gamma rays from the spent fuel interact with the optical fiber scintillation media to produce ionization, which subsequently leads to the emission of fluorescent light ($400 nm) of the doped Ce3+ in the optical fiber. Since the optical fiber scintillator is both highly resistant to radiation and able to withstand high temperature and humidity, more precise and safe measurement is possible in between bundles of a tray. In high radiation fields, the optical fiber cable itself emits light from radiation interactions. Therefore, the signal from the cable has to be compensated by a second optical fiber cable (without scintillation media). The detector can be placed between bundles (1.5 cm gap). The verification system performs gross gamma measurements supporting the re-verification of CANDU spent fuel bundles stored in ponds without requiring movement of the horizontal storage trays. 63.3.6 Unattended NDA Systems Modern nuclear facilities are increasingly automated with the aim of decreasing both personnel exposure and production costs. Owing to automation, direct access to nuclear material may be limited for both the operator and the inspectorate. Unattended NDA systems generate mea- surement data without requiring the presence of an inspector. The data acquired are routinely collected from measurement cabinets after a certain period or are transmitted remotely to an external location (such as an inspector’s office). Main attributes of various unattended NDA systems are listed in > Table 63.6. A combination of containment and surveillance (C/S) data with synchronized NDA data is a very powerful tool to monitor nuclear processes while C/S or NDA data in isolation have less 2928 63 Nuclear Safeguards Verification Measurement Techniques
  • . Table 63.6 Unattended NDA systems System Detector Main measurement task Typical performance values (%)a VIFM VIFB: solid state Si Counts SF bundles ( fuelling machine ! SF pond) Attribute VIFC: solid- state Si & FC Counts SF bundles from core to fuelling machine Attribute SEGM Si-Pin Monitors cask loading into dry storage silo Attribute ISVS Xe-IC + 3 He CoK spent fuel receipts to spent fuel pond Attribute IHVS IC + 3 He CoK of spent fuel from pond to dissolver Attribute PCAS 3 He Pu in PuO2, MOX in canister 1–2 IPCA 3 He Pu in MOX in canister 0.85 3 HPGe 240 Pu-effective (%), U/Pu ratio (A)MAGB 3 He Pu powder/pellets/scrap in process container 4.5 HPGe 240 Pu-effective (%), U/Pu ratio FAAS 3 He Pu in final fresh MOX assemblies 1.8 VWCC 3 He U, Pu in vitrified highly active waste via 244 Cm 5–25 VCAS FC + IC U, Pu in vitrified highly active waste via 244 Cm 5–25 ENGM 24 3 He Pu in fresh MOX (receipt) Attribute CCRM 3 He + NaI CoK of fresh MOX fuel between ENGM and EVRM Attribute EVRM 3 He + NaI CoK of MOX fuel between EVRM and core Attribute EXGM 10 B + 2 IC CoK of spent MOX fuel to and from spent fuel pond Attribute PIMS 142 3 He Pu in powder process area 6 RHMS 3 3 He Pu, U in leached hulls via 244 Cm 5–25 TCVS 3 He Pu in MOX canisters temporary stored in process <6 HMMS 3 He Pu, U in hulls via 244 Cm 5–25 DCPD 3 He CoK of Pu canister movements Attribute DSNC 3 He Pu, U in high active, dry reprocessing material via 244 Cm 5 DSNM 3 He CoK of rear access ports of glove boxes Attribute ATPM Temperature/ flow Power monitoring of research reactors Attribute MUND 3 He Monitoring of fuel transfer casks Attribute SMMS Pressure/ temperature Volume determination/balancing of solutions 0.05 a Average measurement time ~300 s. Nuclear Safeguards Verification Measurement Techniques 63 2929
  • value. These systems help to decrease overall inspection effort and are capable of monitoring routine activities involving nuclear material, such as the complete process of loading, trans- ferring, and storing spent fuel in dry storage silos. The main drawback of these automated systems is their predictability. In most safeguards approaches, this drawback is compensated for by unannounced or short-notice inspections on a random basis. Most unattended NDA systems are part of the HLNC family. Other systems use low-resolution gamma detectors either in combination or as a stand-alone application. Recently, a large reprocessing facility has been commissioned and the safeguards instru- mentation in place has set a new technological standard in terms of the networking and integration of different unattended verification systems and C/S systems covering the complete process and storage areas (Johnson et al. 2001, 2004). Unattended Gamma-Based NDA Systems The VXI integrated fuel monitor (VIFM) is used at CANDU stations to monitor and count fuel bundles discharged from the reactor (Bot et al. 1997; Truong et al. 2001). The system employs radiation hardened solid-state gamma detectors with self-authenticating/tamper-indicating circuitry. The VIFM uses an autonomous data acquisition module (ADAM) to acquire data from the radiation sensors. High operational reliability, great dynamic detection sensitivity (to include all operational possibilities), and insensitivity to power outages are some of the important features of the system. The VIFM has three subsystems: ● A VIFM core discharge monitor (VIFC) detects discharges of spent fuel from a reactor core into the fuelling machine. Both neutron (normal on-power discharge signal) and gamma ray intensities are continuously monitored by solid-state silicon gamma detectors and an array of fission chambers for neutron counting. High dose ($2 Mrad) tolerant electronics are employed to withstand the strong radiation fields. An inspector, evaluating the monitoring data, is able to identify in a straightforward, unambiguous manner the abrupt but characteristic changes in count rate associated with fuel bundle discharges, regardless of the operational status of the reactor. Because of the linear increase in background signal, the system can also track the operating power level of the reactor. ● A VIFM bundle counter (VIFB) counts irradiated fuel bundles as they are transferred between the fuelling machine and the spent fuel bay(s). Collimated gross gamma counters detect each fuel bundle as it passes. The proper placement of detectors and the use of the appropriate evaluation algorithm for the facility enable the device to count the bundles as they pass and, to record the direction in which the bundles are moving. ● A VIFM Yes/No monitor (VIFD) determines if any irradiated fuel has been discharged through access ports that are not part of the normal discharge path. The silo entry gamma monitor (SEGM) monitors the final loading of dry storage con- tainers into a final silo storage location (Zendel et al. 2006). The SEGM consists of a pair of pin- diode gamma silicon detectors, which are inserted at different levels into the verification tubes available for each silo, thereby providing direction-sensitive verification of the silo loading. The detectors are installed before the start of the transfer campaign in all of the silos that will be used for that campaign. The cables from several silos (up to 8) are routed to a common electronics cabinet, where the resulting data is logged. The maximum length of the signal cables connecting the electronic cabinet to the detectors is 200 m. The data can be extracted either 2930 63 Nuclear Safeguards Verification Measurement Techniques
  • locally or remotely via a standard Ethernet link. The detectors are removed by inspectors and replaced by seals after each silo has been completely filled and the top cover welded. The integrated spent fuel verification system (ISVS) consists of time synchronized CCTV cameras and radiation detectors, which verify the receipt of spent fuel assemblies received into a facility, unloaded from the transport cask and transferred into storage ponds (Yokota et al. 1998). The radiation detectors, a high-pressure Xe ionization chamber for gammas and a 3 He tube for neutrons, are assembled with preamplifiers and shielding inside a watertight tube. Each channel is equipped with two tubes at each side to provide redundancy for improved system reliability and to enable verification of the direction of spent fuel moving into the temporary storage pit or returning to the unloading pit. The integrated head-end verification system (IHVS) is intended to maintain continuity of knowledge on spent fuel assemblies moving from the spent fuel storage area to the shear- ing machine, as well as the removal of hulls drums to waste storage (Yoshida et al. 2005). The system employs radiation-tolerant camera(s) and radiation detectors (CRD) contained in a cylindrical body, which is installed into the shielding wall itself and also contains internal shielding to reduce the dose rate to meet requirements at the exterior of the cell. Each CRD has one 3 He neutron detector and two ion chambers to detect the radiation of a spent fuel assembly in a respective cell. The cameras provide identification and monitoring of transfer routes. The IHVS comprises 14 CRD devices covering the feeding, sharing, and hulls drumming cells. The continuous enrichment monitor (CEMO) monitors the absence of HEU production in selected gaseous centrifuge facilities and delivers qualitative Go-No-Go information to the inspectorate (Packer 1991; Packer et al. 1997). The CEMO determines the content of U-235 from the intensity of the 186 keV peak using NaI detectors fixed on the product header pipes and correlating the pressure of the gaseous UF6 with a transmission measurement (radioactive X-ray transmission source: 109 Cd ($20–50 keV)). The enrichment is then calculated based on these two parameters. The system operates continuously and transmits remotely (twice a day) state of health and alarm messages in the event that LEU is not confirmed or when a system fault occurs. Unattended Neutron-Based NDA Systems Continuously operating, unattended and custom-designed NDA systems are being implemented at plutonium handling facilities (e.g., reprocessing, conversion, and fabrication plants) to monitor and determine the plutonium content in various process stages. The main components of these systems are modified HLNC counters tailored to the specification of the material container and process environment. In general, the detectors operate in continuous mode with data dumps every minute. Radiation triggered ID cameras generate time stamped video records, which the inspector can correlate with the raw measurement data. The plutonium canister assay system (PCAS) determines the content of plutonium in MOX and pure oxide powders in cans contained in a specific transport container (four cans per canister) (Menlove et al. 1986). The system can be integrated into an operator’s material handling system and continuous measurement cycles are performed, which are evaluated at the end of a given collection period. Thereby, continuous verification of the flow of canisters can be performed without inspectorate intervention. For inventory verification, an inspector provides an electronic list of canisters to be verified and the operator transfers the selected Nuclear Safeguards Verification Measurement Techniques 63 2931
  • canister to the PCAS during shift work without an inspector being present. At the end of a verification campaign, the inspector collects the verification data for evaluation. The improved plutonium canister assay system (IPCA) is a neutron coincidence detector with provision for the simultaneous measurement of Pu isotopics by HRGS (Abhold et al. 2001). The neutron detector uses a double ring of 3 He tubes for the determination of 240 Pu effective mass contained in MOX canisters. The HRGS includes three Ge detectors that determine isotopic composition including the U/Pu ratio. A digital camera, automatically triggered by neutrons, records the canister ID during these measurements. The IPCA is designed to determine the mass of plutonium and uranium in MOX canisters with an uncertainty better than 0.85% for a range of 5–16 kg of plutonium. This performance significantly reduces the random sample size for destructive analysis. The material accountancy glove box counter (MAGB) provides the plutonium content in specific process containers, handled in the glove boxes of the automated process (Menlove et al. 1993). The system consists of two slab detectors viewing the load cell of a process glove box, where the container filled with either feed powder, pellets, or scrap is placed. A radiation- triggered camera identifies the process container and creates a time stamped video ID record. An upgraded version (AMAGB) employs an HPGe system for the determination of isotopics and U/Pu ratio. The MOX fuel assembly capsule assay system (FAAS) determines plutonium content in the final assembly contained in a storage capsule. Coupled to the automated capsule transfer system, it provides information about the movements of fuel into and out of the product storage (Menlove et al. 1993). It is designed to assay the complete active zone of the assembly with plutonium loadings up to 10 kg and can accommodate 5 m long capsules that contain the fuel assemblies. The unshielded detector body has 12 3 He tubes and an efficiency of $16%. In addition, the continuous mode gives a time history of movements of neutron source material in the vicinity. The FAAS is augmented with a surveillance system to meet verification requirements. The vitrified waste coincidence counter (VWCC) is used at a vitrification facility to quantify plutonium and uranium present in canisters of vitrified highly active nuclear waste that is not suitable for further nuclear use (Beddingfield et al. 1998). The system contains five 3 He tubes shielded by 100 mm of lead. The efficiency is 0.5–1.0% depending on the distance between the detector and the canister. The VWCC calibration is based upon Monte Carlo computer model- ing. The upper limit for Pu as a measured discard in vitrified waste is in the order of 5 g in a 10 kg matrix, which can be easily measured by the VWCC. The individual components are all radiation hardened to allow for proper operation in this strong radiation environment. The coincidence counting rates are used to determine the amount of 244 Cm, which is the dominating neutron contributor. The inspector will then apply the Pu/Cm and U/Cm mass ratios that have been analytically determined for each batch from an aliquot taken from the input tank of the melter. These ratios do not change during the vitrification process and hence Cm can be used as a tag for the mass of plutonium and uranium. The total rates will be used to verify sample-to-sample uniformity. The vitrified waste canister assay system (VCAS) is intended to determine the residual uranium and plutonium content in canisters of vitrified high-level spent-fuel reprocessing waste prior to the termination of safeguards on this material. It consists of five neutron detectors (two 235 U fission chambers, two 238 U chambers and a bare 235 U chamber sensitive to thermal neutrons) and one gamma detector (ionization chamber, meant to authenticate the presence of gamma radiation). In contrast to the VWCC, the VCAS uses 235 U fission chambers 2932 63 Nuclear Safeguards Verification Measurement Techniques
  • to determine the 244 Cm content by singles (totals) neutron counting. The plutonium and uranium contents are calculated from Pu/Cm and U/Cm ratios (determined in the feed solution to the melter). In addition, vitrification of the waste solution is verified by measure- ment of the neutron spectrum (using the ratio of counting rates in 235 U and 238 U fission chambers). The fill level of the canister is determined by a comparison of count rates from top and bottom 235 U fission chambers. The neutron radiation data will be integrated with cameras that monitor the measurement station and verify the canister ID. The entrance gate monitor (ENGM) is a passive neutron coincidence collar detector permanently installed at the entrance to the fresh fuel transfer route of a fast breeder reactor (Hashimoto et al. 1994; Iwamoto et al. 1997). The detector head has four groups of 3 He with six tubes each. Fresh fuel assemblies entering the reactor facility must pass through the ENGM so that their Pu content can be verified. Therefore, the ENGM is the system that verifies the amount of fresh fissile fuel in an assembly and serves as the first detector in a sequence of detector systems that follow the movement of fuel assemblies within the reactor facility. The cask car radiation monitor (CCRM) consists of a 3 He detector and NaI detector installed close to the surface of the cask car. By measuring neutron and gamma doses, the monitor is able to distinguish between fresh fuel and spent fuel as well as between other items (e.g., irradiated reflectors and control rods). The monitor controls the flow of fresh fuel between ENGM and the ex-vessel transfer machine radiation monitor (EVRM) and the flow of spent fuel from the EVRM to the spent fuel storage via the exit gate monitor (EXGM). It further monitors any movements of spent fuel retrieved from the spent fuel pond via the EXGM. The EVRM has the same detection features as the CCRM. It controls the flow of assemblies to and from the core. The EXGM is a watertight detector assembly including a lead shielded 10 B detector and two ionization chambers located underwater, close to the spent fuel transfer equipment, where spent fuel is taken to the storage racks. The direction of the transfer is monitored by two gamma detectors, placed above and below the neutron detector. The plutonium inventory monitoring system (PIMS) is a network of 142 3 He neutron detectors in moderating enclosures, which are installed in a plutonium-powder process area at fixed positions outside of process glove boxes and the ventilation system (Simpson et al. 1998; Parvin 2007; Whitehouse et al. 2004). The collected neutron counts are processed in a matrix approach to image the neutron field of the process area. Any change to the in-process inventory will be detected and can be accounted for on a near real-time basis. The PIMS will also be used to verify the clean out and to measure any residual material. The PIMS is operator-owned equipment used jointly with the inspectorate. Appropriate authentication measures are there- fore in place to validate the measurement results. The Rokkasho hull monitor system (RHMS) consists of three 3 He detectors embedded in the operator’s hull monitor system and determines uranium and plutonium content in canisters of leached hulls at the Rokkasho Reprocessing Plant. The operator uses a sophisticated system with both passive and active interrogation. The safeguards detectors are operated independently of the operator’s system, using the passive phase of the measurement cycle to register total neutrons (singles) from 244 Cm. The plutonium and uranium content can then be calculated from the Pu/Cm and U/Cm ratios derived from the associated dissolver solution. The temporary canister verification system (TCVS) verifies the inventory of plutonium in MOX canisters in temporary storage positions located in glove boxes. It consists of NDA measurement stations on both sides of the storage glove boxes using an array of 3 He tubes. The TCVS is capable of quantifying 0.1–30 kg Pu in MOX powder and continuously records Nuclear Safeguards Verification Measurement Techniques 63 2933
  • data to monitor the movement of canisters (e.g., actual storage position and whether a full canister is coming from a different process step). Quantitative results are needed at specific inventory times achieving measurement uncertainties for partial defect testing (i.e., better than 6% at 1 sigma). The hulls monitoring and measurement system (HMMS) measures the neutron emitted from spontaneous fission of 244 Cm in the hull drum, and calculates the amount of Pu and U contained in the hull drum using the ratios of 244 Cm/Pu and 244 Cm/U arrived at by applying a calculation code. The directional canister passage detector (DCPD) is intended to verify the movement of MOX canisters. Each monitor comprises a double neutron 3 He tube in a polyethylene mod- erator supported by surveillance. The system determines whether a canister follows the normal process route (e.g., when and where the MOX canister came from and when the canister left the process area). All other possible movement combinations should be flagged by the system. The Direct Use of PWR spent fuel in CANDU (DUPIC) safeguards neutron counter (DSNC) is a well-type neutron coincidence counter with 18 3 He detectors configured with appropriate shielding, which measures all types of highly active materials from the dry reprocessing process for CANDU bundles (Menlove et al. 1997). It derives plutonium and uranium contents from the measured 244 Cm contents. Plutonium- and uranium-curium ratios are determined from burnup calculations (Origen code) validated by HRGS scanning. The DSNC is operated remotely in a hot cell. The DUPIC safeguards neutron monitor (DSNM) monitors the rear access ports of the process glove boxes through total neutron monitoring using 3 He detectors (Kim 2007). The mobile unit for neutron detection (MUND) is an all-in-one neutron detection system for data collection and storage capable of running on battery power (Zendel et al. 2006). The unit is based on a 3 He detector mounted inside a polyethylene moderator slab. The preampli- fier, HV supply, discriminator, recording electronics, and battery are all integrated inside a single, sealable enclosure. Two redundant data loggers record the data produced by the detector. The MUND collects data for more than 8 weeks without service. For servicing purposes, the device is removed from its measurement location, data are downloaded from the internal electronics, and batteries are recharged. The system has been specifically designed to perform the unattended monitoring of fuel transfer flasks. It is not intended for use in high- radiation environments. Once installed, an MUND is usually serviced by replacing the unit with a fully recharged one. The mobile monitoring system for container transport (MMCT) consists of radiation monitoring, video surveillance, GPS location equipment, and smart power management to monitor the transfer of spent fuel via railcar (Lucero et al. 2004). The radiation sensors detect the spent fuel assembly during the loading, transfer and unloading process. The detector enclosure contains the detector assembly, which comprises six 3 He neutron tubes and two ionization chambers. The system is capable of acquiring safeguards data continuously for more than 1 week under harsh environmental, outdoor conditions (e.g., À20 C) without requiring recharging of the batteries. Other Unattended NDA Systems The advanced thermo-hydraulic power monitor (ATPM) is a monitoring system that can measure the actual thermal power produced by a reactor to confirm the declared operation in 2934 63 Nuclear Safeguards Verification Measurement Techniques
  • nuclear research reactors (Araujo et al. 2001). The ATPM is a complete system comprising ultrasonic flow and temperature sensors mounted on primary cooling loop elements, together with a data collection and evaluation system housed in a tamper-indicating 1900 industrial cabinet, which can be located up to 300 meters from the sensor location. The ATPM measures the reactor–coolant inlet and outlet temperatures and the coolant flow through the reactor core. These parameters are used to calculate the energy flow rate and the total energy produced in the reactor. The result of this calculation is then used to determine if substantial amounts of fissile material might have been generated in the reactor or, to confirm its declared operation. The solution measurement and monitoring system (SMMS) uses high-accuracy and authenticated pressure and temperature measurement devices installed on selected process tanks to determine the volume, density, and temperature of the respective solution (Ehinger et al. 2004). The instruments are connected directly to the pneumatic dip-tube measurement lines of the tanks. The monitoring data exhibits temporal changes caused, for example, by filling, holding, sparging, sampling, and transferring activities. The system evaluation software dis- criminates between these characteristic changes and other disturbances and alerts the inspector to any unexpected operations. SMMS provides continuity of knowledge for plutonium-bearing materials in the solution process and supports the verification of nuclear material inventory, inventory changes, and other transfers. The SMMS software automatically decides whether a process follows its declared operation based upon the observed volume, density, and temper- ature data. The IPCA Load Cell (IPLC) is an unattended load cell system intended to measure independently, the weight of MOX powder canisters with a relative accuracy of greater than 0.05%. The IPLC includes three load cells located on top of the IPCA system – raw data is collected and processed through an amplifier and dedicated software. The typical weights to be measured are between 200–300 kg. 63.3.7 Other NDA Techniques This category of instrument comprises systems that measure physical properties such as weight, volume, absorption of light, sound velocity and heat. Weighing is performed to determine the net weight of an item or batch that is selected as a sample for verification. It is normally applied in conjunction with other accountancy verification methods in order to determine the quantity of nuclear material within the sample. Volume and concentration measurements, in conjunction with other verification methods, are carried out extensively in reprocessing plants to derive amounts of materials contained in solutions. An ultrasonic thickness gauge determines the thickness of container walls and is needed to correct for gamma ray attenuation. Measurement of decay heat can indicate the presence of spent fuel or, when measured in a calorimeter, can quantify pure plutonium samples. Laser-based technologies are becoming increasingly important, providing new and novel verification and detection tools for current and future safeguards activities (Zendel et al. 2007). Main attributes of various further measurement systems are listed in > Table 63.7. Physical Property Measurement The load cell-based weighing system (LCBS) is used for weighing large uranium hexafluoride (UF6) product cylinders (Dermendjiev et al. 1983). Cylinders are weighed with the LCBS Nuclear Safeguards Verification Measurement Techniques 63 2935
  • attached to a crane. The load cell senses the weight of the suspended product cylinder and results are indicated on a display unit. After assembly and setup (requiring 20–30 minutes), weighing time is typically 3–5 minutes per cylinder with an accuracy of greater than 0.1%. The ultrasonic thickness gauge (ULTG) is a small and simple instrument designed for measuring metal thickness. The measurement principle is based on measuring the time interval . Table 63.7 Other measurement systems System Detector Verification task Typical performance values (%) ICVD Light amplifier for Cherenkov light Spent fuel (cooling time <10 years) Attribute DCVD Digital light amplifier for Cherenkov light Spent fuel (low burnup, long cooling time >10 years) Attribute OFPS 6 Li-doped fibers CANDU spent fuel bundles Attribute LCBS Strain gauge Weighing of UF6 cylinders 0.1 ULTG Ultrasonic sensor Measurement of wall thicknesses 0.3 PPMD Pressure sensor Volume in tanks 0.01 CALO Temperature sensitive resistance sensor Pu in nuclear material 0.4 XRFA Si detector Characterization of metals, alloys, minor and trace elements Attribute KEDG HPGE transmission measurement Pu, U concentration in solutions (>50 g/l) 0.2 HKED XRF + KEDG Pu, U in input solutions (IAT) 0.6 (U), 1.0 (Pu) Pu, U in product solutions (OAT) 0.3 (U), 0.9 (Pu) COMPUCEA LaBr3+ Si U in unirradiated solutions 0.2 HFLS Tunable laser + spectrophotometer Detection of nuclear activities involving UF6 Sensitivity > 0.1 ppm UFLS Tunable laser + spectrophotometer 235 U-Enrichment in UF6 [0.15 (LEU)-0.7 (DU)]a LIBS Laser + spectrophotometer Confirming past nuclear activities and absence of undeclared activities by trace analysis Attribute LIDAR Laser + spectrophotometer Detecting signatures of nuclear activities Attribute XRFS X-ray detector U, Pu, Np, Th, Am concentrations 1 OSAS Alpha spectrometer Pu in diluted solutions (<10 mg/l) OSCA 3 He + HKED Cm/Pu/U ratio in input/HALW solutions OSDM Frequency oscillator Density of solutions 0.04 a Target accuracy. 2936 63 Nuclear Safeguards Verification Measurement Techniques
  • from sound emission to reception of the echo and then multiplying this interval by the material-specific sound velocity. Silicone grease is used as couplant to eliminate any air between the sensor and measurement surfaces. The ULTG is required to measure the wall thickness of containers to allow correction for variations in the attenuation of the 186 keV gamma peak. The ULTG cannot determine multiple layers and measures only the outer layer. Portable pressure measurement device (PPMD) is a lightweight carry-on type instrument comprising digital pressure modules (DPM) packed together with a power supply in a tamper- indicating enclosure (Landat et al. 1997). The PPMD can be connected to the level, density, and reference probes of a tank in parallel to an operator’s own equipment. The sensor data are used to independently verify the volume of solutions in various storage and process tanks. The 3D laser range finder (3D-LRF) is in routine use for design information verification (DIV) activities involving the largest reprocessing plant under safeguards (Agboraw et al. 2006). The system is capable of confirming within an accuracy of millimeters that no structural changes have occurred since the previous scanning and, of highlighting changes that may have occurred, in particular, to maintain continuity of knowledge (CoK) of the interiors of hot cells especially on various piping arrangements. For this purpose, baseline scans – called reference scans – are performed during the plant construction; subsequent verification scans taken during periodic inspection activities are compared to the original references. The ground penetrating radar technology (GPRT) is based on the transmission and reflection of electromagnetic pulses. It represents a nonintrusive means of surveying sites for the identification and/or verification of structural elements. GPRT uses pulses of electromag- netic radiation in the microwave band (UHF/VHF frequencies) of the radio spectrum, and reads the reflected signal to detect subsurface structures and objects without drilling, probing, or otherwise breaking the ground surface. Applications include the detection of hidden objects and structures (Carchon et al. 2006). Calorimetric Techniques Calorimeters measure the heat produced by a sample of nuclear material, whereby the heat originates primarily from the a-decay of the isotopes making up the nuclear material. Calo- rimeters have traditionally been fabricated using a sensor of nickel wire wound around a measurement chamber (Bracken et al. 2002). The nickel wire provides a temperature- sensitive resistance leading to highly accurate and precise electrical measurements of the power produced by a sample. Calorimeters are calibrated with 238 Pu heat sources or plutonium samples with known mass and isotopic composition. Calorimetric assay is the most precise and accurate NDA measurement method for plutonium products (>100 grams). Plutonium samples always contain a mixture of isotopes: 238 Pu, 239 Pu, 240 Pu, 241 Pu, and 242 Pu. The nuclide 241 Am is also present in most plutonium-bearing items. The amount of heat produced per gram of each nuclide is a known physical quantity (‘‘specific power’’) based on the decay rate and decay energy of the isotope. The major contributors to the heat produced are 238 Pu and 241 Am. The measurement of the total heat in a calorimeter coupled with the measurement of the relative abundance of the individual nuclides in the sample (nuclidic/ isotopic composition) provides the elemental mass in the sample. Calorimetry and neutron coincidence counting are complementary assay techniques that provide high assurance of correct and valid measurements. The calorimeter has the advantage being insensitive to geometry, matrix and humidity. However, measurements using calorimeters are very time Nuclear Safeguards Verification Measurement Techniques 63 2937
  • consuming in contrast to neutron measurements. Therefore, calorimetry is barely used by international safeguards but widely used in domestic safeguards to quantify plutonium and americium in plutonium bearing items. X-Ray Measurements X-ray emission is characteristic of the element while gamma-ray emission is characteristic of the isotope. Plutonium and uranium, as atoms with high atomic numbers, absorb energy at distinctive energies (e.g., the K-edge for Pu is at 121.8 keVas discussed below), associated with the energies needed to dislodge electrons from their innermost electron shells. The X-ray fluorescence analyzer (XRFA) is a light, portable, and commercially available instrument (1.4 kg) for the characterization of metals, alloys, and possible dual-use materials by means of X-ray fluorescence (XRF). It is used by safeguards inspectors to verify the correctness and completeness of the presence of declared materials and the absence of undeclared ones. The measurement principle is based on irradiating the sample material by an X-ray photon field generated by an X-ray tube (no radioactive source!) and then measuring the characteristic X-ray fluorescence spectrum emitted by the sample using a high- performance Peltier-cooled Si-PiN detector. XRF can determine semi-quantitatively the rela- tive concentrations of major, minor, and trace elements with atomic masses from 9 (fluorine) through 89–103 (the actinide elements) in various types of samples without requiring any sample preparation (> Fig. 63.4). K-EDGe X-ray densitometry (KEDG) measures the photon transmission through a liquid sample at two energies that bracket as closely as possible the K-absorption edge energy of the element of interest. The K-absorption edge energy represents an element-specific signature. The logarithmic ratio of the photon transmission measured below and above the absorption edge is directly proportional to the volume concentration of the measured element for a well- defined sample cell. For plutonium, radioactive sources from 75 Se (Eg = 121.1 keV) and 57 Co (Eg = 122.2 keV) are used as transmission sources. These g-ray energies very closely bracket the 121.8 keV electron-binding energy of the K shell of plutonium. The majority of the K-edge . Fig. 63.4 X-ray fluorescence analyzer (XRFA) 2938 63 Nuclear Safeguards Verification Measurement Techniques
  • densitometers used for safeguards are now equipped with an X-ray generator that acts as a photon source for the transmission measurement. The high photon strength provided by an X-ray tube allows measurements to be performed on highly radioactive samples. The method is very selective and is one of the most accurate NDA techniques, but the determination of Pu may be biased by the presence of a minor actinide element of lower atomic number, such as U. It is therefore best used for relatively concentrated solutions (>50 g/l) of plutonium in product solutions, input solutions, and process solutions (in-line measurements). Hybrid K-edge (HKED) is a technique used for measuring the concentration of U and Pu in mixed solutions by combining X-ray fluorescence and K-edge densitometry. Depending on the concentration of plutonium and uranium, the higher element concentration is identified by K-edge measurement and the lower concentration from the ratio of Pu/U of the XRF. HKED has become a standard technique to analyze solutions samples containing U and/or Pu over a concentration range from 0.5 to 400 g/l in reprocessing plants. In addition, HKED is also used to determine Np in the presence of fission products. The X-Ray beam in KEDG-mode passes through a sample cell with a known path length and an HPGe detector measures the absorption at the K-edge to determine the concentration, based on the calibration of the system with solution standards. Located at a backward angle of 150 relative to the beam, a second HPGe detector can measure the X-Ray spectrum generated by the elements in the sample. K-edge densitometers equipped with an X-ray generator are usually resident and attached to shielded or unshielded glove boxes for sample handling (> Fig. 63.5). Combined procedure for uranium concentration and enrichment assay (COMPUCEA) is a transportable system used to perform accurate on-site analytical measurements of elemental . Fig. 63.5 The hybrid K-edge densitometer Shielding (W) X-ray tube Beam filter (Cd) PE capsule Glass cuvette for KED Collimator (W) Beam filter (Fe) Beam for KED 180° – q = 30° Conveyor Collimator (W) Beam for XRF PE vial for XRF Tube (SS) ∅0.8mm ∅2.5mm Nuclear Safeguards Verification Measurement Techniques 63 2939
  • assay and enrichment of liquid, unirradiated uranium samples (Ottmar et al. 2007). Solid uranium samples require preparation by quantitative dissolution of the sample. The system is portable and can be used at room temperature. The technique combines absorption edge spectrometry to establish the uranium concentration (L-edge for uranium: 17.17 keV) and a 235 U-enrichment spectrometer based on LaBr3. The L-edge technique uses a small X-ray generator of low energy and an ultrahigh-resolution Si gamma detector operated under modest Peltier cooling. Analytical NDA Techniques at Laboratories Some of the NDA verification techniques on samples of nuclear material taken during a safeguards inspection are carried out at specialized analytical laboratories, for example, the Safeguards Analytical Laboratory (SAL) at Seibersdorf in Austria or the On-Site Laboratory (OSL) at the Rokkasho Reprocessing Plant in Japan. Commercially available and customer- specific equipment is used to determine amounts and isotopic compositions of plutonium and uranium in inspection samples, having various chemical and physical forms, to supplement or replace destructive analysis. A parallel analysis of one and the same sample by NDA and DA serves the quality assurance of the analytical methods and control of the results. High-resolution gamma spectrometry (HRGS) determines enrichment in uranium bulk samples by relating the gamma results to the total U content of the sample that has been assayed by titration or other methods such as IDMS or KEDG. In the absence of radioisotope interferences, the results have a precision and accuracy ranging between 0.5% rel. for natural and 0.2% rel. for enriched U. At OSL, HRGS is applied to Pu sampled from the output accountancy tank (OAT) to determine the isotopic composition, if the sample is not subjected to mass spectrometry. Additionally, Np in U and in Pu product samples can be determined. HRGS is used at SAL to screen all Pu samples as they are received. X-Ray fluorescence spectrometry (XRFS) is a non-portable XRF diagnostics tool, measur- ing well-characterized emissions from various elements (ranging from sodium to the highest elements on the periodic chart) when they are stimulated by x-rays. It provides the amounts of each element present (but not individual isotopes) within microgram detection limits. XRF is used to simultaneously determine concentrations of Th, U, Np, Pu, and Am. The XRF technique is applied if the element concentration to be determined falls below the useful range for K-edge measurements (<40g/l). It provides quantitative concentration measure- ments down to concentration levels of about 0.5 g/l. The quantitative evaluation of XRF measurements for concentrations is not as straightforward as for KEDG, e.g., matrix effects, but the technique is very accurate in performing element ratio measurements to determine the concentration of a minor element relative to the concentration of a major element known from a K-edge measurement. The hybrid K-edge densitometry (HKED) at OSL is the main instrument to analyze solutions samples from the input accountability tank (IAT) and output accountability tank (OAT) containing U and/or Pu over a concentration range from 0.5 to 400 g/l. The KEDG- mode is applied down to 40 g/l and the XRF-mode down to 0.5 g/l U or Pu. Thus, the evaluation of KEDG and XRF spectra of a sample from the IAT yields the concentration of U (KEDG alone), the U/Pu ratio (XRF alone), and the Pu concentration (KEDG and XRF combined). The OSL Alpha Spectrometer (OSAS) estimates plutonium in diluted solutions containing less than 10 mg/l of plutonium. An a-source can be made from the sample through a simplified 2940 63 Nuclear Safeguards Verification Measurement Techniques
  • drop deposition method, followed by combustion. Spectra with higher resolution are obtained by electro-deposition, a method that is also preferred due to safety reasons (namely, using a furnace in a glove box is avoided). The sum activity of 239 Pu + 240 Pu is measured with an a-detector and the Pu concentration of the sample is estimated by comparing the a-count rate of the unknown sample with the count rate of a known standard reference source. The OSL-curium neutron analyzer (OSCA) is a neutron coincidence counter that measures 244 Cm in solutions. At OSL, the system is integrated within a Hybrid-K-Edge (HKED) enclosure. The OSCA associated with HKED or isotope dilution mass spectrometry (IDMS) will provide analytical data to establish the 244 Cm/Pu and 244 Cm/U ratios in IATsolutions or in high active liquid waste solutions. The OSL-density meter (OSDM) measures the density of solutions to estimate accurately the mass of a solution contained in a sampled vessel/container. OSDM measures the frequency of an oscillating tube filled with the solution sample. The instrument is calibrated by measuring the frequency of oscillation using air, water, or liquid density standards. A sample solution is injected, or drawn, into the temperature-stabilized tube and a built-in computer automatically converts the frequency measurement into a density reading. 63.3.8 New and Novel Technologies Emerging and future needs for safeguards verification require state-of-the-art equipment and innovative technological solutions to meet current and future verification challenges (Khlebnikov et al. 2006a; Annese et al. 2009; Tolk et al. 2007). The early detection of undeclared facilities, activities and materials has become a major safeguards task. It plays a primary role in providing independent safeguards conclusions on the completeness and correctness of a state’s nuclear program. The increasing complexity and automation of nuclear facilities as well as new facility types require new approaches and advanced instrumentation. The verification system has to be further developed taking full account of advances in present safeguards verification techniques (‘‘new technologies’’) and through the exploration of innovative technologies that are either already available or investigated in other branches of science and technology fields not traditionally related to safeguards (‘‘novel technologies’’). In particular, capabilities to detect nuclear activities (e.g., reprocessing or enrichment) from a distance are of prime interest and require a new toolbox for the inspectorate. In many cases, safeguards implementation involves unannounced inspections and complementary access (CA) requiring detection devices to search for nontraditional elements/nuclides (such as americium, neptunium, beryl- lium, and tritium) that could indicate the presence of a clandestine nuclear weapons program. Instruments for these specific types of verification activities often need to be made available to inspectors at very short notice. Such equipment must be multipurpose, robust, portable, and easy to operate to allow the inspector to readily perform numerous tasks within the short time span of each particular verification activity, including the search for indicators of undeclared nuclear materials and activities. New Technologies New technologies for safeguards purposes could address missing capabilities in routine verification tools and could enhance the effectiveness and efficiency of present verification systems. Instrumentation based on new technologies might be in its final development stage Nuclear Safeguards Verification Measurement Techniques 63 2941
  • but not yet available for routine safeguards implementation. Such equipment will be autho- rized for routine use subject only to a careful assessment of aspects (including its expected performance, usability, and affordability) and the completion of a successful field test. New sensors for nuclear material detection and characterization, process monitoring equipment/ techniques and analytical equipment are the main new technology drivers. Various efforts are underway to develop techniques for the direct verification of plutonium in spent fuel (Tobin et al. 2008, 2009). The superconducting gamma spectrometer is a cryogenic, ultrahigh energy resolution g-ray spectrometer, operated at temperatures of T % 0.1 K (Ali et al. 2006, 2008). It can be characterized as a cryogenic gamma-ray micro-calorimeter, which measures the energy of radiation from increase in temperature upon absorption of a g-ray. The detector consists of a highly sensitive, cryogenic transition-edge sensor (TES) coupled to a tin g-ray absorber. The TES is a Mo/Cu multilayer sensor capable of determining the energy of a single g-photon with extremely high precision – it operates in the narrow temperature range of the transition from superconducting to normal state. An individual sensor element is 1.5 Â 1.5 mm by 0.25 mm thick. To achieve reasonable counting statistics, detector arrays with many sensors are used (Hoover et al. 2009). The system offers an order of magnitude improvement in energy resolution over conventional high-purity Ge (HPGe) detectors. It could be used for accurate enrichment measurements and Pu-isotopics. The differential die-away self-interrogation (DDSI) technique is a recently conceived NDA technique for quantifying the total fissile and fertile content in spent fuel (Menlove et al. 2009) using a modified 3 He coincidence counter. In DDSI, the spontaneous fission neutrons from 244 Cm that are present in the assembly act as the interrogating source. After each spontaneous fission event, the neutron time distribution is analyzed to determine the fast and slow neutron distributions from the spontaneous and induced fissions in the sample, respectively. Fissile mass is determined from the fissions induced by reflected thermal neutrons originating from the spontaneous fission reaction. The sensitivity of the fissile mass measurement is enhanced by measuring the sample with and without a cadmium liner between the sample and the surrounding moderator (‘‘passive-neutron albedo reactivity (PNAR)’’). The fertile mass is determined from the multiplicity analysis of the neutrons detected soon after the initial triggering neutron is detected. Self-interrogation neutron resonance densitometry (SINRD) uses the unique neutron- resonance cross-sectional structure for fissionable nuclides such as 235 U, 233 U, 239 Pu, and 241 Pu. Its sensitivity is based on using the same fissile materials in the sample and in the fission chamber because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber (LaFleur et al. 2008, 2009). The amount of resonance absorption of these neutrons in the spent fuel can be measured using 235 U and 239 Pu fission chambers placed adjacent to the assembly. 252 Cf interrogation with prompt neutron detection (CIPN) is a lightweight, active interrogation instrument/detector, similar in size and shape to the FORK detector (Schear et al. 2009). The main difference between CIPN and a FORK detector is the introduction of an interrogating 252 Cf source adjacent to the fuel and opposite to the fission-chamber detector region. Two gross-neutron measurements are performed: a background measurement (with- out 252 Cf) and an active assay with the 252 Cf source next to the fuel. Fissile content is quantified using the measured neutron multiplication in the assembly. The delayed neutron technique quantifies the fissile content in a spent fuel assembly by measuring the delayed neutrons emitted by fission fragments after active interrogation by 2942 63 Nuclear Safeguards Verification Measurement Techniques
  • a neutron burst ($1 s) from a neutron generator (Schear et al. 2009). The generator stops after 1 s and starts again after another second. In between, the delayed neutrons are being measured and multiple interrogation/measurement cycles are performed to gain sufficient counting statistics. Self-induced X-ray fluorescence (XRF) can be used to quantify the Pu content in spent fuel (Charlton et al. 2009). The radioactive decay in the spent fuel leads to self-induced XRF of uranium and plutonium atoms in the fuel and resulting X-rays are then emitted by the fuel. It has been demonstrated by measurements and simulations that the Pu/U atom ratio at the surface of a fuel pin can be measured in an appropriately designed and implemented instru- ment using the 103.7 keV K X-ray from Pu. Lead slowing-down spectroscopy (LSDS) is an active interrogation technique that has been used for several decades in cross-sectional measurements. It can provide independent Pu isotopic mass without operator-declared information about the spent fuel (Gavron et al. 2009). In typical operation, a neutron source provides a pulse of neutrons close to the center of a large block of lead. The neutrons are moderated slowing down gradually. The incoming neutrons from the slowing down process will induce fission in the fuel placed in the center of the lead block and high-energy fission- neutrons are emitted. An annular ring of fission chambers selectively measures the higher-energy fission neutrons as a function of the time after the neutron pulse. Potential threshold fission chamber types include those lined with ultra pure 238 U (to minimize the contributions from 235 U) and those based on 232 Th. New time- spectra analysis methods are applied to determine the fissile mass of 239,241 Pu and 235 U. A new on-line monitoring system featuring Raman and ultraviolet-visible-near infrared (UV-vis-NIR) spectroscopy methods combined with a Coriolis and conductivity probes pro- vides immediate chemical data and flow parameters of high-level radioactive waste streams (Bryan et al. 2009). This allows real time monitoring of the solvent extraction flow sheets for quick detection of undeclared activities with fissile isotopes present in the radiochem- ical streams during reprocessing activities. The proof-of-principle has been successfully demonstrated for on-line monitoring of U (VI), nitrate, and HNO3 by using Raman spectros- copy and on-line monitoring of Np, Pu, Nd, for both aqueous and organic phases, using Vis/NIR. The multi-isotope process (MIP) monitors spent nuclear fuel reprocessing facilities on- line, nondestructively, and in near real time (Orton et al. 2009). The method is based upon the measurement of distribution patterns of a suite of indicator (radioactive) isotopes present within particular process streams. Distribution patterns, monitored on-line by gamma spec- trometry, are compared in near real time to patterns representing ‘‘normal’’ process conditions using multivariate pattern recognition software. The MIP monitor is sensitive to minor alterations in major process variables including acid and organic ligand concentration. There- fore, it might be used to confirm the declared operation of a reprocessing plant, providing additional information on reactor type, burnup, and cooling time. The laser-based optical and chemical imager (LOCI) is a unique instrument that combines accurate isotope ratio analyses obtained both by laser desorption Fourier transform ion cyclotron resonance mass spectrometry (FTICR-MS) and by LIBS without any sample prep- aration (Scott and McJunkin 2009). A single photon ionization (SPI) process is implemented allowing near 100% ionization efficiency for elements and compounds with ionization energies less than 10.5 eV. The FTICR-MS and LIBS isotope capability coupled with LOCI’s wide mass range, mapping capability, high resolution, and automated data collection as well as data interpretation offers an alternative to the labor-intensive bulk analysis of traditional methods, Nuclear Safeguards Verification Measurement Techniques 63 2943
  • such as TIMS. This capability will be very useful for environmental samples and for determin- ing the burnup rates within fuel elements. Electrochemically modulated separations (EMS) is a technique that relies on electrochem- ical redox adjustment and surface chemistry to effect isolation and accumulation of Pu at a target electrode (Clark et al. 2006; Duckworth et al. 2009). EMS may be a rapid and cost- effective means of performing Pu separations that has implications for DA and NDA, simpli- fying laboratory separations and allowing NDA for input accountancy tanks. Automated radioanalytical chemistry can provide near real time monitoring of reprocessing plant operations (O’Hara et al. 2009). A sequential injection chromatography system for the separation and analysis of Am, Pu, and Np isotopes is integrated in a modular system that automates the complete sample analysis process, from initial sample preparation to final data reporting. The laser item identification system (L2IS) is capable of monitoring all transfers of UF6 cylinders between process areas (Poirier et al. 2009). L2IS uniquely identifies each cylinder through exploring the unique microstructure of each cylinder’s surface with different lasers. It has been demonstrated that every cylinder has a unique ‘‘fingerprint’’ that remains intact even under extreme environmental conditions. The L2IS system is composed of a portable unit, operated in attended mode, and a fixed installed unit, operated without inspector presence. The portable unit acquires the fingerprints of a given set of feed cylinders intended to be used over the coming months and the fixed system monitors the flow of previously identified cylinders in a transfer corridor. Novel Technologies Novel Technologies aim to provide access to a wider range of methods and instruments in support of safeguards implementation by adapting technologies already available or being developed in other branches of science and technology fields but not used for safeguards. Such technologies could significantly contribute to the early detection of undeclared nuclear activities and material. Typical examples include the determination of an undeclared location previously used for storing radioactive material, forensics of materials found on-site and evidence of nuclear fuel-cycle process activities at suspected locations. In contrast to new technologies, safeguards equipment based on novel technologies is in its early stage of development and may not be readily available within a short time frame. Several promising technologies based on laser spectrometry and optically stimulated luminescence (OSL), anti neutrino detection, nuclear magnetic reso- nance (NMR), atmospheric gas sampling, and analysis and various electromagnetic systems have been identified (Khlebnikov et al. 2006b; Whichello et al. 2009). Nanotechnology in its various forms such as nano-electronics, nano-electromechanical systems, ultra-small, highly sensitive, and selective sensors (Morrison et al. 2007) could significantly contribute to novel technolo- gies. A novel suite of smart safeguards verification systems can be expected with unique detection features allowing at the same time further miniaturization of the equipment. The most impressive advance allowed for by nanotechnology development is the possibility of autonomous smart sensor networks. These networks could support safeguards monitoring functions by capturing data, processing and transmitting information, and communicating with other sensors in potentially hostile environments. Optically stimulated luminescence (OSL) is capable of de-trapping radiation-induced excitation energy accumulated during the irradiation of a surface (Kosierb 2007). The release 2944 63 Nuclear Safeguards Verification Measurement Techniques
  • of this energy could be stimulated by various types of lasers in the visible and infrared frequency ranges. The intensity of the resulting photon emission is proportional to the radiation dose absorbed by the material, unveiling information on the former presence of a radioactive source. The detection system uses a charge-coupled device (CCD) camera attached to a photomultiplier. The optical part of the detector has to be optimized to decouple the larger signal, emanating from the optical pump, from the smaller signal emitted by the sample. Samples collected by the inspectors are analyzed by OSL, indicating the previous presence of stored nuclear materials and thereby presenting the potential to disclose undeclared activities. OSL could also be used for container verification. Luminescent phosphor additives, ionized by radiation, are mixed with paints or clear coats and are applied to the surface of a container (Miller et al. 2009). The OSL phosphors luminesce in proportion to the ionization radiation dose and the intensity of excitation light. The OSL coatings/additives will be invisible to the naked eye, but could be seen using an InGaAs infrared detector. The OSL additives would tag the container and reveal an attempt to tamper with the container and therefore increase the confidence that its integrity had not been compromised. Tunable diode laser spectroscopy (TDLS) systems are tuned to access specific regions of the mid-infrared spectrum where most gases of interest such as UF6 have strong absorption while common gases, such as oxygen and nitrogen, do not. TDLS systems have the potential to determine 235 U enrichment in UF6 gas and to indicate the presence of HF gas, a by-product of enrichment activities. The HF detector laser system (HFLS) is a portable instrument for HF gas detection, designed for easy operation in airborne and ground-based mobile searches for enrichment activities. The HFLS is built as a backpack unit allowing continuous air monitoring while leaving the inspector’s hands free. A tunable diode laser shines through a multipass cell, which is continuously collecting air gas. The detector then analyzes the unique absorption lines caused by the HF gas in the cell. The instrument is very sensitive and can measure HF concentrations of >0.1 ppb. The system provides a very quick measurement and identification with high spectral resolution in the infrared range. The UF6 detector based on laser spectrometry (UFLS) is an on-site analytical instrument based on TDLS. It measures the enrichment of UF6 samples (Lebrun et al. 2008). The system has passed feasibility study and is now under development for field use. It determines the concentration of 235 U and 238 U in UF6 on-site with an accuracy of greater than 1% for 235 U enrichment. The precise measurement of the isotopically broadened absorption peaks of 235 U and 238 U requires a mid-IR laser with wide single-mode tuning ranges, better than 4 cmÀ1 of continuous tuning at 1,290 cmÀ1 and less tuning at 852 cmÀ1 . In comparison with mass spectrometry, UFLS does not require a highly trained specialist to perform measurements. It is hoped that the instrument will partly replace the need for DA and thereby will improve verification timeliness and reduce inspection resources. Laser-induced breakdown spectroscopy (LIBS) is being applied for a novel complementary access instrument based on the detection of gaseous and solid signatures and indicators of nuclear fuel cycle processes (Kosierb 2007; IAEA 2009b). LIBS is an atomic emission spectros- copy technique that utilizes a pulsed, well-focused laser to create a micro-plasma on the sample surface (Salmon et al. 2008). The resulting light emission spectrum of the decaying vapor plume has well-known specific emission lines that are analyzed by an integrated spectrometer. The spectroscopic profile is compared to those in its library to determine the composition of the material. LIBS can perform the analysis of elemental composition and trace analysis of solid Nuclear Safeguards Verification Measurement Techniques 63 2945
  • materials to confirm past nuclear activities and the absence of undeclared activities, and is therefore considered as a possible screening device to reduce the number of environmental samples (> Fig. 63.6). The light detection and ranging (LIDAR) system may remotely sense the presence of characteristic gaseous compounds emanating from nuclear fuel cycle (NFC) processes into the atmosphere (IAEA 2006b). LIDAR might be capable of analyzing emissions and detecting fingerprints of undeclared nuclear activities from a distance of some kilometers of a suspected site by using a mobile LIDAR laboratory in its vicinity. A laser, tunable to precise wavelengths, selectively stimulates such airborne molecules. A light-sensitive telescope scans the atmo- sphere, detecting the presence of the stimulated molecules. For safeguards purposes, differen- tial absorption LIDAR (DIAL) is particularly interesting. This technology sends laser pulses tuned on two different wavelengths to the atmosphere – one specific to the strongly absorbing molecule, the other less absorptive as a reference – and then analyzes the intensity of light scattered back over time. The signals are processed providing the transmission both on and off a molecular absorption feature to give a measure of the concentration. The antineutrino detector can measure the plutonium content of an online nuclear reactor core, may provide effective power and burn-up to confirm that reactors are operating as declared (Bernstein et al. 2003; Bowden et al. 2007; IAEA 2009c). Antineutrinos are produced in nuclear reactors when uranium and plutonium atoms fission into neutron-rich fragments that undergo successive beta decays. A common detection scheme uses the inverse beta decay reaction producing a closely timed coincident neutron and positron – these are detectable in a scintillator that is shielded from background radiation. Despite the small cross section, the abundant antineutrino production of the core allows for high statistical detection with modestly sized detectors (1 m3 ) at practical standoff distances (tens of meters). When an antineutrino collides with a proton, it produces a positron and a neutron. The interaction of these two particles creates the antineutrino signature – two relatively intense flashes of light that occur so close in time to one another that they appear to be almost simultaneous (> Fig. 63.7). . Fig. 63.6 Laser-induced breakdown spectroscopy (LIBS) 2946 63 Nuclear Safeguards Verification Measurement Techniques
  • The Fourier-transform infrared system (FTIR) is a well-known spectroscopic technique based on the absorption of infrared photons that excite vibrations of molecular bonds. Molecules such as U3O8, UO2, UO3, ThO2, have characteristic absorption bands in the infrared region that can be used like a fingerprint to detect their respective presence. FTIR radiometry has become a relatively mature and reliable method for the identification and measurement of chemicals emitted from stacks and its potential for passive standoff detection of nuclear material is under investigation (Puckrin and The´riault 2006). Electromagnetic wave gradiometer technology can be used to detect underground electri- cally conductive structures or paths such as clandestine tunnels (Matter 2003). Primary low- frequency electromagnetic (EM) waves induce electrical current flow in conductive paths located underground. Conductive paths are abundant in underground operating structures, such as electrical power cables or air ventilation pipes, or conductive paths such as the moist conductive soil layer that commonly surrounds tunnels with high humidity and condensation. The transmitter can be located above ground (either in a fixed location or carried by an inspector) or underground in a borehole. The induced current flow generates a secondary magnetic wave that is retransmitted through the ground and measurable on the surface above the underground structure using a gradiometer antenna array with a narrow band synchronous receiver. The signal is scanned on the surface and a plot of its magnitude yields a profile showing the structure’s centerline and depth information. Nuclear magnetic resonance (NMR) could measure the 235 U/238 U ratio accurately and therefore it could be used to detect clandestine enrichment at a declared facility, applying mature NMR technology (Pepper et al. 2007; Magnelind et al. 2009). NMR provides . Fig. 63.7 The antineutrino detector Water/polyethylene shielding Liquid scintillator filled cells Muon veto paddles Nuclear Safeguards Verification Measurement Techniques 63 2947
  • a quantitative measure of the number of nuclei of a given isotope in a given sample. The NMR signature of materials is measured in ultralow magnetic fields and at ultralow frequencies. The method has the potential of quickly and accurately (possibly at the fraction of a percent level) measuring the amounts of 238 U and 235 U in each and every cylinder containing feed, tails, and product, as well as monitoring the flow of UF6 in the cascade area. The benefits of NMR include that it is nonintrusive and no source is required. Acoustic wave analysis technology may provide a viable containment verification tool (Goldfarb 2007). Acoustic waves are disturbances in mechanical vibrations in solids, liquids, or gases. The interpretation of acoustic waves can unveil subtle structural changes in materials, including cracks, pits, voids, gaps, bends, and changes in density and elasticity. A monitoring system with a network of acoustic sensors may be useful to detect abnormal process operations. The speckle laser interferometric application could detect hidden underground structures or other non-visible items (Mersch 2007; Nothdurft and Yao 2005). The system uses subtle differences in surface response to coherent light and requires significant interpretive analysis. It subtracts a picture taken from a surface illuminated by laser light (background picture) from subsequent pictures taken after ‘‘disturbing the surface,’’ e.g., by a small underground detona- tion. The effect of the observed subsurface disturbance on the surface is a function of the hidden subsurface features. Remote sensing in satellite imagery and ‘‘stereo mapping’’ could be further developed to enhance interpretation of site activities using the broad potential of satellite imagery data ranging from panchromatic, multispectral, hyperspectral to radar for site description and change detection (Niemayer 2009). A satellite could view the earth in different spectral bands covering the visible and infrared band. It could measure local ground and water temperatures to an absolute accuracy of 1 K. By further investigating other portions of the electromagnetic spectrum, it may be possible to detect and identify the location of an undeclared nuclear activity, e.g., to detect the waste heat from a clandestine plutonium production reactor. Hyperspectral data allow for a quantitative estimation of geophysical and geochemical characteristics of the earth’s surface and is therefore useful for assessing, e.g., surface cover changes due to drilling, mining, and milling activities. Atmospheric gases sampling and analysis can provide useful information about the existence and nature of ongoing nuclear activities. 85 Kr could be indicative of reprocessing activities (Kalinowski et al. 2006; IAEA 2006a). For each kilogram of plutonium reprocessed, some 10–35 TBq of 85 Kr are typically released into the atmosphere. This evidence can be detected from a distance, varying from hundreds of meters to some kilometers, depending on meteo- rological conditions. The technique uses air sampling with a cryo-absorption device to concentrate the fraction of noble gases. The amount of 85 Kr is determined either by low-level counting of the b-radiation, by accelerator driven mass spectrometry or by atom-trap trace analysis (ATTA). Micro seismic monitoring can detect any abnormal underground activities that could indicate unauthorized design changes and containment breaches in final nuclear depositories (Saari and Lakio 2008). It consists of several seismic sensors grouped in a network to monitor remotely excavation-induced micro-earthquakes and explosions occurring inside the local geology of nuclear repositories. The system can also detect early the drilling activity of a tunnel-boring machine trying to excavate an access tunnel to the depository from outside (Saari and Lakio 2009). Microelectromechanical systems (MEMS) have the potential for creating miniature bench- top laboratories on a chip (Janssens-Maenhout and Nucifora 2007). Such miniaturized 2948 63 Nuclear Safeguards Verification Measurement Techniques
  • chemical laboratories could provide real-time chemical detection of trace elements in vapor and they could be tailored to detect many classes of chemicals, providing a quick indication of the presence of a class of selected chemicals in the field. MEMS would work with sensors that would use polymer or gel-coated silicon devices to trap targeted chemicals, and then send the agents through fluidic channels to on-chip arrays of surface-acoustic-wave detectors (SAWs). The sensors can be ‘‘pre-tuned’’ to targets of interest, during fabrication, by careful selection of absorption layer and can perform spectroscopic measurements. A follow-on device would integrate the fluidics, sensors, and support electronics on a single device. Detectors, sensitive to hydrogen fluorine and other fluorine-containing gases (Moritz et al. 1999; Filippov et al. 2007) are of safeguards interest for indicating possible enrichment activities. These types of sensors are based on the Pd-LaF3-SiO2-Si structure that generates a signal resulting from the electrochemical reaction with the fluoride on the surface of the sensor. The lower detection limit offered by this technology has been estimated at around 0.05 ppm of HF. It can be assumed that a device based on this semiconductor sensor is able to register directly both UF6 and further products resulting from its degradation. Other sensors are composed of nanostructured materials such as carbon nanotubes and metal oxide nanowires. These materials have a controlled pore size and an increased adsorptive capacity due to the large surface area enabling the selective uptake of gaseous species. The absorption can lead to a measurable change in some specific properties, e.g., conductance, capacitance, etc. (Li et al. 2003). Nanocomposite scintillators utilizing nanocrystals of known scintillator and detector mate- rials will allow for the production of large composites of the nanocrystals with an expanded variety of attainable shapes and sizes (Del Sesto et al. 2007; McKigney et al. 2007). These nanocomposite materials are expected to have improved properties with respect to the properties of the bulk single crystal scintillators from which they are derived. Preliminary measurements show that nanophosphor materials have a significantly higher light output (by a factor of 3) than single-crystal materials of the same weight. This translates to greater scintillation efficiency. In addition, a shift toward longer wavelengths is observed in both the photoluminescence excitation and emission of the nanomaterial with a smaller overlap of the excitation and emission spectra, enhancing the energy resolution of the nanocrystal material. Nanocomposite semiconductors such as nanowires arrays of CdZnTe can be used for detecting low-energy gamma rays (Gandhi et al. 2008). The CdZnTe compound semiconduc- tor is electrodeposited in the form of nanowires onto a TiO2 nanotubular template. The preliminary results indicate that the CZT nanowire arrays can be used as radiation detector materials at room temperature with a much lower bias potential (0.7–2.3 V) as compared to the 300–500 V applied to bulk detector materials. Solid-state neutron detectors using silicon nanopillars (Cheung et al. 2006) are being developed. The space between the silicon nanopillars is completely filled with 10 B-enriched material using chemical vapor deposition (CVD). The incident neutrons are converted by the nuclear reaction 10 B(n, a)7 Li to a-particles, which are collected and detected by the silicon (Nikolic et al. 2007). The coin-sized detector has achieved a thermal neutron detection efficiency of 20% and neutron/gamma discrimination of 105 , which are comparable to the parameters of a 3 He-based detector. Such detectors could be arranged in flexible arrays and could be directly attached to a moderator. They could also replace 3 He tubes. Another attractive application combines recent advancements in ultrawide band radio frequency identification (RFID) technology with such neutron detectors (Nekoogar et al. 2009). The RFID neutron tags can be used for neutron monitoring. They are totally passive and will operate indefinitely Nuclear Safeguards Verification Measurement Techniques 63 2949
  • without battery power. The tag is compact; it can be directly mounted on metal, and has high performance in dense and cluttered environments. Miniaturized wireless semiconductor devices can form sensor networks – also known as smart dust (Sailor and Link 2005). These devices are millimeter- to micrometer-sized multifunctional packages, combining a sensor (e.g., optical, chemical, vibration, or radiation) with computing and communication capabilities. A large number of these packages can be integrated to form an ‘‘Internet-like’’ network. The devices could be inexpensively mass- produced using conventional silicon fabrication techniques coupled with nanotechnology. Potential applications for safeguards may include the monitoring of facilities and inventories, measuring temperature profiles and detecting undeclared process activities via chemical signatures. 63.4 Laboratory Analysis for Nuclear Material Accountability Verifications 63.4.1 Introduction One diversion scenario is the removal of small amounts of material over a longer period, masking under the measurement uncertainties the slow accumulation of ‘‘significant quanti- ties’’ of material in a ‘‘protracted diversion.’’ This can involve small underestimates in the declaration of inputs, small overestimates of the outputs, or inflated estimates of measurement uncertainties by the operator. Destructive analysis (DA) remains the basic tool to detect such potential actions by analyzing independent samples obtained by the inspectors during physical inventory or material flow verifications, based on a random sampling plan. The inspection samples are most often shipped for analysis to specialized laboratories operated by or on behalf of the inspectorate. In the case of some large bulk-handling facilities, inspectorate specialists will analyze the inspection samples locally or on the plant site at a dedicated laboratory. Inspectors may perform additional inspections and request other samples for the purpose of verifying the quality of the measurement systems and providing independent estimates of the measurement uncertainties. The inspectorate laboratories supply also certified materials, which are submitted to the operator laboratory as part of external analytical quality control measures. They also prepare and/or certify spikes for isotope dilution assays and physical or chemical standards for the calibration of portable or installed instruments used for on-site verification measurements. The accountability measurements of the facility operators and the verification measure- ments by the inspectorate are expected to meet internationally accepted standards of precision and accuracy (IAEA 2001), which were defined by an international panel of analytical chemists and data evaluation specialists and are reviewed at regular intervals. 63.4.2 Bulk Measurement, Sampling, Conditioning, and Shipment of Safeguards Inspection Samples A prerequisite for accurate destructive analysis is a strict representativity of the sample, when it is taken, and its proper conservation during its handling at the plant and its shipment to the analytical laboratory. If the purpose of the sampling is to verify material accountability, the 2950 63 Nuclear Safeguards Verification Measurement Techniques
  • weight or volume of the material to be sampled must be measured just before sampling and recorded. Operator’s installations and procedures are normally used for bulk measurements of weights or volumes and sampling for verifications. Avalidation of operator sampling installations and procedures is therefore done during the initial design information verifications and associate inspectors, sampling specialists, and analysts of the inspectorate laboratory. The sampling installations and procedures should comply with internationally accepted standards and technical norms. Inspector’s samples of plutonium and spent fuel samples must frequently be pretreated, diluted, conditioned, and packaged so that the amount of radioactive material per package complies with international air transport regulations (IATA 2003). Because of their small size, the composition of inspec- tion samples is prone to be affected by humidity, air oxidation or evaporation. The inspector- ates select and qualify, therefore, carefully the vessels in which inspection samples are taken, conditioned, and shipped in a way to ensure their integrity until they reach the laboratory. The vessels should be chemically inert and should not break during transport. Their tare should be stable over time and varying atmospheric conditions, and it must be possible to close or cap them hermetically, so that a weight control may detect potential changes in the composition of the sample between its taking and its analysis and permit their correction, when necessary. Plastic vessels are used for samples of solutions or sintered pellets and dense metal. Glass vessels are chosen for powder and most other samples. The sampling vessels, the sample vials and their handling must always remain under inspectors’ supervision and control. Document STR69 (IAEA 2004) describes the precautions and procedures adopted for obtaining reliable DA inspection samples from most types of materials subject to nuclear safeguards. Sampling is not recommended in the case of finished or intermediate products from which analytical samples cannot be taken without destroying a valuable manufactured item (e.g., rods, bundles, assemblies). It is also undesirable on very heterogeneous materials in small quantities from which representative samples cannot be taken without unreasonably great effort and expense, and therefore sampling may not lead to meaningful results. Verifications of the accountability of the above materials will rely on NDA methods, combined with C/S measures. However, when the weight or volume of the material cannot be measured, the taking and analysis of a sample may still be justified for verifying the operator’s measurement system. The procedures are designed to yield at least duplicate results, which allow estimating the uncertainties resulting from subsampling and conditioning steps. Four examples are outlined below. Spent Fuel Solutions In spent fuel reprocessing plants, inspection samples of spent fuel (SF) solutions are taken from tanks equipped with means to homogenize the solution, by air sparging or mechanical stirring, and to measure the weight of the solution, using strain gauges, or its level and volume, usually with dip tube systems and differential manometers (ISO 1996a). During normal plant oper- ation, this is done solely at the input accountability tank for material flow verification. Inventory verifications are scheduled when the plant is idle and most fuel material is removed from the process areas. Only storage tanks may need then to be measured and sampled. The norm ISO18213 gives detailed recommendations concerning accountability tank installations and procedures for volume calibration and measurement (ISO 2007b). It reflects the results of extensive experience collected at a number of spent fuel reprocessing plants Nuclear Safeguards Verification Measurement Techniques 63 2951
  • (CETAMA 1997; Hunt et al. 1994; Janin 2006). For accurate bulk measurements and repre- sentative sampling, the ‘‘dead volumes’’ of ancillary pipes should be negligible compared to the volume of the bulk solution, and the heel remaining at the bottom of the tank after it is emptied should be minimized. Such tanks are subject to initial calibration before the start of plant operation and typically verified once a year (Janin 2006). Primary samples of 3–5 ml are withdrawn from a sampling pot through which an airlift circulates the bulk solution, while the tank is stirred (ISO 1995a). The operator collects a sequence of primary samples in evacuated 5–10 ml vials, capped with a synthetic rubber septum. The inspector selects one of the primary sampling vials and witnesses its subsequent subsampling and conditioning. Occasionally, the inspector may request replicate primary sam- ples to check their representativity. This can be done on the spot by witnessing a measurement of their density and comparing the results with the density of the bulk solution measured with the dip tubes in the tank, after referring all values to the same temperature (ISO 1995b). The large-size-dried (LSD) spike procedure is the standard IAEA procedure for the preparation and analysis (see section ‘‘Isotope Dilution Mass Spectrometry’’) of high burnup spent fuel samples with a 239 Pu abundance lower than 80% by isotope dilution mass spec- trometry (IDMS). The procedure for sample conditioning at the facility and the measurement scheme followed in the laboratory are outlined in > Fig. 63.8. The first of the three subsamples is intended for isotopic analysis and is transferred into an empty 10 ml penicillin vial. The second and third subsamples, intended for elemental assay, are transferred into separate 10 ml penicillin vials, which contain certified amounts of a mixed U/Pu tracer. The latter two subsamples are weighed to an accuracy of Æ0.0005 g or better. The tracers are re-dissolved with nitric acid; the subsamples are diluted to about 7 ml and homogenized. An aliquot of about 50 ml of each diluted subsample is transferred into a clean and empty penicillin vial of 5 ml and dried carefully. The penicillin vials are capped and inserted in a small lead pot. The lead pots are bagged out and packed in a tin can. Several tin cans may be packaged in a single Type A container, which is air flown to the inspectorate laboratory. Uranium Hexafluoride in Pressurized Cylinders Bulk measurement of UF6 in isolated UF6 cylinders consists of determining the gross weight of the cylinder on a suitable scale just before sampling and then subtracting the known tare. To obtain a representative sample suitable for elemental uranium assay, the cylinder is placed in an autoclave, connected to a stainless steel vacuum rig (> Fig. 63.9), heated and kept at 93–113 C throughout a prescribed period and during sampling. After complete homogenization, a measured portion of UF6 is withdrawn from the liquid phase via a ‘‘pigtail’’ and transferred into a large primary sample bottle. The convection currents in the liquid UF6 readily achieve isotopic homogeneity. However, chemical homogeneity is more difficult to reach in the presence of insoluble particles or volatile impurities. Excessive volatile impurities are vented to the process before weighing and sampling the cylinder. The primary sample bottle is transferred to the operator’s laboratory and mounted on a subsampling station (> Fig. 63.10). The UF6 is liquefied and a subsample of 1 to 5 g is collected again from the liquid phase (ANSI 1972; ISO 1996b) into a finger tube (> Fig. 63.11), which is weighed and shipped to the inspectorate laboratory. If the subsample is not issued from homogeneous liquid phases, only the uranium isotopic analysis will provide a meaningful result. In such a case, it is recommended to skip the elemental uranium assay. 2952 63 Nuclear Safeguards Verification Measurement Techniques
  • . Fig. 63.8 Standard sample preparation and measurement scheme for spent fuel samples of high burnup (IAEA 2004) Concertration 3 50 μl 1 min 30 min, 90–100°C 7 ml 3NHNO3 LSD-2 Subs. 1 ml Pu U 311312 Concertration 2 50 μl 1 min 30 min, 90–100°C 7 ml 3NHNO3 LSD-1 Subs. 1 ml Spent fuel input sample Facility 10 ml Add Stir w. magn bar Aliqucting into 5 ml vials Evaporation Ship to SAL SAL Redissolution Chem treatment Penicillin vials Spike Add Weighing Heat Stir w. magn bar Aliqucting into 5 ml vials Evaporation Ship to SAL Redissolution Chem treatment Pu U 211212 Isotopic comp. 1 50 μl 1 min 7 ml 3NHNO3 Subs. 1 ml Pu U 101102 103 Nuclear Safeguards Verification Measurement Techniques 63 2953
  • Plutonium Oxide Powders PuO2 powders readily exchange atmospheric moisture and carbon dioxide, particularly on their surface. With the precautions described below, the shipper’s and the receiver’s data were in . Fig. 63.9 Outline of a system for the sampling of UF6 cylinders (IAEA 2004) Autoclave UF6 stock cylinder Fitting Valve LN Liquid nitrogen Dry air or nitrogen Sample cylinder LN Cold trap Chem. trap Pump Pressure gauge Dry air or nitrogen Pig tail . Fig. 63.10 System for subsampling Liquid UF6 from a primary sample cylinder (ANSI 1972) Bulk sample container Fitting for spectrometer copper tube Fitting LN Valve Liquid nitrogen Fluorothene subsample tube 1/2 inch flare female 1/2 inch flare male Dry air or nitrogen Cold trap chem trap vacuum pump LN 2954 63 Nuclear Safeguards Verification Measurement Techniques
  • excellent agreement, with uncertainties of the order of 0.07% over a large number of shipments of plutonium oxide between the plants of La Hague in France and Sellafield in the United Kingdom (Swinburn and McGowan 1976). The operator’s sampling and subsampling pro- cedures must conform to internationally accepted standards (ISO 1993). Special care is further required in the conditioning of inspector’s subsamples to minimize the atmospheric effects and to allow a correction for changes in the chemical composition before the sample reaches the inspectorate laboratory. The selected can of powder is weighed just before sampling. A large primary sample is collected in a tared and gas-tight flask, under a dry controlled atmosphere in the glove box where the material is processed or stored. The primary sample flask should be transferred immediately to a subsampling glove box where its gross weight is measured on an analytical balance and recorded. Just prior to subsampling, the sample flask is shaken vigorously; the gross weight is checked and recorded. Duplicate subsamples of about 1 g are weighed to Æ0.5 mg or less into tared vessels. These vessels can be small glass tubes, like the inserts of the BC4 and BC5 brass containers (> Fig. 63.12), whenever one can use a Type B package carrying several grams of plutonium to ship a load of samples. These containers were actually designed and qualified for use with the PAT2 container, which had received approval for the shipment of Type B quantities of Pu samples to or from the USA (Kuhn et al. 1982). Their compact size make them very attractive for shipments of gram size Pu samples in any Type B container. . Fig. 63.11 UF6 subsampling tubes (IAEA 2004) Nuclear Safeguards Verification Measurement Techniques 63 2955
  • In other cases, the subsamples are dissolved in nitric acid in the operator’s analytical laboratory. An aliquant of each subsample, carrying about 4 mg of plutonium, is taken into a 5 ml penicillin vial and dried carefully to form an adherent film of nitrate salt on the bottom of the vial. The two aliquants are intended for elemental assays. Another fraction of at least one subsample is usually prepared for isotopic analysis and dried as above. With current products issued from high burnup spent fuel, some 20 penicillin vials could be packed together into a single Type A container and flown to SAL. Uranium Dirty Scrap Materials Uranium scrap materials in cans or hoppers are expected to be heterogeneous in element concentration and isotopic composition, and tend to change in composition upon exposure to atmosphere. As far as possible, the content of the container to be verified should be carefully mixed just before sampling. The item is weighed on a suitable scale immediately before duplicate samples of relatively large size are taken by collecting material from various spots in the bulk of the material. Each sample of about 30 g is introduced into a tared glass bottle, which is immediately capped tightly and weighed. Metal-capped bottles used in the pharmaceutical industry (> Fig. 63.13) are ideal for this purpose as their tare is very stable and the cap inner PE lining insures a very tight closure. The same bottles are used to take and ship gram size samples of uranium compounds in powder form, like UO2 products or yellow cake. The samples are packed without further handling in an ‘‘Excepted’’ or ‘‘Type A’’ package and flown to the inspectorate laboratory. . Fig. 63.12 BC5 and BC4 brass vials with glass tube insert (IAEA 2004) 2956 63 Nuclear Safeguards Verification Measurement Techniques
  • 63.4.3 Safeguards Analytical Laboratories Off-Site Laboratories The concept of IAEA’s safeguards analytical services (Lopez-Menchero et al. 1976) foresaw that the agency would establish and operate a fully equipped safeguards analytical laboratory (SAL). The analytical capability of SAL was to be such that samples taken from any key measurement point of the fuel cycle could be analyzed and that the results of these analyses would meet the requirements of safeguards accounting verification. To accommodate the 5,000 samples that were anticipated annually, SALwas to become part of a network of analytical laboratories (NWAL), along with existing laboratories nominated by member states for this purpose at the request of the IAEA. This approach was selected for reasons of economy and flexibility of services, but also because it provided the possibility of checking the accuracy of the analyses of actual inspection samples by comparing results obtained by different verification laboratories. Such an intercomparison was to become an important feature of the quality control program that the IAEA maintains to ensure the quality of measurements made by SAL and the network. Other safeguards systems operate in a similar way, which indeed strengthen the credibility of the verification measurements. After the successful completion of the PAFEX-l and PAFEX-2 tests, IAEA NWAL became operational in 1975 (> Fig. 63.14). In order to preserve the confidentiality of their origin, all IAEA inspection samples are sent first to the IAEA Laboratories in Seibersdorf (Austria), from where a number of them are redispatched to NWAL, in accordance with instructions of the inspectorate. In 1975, about 480 samples, mostly uranium products but also a few of plutonium dioxide, were analyzed. A maximum was reached in 1990 with 1,600 samples, including uranium, plutonium, and spent fuel samples received and analyzed (Deron et al. 1994). Safeguards analytical laboratories use essentially the same sorts of techniques and equip- ment as research and plant laboratories for nuclear material isotopic and elemental assay and . Fig. 63.13 Glass bottle for samples of uranium oxide powders and scrap (IAEA 2004) Nuclear Safeguards Verification Measurement Techniques 63 2957
  • material accountability. DA has the advantage that the sample is treated in a way to optimize the conditions of its final measurement. However, advances made in NDA measurements have found many applications in laboratories too, as can be seen in > Sects. 63.4.3, > 63.4.4, and > 63.5. On-Site Laboratories With the planning and construction of large spent fuel reprocessing or mixed U/Pu oxide fuel fabrication plants with annual throughputs close to 1,000 ‘‘Significant Quantities,’’ the safe- guards inspectorates planned minimal laboratory facilities directly on site. Their justification is to reduce strongly the number of samples of sensitive materials, which need to be shipped to distant laboratories, and to obtain analytical results faster, by eliminating long shipment times. Much research went into developing and testing robotized equipment to perform DA treat- ment and measurements automatically (Brandalise et al. 1994; Takahashi et al. 1994; Zahradnik and Swietly 1996) for their eventual installation at on-site laboratories. EURATOM set up on-site laboratories in La Hague in France (Regnier et al. 1994) and in Sellafield in the UK (Schneider et al. 1999). The Japanese state system and the IAEA operate jointly an On-Site Safeguards Analytical Laboratory (OSL) at the Rokkasho-Mura Reprocessing Plant (RRP) in Japan (Zahradnik-Gueizelar et al. 2007; Iwamoto et al. 2006). The latter includes (Matsuo 2008) a high activity laboratory handling spent fuel and high active waste samples in hot cells, a medium activity laboratory equipped with shielded glove boxes where plutonium and uranium products are measured and subsampled (> Fig. 63.15), a low . Fig. 63.14 The IAEA Network of Analytical Laboratories in the 1980s AEAT Harwell CEA Fontenay AEC Chalk river NBL Argonne ORNL Oakridge KRI St. Petersburg JAERI Tokai IRMM Geel CEN Mol ECN Petten EUREX Saluggia ITREC Bari NRI Prague Nuclear material analysis NWAL BAM Berlin NMCC Tokyo BHABBA Mumbai CRIP Budapest OeFZS Vienna Providers of reference material 2958 63 Nuclear Safeguards Verification Measurement Techniques
  • activity laboratory for chemical separations and alpha spectrometry, and a mass spectrometry laboratory (> Table 63.8). The pneumatic sample transfer system within OSL/RRP is connected to the operator’s automatic sampling and sample transfer systems. Inspection sampling jugs are fed into the . Fig. 63.15 OSL facilities at RRP. (a) Hot cells; (b) Glove Boxes (Matsuo 2008) a b . Table 63.8 OSL/RRP analytical equipment (Midorikawa et al. 2006) Laboratory area Sample handling equipment Analytical instruments Treatment/measurement See section High activity Hot cells HKED U, Pu, Np concentration 14.3.1 OSCA 244 Cm 14.3.1 Density meter Solution density Balance Subsampling/spiking Pu(VI) Spectrophotometer Pu concentration < 10 g/l FP separation unit Free U, Pu from FP Medium activity Shielded glove boxes HKED U, Pu concentration 14.3.1 Density meter Solution density Balance Subsampling/spiking HRGS Pu isotopics 14.3.1 Low activity Glove boxes Separation robot U, Pu separation Alpha spectrometer 238 Pu/(239 Pu + 240 Pu) and Pu concentration < 10 mg/l Mass spectrometry Glove boxes TIMS U, Pu isotopics and content by IDMS, 14.E4.5.5.2 Nuclear Safeguards Verification Measurement Techniques 63 2959
  • system from a jug-feeding machine installed in OSL/RRP. Authentication of inspectors’ jugs is further ensured by inspector passage detectors tracking the inspectors’ jugs between their departure from OSL, their arrival at the automatic sampling bench, and their return to OSL. Correlation with information from the operator-declared data and the automatic sampling authentication system confirm the authenticity of the inspector samples (Duhamel et al. 2004). 63.4.4 Isotopic Analysis Isotopic Analysis by Mass Spectrometry A mass spectrometer consists of three main components, which are kept under high vacuum: an ionization source, a mass analyzer, and a detector assembly (> Fig. 63.16). Ionization of the analyte may be produced by thermal ionization, inductively coupled plasma, or other means. The mass analyzer separates the ions according to their electrical charge and kinetic energy. It can be a time-of-flight system, a quadrupole, an electromagnetic sector, an accelerator, or another kind of mass filter. Electromagnetic sectors offer better mass resolutions and isotope abundance sensitivities than cheaper quadrupole analyzers and are therefore preferred to obtain the accuracies required for safeguards isotopic DA measurements. Isotopic Analysis by Thermal Ionization Mass Spectrometry Thermal ionization mass spectrometers (TIMS) with magnetic sector are thus the basic instru- ments for Pu and U isotopic analysis in safeguards inspection samples. The performance of commercial instruments has improved tremendously, particularly in the last 30 years, in terms of vacuum capability, design of the ionization source and detector assembly, stability of electrical supplies, sensitivity and linearity of ion current amplifiers, and includes full autom- atization of the measurement and data reduction. In modern instruments, the ionization source is equipped with a turret, which may hold 13 to MS filament assemblies (Tuttas et al. 1998). Each assembly carries one or two side filaments, usually in platinum or tungsten, and a center rhenium ionization filament. A 1 ml or less of sample solution containing 0.05 to 1 mg of uranium or 5–100 ng of plutonium is loaded onto a side filament, and dried electrically under controlled conditions to obtain a reproducible . Fig. 63.16 Main components of a mass spectrometer Sample Ion source Mass analyzer Detector Data analysis 2960 63 Nuclear Safeguards Verification Measurement Techniques
  • oxide deposit. The loaded turret is mounted inside the ionization source and the ionization source cavity is evacuated to 10À4 Pa or less. The samples are degassed in sequence before starting a cycle of measurements as follows. The temperature of the center ionization filament is raised to about 2,500 C and maintained at a constant value within 10 C, preferably under optical pyrometer control. The temperature of the sample is raised stepwise; molecular species volatilize and break down partially to produce U or Pu atoms. Impacts with electrons emitted by the center filament produce positive ions, which are extracted by a set of electrostatic focusing lenses under high voltage, and injected into the mass analyzer. The magnetic field of the analyzer is adjusted to collect the major analyte isotope ions in the center of the selected ion current detector. Focusing of the ion beam is adjusted to obtain a maximum intensity of the major isotope ion current. The instrument and its tuning are optimized to obtain a very flat top with sharp edges when scanning across an isotope ion peak (> Fig. 63.17). Various approaches have been devised to increase the thermoionization efficiency, such as electro-codeposition of actinides and platinum on the sample filament (Rokop et al. 1982), the resin bead loading technique (Walker and Smith 1979), or carburization (Jakopic et al. 2008). All these methods are said to facilitate the reduction of the sample species into atoms and enhance their ionization. An electrically heated cavity as sample holder installed in the source turret of a commercial mass spectrometer (Riciputi et al. 2002) yields also amazing increases of the ionization efficiency up to 5 and 7%. Ions captured in a faraday cup detector produce very low electrical currents, typically below 1 pA, which flow across a very large resistor to yield voltage signals in the range of 0.1–100 mV. The linearity of the response of faraday cups, the stability of the signal and the signal-to-noise ratio were greatly increased with the introduction and improvements of digital voltage amplifiers, the use of 106 MΩ voltage drop resistors, instead of the 105 MΩ resistors used earlier (Tuttas et al. 1998). In standard designs, the image of the exit slit of the ionization source, focused on the ion detector entry slit, has the same size than the source slit. Operating . Fig. 63.17 Typical scan over isotope ion peaks measured with compact discrete-dynode detectors 251.50 251.65 Mass (u) 251.80 IC2: 185ReO3 IC6: 192 OsO3 IC4: 188OsO3 IC3: 187 OsO3 0 50,000 100,000 150,000 200,000 cps Nuclear Safeguards Verification Measurement Techniques 63 2961
  • with a magnification of Â2 instead of Â1, and installing deeper faraday cups (Thermo Fisher Scientific: Triton 2005) provide greater space to install multidetectors and decrease the loss of secondary ions, minimizing thereby between-cup differences (> Fig. 63.18). Minor isotopes or samples of very small size, typically in ng range or below, are measured with ion multipliers rather than with faraday cups. Three types of ion multipliers may be used: Daly scintillation detectors (Daly 1960), discrete-dynode multipliers (DNM), or continuous electron multipliers (‘‘channeltrons’’), with amplifications of 104 –108 . Channeltrons are very compact but have a narrower dynamic range than DNMs. The output signal of the ion multiplier can be used to measure the instantaneous intensity of the ion beam as it is done with the faraday cup signal. Alternatively it serves to count the number of ions hitting the detector, using pulse counting technology. Channeltrons with a diameter of 2 mm, faraday cups, 3.5 mm in diameter, or 7 mm wide discrete-dynode detectors exist now. Multidetector assemblies with 8–12 detectors can be built with such devices (Tuttas et al. 2005). The individual detectors are spaced in a way to allow a simultaneous collection of the ion currents of isotopes of interest, namely 233 U, 234 U, 235 U, 236 U, 238 U and/or 238 Pu, 239 Pu, 240 Pu, 241 Pu, 242 Pu and 244 Pu. The ion synchronous collection eliminates the effects of changes in thermionic emission and time drift of the ion currents. This improves the reproducibility of the results and shortens the duration of a measurement, but raises the question of checking the relative ion sensitivity of the various detectors when using a multi-ion-detector thermal ionization mass spectrometer (MIC/TIMS). Highest isotope abundance sensitivity is needed to measure well very minor isotopes, e.g., to determine the age of a uranium sample (Wallenius et al. 2001). Use of tandem magnetic sectors or addition of an energy filter (Thermo Fisher Scientific, Patent 1993) like a retarding potential quadrupole lens (RPQ) (> Fig. 63.19) would do the job. All processes in the course of the loading of the sample and its TIMS measurement may entail small isotopic fractionation effects, which bias isotope ratio measurements. This is particularly the case during sample evaporation and ionization within the ionization source, which favor the lighter isotopes. These effects are globally corrected via a mass discrimination factor determined by measuring U and Pu isotopic standards, such as the ones listed in > Table 63.9, under exactly the same conditions than actual samples are. . Fig. 63.18 Increased magnification reduces losses of secondary ions within faraday cups Larger effective depth of cups: less emission of secondaries Secondary particles are released close to cup entrance Faraday cup Faraday cup Magnet lens Magnet lens Source Source 2962 63 Nuclear Safeguards Verification Measurement Techniques
  • . Fig. 63.19 Isotope abundance sensitivity with and without a retarding potential quadrupole 229.5 0 0 30 90 150 2 4 231.5Mass (u) A < 2 x 10–6 A < 5 x 10–8 232 Th/230 Th = 2900000 RPQ Peak tail of 232 Th . Table 63.9 Recommended uranium and plutonium isotopic reference materials for the determination of mass discrimination in isotopic analyses Supplier Reference Certified ratio Certified value (a) IRMM Geel (Belgium) (IRMM 2008) EC-NRM-199 233 U/238 U 1.00001 (30) 235 U/238 U 1.00015 (20) IRMM-074/1 233 U/238 U 1.02711 (26) 235 U/238 U 1.000254 (15) IRMM-3636 233 U/236 U 1.01906 (16) IRMM-290a/A2 239 Pu/242 Pu 1.013450 (82) IRMM-290b/A3 239 Pu/242 Pu 1.000730 (82) NBL Argonne (USA) (NBL 2008) NBL-U500 235 U/TotU 0.49696 (50) 238 U/TotU 0.49711 (50) NBL-128 239 Pu/242 Pu 0.99937 (26) NBL-137 239 Pu/240 Pu 0.24131 (39) CETAMA Bagnols (France) (CETAMA 2009) MIRF-01 239 Pu/242 Pu 0.9783 (5) MIRF-02 233 U/236 U 0.9681 (10) Note (a): The figures in parentheses represent the expanded standard uncertainties on the last decimal figures of the certified values, with k = 2. Nuclear Safeguards Verification Measurement Techniques 63 2963
  • In the standard data collection scheme, the isotopic ion beams are collected sequentially on a single detector by changing the magnetic field of the mass analyzer or the accelerating voltage of the ionization source. The mass spectrum of interest is scanned repeatedly in a way that allows to account for the changes in the rate of ion emission with time. Peak jumping is the preferred scanning scheme when the measurement is done in automatic mode. With multidetection the magnetic field and accelerating voltage are kept constant and the various isotope ions are collected simultaneously but in separate detectors. Ion current changes with time no longer perturb the measurement. However, as said above, it is necessary to verify that all detectors have the same response, or ion sensitivity, or otherwise to correct for differences (Ramakumar and Fiedler 1999). Total evaporation is very effective in minimizing the isotope discrimination effect. In this scheme, the sample is heated progressively and the isotope ion currents are collected as early as possible after the beginning of the ion emission. Data are collected until the sample has been totally exhausted. Total evaporation is readily applied in multi-detection mode (Callis and Cappis 1996; Aggarwal et al. 1999), but can also be used in single detection mode (Fiedler 1995). The method permits the analysis of samples much smaller than required for conven- tional techniques using Faraday collectors. Reproducibilities of 0.02% RSD and less have been obtained for U and Pu isotope ratios (Callis and Abernathey 1991). The value of the isotope ratio is derived from the ratio of the measured ion currents, or their integral, corrected for the isotope discrimination effects as shown in the example below: n 236 U À Á n 238 U À Á ¼ 1 À 2kð ÞI 236 U À Á I 238 U À Á ð63:1Þ where n(236 U)/n(238 U) is the ratio of 236 U and 238 U isotopes to be determined, I(236 U)/I(238 U) is the ratio of the measured ion currents, or their integral, k is the mass discrimination factor per mass unit. As said above, the factor k is obtained by measuring for example an isotopic reference material with certified 235 U/238 U ratio k ¼ R RðIÞ À 1=ð Þ 3= ð63:2Þ where R is the certified value of the isotope ratio n(235 U)/n(238 U) of the reference material R(I) is the ratio I(235 U)/I(238 U) of the measured ion currents. Ion fractionation can be very effectively corrected by an internal calibration after spiking a uranium sample with a mixture of isotopes of certified ratio, like IRMM-3636 (Richter et al. 2008) or CETAMA-MIRF-02, or spiking a plutonium sample with a certified mixture of 242 Pu and 244 Pu isotopes. Several batches of mixtures of 236 U, 233 U, 242 Pu, and 244 Pu isotopes were prepared for isotopic and IDMS assays of spent fuel solution samples using the method of internal calibration (Stepanov et al. 1990b). Calibration of such spikes could now be done to an accuracy of Æ0.05% (1s) against isotopic standards certified to Æ0.03% or less, like the 233 U/235 U/238 U reference materials IRMM-74/7-10 with low 233 U abundance. Unfortunately, there exists yet no mixture of 239 Pu and 240 Pu isotopes certified on a purely gravimetric basis. The results reported in Stepanov et al. (1990b) led to suspect that some small differences between laboratories may be due to uncorrected differences in detector sensitivities or errors therein, at the laboratories using MIC/TIMS instruments. On commercial MIC/TIMS, a program measures the relative gains of the various amplifier chains. A program connecting a given amplifier chain sequentially with the various ion detectors, called virtual amplifier, corrects for differences between amplifier chains. Yet it is also necessary to measure the ratios of the ion capture efficiencies of the various cups or ion multipliers. This may be done by 2964 63 Nuclear Safeguards Verification Measurement Techniques
  • producing a very steady ion beam, often the rhenium ion beam emitted by the ionization filament, and programming the instrument to peak jump from one collector to the next. More convincing but more tricky is to load a very accurately characterized mixture of three or more isotopes of the element of concern (Fiedler and Donohue 1988). The isotope ratios are measured in multidetection mode while the isotope ion spectrum is shifted by one detector in successive steps. > Table 63.10 describes a sequence of steps which may be used with a mixture of 239 Pu, 240 Pu, 242 Pu and 244 Pu isotopes, available with reference material UK Pu5/92138 issued by AERE, Harwell (Hamilton et al. 1989). The isotope fractionation at each step is corrected by reference to the certified isotope ratios. The idea was further developed by Dubois et al. into a dynamic multidetection measurement mode (Dubois et al. 1989), which practically eliminates mass fractionation effects and possible mismatches of cup efficiencies with a 2-isotope internal standard and MIC/TIMS. According to the results presented, pre- cisions and accuracies of 0.01% are achievable with this procedure. An ultimate refinement has been introduced by performing total evaporation measurements with peak tailing correction in dynamic multicollection mode, using a MIC/TIMS with magnetic sector equipped with a dispersion quadrupole (Goldberg et al. 2002; Richter and Goldberg 2003). High-accuracy measurements require very sophisticated instruments and measurement procedures, but isobaric interference of 238 U on 238 Pu measurement, and that of 241 Am on 241 Pu also deserve great attention. These can be minimized by taking advantage of different atom- ization kinetics. Actually in resin bead measurements the 238 U interference is corrected by collecting 235 U isotopes during the Pu measurement (Walker and Smith 1979). It is now becoming practical to do so with commercial MIC/TIMS (Aggarwal and Alamelu 2005). Yet in order to achieve best results it is customary to perform a careful chemical separation and to eliminate uranium contaminations. Chemical treatment is unavoidable in IDA to ensure a complete isotope exchange (see section ‘‘Isotope Dilution Mass Spectrometry’’). A chemical separation follows to provide U and Pu fractions in a clean and controlled matrix. When analyzing spent fuel samples the separation serves also to yield samples free from fission products, which do not need to be handled in a high activity laboratory (see > Sect. . Table 63.10 Example of a sequence of measurements of a 4-Pu isotope mixture for the determination of relative ion efficiencies of the 7 faraday cups of a MAT 261 mass spectrometer Step Faraday cup number 8 7 6 5 4 3 2 1 239 240 242 244 2 239 240 242 3 239 240 242 244 4 239 240 5 239 240 242 6 239 240 242 7 239 240 8 239 240 242 244 9 239 240 242 10 239 240 242 244 Nuclear Safeguards Verification Measurement Techniques 63 2965
  • Proven valency adjustment and separation procedures are outlined below. (a) ISO 8299 procedure (ISO 2005b): Reduce plutonium to Pu(III) and Pu(IV) in 7 M nitric acid with Fe(II), reoxidize to Pu(IV) with nitrite, then adsorb U and Pu in 7 M nitric acid on a quaternary ammonium anion exchange column, elute fission products, Am and other potential interferences with 7 M nitric acid before collecting the uranium fraction; finally reduce Pu on the column to Pu(III) with hydroxylamine and elute it with the same reagent. (b) ISO 15366 procedure (ISO 1999): Reduce and reoxidize Pu in 3 M nitric acid as in the previous procedure, adsorb Pu and U on a TOPO impregnated silicagel column, wash out fission products, Am and other potential interfering species with 3 M nitric acid, then condition the column with 1.7 M nitric acid before reducing Pu on the column with a freshly prepared solution of 0.1 M HI/1.7 M HNO3; elute Pu with the HI/HNO3 reagent, and finally wash out U with water. (c) LANL procedure (Marsh et al. 1974): Oxidize Pu to Pu(VI) by fuming the sample with HClO4, re-dissolve the sample in 12 M HCl and adsorb U and Pu on a quaternary ammonium anion exchange column, elute fission products, Am and other interfering species with 12 M HCl; then reduce Pu on the column and elute it with a reducing solution of 0.1 M HI/12 M HCl; finally wash out U with dilute 0.1 M HCl. (d) UTEVA procedure (Morgenstern et al. 2002): In 6 M HNO3/0.3% H2O2 a sample of spent nuclear fuel is expected to contain Am(III), Pu(IV) with 13% Pu(III), Np(IV), and U(VI) along with the fission products. In this medium Pu, Np and U are adsorbed on a liquid ion exchanger column (UTEVA), while the fission product and Am are not retained. Pu is fully reduced to Pu(III) and eluted with 2 M HNO3/0.002 M ascorbic acid. Then Np is recovered with a 2 M HNO3/0.1 M oxalic acid eluent. Finally, U is eluted with 0.007 M ammonium oxalate. (e) The resin bead procedure (Marsh et al. 1981): An aliquot of sample containing about 1 mg of U and/or 10 mg of Pu is adjusted to 8 M HNO3 and subjected to a redox valency adjustment with Fe(II) and nitrite as in procedures (a) and (b). The sample is equilibrated with about 10 anion exchange resin beads of around 300 mm diameter. Each bead collects about 10 ng of U and/or Pu and is mounted on a mass spectrometer filament for a sequential measurement of the Pu and U isotopic compositions. Isotopic Analysis by Inductively Coupled Plasma Mass Spectrometry While the ionization filament in a TIMS may be heated to a maximum temperature of 2,500 C, inductively coupled plasma reach temperatures of 15,000–20,000 C and have much higher ionization efficiency. A pneumatic nebulizer serves to inject sample solutions directly into the plasma. Inductively coupled plasma mass spectrometers (ICP-MS) have thus become powerful tools in trace and ultra-trace analyses with detection limits down to 1 ppq or less. > Section refers to ICP-MS applications in measuring alternative nuclear mate- rials, Am, Np, and Cm, and > Sects. 63.5.2 and dwell upon their use in environmental analyses for safeguards. Commercial ICP-MS equipped with magnetic sectors and multidetection could meet the requirements of safeguards accountancy verifications (Pereira de Olivera et al. 2008). Prior chemical separations of the nuclear elements are, however, of benefit to reduce isobaric interferences by the numerous polyatomic ions generated by the plasma source. An elegant solution is to attach a liquid chromatography unit in line with the ICP source (Guenther- Leopold et al. 2005). Double focusing mass filters or diverse energy filter components 2966 63 Nuclear Safeguards Verification Measurement Techniques
  • overcome some of the interferences. Collision reaction cells (Varian 2009) are also common accessories to segregate polyatomic ions and might improve the isotopic analysis of actinides too. Ion multiplier detectors should be protected against impacts from high-energy photons and neutral particles, as done with the 90 deflector of Varian (Elliott et al. 2005) (> Fig. 63.20). Active research in safeguards applications of ICP-MIC-SFMS is thus to be expected in DA of nuclear fuels as well as in environmental analyses. 238 Pu Abundance by Alpha Spectrometry Isobar 238 U can be a source of positive bias in the determination of 238 Pu isotopic abundance by mass spectrometry despite prior chemical separation of the two elements. In this respect, eluting Pu before U is advantageous (see section ‘‘Isotopic Analysis by Thermal Ionization Mass Spectrometry,’’ procedures (b)–(d)). Great attention must be placed anyway to minimize sources of blanks due to low levels of natural uranium in most chemical reagents and mass spectrometry filaments. It is therefore still customary to perform 238 Pu measurement by alpha spectrometry in parallel with mass spectrometry. 238 Pu abundance in safeguards samples varies between 0.001% in so called weapon grade plutonium to over 1.5% in plutonium from high burnup spent fuel. A typical alpha spectrum (Bagliano et al. 1991) obtained with an ion-impact detector with a resolution of 12 keV is shown in > Fig. 63.21. The presence of Am-241 with alpha rays in the range of 5.39–5.49 keV will yield high results. > Table 63.11 lists the relevant nuclear data. The accuracy of the alpha spectrometry of plutonium depends strongly upon the chemical preparation of the sample, the production of carrier-free alpha sources, the counting geometry, the detector resolution, and the method used to correct for the tailing of the 239 Pu and 240 Pu alpha peaks. . Fig. 63.20 The Varian 90 degree deflector (Elliott et al. 2005) Skimmer core Photons and neutrals X Y Electrostatic field Mass analyzer entrance Nuclear Safeguards Verification Measurement Techniques 63 2967
  • . Table 63.11 Relevant nuclear data (Nichols et al. 2008) Nuclide Half-life [year] Radiation Energy [keV] Emission probability [%] 238Pu 87.74 a 5,456.3 28.29 Æ0.03 5,499.03 69.3 239Pu 24,110 a 5,105.5 11.94 Æ30 5,144.3 17.11 5,156.6 70.77 240Pu 6,561 a 5,123.68 27.16 Æ7 5,168.17 72.74 bÀ 20.78 100.00 241Pu 14.290 a 4,853.0 0.0003 Æ0.006 4,896.3 0.002 242Pu 373,000 a 4,858.1 23.49 Æ3,000 4,902.2 76.48 244Pu 8.00 Â 107 a 4,546 19.4 4,589 80.5 241Am 432.6 a 5,388.2 1.660 Æ0.6 5,442.8 13.1 5,485.56 84.45 g 59.54 35.9 26.34 2.27 . Fig. 63.21 Typical plutonium alpha spectrum (Bagliano et al. 1991) 100 10,000 D C B Exp. fit tail correction 4.94 4.90 5.10 5.17 5.46 5.50 4.874.614.544.28 5.20 5.27 5.53 a b c d A E x p • f i t MeV 238 Pu 239 Pu 240 Pu 241 Pu 242 Pu 2968 63 Nuclear Safeguards Verification Measurement Techniques
  • The separation of 241 Am deserves particular attention, when the 241 Pu abundance is high and when a long time has elapsed since the plutonium fuel material was produced. This is the case with samples of high burn-up spent fuel. The standard anion exchange method described in section ‘‘Isotopic Analysis by Thermal Ionization Mass Spectrometry,’’ para. (a) is well suited for the purpose, as Pu is eluted long after Am, fission products and U are. It is a good practice, nevertheless, to measure the 241 Am gamma emission at 60 keV in the separated Pu fraction to verify that it will not bias the results. The aim of the source preparation is to obtain very thin, smooth, and uniform deposit to minimize the effects of self-absorption and retrodiffussion, which alter the resolution of the alpha spectrum. Some of the methods are listed below: Electrodeposition on polished platinum or stainless steel disks, with a variety of supporting electrolytes (ISO 2007d; Ingelbrecht et al. 1997; Amoudry and Silly 1984; Hallstadius 1984; Lally and Glover 1984). Vacuum sublimation on glass (Amoudry and Eloy 1984). Drop deposition, usually with addition of spreading agents such as tetraethyleneglycol (ISO 2005c; Miguel et al. 1984). Deposition on a mass spectrometry filament to perform alpha and mass spectrometry measure- ment on the same filament (Strebin and Robertson 1977). The deposit should not carry more than 10–20 mg/cm2 of matter to minimize self- absorption (Amoudry and Silly 1981). The support material must have a finely polished surface and be inert to corrosion. Support material of high atomic weight reduces retrodiffusion of alpha rays. Gold-plated copper or steel support produces sources of resolu- tion down to 12 keV. Several publications report comparisons of different methods of source preparation (Aggarwal et al. 1985; Lally and Glover 1984). The spectrum of alpha radiation may be measured in a gas proportional counter or a gridded ionization chamber (Bortels et al. 1985), but semiconductor detectors are more commonly utilized in nuclear fuel analyses. The properties of Si detectors for alpha spectrom- etry have been studied extensively (Bortels et al. 1994, 1993; Bortels 1991; Bortels and Bauer 1990a, 1990b; Aggarwal et al. 1988). Particle implanted passivated silicon (PIPS) detectors nowadays present resolutions down to 9 keV (> Fig. 63.22), which permit, with a favorable geometry (Amoudry and Silly 1981), a deconvolution of the 239 Pu and 240 Pu alpha peaks in the range of 5.0–5.2 keV (Amoudry and Burger 1984). The counting chamber is kept under reduced pressure of 0.5 Pa or less during measurement. A number of elaborate peak fitting algorithms have been proposed to correct for the tailings and deconvolute overlapping peaks (Raab and Parus 1994; Babeliowsky and Bortels 1993; Garcia-Torano 1990; Amoudry 1983). Deconvolution of the 239 Pu and 240 Pu peaks is not normally done to measure the 238 Pu abundance in safeguards samples, as more precise results are obtained by combining the alpha activity ratio 238 Pu/(239 Pu + 240 Pu) with the 240 Pu/239 Pu isotope ratio measured by mass spectrometry. The 238 Pu/239 Pu isotope ratio is calculated according to > Eq. (63.3): R8 ¼ RaT8 TÀ1 9 þ R0 T0= À Á ð63:3Þ where R8 is the 238 Pu/239 Pu isotope ratio, Ra the 238 Pu/(239 Pu + 240 Pu) alpha activity ratio, R0 the 240 Pu/239 Pu isotope ratio measured by mass spectrometry, T8, T9, T0 are the halflives of isotopes 238 Pu, 239 Pu, 240 Pu, respectively. Nuclear Safeguards Verification Measurement Techniques 63 2969
  • A simple algorithm is sufficient to correct the (239 Pu + 240 Pu) alpha activity for the tailing of the 238 Pu peaks, such as a ‘‘linear decrease,’’ an ‘‘exponential decrease’’ (ED), or a GP (geometric progression) tail correction. Results obtained using the GPRANL code (Gunnink et al. 1984), the ED or GP correction methods agree within 0.3% of the mass spectrometry results, as long as the 238 Pu/(239 Pu + 240 Pu) alpha activity ratio stays below 5 (Bagliano et al. 1991; Aggarwal et al. 1980; Ramaniah et al. 1980). In these cases, and with sources of suitable quality, the counts around 5.17 MeV, at the 239 Pu and 240 Pu peaks, should be at least 300 to 500 times larger than the counts in the valley around 5.25 MeV. Then tailing corrections will not exceed 0.2 to 0.5%. At present, IAEA SAL estimates at 0.4% the total uncertainty (1s) of its determinations of 238 Pu abundance in safeguards samples by a combination of alpha and mass spectrometry taking into account the contributions of short and long term measurement reproducibility, calibration, and nuclear data uncertainties. Gamma Spectrometry of Nuclear Material Samples Gamma ray spectrometry is widely used in safeguards laboratories to measure the isotopic abundance of various U and Pu isotopes but also their concentration, as well as the concen- tration of ‘‘alternative nuclear elements’’ (see Sect. fission and decay isotopes (Parus and Raab 1996; Parus et al. 1987). Laboratory applications make use of the developments of NDA technology mentioned in > Sect. 63.3.1, but are usually part of DA procedures. The 235 U abundance of LEU products is determined with a total uncertainty of 0.13% (1s) by measuring the 186 keV gamma rays with a NaI or GeLi well detector on 5 ml solutions containing a well-known amount of uranium, close to 0.5 g. The volume of the sample and synthetic calibration test solutions, their acid and U concentrations are kept in a very narrow range so that the self-absorption and counting geometry be constant factors or minor variations be accurately . Fig. 63.22 Alpha spectrum of a 239 Pu/240 Pu mixture obtained with a PIPS detector of 9 keV resolution 10 100 1396 1896 Channels 239 Pu (5.10) 239 Pu (5.14) 240 Pu (5.12) 240 Pu (5.17) 239 Pu (5.16) 1,000 10,000 Counts 2970 63 Nuclear Safeguards Verification Measurement Techniques
  • corrected. The technique is a good back up to mass spectrometry in case of a pile up of samples. Alternatively, the procedure can be used to measure U concentration of samples with known 235 U abundance, when the accuracy of a titration is not required. High-resolution gamma spectrometry (HRGS) is routinely used as screening and quality control tool in processing nuclear material as well as environmental samples (see > Sect. In IAEA/SAL vials of Pu product samples are all subject to gamma spec- trometry before any wet chemical treatment is undertaken. Estimates of elemental and isotopic composition of U, Pu, Am, or other relevant radioisotopes serve to orient further processing and are recorded for later use. The gamma spectometry data help to select the optimal spike for IDMS or IDAS, when spiking is to be done in the laboratory. On the whole, U and Pu isotopic gamma results will be matched with the mass spectrometric assays to exclude any gross error in the analytical process. Gamma rays at 60 keV will detect 241 Am, which may remain in the separated Pu fractions, before alpha and mass spectrometry (see > Sects. and > Fission products are also measured in spent fuel samples when inspectors want to verify burn-up data. The MGA (Wang et al. 1999; Buckley et al. 2000) and MGAU (Gunnink et al. 1994) gamma spectroscopy softwares, designed for the low-energy region, are used commonly for measure- ments of Pu and U in product or scrap samples, while general surveys of fission products require measurement over the entire energy range and call usually upon commercial softwares (Canberra 2009a, 2009b). 63.4.5 Elemental Assay Ignition Gravimetry of U, Pu, Th Ignition gravimetry remains the most reliable and accurate analytical method for elemental assay, especially if gram size samples are available. It is most easily applied for the analysis of nuclear grade oxides of uranium, plutonium or thorium, or to samples that can be readily transformed into an oxide, such as powders and solutions of nitrate salts (Rodden 1953; ISO 1987a, 1987b, 2003, 2004b, 2007). The sample in oxide form is ignited, usually in a platinum crucible, to a temperature between 870 and 1,250 C, depending on the element, for 1 hr periods to reach a constant weight and the formation of a stoechiometric oxide. The average relative atomic mass of uranium or plutonium, W, must be derived from their measured or known isotopic composition. The total mass fraction of nonvolatile impurities, e, remaining in the ignited oxide must also be measured or known with a suitable precision and accuracy. The mass fraction, C, of the analyte, M, (M = U, Pu or Th), is derived from > Eq. (63.4): C ¼ mo 1 À eð ÞG ms= ð63:4Þ where mo is the mass of ignited oxide, ms the mass of the sample, G the gravimetric factor calculated with > Eq. (63.5): G ¼ xW xW þ yWoð Þ= ð63:5Þ where Wo = 15.9994 is the relative atomic mass of oxygen, and MxOy is the stoichiometric molecular formula of the ignited oxide, with M = U, Pu or Th. A precision of 0.05% or less is expected if the sample contains less than 0.05% nonvolatile impurities. Random uncertainties may increase to 0.15% if the impurity content reaches 0.5%. Nuclear Safeguards Verification Measurement Techniques 63 2971
  • Deviations of the ignited oxide from stoechiometry lead to systematic errors, which should not exceed 0.1%. Systematic errors may also come from calibration errors in the impurity analyses and from systematic errors due, e.g., to uncertainties in their chemical form. The result carries a positive bias, if an impurity, although present, is not measured or detected. Less pure materials may be analyzed by ignition gravimetry after a suitable chemical purification. For example, thorium in ores, alloys or other materials may be separated by precipitation as oxalate (Scott 1939; Ewing and Banks 1948; Willard and Gordon 1948; Kall and Gordon 1953), which is then ignited to ThO2 at 95 C. However, the extra steps bring additional random as well as systematic errors. Uranium Titration Most uranium titration procedures rely on the redox properties of the U(VI)/U(IV) couple. In early redox titration methods, U(VI) was reduced with Ti(III), Cr(II), Pb, Cd amalgam or a Jones reductor. Most of these procedures are not very selective, require a prior removal of nitrate, or operate at elevated temperatures and under inert atmosphere. Samples of nitric solutions, adjusted to 1N HNO3 and 2N H2SO4, can be titrated directly by Ce(IV) with o-phenanthrolin as colorimetric end point indicator, after Ti(III) reduction (Corpel and Regnaud 1962). The method was adapted to a gravimetric titration by Cr(VI) with ampero- metric end point (Nair et al. 1996) and applied to diluted dissolver solutions, containing 100–300 mg of uranium, with burnup of up to 10,000 MWd/t. Uranium can be titrated potentiometrically with Fe(II) from strong phosphoric acid medium (Rao and Sagi 1962). Fe(II) and W(VI) do not interfere, but V(V), Mo(VI), and Cr(II) do. A fast and very selective procedure was derived from the latter, based on the reduction of U(VI) by an excess Fe(II) in a phosphoric acid medium, the catalytic destruction of the excess Fe(II) by nitric acid in the presence of molybdate, followed by a dilution and a titration with a standardized solution of dichromate with a colorimetric end point detection (Davies and Gray 1964). Out of 23 species, only Ag(I), V(IV), or V(V) appear to interfere appreciably. In the Davies & Gray/ NBL-modified procedure a potentiometric titration with vanadyl ion as electrochemical indicator replaced the colorimetric titration (Eberle et al. 1970). This procedure is widely used for accountability and safeguards verification measurements. ISO 7097-1 describes its application to a gravimetric titration of aliquants of samples of uranium nitrate solutions, metal, oxides and hexafluoride of nuclear grade, containing about 50 mg of uranium (ISO 2004a). The overall process is equivalent to a titration of U(IV) with dichromate according to > Eq. (63.6) below: Cr2O7 2À þ 3U4þ þ 2Hþ ! 2Cr3þ þ 3UO2 2þ þ H2O ð63:6Þ The mass fraction of uranium in the sample, CU, in g/g, is equal to: CU ¼ WT mc 2ma=ð Þ ð63:7Þ where W is the relative atomic mass of the uranium in the sample, T the titer of the dichromate solution in equivalent/gram, mc the mass of dichromate solution used to reach the end point, in gram, ma the mass of sample aliquant used in the titration, in gram. The final titration is usually performed with a commercial automatic volumetric titrator. Gravimetrically calibrated burettes yield coefficients of variation of 0.05% or less for the overall uncertainties of random or systematic nature in the assays of samples of uranium and mixed 2972 63 Nuclear Safeguards Verification Measurement Techniques
  • uranium/plutonium products. Robot systems (> Fig. 63.23) automatize the entire process, including the redox steps preceding the titration, and run reliably unattended overnight with the same precision and accuracy than with the semi automatic procedure (Aigner and Aigner 1994). A scaled down procedure processes 5–25 mg aliquants of uranium with no loss of precision (Slanina et al. 1976). The Davies & Gray procedure is sufficiently selective to be applicable to the determination of uranium in dissolver solutions from spent fuel reprocessing plants. The dichromate titrant is particularly attractive as its solutions are very stable as long as they are stored in hermetic vessels, and can be calibrated directly versus primary standards, namely, NIST 136f dichromate (Montgomery and Sanerwein 2008) and NBL 112A (NBL 2008) or CETAMA MU2 (CETAMA 2009) uranium metal. Nonetheless, NBL felt compelled to replace dichromate by Ce(IV), which is not listed as a toxic chemical (Zebrowski et al. 1995; ISO 2007a). Plutonium Titration Plutonium titrations take general advantage of the properties of the Pu(III)/Pu(IV) or Pu(IV)/ Pu(VI) electrochemical couples. Microvolumetric titrations of mg amounts of plutonium with potentiometric end point detection were widely used in the past (Metz 1957; Fudge et al. 1960). Plutonium was often reduced to Pu(III) with an excess Ti(III), which was titrated color- imetrically (Corpel and Regnaud 1962) to Pu(IV) with Ce(IV), after destroying the excess Ti(III). The method was adapted to a sequential amperometric titration of U and Pu in samples of breeder reactor fuel materials (Chadwick and McGowan 1972). 500 mg samples of Pu metal . Fig. 63.23 IAEA-SAL robotized system for the treatment and titration of uranium samples by the NBL modified Davies & Gray method (Aigner and Aigner 1994) Nuclear Safeguards Verification Measurement Techniques 63 2973
  • can be dissolved in 6N HCl to obtain Pu(III), which is titrated spectrophotometrically with Ce(IV) in presence of ferroin with uncertainties of Æ0.07% (Caldwell et al. 1962). 50–140 mg samples of Pu metal, dissolved in 4N sulfuric acid and reduced to Pu(III) on a Jones reductor, may be titrated gravimetrically with Cr(VI) using two polarized gold electrodes to detect the end point, with a coefficient of variation of 0.05% and 100.01% recovery (Pietri and Baglio 1960). In a more selective method, Pu is reduced to Pu(III) with an excess Cu(I) in a solution of HCl/sulfamic acid and aluminum chloride. Fe(II) and the excess Cu(I) are reoxidized by Cr(VI) up to a potentiometric end point. Sulfuric and phosphoric acid are then added to lower the Pu(III)/Pu(IV) redox potential, so that Pu(III) can be titrated with Cr(VI) without oxidizing chloride (Davies and Towsend 1974). Another group of methods relies on a prior oxidation of plutonium to Pu(VI) by fuming perchloric acid (Waterbury and Metz 1959), argentic oxide (Drummond and Grant 1966), sodium bismuthate (Charyulu et al. 1984), or Ce(IV) (MacDonald and Savage 1979). In the so- called AgO-method, a longtime accountability method at many bulk handling plants, the excess AgO is destroyed by sulfamic acid, an excess standardized ferrous sulfate is added to reduce Pu(VI) to Pu(IV) and the excess Fe(II) backtitrated with Ce(IV) by potentiometry with polarized electrodes (Corpel and Regnaud 1966) or by amperometry with Cr(VI) (Cherry et al. 1968). However, Ag+ , coming from the destruction of AgO, oxidizes a fraction of the excess Fe(II) and introduce a very reproducible bias of 0.2%. The titrant therefore must be standard- ized against a Pu standard solution treated in the same way than the samples (Cauchetier et al. 1975). 241 Am gives also a positive bias for Am/Pu ratios above 0.02 because of a partial oxidation of Am or the presence of radiolytic H2O2 (Spevackova et al. 1978). Unfortunately, Cr interferes, yielding high results in case of corrosion of process tanks. A method, free from Fe or Cr interference, (MacDonald and Savage 1979) involves a prior oxidation of plutonium to Pu(VI) with an excess of Ce(IV). Sulfamic acid is added to avoid side reactions with nitrites. Fluoride residues from the dissolution of solid samples are complexed by additon of Al(III). The excess Ce(IV) is then reduced by a slight excess of arsenite in the presence of Os(VIII) as catalyst. Excess As(III) is oxidized by permanganate, also in a very quantity. Finally, oxalic acid is added to reduce the excess Mn(VII). Pu(VI) is then reduced quantitatively to Pu(IV) with Fe(II) and the excess Fe(II) backtitrated with Cr(VI). All steps are followed potentiometrically and carefully timed. The Pu mass fraction, CPu, in g/g, in the sample aliquant submitted to titration is calculated according to the following equation: CPu ¼ W Tf mf À Tcmcð Þ 2ma= ð63:8Þ where W is the relative atomic mass of the plutonium in the sample, Tf the titer of the Fe(II) solution, in equivalent/g, mf the mass of Fe(II) solution, in g, Tc the titer of the dichromate solution in equivalent/g, mc the mass of dichromate solution used to reach the end point, in g, ma the mass of sample aliquant used in the titration, in g. A scaled down procedure, applicable to the titration of 4 mg of plutonium, was set up at IAEA/SAL with the assistance of Dounreay (MacDonald and Savage 1987) and of the Nuclear Research Institute in R´ ez´ (CSSR) (Kuvik 1991). This made possible to receive and titrate samples of Pu products taken in Japanese facilities, as only mg amounts of plutonium can be air flown from Japan in Type A packages. The process including the preliminary redox steps, was automatized to achieve coefficients of variation of 0.05% (Kuvik et al. 1992; Ronesch et al. 1992). 23 species do not interfere, but vanadium does quantitatively, and Np partially. The test sample should carry less than mg-amounts of nitrite, fluorosilicate, and iodate. The effect of americium remains to be studied (ISO 2000). 2974 63 Nuclear Safeguards Verification Measurement Techniques
  • Controlled Potential Coulometry of Plutonium Before operational amplifiers became available, constant current coulometry (CCC) was used more often than controlled potential coulometry (CPC) because the instrumentation was so much simpler. CCC continued to be used for highly precise U determinations in very pure materials (Goode et al. 1967; Malinowski 1967; Merciny et al. 1981). In CPC, the measurement cell (> Fig. 63.24) consists of a sample cell [1], a working electrode [4], a counter electrode compartment [5], and a reference electrode compartment [6]. A potentiostat supplies an electrolysis current between the working and the counter electrode in a way to maintain a chosen difference of potential between the working and the reference electrodes. Normally, a Pu assay involves two steps: first, the potential of the working electrode is set to a value leading to the reduction of Pu to Pu(III); the electrode potential is then changed to a value required to reoxidize Pu(III) to Pu(IV). The reduction and reoxidation potentials are selected to reach near quantitative electrolysis of plutonium while minimizing the electrolysis of possible interfering species. A current integrator measures the quantity of electricity needed to achieve the reoxidation of Pu(III) to Pu(IV) and relates it to the amount of plutonium. Dilute sulfuric (Shults et al. 1959; Jackson et al. 1980), nitric (Holland et al. 1998), perchloric (Metz and Waterbury 1962), hydrochloric (Shults 1963), or phosphoric acid (Ba¨rring and Jo¨nsson 1970) have been used as supporting electrolyte. The choice of the supporting electrolyte determines the magnitude of interferences from potential impurities (Milner and Phillips 1974): iron interferes stronger in nitric acid than in sulfuric acid, but little or not in hydrochloric acid, as indicated by the Pu(III)/Pu(IV) and Fe(II)/Fe(III) formal potentials (> Table 63.12). Oxygen removal is also a greater challenge with sulfuric acid than with nitric acid. . Fig. 63.24 CPC cell 1 Sample cell 2 Stirrer 6 Reference electrode 3 Inert gas Counter electrode 5 4 Working electrode Nuclear Safeguards Verification Measurement Techniques 63 2975
  • Chemistry may be combined with CPC to improve selectivity. For example (Jackson et al. 1980) applying the chemistry of the Davies and Townsend titration (see > Sect., Pu is first reduced electrolytically to Pu(III) in an hydrochloric/sulfamic acid mixture along with iron and other potential interfering species. Iron and a number of the other impurities are reoxidized next without reoxidizing Pu(III). Finally, phosphate is added to lower the redox potential of the Pu(III)/Pu(IV) couple to permit a quantitative electrooxidation of Pu(III) to Pu(IV), without oxidizing chloride ions. The procedure is very selective and is applicable to Pu determinations in the presence of uranium in 10:1 U/Pu molar ratio. CPC at a solid electrode has also been applied to determine uranium (Davies et al. 1970; Phillips and Crossley 1978), which is reduced from U(VI) to U(IV) with electrogenerated Fe(II). The approach has been used for sequential coulometry of U and Pu (Angeletti et al. 1969; Davies et al. 1970; Fardon and McGowan 1972; Phillips and Crossley 1978). Controlled potential reduction of Pu(VI) to Pu (IV) has also been used to avoid iron interference, but plutonium must be oxidized chemically by AgO, Ce(IV) or fuming HClO4 prior to its coulometric determination (Shults 1963). The characteristics of a coulometric method depend also upon the choice of the working electrode. Electrode double-layer charging, surface oxide formation, hydrogen evolution at low potential, and electrode or electrolyte oxidation at higher potentials contribute to the back- ground current and depend upon the electrode material and the electrolyte. Mercury pool electrodes have been used for uranium coulometry, but efficient stirring of the solution/ electrode interface is difficult to achieve. Long electrolysis times lead to uncertainties in the background current correction. Faster electrolysis is achieved with solid working electrodes such as a Pt gauze electrode (Shults et al. 1959; Shults 1960; Jackson et al. 1980), or a gold mesh electrode (Holland et al. 1998). Glassy carbon was used as indicating electrode in Pu titrations and CPC of iron, chromium, uranium, and neptunium was done with a glassy carbon crucible as working electrode (Plock and Vasquez 1969). Electrochemical reactions at solid electrodes may be slowed down considerably by adsorption or deposition of trace species in solution. Pretreatment and storage of the electrode must be subject to well-validated procedures in order to maintain or restore the electroactive properties of the working electrode. Samples having undergone an ion exchange purification carry resin degradation species, which accelerate electrode ‘‘poisoning.’’ This effect is overcome by fuming the purified sample in concentrated sulfuric acid before performing the CPC measurement. The design of the electrolysis cell is a critical parameter. The objective is to achieve a fast and rigorous removal of dissolved oxygen, to maintain accurate control of the potential of the . Table 63.12 Formal potentials of Fe(II)/Fe(III) and Pu(III)/Pu(IV) redox couples, volt versus standard calomel electrode (SCE) Electrolyte Concentration Potential Fe(II)/Fe(III) Potential Pu(III)/Pu(IV) Difference H2SO4 0.5M 0.492 HNO3 0.9M 0.66 0.668 0.01 HClO4 1M 0.53 0.73 0.20 HCl 2.5M 0.446 0.705 0.259 5.5M 0.372 0.686 0.315 7.4M 0.323 0.640 0.317 2976 63 Nuclear Safeguards Verification Measurement Techniques
  • working electrode, to achieve near-quantitative electrolysis in the shortest possible time, and to minimize diffusion and leakages between the working electrode compartment and the counter electrode and reference electrode compartments. The working electrode compartment being usually a small cylindrical glass vessel, ideal control of the electrode potential would require a perfect cylindrical symmetry of the working electrode, the counterelectrode, and the mechan- ical stirrer. Full symmetry never exists in practice, as either the counterelectrode (Holland et al. 1998) or the stirrer is off center (Jackson et al. 1980). It is essential to locate the tip of the reference electrode compartment in a region of highest current density and as close as possible to the surface of the working electrode, in order to best control the electrochemical reactions occurring there (Harrar 1968; Harrar 1987; Harrar and Shain 1996). The larger the ratio of the area of the working electrode to the volume of the test solution, the faster is the electrolysis. Wrapping several layers of platinum or gold gauze as the working electrode is probably not so much increasing the effective area of the electrode but decreases the volume of the test solution. Yet the benefit is partially offset by the increase of the capacitive component of the background current. Analysts deploy much ingenuity in devising special designs of the mechanical stirrer to reach the fastest stirring rate without solution splashing. Modern electronics will meet the following essential instrumental requirements (Holland et al. 1998). For assays of up to 15 mg of plutonium, the potentiostat should be able to supply a current output of 250 mA or more, in order to observe an exponential decay of the coulometric current at the outset of the electrolysis. The potential control circuitry should be fast enough to raise the control potential by 1 V within a millisecond with less than 1 mV overshoot. The selected potential should remain stable to Æ1 mV during measurements. The meter measuring the difference of potential between the working and reference electrodes should have an input impedance of 1011 O or more. The current integrator should cover a range of 50 mA to 100 mA with a linearity of 0.005% or less. It is recommended to include in the equipment an adjustable constant current supply with a stability of 0.002% to perform an electrical calibration of the current integrator with an accuracy of 0.01%. If room temperature varies excessively, the instrument should be located in a temperature-controlled cabinet to limit electronic drift to the above values. The norm ISO 12183 incorporates the latest technological experience (ISO 2005d). It describes the coulometric assay of 4 to 15 mg aliquants of plutonium in 0.9M nitric acid, using a gold working electrode. The current integrator is calibrated electrically. A blank is measured first under conditions similar to those used for a plutonium sample. The plutonium test aliquant is weighed into a coulometric cell, fumed to dryness with sulfuric acid and re- dissolved with the blank solution. The test solution is degassed and about 99.9% of the plutonium reduced to Pu(III). The rest potential of the working electrode is measured. The potential of the working electrode is raised to oxidize Pu(III) to Pu(IV). The electrooxidation current is integrated until about 99.9% of the electrolysis is completed. The quantity of electricity used in the oxidation and the rest potential of the working electrode at the end of the oxidation are recorded. The fraction, f, of plutonium electrolyzed is calculated according to the Nernst equation. Stopping the electrolyses before their completion shortens the process and reduces the background current correction and its uncertainty. With automatic coulometers, like the ones described by Jackson et al. (1980) and Phillips et al. (1977), uncomplete electrolysis is not usually corrected mathematically but via a chemical calibration against plutonium reference materials. The Harwell instrument (Phillips et al. 1977) was equipped with an automatic sample changer. Pd, Ir, Pt, Au interfere but are normally not of concern in nuclear fuel materials. Interfer- ence of Fe or Np can be corrected if their content is measured or known and their atom ratio to Nuclear Safeguards Verification Measurement Techniques 63 2977
  • plutonium does not exceed 0.005. A confidence interval of 0.1 to 0.2% is achievable for a single detetermination of plutonium at a confidence level of 0.95. This includes the contribution of uncertainties of systematic character, which should not exceed 0.03%. The uncertainties will increase if the sample must be purified by a suitable chemical separation before applying the coulometric measurement (Pietri et al. 1981; Mitchell et al. 1990). The accuracy of the procedure is verified by measuring aliquants of standard plutonium solutions. With increasing concerns about the treatment of radioactive wastes, CPC of plutonium or uranium in sulfuric or nitric acid is becoming more attractive, as titration procedures like the Davies&Gray or the McDonald&Savage methods yield complex liquid wastes, which are difficult to condition. Isotope Dilution Assays Isotope dilution analysis (IDA) involves mixing an aliquant of the sample with an aliquant of an artificially enriched isotope of the element to be analyzed. The content of the latter in the original sample is derived from the results of the isotopic analysis of the mixture compared to the isotopic compositions of the sample and of the enriched isotope. The term ‘‘spiking’’ is often used to refer to the mixing step. The aliquant of enriched isotope added to the sample is also commonly called a ‘‘spike.’’ IDA is used in particular for the determination of uranium, plutonium, or thorium in solutions of irradiated nuclear fuels. It has also been used for an independent measurement of the total volume or mass of solution in accountability tanks of chemical processing plants. Radiometric Isotope Dilution Assay Techniques Gamma spectrometry could readily be used to determine uranium with good accuracy by spiking the sample with a uranium spike of sufficiently different 235 U abundance, and mea- suring the 185 keV gamma ray emission of the 235 U isotope before and after spiking with enriched 235 U. Alpha and gamma spectrometry were shown to be applicable to the IDA of plutonium (Mayer et al. 1995; Parus and Raab 1996) measuring the ratio of the 238 Pu activity to the sum of the 239 Pu and 240 Pu activities. The 238 Pu/(239 Pu + 240 Pu) alpha activity ratios should not exceed 5 in the spiked and unspiked samples (Ramaniah et al. 1980; Parus et al. 1992), when determining Pu concentrations by isotope dilution alpha spectrometry (IDAS). A plutonium containing 95% of 239 Pu or more and less than 0.01% of 238 Pu is a convenient spike for IDAS of high burnup plutonium samples holding, e.g., 1% of 238 Pu and 50% of 239 Pu, just as it is too for isotope dilution mass spectrometry (IDMS). Use of a 238 Pu spike is recommended only when the 238 Pu isotope abundance does not exceed 0.1% in the unspiked sample. Precautions required to reach accuracies of a few percent by IDAS were described in > Sect. Isotope Dilution Mass Spectrometry Although the technique is more expensive than radiometry, mass spectrometry is, however, the preferred method for IDA of samples of nuclear materials taken for accountability verifications. Traditionally, separated or highly enriched isotopes, absent from the samples or only present as minor isotopes, such as 233 U, 242 Pu, 244 Pu or 230 Th were used as spike material in isotope dilution mass spectrometry. IDMS is the standard operator method for the accountability of nuclear materials in input solutions of irradiated nuclear fuels. The method involves taking a representative sample from the input accountability tank and diluting it accurately 100– 1,000-fold in a heavily shielded cell. A portion of the diluted solution is transferred to 2978 63 Nuclear Safeguards Verification Measurement Techniques
  • a medium activity laboratory where it is subsampled and mixed with an accurately measured aliquot of a calibrated dilute spike solution. IDMS is also the basic technique for safeguards verification measurements on spent fuel solutions and active liquid wastes from reprocessing plants, although it may be used too for the verification of the composition of any sample taken in the nuclear fuel cycle. At first, the IAEA used 233 U and 242 Pu spikes in the form of a dried nitrate carrying about 1 mg of uranium and 10 mg of plutonium in a sealed penicillin vial. The inspector witnessed the taking of the primary sample, its dilution and requested the weighing of a milliliter of the diluted input solution into duplicate dried spike vials. The inspector witnessed further the heating of the subsamples with nitric acid to re-dissolve the mixed spike, their drying, the capping of the penicillin vials, their bagging out and packaging in Type A containers, which were air flown to Seibersdorf. (a) The resin bead method. In the resin bead method (Carter et al. 1976) a few beads of anion exchange resin are used to retrieve minute amounts of the unspiked and spiked samples, and transferred to the safeguards laboratory. At the analytical laboratory, a single bead of each sample is mounted on the filament of a thermal ionization mass spectrometer to measure plutonium and uranium sequentially. It was hoped that such resin bead samples may be air mailed to SAL in Seibersdorf in accordance with the regulations of the international air transport association (IATA) and the universal postal union (UPU) valid in 1985. It is essential to achieve full chemical equilibrium between the sample and the spike before adsorbing plutonium on resin beads. According to (Marsh et al. 1981) ferrous sulfamate and nitrite are the most effective reagents to do this. The optimized procedure was tested successfully at La Hague (Cesario et al. 1987) and at the reprocessing plant in Tokai Mura in the frame of the TASTEX program (Walker et al. 1981); the test samples having been sent by registered air-mail or shipped as ‘‘excepted’’ air cargo. Current international air mailing of radioisotopes (UPU 2009) have become more restrictive but shipping resin bead samples of irradiated fuel samples as ‘‘excepted’’ air cargo remains possible (IATA 2010). Yet, this attractive approach has not been implemented further, mainly because the preparation of the samples, although simple, requires more handling by the plant operator and therefore more efforts from the inspector for witnessing or authenticating these operations. (b) Spiking undiluted spent solution samples. Actually, enriched 235 U and low burnup pluto- nium with a 239 Pu abundance of 95% or higher are very suitable spikes for the assay of samples of industrial nuclear materials. Alloy chips containing about 50 mg of 20–90% enriched 235 U and 2 to 4 mg of plutonium with a high 239 Pu abundance, known as UPu metal spikes, were used by EURATOM inspectors to spike samples of concentrated input solutions directly (Debievre and Van Audenhove 1976). The homogeneity of the alloys is considered a limiting factor in the accuracy of their certification. A similar method based on the use of Al capsules holding a dried mixture of uranyle and plutonium nitrate suitable for spiking a sample of the concentrated input solution (Koch et al. 1976). Both techniques avoid the need for an accurately measured dilution and its independent verification. It is also claimed that an effective valency cycling and isotopic equilibration of the plutonium is achieved during the re-dissolution of the metal alloy or the aluminum capsule. A small fraction of the spiked solutions containing 1–10 mg of plutonium are taken, dried, and shipped to the safeguards laboratory. (c) The large size dried spike method. In the large-size dried spike (LSD) procedure, selected by the Nuclear Material Control Centre Laboratory (NMCC) in Tokai Mura and the Nuclear Safeguards Verification Measurement Techniques 63 2979
  • IAEA (Kuno et al. 1989), the spike is a mixture of uranyle and plutonium nitrate carefully dried in a 10 ml penicillin vial. For the assay of spent fuel solutions and MOX products with U/Pu ratios around 100:1, each vial contains about 50 mg of uranium-enriched just below 20% in 235 U and 2 mg of plutonium with a 239 Pu abundance of 95% or more. Calculations (Laszlo et al. 1991) and experience show that these amounts are appropriate to reach relative uncertainties around 0.1% in a single measurement of samples carrying about 100 mg of depleted uranium and 1 mg of high burn plutonium with a 239 Pu abundance of 50%. The mass of uranium or plutonium, m, in the sample aliquant, in g, is derived from > Eq. (63.9) below: m ¼ S Is I=ð Þ Rs À Rmð Þ Rm À Rð Þ= ð63:9Þ where S is the mass of spike plutonium added to the sample aliquant, in g, Is is the 235 U or 239 Pu isotopic abundance in the spike, in weight %, I is the 235 U or 239 Pu isotopic abundance in the sample, in weight %, R, Rs, Rm are the 235 U/238 U or 240 Pu/239 Pu isotope ratios, respectively in the sample, spike, and their mixture. The diagrams in > Fig. 63.25 assume that the 240 Pu/239 Pu atom ratios in the unspiked and spiked samples can be measured with a relative uncertainty of 0.05%. The uranium enrichment should not exceed 20% so that the spike does not come in the category of high enriched uranium (HEU), subject to stricter safeguards controls, and will not be allowed in facilities, which are not licensed for the processing of HEU materials. (d) Preparation and certification of LSD spikes. NMCC and IAEA SAL jointly prepare a stock solution by dissolving metal standards of natural or depleted uranium (NBL112A, EC101, CEA-MU2 or JAERI-U4), HEU (NBL-NRM116) and plutoniumwith high 239 Pu abundance (NBL 126, EC 201, CEA-MP2 or NBS 949) and mixing them to obtain a concentration of about 2 mg/g of plutonium and 50 mg/g of uranium with a 235 U abundance just below 20%. The titers are calculated from the ‘‘make-up’’ data and the certificates of the source . Fig. 63.25 Random measurement uncertainty of the plutonium assay using the LSD spike, as a function of the ratio of mass of 239 Pu spike and mass of Pu in sample 1 2 0.0 Puassaymeasurementuncertainty(%) 0.2 0.4 239 Pu abundance in spike 95% 98% 4 Ratio (239Pu)spike/(Pu)sample 2980 63 Nuclear Safeguards Verification Measurement Techniques
  • metals. The titers are verified (Doubek et al. 1991) by mass spectrometric analyses of the isotopic compositions, potentiometric titration of uranium and potentiometric titra- tion, or controlled potential coulometry of plutonium in accordance to ISO 10980 (ISO 2005b). The advantage of the LSD spikes is that they can be prepared from primary reference materials more easily than UPu alloy spikes. On the other hand, the dissolution of the LSD spike with the spiked sample does not bring the benefit of a redoxequilibration of the sample and spike plutonium, as the dissolution of metallic spikes does. It is therefore essential to avoid losses of LSD spiked samples after their drying in the plant laboratory and until their chemical treatment at the safeguards laboratory. As a quality assurance measure, a number of LSD spikes shipped to a plant under safeguards are returned unused to the verification laboratory in order to check that the spikes are not damaged during shipments or storage. LSD spikes of certified isotopic and element compositions are available from IRMM since the 1970s (IRMM 2008; Verbruggen et al. 2008a). Automatic equipment (Verbruggen et al. 2008b) was installed to boost the production, also starting from certified metal reference materials. (e) Analytical measurement procedures. At the safeguards laboratory, the unspiked and spiked samples are re-dissolved in nitric acid and undergo a chemical valency adjustment to achieve isotopic equilibration between the sample and the spike. Uranium and plutonium are separated from each other and from interfering isobares or species altering the ionization and/or fractionation pattern of uranium or plutonium during their measure- ment by thermal ionization mass spectrometry. Alternative chemical treatments were described in > Sect. about isotopic analysis by mass spectrometry. The procedure follows essentially the international standard ISO 8299 (ISO 2005a) with an expected precision and accuracy of 0.1% or better. The calculations are also in line with the mathematical model described by (Kipphardt et al. 2004). (f) Verifications of tank volumes. For an independent verification of the total mass or volume of solution of nuclear material in accountability tanks of chemical processing plants a known amount, S, of a tracer element is added to the nuclear fuel solution in the tank. Only elements which do not disturb the plant process can be used as tracers. They should also be absent or present only in low concentrations in the nuclear material solution. Natural lithium (Bokelund 1970), magnesium (Mathew et al. 1976), and lutetium (Bokelund et al. 1996) have been used as tracers in demonstration exercises. The tank is carefully mixed after the addition of the tracer, the solution is sampled, the sample is diluted by a factor K, and aliquants of the diluted sample are then spiked with 6 Li, 26 Mg or 176 Lu enriched isotope. Li, Mg, or Lu is separated after a chemical equilibration step. A blank sample, taken prior to the addition of the tracer in the tank, is treated in the same way to determine the isotopic composition and the concentration of Li, Mg, or Lu, which may be present initially in the tank solution. The concentration of added tracer, T, observed in the tank solution, is derived from the dilution factor, K, TIMS measurements of the isotope ratios 6 Li/7 Li, 26 Mg/24 Mg or 175 Lu/176 Lu in the blank sample, in the tracer and in the sample collected after the addition of the tracer in the tank, and the isotopic compositions of the blank and the tracer. The mass of the solution, M, in the tank is equal to: M ¼ S T= ð63:10Þ The diluted sample can also be spiked with an enriched isotope of uranium, pluto- nium, or thorium to determine by IDMS its concentration in nuclear fuel element, Cd, Nuclear Safeguards Verification Measurement Techniques 63 2981
  • along with its concentration in added tracer, Td. The total mass of nuclear fuel element present in the tank, NF, is readily calculated according to > Eq. (63.11): NF ¼ S Â Cd Td= ð63:11Þ This estimate is independent of the dilution factor, K, which does not need to be measured. Risks of sample contamination are high with Li or Mg tracers and isotopic fraction- ation is large during lithium TIMS measurements. Neodymium has been considered as an alternative tracer (Debievre 1997), as it is a fission product present in measurable concentrations in spent fuel solutions and can serve as in situ spike. The accuracy of TIMS measurements is now so good that it could be practical to use enriched 235 U or 233 U as a tracer for the direct measurement of the total content of fissile material in a tank, skipping altogether the verification of the volume or mass of solution in the tank (Debievre and Perrin 1991). Nonetheless, the efficiency of the mixing of the tracer and the fuel solution, the impact of solution trapped in ancillary tubes and a number of other points remain to be investigated, so that the procedure has rarely been implemented in safeguards verifications up to now. Spectrophotometric Determination of Hexavalent Plutonium The absorption peak by PuO2 2+ ions at 831 nm is the basis of selective Pu determinations (ISO 2009; Cauchetier 1981). Pu in nitric acid solutions is quantitatively oxidized to Pu(VI) with ceric nitrate or argentic oxide AgO. A recording double-beam spectrophotometer is used to scan the absorption spectrum between 800 and 860 nm (Schmieder et al. 1970). The absorption peak at 831 nm is very sharp with a half height width of 26 A˚ (> Fig. 63.26). The peak height is proportional to the concentration of plutonium in the sample up to 0.55 g/l, provided that the spectrophotometer has a resolution of 4 A˚ or less and the scanning speed is slow enough to allow the recorder to complete a full scale excursion while scanning across the absorption peak. . Fig. 63.26 Pu(VI) absorption peak at 831 nm (Schmieder et al. 1970) 800 0.0 0.2 0.4 0.6 0.8 E 900 nm Absorption spectrum of Pu(VI) 0.43 mg/ml in (a) 0.56 M HNO3 and (b) 14.1 M HNO3 (a) (b) 2982 63 Nuclear Safeguards Verification Measurement Techniques
  • The method is calibrated against Pu standard solutions treated in the same way than the unknown samples. Acidity, anions and temperature affect the oxidation of plutonium and the Pu(VI) extinc- tion coefficient (Savage et al. 1986). Oxidation by Ce(IV) is less prone to anion interferences than AgO oxidation. The acidity should preferably not exceed 2 N when using Ce(IV) as the oxidant, except in the presence of phosphates. Yet for a long time AgO was the preferred oxidizing agent at La Hague, particularly in the assay of dissolver solutions containing always a significant fraction of fine solids (Patigny 1982). The extinction coefficient of Pu(VI) decreases by 0.7% with an increase of 0.1 M in nitrate concentration. It decreases too with temperature by 0.5% per  C. All these parameters must be carefully controlled and be the same for the treatment and measurement of the samples and calibration standards. Pu(VI) spectrophotometry is actually used in many instances at spent reprocessing plants, including in their accountability programs (Grison 1980; Holland and Dorsett 1986). Fiber optics has eased its applications to analyses in glove boxes and hot cells (Schott 1984). Diode- array spectrophotometers should find applications in process control and safeguards (Cauchetier et al. 1985), as the optical spectra give abundant information on the species present in process solutions. Pu and U spectrophotometry is amenable to automation (Jackson et al. 1976). Use of Nd(III) as internal standard and authentication tag was validated for on-site safeguards measurements of plutonium in highly active waste solutions in spent fuel reprocessing plants by Pu(VI) spectrophotometry (Surugaya et al. 2008). A reproducibility of 10% should be achievable for plutonium concentrations around 10 mg/l. The limit of detection varies between 0.1 mg/l and 2mg/l depending upon the authors. Higher reproducibilities and accuracies of 0.06–0.1% were obtained (Nikitina et al. 2007) with a two-channel photometer for the characterization of 240 Pu, 242 Pu or 244 Pu spikes or their mixtures. The two channels are centered on the Pu(VI) absorption peak at 831nm and in a region of low absorbance at 720 nm. The ratio of the light intensities in the two channels is taken as a measure of the plutonium light absorption. The technique was proposed for the determination of U, Pu, and fission products such as Nd and Rh in dissolver solutions (Stepanov et al. 2002) with expected reproducibilities of 0.1 to 0.25%. LSD spikes containing around 50 mg of 20% enriched uranium and 2 mg of plutonium have been verified with the same technique and precision. With such a capability, the technique would qualify as a bias defect detection method. X-ray Absorption and Fluorescence Spectrometry In the 1970s (Pella and von Baeckman 1969; von Baeckman 1971), an instrument was developed for the automatic dilution and L-X-ray fluorescence analysis (XRFA) of U and Pu in spent fuel solutions with a precision and accuracy of about 1%. The hybrid K-edge densitometer (HKED) turned out to find quicker applications for safeguards verifications (Ottmar et al. 1986; Ottmar and Eberle 1991) at the input of spent fuel reprocessing plants, as the sample needs neither a dilution nor any treatment before measurement. The technique, considered as an NDA method and described in > Sect. 63.3.1, is installed in particular in on-site analytical laboratories and operated by analytical chemists on duty there. Operators have found the instrument so practical and reliable that they use it in their nuclear material accountability program (Brousse et al. 1993). The HKED fluorescence channel is also used for DA of Th or mg-size Pu samples, Nuclear Safeguards Verification Measurement Techniques 63 2983
  • measuring the Th/U or Pu/U ratios against a uranium ‘‘internal standard’’ (Ottmar et al. 1991; Doubek et al. 1994). L X-ray fluorimetry with commercial XRF analyzers has been applied to determine U, Th and Pu in oxides and mixed oxides dissolved in borax or polyphosphate glass beads with 0.3% reproducibility. Average relative differences of Æ0.2% were observed with results of titrations or gravimetric assays with standard deviations of 0.5% on the differences (Bagliano et al. 1991). The technique has the advantage of yielding only solid wastes and no liquid waste. These instruments are also powerful tools to determine the overall chemical composition of mixed fuels, alloys, nuclear scrap, sludges, fuel cladding, and structural components, etc. Analyses of impurities in ignited oxides are an important component of U, Th, and Pu gravimetric determinations (> Sect. Checks of the chemical purity of reagents and samples at key analytical steps are another application of X-ray fluorimetry to the control of the quality of DA procedures. Assay of Alternative Nuclear Materials The irradiation of nuclear fuels in the reactors produce important amounts of fissile isotopes such as 237 Np, 241 Am, and 243 Cm, besides Pu. Since 1993, the IAEA is called to take these isotopes into consideration in its safeguards duties, as a category of ‘‘alternate nuclear mate- rials’’ (ANM), which might be used to fabricate nuclear explosives (Charlton and Stanbro 2001; Bathke et al. 2009). To guard against such misuses, it was agreed at present to verify that undeclared production of these isotopes does not remain concealed. This is done through an evaluation of the flow sheet of the chemical processing plants (Koch et al. 1997; Rance et al. 2007) and the verification that the process is not modified to allow a separation of ANMs. To perform such ‘‘flow sheet verifications’’ the inspectors must be able to use at least some qualitative or semiquantitative tools to confirm that ANMs are not present in unexpected steps or amounts in the processes. Safeguards focuses particularly on 237 Np, which is a long-lived nuclide, as it is expected to be easier to use to produce alternate nuclear fuels or nuclear explosives. An attractive proposal is to combine U, Pu, Am verifications by HKED with a simple DA measurement of Np/Pu ratios (Ottmar et al. 2001) at spent fuel reprocessing and Pu fuel fabrication plants (> Fig. 63.27). A large arsenal of ‘‘conventional’’ methods exists for measurements that are more elaborate whenever this would be needed. Separated ANMs can be assayed by gravimetry, titration, or coulometry (Crossley 1988). Smaller or more dilute materials may be analyzed by spectropho- tometry (ISO 1997; Kageyma et al. 2001), L X-ray fluorimetry (Benony et al. 1994), IDMS (Efurd et al. 1986), IDAS (Aggarwal et al. 2005) or isotope dilution gamma spectrometry (IDGS) (Sus et al. 1996). Gamma (Ludwig et al. 2008) or alpha spectrometry (ISO 2007c; Sus et al. 1998), differential pulse polarography (Cauchetier 1978), laser-induced fluorimetry (Stepanov et al. 1990a; Aleksandruk et al. 1991) are applicable to very dilute materials. The sensitivity of laser-induced photoacoustic spectrophotometry (LIPAS), developed to investi- gate actinide chemistry in dilute solutions, would allow to determine Am down to 10À8 M and Np down to 10À7 M (Ewart et al. 1987), with careful control of the oxidation state and chemical forms. A review of ‘‘nonconventional’’ methods (Rosenberg 1992) concludes that ICPMS can be the most attractive and versatile instrument as it already is in environmental analyses (Varga et al. 2006), when combined with modern chromatographic separations (Vajda et al. 2009). 2984 63 Nuclear Safeguards Verification Measurement Techniques
  • 63.5 Environmental Sampling and Analysis to Verify the Completeness of State Declarations 63.5.1 Introduction Starting from the inception of international safeguards under INFCIRC-66 and -153 type agreements, the main emphasis of IAEA safeguards inspections was to verify that declared nuclear material placed under safeguards was accounted for and had not been diverted for non- peaceful purposes. Concerns about the existence of undeclared nuclear materials or activities in states subject to comprehensive safeguards agreements were not directly addressed by the IAEA, with the assumption being that any such activities would be detected by other means. The IAEA’s involvement in such investigations began with the implementation of United Nations Security Council Resolution 687 (United Nations Security Council 1991) following the expulsion of Iraqi armed forces from the territory of Kuwait in early 1991. UNSCR 687 entered into force on April 3, 1991, and specified that the IAEA, in conjunction with a UN Special Commission (UNSCOM), would carry out inspections inside Iraq for the destruction, removal, or rendering harmless of all ‘‘nuclear weapons or nuclear-weapons-usable material or any subsystems or components or any research, development, support or manufacturing facilities related to the above.’’ The IAEA, in turn, set up an Action Team consisting of safeguards inspectors and experts, supplemented by experts supplied by its member states, to carry out these inspections aimed at understanding all aspects of the Iraqi nuclear activities carried out prior to 1991 in breach of their safeguards undertakings under the NPT. . Fig. 63.27 Proposed scheme for the control of alternative nuclear materials at a spent fuel reprocessing plant with the HKED (Ottmar et al. 2001) HKED [Np], [Am] HLLW Process step Measurement (analyte) HRGS [Np]/[Pu] FP separation HKED [U], [Pu], [Am] Dissolved spent fuel Accountancy tank Np, Am, FP separation U-Pu purification Pu product HRGS [Np], [Am] Nuclear Safeguards Verification Measurement Techniques 63 2985
  • The IAEA Action Team inspections in Iraq began in May, 1991, and continued until December 1998 during which time a number of previously unknown nuclear facilities and processes came to light, including the uranium recovery from phosphate ores at Al Jesira and Al Qaim, the electromagnetic isotope separation facilities at Al Tarmiya and Ash Sharqat, the centrifuge facility at Al Furat and the nuclear weapon design and explosive test facility at Al Atheer (Dillon 2002). The details of this undeclared nuclear development program were painstakingly pieced together from physical evidence (buildings and equipment), interviews with Iraqi scientists, documents seized by the inspectors and, last but not least, by measuring small traces of nuclear material in so-called ‘‘environmental samples’’ taken at known and suspected nuclear facilities in Iraq (Donohue and Zeissler 1993; Zeissler and Donohue 1995). The collection and analysis of such environmental samples were a completely new activity for the IAEA. It exposed the agency to a number of developments that had been carried out in the member states prior to that time. Among these were methods for sampling radioactive effluents in waterways, for taking swipe samples of dust in buildings and from equipment as well as the analysis of such samples by highly selective and sensitive methods such as the fission- track method combined with thermal ionization mass spectrometry (FT-TIMS). The uncovering of the Iraqi undeclared nuclear program starting in 1991 demonstrated that the traditional nuclear accountancy-based safeguard system applied up to that time needed to be strengthened. Addressing the correctness of the states’ declarations was not a protection against a state that was determined to carry out undeclared nuclear activities for the production of special nuclear materials and with the intention of hiding this from IAEA safeguards inspectors. This revelation convinced the IAEA and its member states of the need to develop new approaches and tools to provide additional assurances about the completeness of the states’ declarations and the absence of undeclared materials and activities. The efforts to strengthen agency safeguards in light of these events was embodied in the Program 93+2, which was initiated in 1993 and was planned to finish in time to provide input to the NPT Review and Extension Conference planned for 1995 (IAEA 1994, 1995, 1996). Program 93+2 studied a number of strengthening measures such as the use of third-party or open-source information, use of satellite imagery and import-export information, as well as the logistics and optimized use of environmental sampling and analysis. The environmental sampling activities under Program 93+2 focused on 12 field trials in which a wide variety of environmental samples were taken around known nuclear facilities in 11 member states (see > Table 63.13). The main conclusions from these field trials were that nuclear activities, especially those involving large amounts of nuclear or radioactive materials were detectable in the environment at distances of 10–20 kilometers by use of water, sediment, biota, and vegetation samples. However, the dilution effects were found to be large and the signatures were often mixed with global fallout or releases from other facilities. These effects tend to reduce the usefulness of such sampling for safeguards purposes and to raise the question of the cost-effectiveness of such sampling on a regular basis to detect undeclared activities in a wide-area approach (i.e., where the exact source of effluents is not known a priori). However, the field trials did demonstrate the utility and cost-effectiveness of swipe sampling inside facilities; the complete details of declared activities could be revealed by sensitive and selective analysis of such samples (see > Sects. 63.5.5 and > 63.5.6). In parallel with the 93+2 field trials, the IAEA designed and constructed a special Clean Laboratory for Safeguards in Seibersdorf, Austria (see > Sect., which went into full operation in 1996. In the years since 1996, the environmental sampling for safeguards program 2986 63 Nuclear Safeguards Verification Measurement Techniques
  • has grown in terms of the number of samples taken and the number of facilities that have been sampled. Protocols have been implemented for all aspects of the program, including sample kit production, sample taking and shipment, screening, analysis at SAL/CL or in the network of analytical laboratories (NWAL) in the agency’s member states, as well as the evaluation of the data and drawing of safeguards conclusions (IAEA 2005). Between 300 and 700 swipe samples . Table 63.13 Field trials of environmental sampling for safeguards under Program 93+2 Country Facility Date Conclusions Sweden Power reactors, Research center Sept 1993 Fission and activation products were seen in effluent water released in a coastal area Hungary Power reactor Oct 1993 Elevated fission and activation products in a river were possibly due to Chernobyl accident USA Enrichment plant Mar 1994 Enriched uranium could be seen in vegetation, water and sediment at a distance of several km Japan Reprocessing plant Apr 1994 No clear indication of reprocessing could be seen in water released in a coastal area South Africa Enrichment plant Apr 1994 Enriched uranium particles were seen in vegetation and swipes from vegetation as well as in swipes from inside buildings Australia Research center Apr 1994 Enriched uranium used as targets for medical isotope production could be seen in swipe samples as well as depleted uranium used in research activities Argentina Enrichment plant May 1994 Swipe sampling in buildings showed depleted and enriched uranium particles typical of enrichment operations Indonesia Research center May 1994 Low levels of fission and activation products were seen in water and vegetation. Swipe samples taken in buildings showed low and high-enriched uranium particles from research and medical isotope production activities Republic of Korea Research center June 1994 Water samples showed both depleted and enriched uranium associated with research and fuel fabrication activities. Swipe samples from buildings showed depleted and low-enriched uranium particles UK Reprocessing Sept 1994 Fission products associated with fuel reprocessing were seen in water and vegetation; uranium particles found in swipe samples from process buildings were diagnostic of the processes carried out Netherlands Enrichment Plant Dec 1994 No signatures of enrichment could be found in water, sediment, biota, or vegetation, but swipe samples from inside buildings showed depleted and enriched uranium particles representative of enrichment operations. Japan Enrichment Mar 1995 No signatures of enrichment were found outside the facility, but swipe samples from inside buildings showed clear signatures of enrichment operations Nuclear Safeguards Verification Measurement Techniques 63 2987
  • are taken each year by agency inspectors and each sample is subjected to screening and detailed analysis at two separate laboratories in order to obtain cross confirmation about the presence of any unexpected or undeclared nuclear materials present. The following sections give a detailed picture of the environmental sampling process. 63.5.2 Sampling, Conditioning, and Shipment of IAEA Safeguards Environmental Inspection Samples Cotton Swipe and Other Swipe Materials The basic IAEA swipe sampling kit (see > Fig. 63.28) consists of six cotton wipers, which are 10 Â 10 cm in size. Mini-grip bags of two different sizes are included to individually bag and double-bag the swipes after sampling; this double-bagging is the principle means to stop cross- contamination of the sample from that point onward. Each sampling kit also contains two pairs of clean-room latex gloves, a sample data sheet, pen, and extra labels. The kits are prepared in a Class 100/ISO Class 5 work area that is monitored before and after each batch . Fig. 63.28 Standard cotton swipe sampling kit 2988 63 Nuclear Safeguards Verification Measurement Techniques
  • of kits is prepared to ensure that the cleanliness class is not exceeded. In addition, swipes are taken of the workbench before and after sample kit preparation and these swipes are analyzed by bulk and particle methods to detect unusually high levels of U or Pu or the presence of particles containing these elements. A certain number of unused kits from each production batch are also archived in case of future questions about their cleanliness. Finally, the cotton wipers used in making the kits are subjected to destructive (bulk) analysis whenever a new batch is received; typical levels of U in the wipers are from 1–3 ng whereas Pu is below the detection limits of the most sensitive techniques available (less than approximately 1 fg). A second type of swipe material (designated Type-J) consists of cellulose material similar to filter paper with a circular center part of 2.5 cm diameter. These cellulose wipers are used to take swipe samples inside hot cells. Four wipers, attached to a plastic holder, are introduced into the hot cell of interest and used by the remote manipulators to take a swipe sample. Upon exiting the bank of hot cells, usually through a glove box, one of the wipers is detached and given to the facility operator for making radioactivity measurements that will determine the shipment conditions for the remaining three wipers that are sent back to the IAEA in Austria. Each of the three hot cell swipes is packaged in a plastic bottle that is bagged in a polyethylene mini-grip bag or heat-sealed in plastic (so-called bag-out) before being placed in a Pb or stainless steel container to provide a minimum of shielding for energetic beta or gamma radiation. The three shielded samples are then placed in a special sealable aluminum can and eventually in a steel drum or other container that is licensed for transport of excepted or type-A quantities of radioactive materials. The Type-J wipers are also used to take samples in a certain enrichment facility under safeguards in a nuclear weapons state. The wiper is inserted in a special metal fitting (a so- called Koshelev fitting) that is part of the pipe-work connected to the enrichment cascade. Therefore, this sample comes into contact with the UF6 gas in the pipe-work and can be used to detect the presence of material with higher 235 U enrichment than declared. Air Filters During the IAEA Action Team inspections in Iraq (see > Sect. 63.5.1), a variety of sampling media were tested, in the context of non-site-specific or wide-area environmental sampling. In addition, various field trials of wide-area environmental sampling were carried out in which high-volume air samplers were employed. One such sampler is shown in > Fig. 63.29. It consists of a high-volume air blower system pulling ambient air through a 20 Â 30 cm paper filter. Typical air volumes sampled by such devices are in the range 40–900 m3 per hour. Deployment of air filter samplers in a safeguards role presents the following challenges: Each sampler must be supplied with electrical power and is subject to interruptions if used in an unattended mode. Cameras or other systems would be needed to ensure that the sampler is not tampered with. Clogging of the filter medium with sand or high dust loading would limit the duty cycle between filter changes. Analysis of the resulting filter samples would be possible by nondestructive methods such as gamma spectrometry or X-ray fluorescence but these do not represent the most sensitive methods. Bulk analysis would be affected by the high dust loading of the filter and perhaps by the filter material itself, e.g., glass-fiber filters would be difficult to destroy chemically Nuclear Safeguards Verification Measurement Techniques 63 2989
  • and would add unwanted elements. Particle analysis would present similar difficulties as with soil sampling. Use of high-volume air samplers in a wide-area sampling scheme would be prohibitively expensive because a ‘‘grid’’ of samplers would be needed with samples collected every week, representing many hundreds or thousands of samples per year for a medium-sized area (for example, Iraq has an area of 438,000 km2 ). For the reasons listed above, as well as owing to the large dilution effects that can be expected as a plume of effluent leaves a nuclear facility and travels down-wind, high-volume air sampling was not chosen as a routine sampling method for IAEA safeguards. The IAEA reserves the right to implement such wide-area sampling under the additional protocol (IAEA 1997) in those cases where it is both feasible and cost-effective. Water, Soil, Vegetation, and Biota Samples Samples of the true ‘‘environment’’ such as water from rivers and lakes, sediment, biota, vegetation, and soil, were taken in Iraq by the IAEA Action Team in the period 1991–1998 and during the field trials carried out under Program 93+2. Protocols and sampling equipment were developed and tested for these different sample types. One such device was a high-volume water sampling system based on a sample collection cartridge connected to a water pump powered by a 12 V battery, as shown in > Fig. 63.30. The cartridge contains a paper filter to trap particulates and radionuclides adsorbed on particles and a mixed ion exchange bed to remove dissolved species such as 137 Cs. The flow rate through such a sampler is in the range of 200–300 liters per hour. . Fig. 63.29 High-volume air sampler 2990 63 Nuclear Safeguards Verification Measurement Techniques
  • Sampling of soil, vegetation, and water-borne biota for safeguards purposes was performed during the 93+2 field trials using ad hoc methods developed at the time. Plastic mini-grip bags were used for soil and vegetation and 1 liter wide-mouth plastic bottles were used for biota sampling. A more systematic study of such sampling methods was carried out and special protocols were developed (Rose et al. 1997). Since that time, these types of samples have rarely been taken by IAEA inspectors and, therefore, will not be covered here in detail. Process and Structural Materials There is no limit to the types of materials that might be taken as safeguards samples in the search for undeclared activities. Elemental impurities can be used to trace the origin of nuclear materials through various stages of a technological process such as mining of uranium ore, concentration of the uranium in the form of yellowcake, conversion of yellowcake to UO2, further conversion to UF6 for enrichment, etc. Elemental impurities are removed at each stage, until nuclear grade purity is achieved, but certain elements may be reintroduced coming from the reagents or vessels used. Although such information may be used to match two or more known materials, the sampling uncertainties and batch-to-batch nonreproducibility will make absolute source attribution much more difficult. 63.5.3 Safeguards Analytical Laboratories Clean Laboratory for Safeguards The role of the clean laboratory (CL) for safeguards as well as the structure of IAEA’s environmental sampling for safeguards (ESS) program are described elsewhere . Fig. 63.30 High-volume water sampler and cartridge Nuclear Safeguards Verification Measurement Techniques 63 2991
  • (Deron et al. 1996; Donohue 1998, 2002; IAEA 2003). The clean laboratory (CL) consists of two zones; the first zone of approximately 100 m2 is maintained at a cleanliness level of ‘‘Class- 100,000’’ according to the US Federal Standard 209E or ISO Class 8 according to the ISO Standard 14644-1. The CL is equipped for the measurement of samples using instrumental methods such as thermal ionization mass spectrometry (TIMS), inductively coupled plasma mass spectrometry (ICP-MS), scanning electron microscopy (SEM) in combination with X-ray spectrometry (XRS) and high-resolution gamma spectrometry (HRGS). A higher class of cleanliness is maintained in the second zone of the CL, also of approximately 100 m2 area, where work surfaces are kept at US Class 100/ISO Class 5 for the chemical treatment of swipe samples and the production of certified clean sampling kits. The CL was constructed in 1995 and entered into full operation in 1996 at the same time that the agency began to implement environmental sampling for safeguards (ESS) as a routine tool under existing legal authority. A floor plan of the CL is shown in > Fig. 63.31. The CL was designed to support the following activities: 1. Preparation and certification of clean sampling kits and supplies for taking environmental samples. The number of sampling kits produced per year is approximately 1,000. 2. The receipt, book-in and recoding of up to 1,000 environmental samples per year. After initial cleaning of the outer sample bags and screening by high-resolution gamma . Fig. 63.31 Floor plan of the IAEA Clean Laboratory for Safeguards Chemistry 1 Room 11 Chemistry 2 Room 12 Room 24 Chemistry 3 Room 13 Chemistry 4 Sampling kit preparation Room 14 Room 15 SEM particle preparation Room 16 Electron microscopy SEM- EDX/WDXHRGS & XRF screening Room 17 Room 5 Room 2 Room 4 Room 6 Filament preparation Room 7 Ladies changing area Corridor Room 10 Room 8 Mens changing area Sample preparation screening and TXRF Room 9 Room22 Data acquisition Mass spectrometry TIMS Room 21 Ventilation control PC Room23 #28685 #28694 #28691 #28695 #28692 #28608 2992 63 Nuclear Safeguards Verification Measurement Techniques
  • spectrometry (HRGS) and X-ray fluorescence spectrometry, the samples are kept in archival storage until a detailed analytical request is received. It is IAEA policy to retain in the archive at least one subsample of every environmental sample received for an indefinite period. For this reason as well as others, it is sometimes necessary to split subsamples; this is done in a glove bag inside a Class-100 clean-air bench in order to avoid cross contamination. 3. Detailed bulk analysis of environmental samples according to an analytical request. Bulk analysis (see > Sect. 63.5.5) refers to the complete dissolution of the subsample (usually a cotton or cellulose swipe), chemical treatment and measurement of U and Pu by thermal ionization mass spectrometry (TIMS, see > Sect. or inductively coupled plasma mass spectrometry (ICP-MS, see > Sect. Element concentrations are measured using the isotope dilution mass spectrometry (IDMS) method. In addition to U and Pu, the element Am is sometimes measured using either ICP-MS or alpha spectrometry. 4. Particle analysis of environmental samples by secondary ion mass spectrometry (SIMS, see > Sect. or scanning electron microscopy in combination with X-ray spectrometry (SEM-XRS, see > Sect. The SEM-XRS instrument and its sample-preparation area are located in the clean laboratory, but the SIMS instrument is not. Sample prepara- tion for SIMS can be carried out in the CL, but a small dedicated clean-room at US Class- 100/ISO-Class 5 cleanliness level is colocated with the instrument to avoid potential cross contamination when transporting the prepared sample planchets to the instrument. The SEM-XRS method is also supported in the CL by an array of optical microscopes equipped with micromanipulation systems for picking up particles of interest prior to further treatment such as chemical analysis and isotopic measurement by TIMS or ICP-MS. In general, the CL is operated as a mixed-flow clean-room facility, meaning that the Class- 100/ISO Class 5 areas are the workbenches and fume cupboards in the various rooms, rather than the entire room. A laminar flow of clean air falls from high-performance particulate (HEPA) filters in the suspended ceiling of the room and is directed to the work area by plastic curtains. Certain clean laboratories in the network employ full laminar flow in the rooms at US Class 10/ISO Class 4 cleanliness conditions. The most sensitive measure of effectiveness is the blank levels of U and Pu that can be achieved, rather than the chosen cleanliness class. Blanks consist of the following types: 1. Reagents used in the chemical treatment of the samples for bulk analysis (‘‘process blank’’). These are typically ultra-pure mineral acids such as HNO3 or HCl as well as reducing agents such as HBr or HI and highly pure water (18 MO). 2. Blanks of the swipe sample matrix (‘‘swipe blank’’) after each processing step. The swipe blank contains the process blank as well as any U, Pu or interfering elements contained in the swipe material. The cotton wipers chosen for swipe sampling are generally quite low in U content – 1–5 ng per wiper – and the Pu content is not measurable by bulk analysis: it is believed to be below 1 fg (the most likely source would be Pu from nuclear weapons fallout). 3. Blanks of the air in a clean-room laboratory or other working area (‘‘room blank’’). These are collected in a petri dish of approximately 100 cm2 area that is left exposed on a surface for 1–2 weeks. Then, U and Pu are recovered with concentrated HNO3, spiked with tracers of 233 U and 242 Pu, and subjected to bulk analysis by ICP-MS. Typical values in a Class 100/ ISO Class 5 working area are in the range 0.1–0.2 fg/cm2 – hour for U and below the detection limit of the ICP-MS method for Pu (estimated to be 104 times lower). Nuclear Safeguards Verification Measurement Techniques 63 2993
  • The IAEA Clean Laboratory for Safeguards is certified annually to meet the specified cleanliness standards by an outside organization. The Network of Analytical Laboratories (NWAL) of the IAEA The IAEA collaborates with a number of laboratories in the member states that form a network dedicated to the analysis of environmental samples for safeguards. > Table 63.14 lists the members of this network and the analytical services they offer. 63.5.4 Sample Screening Methods Screening of incoming environmental samples is performed to obtain information that will guide the further detailed analysis and to assist in the shipment of samples to the NWAL. Samples that are known to be radioactive for the purposes of shipment from the field to the IAEA are delivered directly to the nuclear laboratory of SAL for screening and archival storage because the CL is not licensed to handle such materials. All samples known to be below the limits for radioactive shipment purposes are delivered to the CL where they are screened, . Table 63.14 Member laboratories of the IAEA NWAL for environmental sample analyses. Techniques in use by the NWAL include FT-TIMS = fission-track thermal ionization mass spectrometry, AMS = acceler- ator mass spectrometry, SIMS = secondary ion mass spectrometry, HRGS = high-resolution gamma spectrometry, TIMS = thermal ionization mass spectrometry, ICP-MS = inductively coupled plasma mass spectrometry, SEM = scanning electron microscopy Laboratory Particle analysis Bulk analysis U.K. Atomic Energy Authority – Aldermaston, U.K. FT-TIMS QinetiQ – Malvern, U.K. SIMS Australian Nuclear Science and Technology Organization, Lucas Heights, Australia AMS Air Force Technical Applications Center – Patrick Air Force Base, Florida, U.S.A FT-TIMS U.S. Department of Energy – Oak Ridge National Laboratory, Los Alamos National Laboratory, Pacific Northwest National Laboratory, Lawrence Livermore National Laboratory HRGS, TIMS and ICP-MS Commissariat a` l’Energie Atomique, Bruye`res-le Chatel, France FT-TIMS European Commission Joint Research Centre Institute for Transuranium Elements, Karlsruhe, Germany SIMS HRGS, TIMS and ICP-MS Khlopin Radium Institute, St. Petersburg, Russian Federation HRGS, TIMS Laboratory for Microparticle Analysis, Moscow, Russian Federation SIMS Japan Atomic Energy Research Institute, Tokai-mura, Japan SIMS, SEM, FT-TIMS HRGS, TIMS 2994 63 Nuclear Safeguards Verification Measurement Techniques
  • archived, analyzed, or transferred to the NWAL. The limits for declaring a sample as radioactive for the purposes of shipment are described in the International Air Transport Authority (IATA 2010) regulations, which are based on the IAEA Regulations for the Safe Transport of Radioactive Materials (IAEA 2009a). So-called ‘‘cold’’ or non-radioactive environmental samples can actually contain a measurable amount of alpha, beta, or gamma radioactivity as long as they fall below the limits given in the above regulations. This is why HRGS screening is performed on all environmental samples received, regardless of their status under the shipping regulations. Gamma Spectrometry Screening of all environmental samples is performed by HRGS either in the CL for ‘‘cold’’ samples or in the nuclear laboratory of SAL for ‘‘radioactive’’ samples (Parus et al. 2003; Carchon et al. 2007). Cotton swipes and cellulose (Type-J) swipes are placed in a Marinelli beaker to maximize collection efficiency. The samples are measured with a coaxial Ge detector system in the CL for a typical measurement time of 6 hours. Samples that are known to be more active are measured in SAL at greater distance from the detector or for shorter periods using gamma spectrometers with high-purity planar Ge detectors. The range of gamma-ray energies measured with these systems is from 20 keV to 2,620 keV, which allows the measurement of most fission and activation products expected to be found in environmental samples. The nuclides of interest, their half-lives and main gamma-ray energies are shown in > Table 63.15. The data from HRGS screening are evaluated with commercial software using energy and efficiency calibration data from the measurement of reference material samples. The result is a report for each sample of those nuclides that were detected above the detection limit or the minimum detectable amount for all other nuclides of interest. X-Ray Fluorescence Spectrometry In addition to the HRGS screening that is performed on all environmental samples received by the IAEA, X-ray fluorescence (XRF) screening is performed on all ‘‘cold’’ samples in the CL to detect significant amounts of U. This screening is useful for the CL and network laboratories in their further processing of the samples in case the amount of U present represents an increased risk of cross contamination. The equipment for this screening consists of a 100 W X-ray generator with Rh anode and end-window configuration, a robotic arm for manipulation of the samples (usually a single cotton swipe in a plastic bag held in a metal frame), and two X-ray detectors (a high-purity Ge detector and a Si(Li) detector). The Si(Li) detector is equipped with an energy filter (Bragg filter using single-crystal graphite strips) tuned to the L-alpha energy of U (13.6 KeV). This system, referred to as ‘‘Tripod I,’’ has a detection capability of approximately 35 ng/cm2 for U with a 4 hour measurement time and the data can be displayed as an elemental map. 63.5.5 Bulk Sample Analysis The purpose of bulk analysis is to detect U and Pu at the lowest possible levels and to measure both their concentration and isotopic composition. This process involves destruction of the Nuclear Safeguards Verification Measurement Techniques 63 2995
  • sample matrix (usually cotton or cellulose swipes) by high-temperature ashing followed by dissolution in strong mineral acid. The resulting solution is split in two equal parts and one part is kept as an archive in case problems should occur with the other fraction. The fraction of the original solution to be measured is split again into portions that will result in U and Pu concentration determination by use of the IDMS method and fractions that will be processed for U and Pu isotopic composition measurements. Tracers The primary tracers used for bulk analysis are shown in > Table 63.16. They are produced by the Institute for Reference Materials and Measurements in Geel, Belgium. . Table 63.15 List of radionuclides expected in environmental samples Nuclide Half-life g-line (keV) Nuclide Half-life g-line (keV) 51 Cr 27.7 days 320.1 124 Sb 60.2 days 602.7(1,691) 54 Mn 312.1 days 834.8 125 I 59.41 days 35.5 57 Co 271.8 days 122.1 125 Sb 2.758 a 427.9 58 Co 70.8 days 810.8 125m Te 57.4 days 35.5(109.3) 59 Fe 44.5 days 1,099.3 127m Te 109 days 88.3 60 Co 5.27 a 1,332.5 129m Te 33.6 days 459.6 65 Zn 244.3 days 1,115.5 131 I 8.02 days 364.5 75 Se 119.8 days 264.7 134 Cs 2.062 a 604.7 91m Nb 60.9 days 1,204.7 137 Cs 30.017 a 661.6 92m Nb 10.15 days 934.4 140 Ba 12.75 days 537.3 95m Nb 86.6 h 235.7 140 La 1.678 days 1,596.2 95 Nb 34.97 days 765.8 141 Ce 32.5 days 145.4 95 Zr 64.02 days 756.7 144 Ce, 144 Pr 284.89 days 696.5 99 Mo 65.94 h 739.5 152 Eu 13.54 a 121.78 99m Tc 6.01 h 140.5 154 Eu 8.59 a 1,274.4 102m Rh 2.9 a 475.1 155 Eu 4.76 a 86.5(105.3) 103 Ru 39.26 days 497.1 192 Ir 73.83 days 205.8(484.6) 106 Ru,106 Rh 373.6 days 621.9(511.9) 203 Hg 46.6 days 279.2 108m Ag 418 a 722.9(433.9) 231 Th 25.52 h 25.64 109 Cd 462.6 days 88.03 234m Pa 1.17 min 1,001.03 110m Ag 249.8 days 657.8 234Th 24.1 days 63.29 121m Te 154 days 212.2 234 U 2.455E+5 a 53.2 121 Te 16.78 days 573.1 235 U 7.038E+8 a 185.71 122 Sb 2.70 days 564.2 237 Np 2.14E+6 a 86.48 123m Te 119.7 days 159.0 239 Pu 24,110 a 129.30 124 I 4.18 days 602.7 241 Am 432.2 a 59.54 2996 63 Nuclear Safeguards Verification Measurement Techniques
  • Sample Preparations and Separations The procedure for chemical separation of these various aliquots is based on anion exchange and solvent extraction chromatography using AG MP-2 anion exchange resin in HNO3 medium and UTEVA (Eichrom Corp.) in HNO3 and HCl media. The basic steps in a bulk- analysis preparation scheme are listed below (Shinonaga 2008): 1. The swipe is placed in a covered quartz tube and ashed in an oven at 600 C for 8–10 h. A blank swipe and an empty quartz tube are treated at the same time to provide the swipe and process blank values. 2. The ash is dissolved in 16 M HNO3 and dried and re-dissolved several times. It is then treated with H2O2 and aqua regia to complete the oxidation and sample destruction process and a ‘‘mother’’ solution is re-dissolved in 8 M HNO3. 3. The mother solution is split into 3 parts; 20% will be used for U IDMS and 238 Pu analysis by TIMS, and alpha spectrometry, respectively. Forty percent will be used for U isotopic analysis, and Pu elemental and isotopic analysis by IDMS (there is no un-spiked Pu fraction and Pu isotopic composition will be estimated by isotopic stripping of the 242 Pu spike). The rest (40%) of the solution is kept as an archive. 4. High-purity isotopic spikes of 233 U (IRMM-057) and 242 Pu (IRMM-044) are added to the 20% and 40% fractions, respectively, and the isotopes are equilibrated by repeated drying and re-dissolution in 16 M HNO3 and treatment with H2O2. 5. Resin columns containing AG MP-2 and UTEVA are prepared in nitrate form in 8 M HNO3. The sample fraction is loaded on the MP-2 column in 8 M HNO3 and rinsed further with 8 M HNO3 with the eluate going into the UTEVA column. The MP-2 column is further processed for Pu analysis and the UTEVA column is further processed for U analysis. 6. The Pu is eluted from the MP-2 column with a mixture of 9 M HCl and 1 M HI, then the Iodine is removed by repeated fuming with 10 M HCl. This Pu fraction is further cleaned using an MP-2 column in the chloride form and elution with HCl-H2O2 and HBr. This fraction is finally taken up in 16 M HNO3. . Table 63.16 Tracers used for isotope dilution mass spectrometry Tracer Isotope ratio Isotope amount ratio IRMM-057 (233 U Spike) 234 U/233 U 0.000352 (14) 235 U/233 U 0.000004124 (29) 236 U/233 U 0.0000000434 (14) 238 U/233 U 0.00001043 (21) IRMM-044 (242 Pu Spike) 238 Pu/242 Pu 0.000009 (6) 239 Pu/242 Pu 0.000827 (4) 240 Pu/242 Pu 0.000108 (4) 241 Pu/242 Pu 0.000009 (4) 244 Pu/242 Pu 0.000015 (4) Note: IRMM-057 certificate date February 2003; IRMM-044 validity date 30.06.1989. Nuclear Safeguards Verification Measurement Techniques 63 2997
  • 7. The U on the UTEVA column is eluted with 0.05 M HCl, dried and re-dissolved in 10 M HCl. This fraction is then further purified with an MP-2 column in the chloride form at a molarity of 10 M HCl. The U is finally eluted with 0.05 M HCl, dried, and converted to the nitrate form with 16 M HNO3. Thermal Ionization Mass Spectrometry The measurement of U IDMS and U isotopic fractions from the chemical separation shown above are performed on a Thermo-Fisher Triton mass spectrometer equipped with a single-ion counting detection system (Thermo Fischer Scientific: Triton 2005). Therefore, a peak- jumping measurement scheme is used along with manual sample filament heating procedure to optimize the signal intensity prior to data acquisition. Depending on signal intensity, one or two blocks of data are taken – consisting of 20 scans each. A single block takes approximately 20 minutes. The isotopes of U from 233 to 238 are symmetrically scanned to cancel out drift in signal intensity during the scan. Typical count rates for a 1–2 ng filament loading are 20,000– 50,000 ion counts per second for the major isotope present. Inductively Coupled Plasma Mass Spectrometry In the past decade, ICP-MS instruments have become sufficiently sensitive and stable to offer an attractive alternative to the traditional TIMS measurements for bulk analysis of U and, especially, Pu. The ICP-MS method offers the advantage that samples are presented to the instrument in solution form which eliminated the filament-loading step and fila- ment-blank considerations that are seen with TIMS. An additional advantage is the more constant sensitivity of the ICP plasma ion source and the absence of interference effects that suppress ionization of the element of interest. On the other hand, the ICP source produces a much larger number of molecular ion interferences that affect the accuracy of a Pu measurement, especially at near the detection limit (low parts per quadrillion, ppq). The best approach for these interferences is to remove all possible interfering elements during the chemical preparation of the samples. In any case, the potential interferences must be measured at some point during the analysis of a sample to ensure that they are below the threshold for significant contribution to the masses of interest (238 Pu through 244 Pu). In those cases where the interfering elements are above this threshold, a correction can be made for them, based on the ‘‘formation co-efficient’’ that relates the ion signal of the interfering element to the production of molecular species at the mass of a plutonium isotope. This correction procedure will naturally increase the uncertainty of the final Pu isotopic data, in some cases significantly. The ICP-MS instrument used in the IAEA Clean Laboratory is the Thermo-Finnigan Element 2. It uses a magnetic sector double-focusing mass spectrometer with single-ion counting detector system and peak-jumping data collection. The quoted sensitivity with a concentric nebulizer is 109 cps per ppm, which gives a theoretical sensitivity of 1 ppq for elements that do not suffer molecular interference effects. Special de-solvating nebulizers such as the Apex Q (Elemental Scientific Corp.) can improve the sensitivity further. 2998 63 Nuclear Safeguards Verification Measurement Techniques
  • 63.5.6 Particle Analysis Early in the environmental sampling program of the IAEA it was recognized that the analysis of individual micrometer-sized particles was a source of unique information about nuclear materials and activities. > Table 63.17 shows the calculated composition of 1 mm diameter particles coming from various nuclear processes. Thus, it can be seen that a pure particle of natural U oxide (‘‘NU’’) contains about 1010 U atoms in total and that when this particle is irradiated in a reactor, approximately 5 million atoms of 239 Pu would be created. Furthermore, a particle of high-enriched uranium (‘‘HEU’’) would produce only small numbers of 231 Pa and 232 Th daughter atoms in 10 years of decay. To be able to ‘‘age-date’’ such a particle would involve measurement of these small components, something which is currently not possible with the most sensitive techniques. The above considerations show that particles contain small amounts of material, but that the major isotopes would be easily measurable by sophisticated methods such as TIMS or ICP- MS. One of the most challenging aspects of particle analysis for safeguards is the methods used to locate the particles of interest in a ‘‘sea’’ of uninteresting environmental materials such as minerals and organic particles like pollen and fibers. The detection and location of particles containing fissile nuclides such as 235 U and 239 Pu can be accomplished by the fission-track technique (Fleischer et al. 1975). This method is based on creating a close contact between the particles and a polycarbonate plastic surface (Lexan) which is then irradiated in a nuclear reactor at high thermal neutron fluence (approximately 1 Â 1015 n/cm2 ) for several minutes. During this irradiation, atoms of the fissile isotopes will undergo fission and the energetic fission fragments will leave damage tracks in the plastic material. After separation of the particle layer from the Lexan, it is etched in NaOH solution to make the damage tracks visible under an optical microscope. The selectivity of this method is very high because no other particles will leave such tracks and the number of tracks compared to the size of the particle gives a clue to the presence of enriched U or Pu because natural U with a 235 U content of 0.7% only leaves a small number of tracks in comparison. The fission-track method has been . Table 63.17 Composition of typical 1 mm diameter particles found in nuclear facilities Nuclide NU Irradiated NU (700 MWD/T) LEU (4%) Irradiated LEU (30 GWD/T) HEU (93%) Decay of HEU (10 years) U-238 9,900 M 9,900 M 9,600 M 9,310 M 600 M U-236 <1 1.2 M <1 45 M <1 U-235 72 M 64 M 400 M 155 M 9,300 M U-234 550 K 530 K 3,600 K 2,500 K 100 M Pu-239 5.5 M 57 M Pa-231 92 Th-230 2,820 Note: GWD/T = gigawatt-days per ton; MWD/T = megawatt days per ton; LEU 4% = low-enriched uranium with 4% abundance of 235 U. Nuclear Safeguards Verification Measurement Techniques 63 2999
  • combined with thermal ionization mass spectrometry (FT-TIMS) to provide a powerful method to locate particles containing U or Pu and then to measure their isotopic composition with high sensitivity and accuracy. A calculation of the number of U atoms needed to produce one fission track in Lexan under nominal irradiation conditions is about 107 or 4 fg, keeping in mind that a 1 mm diameter particle would contain about 5 pg of U. Furthermore, if such a particle can be manipulated and mounted on a filament for measurement by TIMS, the efficiency of ion collection would be in the range of 0.1–1%, i.e., one ion detected for every 100–1,000 atoms loaded. A particle containing 1010 atoms would therefore provide a minimum of 107 ions at the mass spectrometer output. Thus, minor isotopes such as 234 U or 236 U could be measured with counting-statistics limited accuracy of several percent relative and the 235 U could be measured with less than 1% relative uncertainty. These calculations are borne out by actual measurement data from the network laboratories that perform FT-TIMS measurements (Stetzer et al. 2004; Park et al. 2006; Usuda et al. 2006). Sample Preparation Most particle-analysis methods (SEM, SIMS and FT-TIMS) require the removal of particles from the swipe substrate and deposition on a flat surface. In the case of FT-TIMS, the particles are removed by ultrasonic treatment in a suitable suspension medium such as ethanol or siloxane. The suspension is then mixed with collodion and dried as a thin layer on the Lexan plastic for irradiation. After irradiation, this collodion layer can be peeled off to allow chemical etching of the fission tracks in the plastic. The collodion layer can also be replaced on the Lexan after etching with a slight offset so that the particles and tracks are visible under a light microscope at magnification 250–500. Replacement of the collodion layer is not necessary if a comparator microscope is available that allows viewing two objects (i.e., the collodion layer and the Lexan with tracks) simultaneously. For both SEM and SIMS, the only requirement is that particles should be deposited on a polished flat and conducting surface; pyrolytic graphite planchets with a diameter of 2.5 cm are frequently used. The particles are deposited from suspension or can be deposited using a vacuum extractor system in which the planchet acts as an impactor in a flow of gas with the particles entrained. The one disadvantage of the vacuum extractor is that many particles bounce off the planchet and often a type of sticking agent (polyisobutylene in nonane) is applied and later removed by heating at 250 C. In all procedures where ultrasonic removal and suspension in an organic liquid are used, care must be taken to avoid cutting the swipe material because it results in many organic fibers on the planchet that can accumulate charge and deflect the electron or ion beam of the final measurement instrument. Care is also needed in all methods of particle deposition to avoid too thick a deposit where particles are on top of each other or too close together to allow unambiguous measurements. An optimal spacing between particles is 10–20 mm. Thermal Ionization Mass Spectrometry The basic processes in TIMS, which apply equally to the measurement of particles, have been published elsewhere (Duckworth et al. 1986). In TIMS, the particle is deposited on a pure metal filament; typical filament materials are W, Ta, and Re which have been purified by 3000 63 Nuclear Safeguards Verification Measurement Techniques
  • zone-refining to greater than 99.99% purity in order to remove possible U background. In the vacuum system of the mass spectrometer, the filament is heated by passing a current of several amperes through it until it reaches a temperature of 1,400–1,800 C. At these temperatures, the elements Pu and U tend to evaporate from the particle in contact with the filament surface and a certain fraction of these atoms will leave as positive ions. The ratio between neutral and ionic species is given by the Saha-Langmuir equation (Duckworth et al. 1986): Nþ ¼ N0 exp e o À ’ð Þ½ Š kT= ; ð63:12Þ where N+ is the number of positive ions produced, N0 is the number of neutral atoms produced, e’ is the ionization potential of the element, eo is the work function of the surface, k is the Boltzmann constant, and T is the absolute temperature of the filament surface. The main challenge of single-filament TIMS is that there is a competition between evaporation of the analyte atoms and their ionization; a higher temperature favors ionization but accelerates the loss of material from the filament. In situations where the amount of sample element is severely limited, this competition is a serious challenge. Use of separate evaporation and ionization filaments allows the analyst to better control both processes but the geometrical coupling between the two filaments separated by about 1 cm means that the overall ionization efficiency is reduced. Under the best of circumstances, the ion production efficiency of a thermal ion source is in the range 0.1 to 1% (ions collected per atom loaded). The sensitivity of TIMS for low levels of actinide elements has been demonstrated in many laboratories (Elliott et al. 2006; Buerger et al. 2009; Kraiem et al. 2010). Modern TIMS instruments combine a single magnetic sector with multi-ion detection systems, although the use of multiple ion counters is not yet mature and many laboratories continue to use a single-ion counting detector and peak jumping. Use of multiple Faraday collectors is a mature technology with excellent methods to inter-calibrate the collectors, but the ion signals and consequent sample amounts must be high for this method and it is not commonly used with ultra-low amounts (pg-ng) as found in environmental samples. Use of a reverse polarity quadrupole filter before the ion-counting detector is also used to reduce the tailing of ions from major isotopes, thus improving the abundance sensitivity from a value of 106 to around 108 . Secondary Ion Mass Spectrometry A description of the basic operation of a SIMS instrument is also found elsewhere (Duckworth et al. 1986). The basic principle is the bombardment of the sample surface with energetic ions (typically O2 + at 10–15 keV of energy). The interaction of these ions with the sample atoms results in sputtering of material from the surface. A small fraction of the sputtered material is in the form of positive secondary ions, which can be accelerated into an ion lens and analyzed in a double-focusing mass spectrometer. Raster scanning of the primary ion beam and synchro- nous display of the detected ions will provide an ion image showing local concentrations of that isotope. The ion microscope produces a similar image through stigmatic imaging of the secondary ions and their detection by a position-sensitive ion detector system. In the case of particle analysis, the particles removed from a swipe sample are deposited on a flat conductive surface and illuminated by the primary ion beam. The ion image of this field (typically 150  150 mm2 ) will show the presence of particles in the 0.5–10 mm diameter range that contain, e.g., either 235 U or 238 U. Mathematically comparing these two images will yield the Nuclear Safeguards Verification Measurement Techniques 63 3001
  • ‘‘enrichment’’ of the particles. Measurement of an area on the planchet of 1 cm2 using fields of 150 Â 150 mm in size takes approximately 6–8 hours. The sensitivity of SIMS for U particles is limited by the secondary-ion production and extraction efficiency, which is approximately 0.1% under typical measurement conditions (Tamborini et al. 2004). Therefore, a U particle of 1 mm diameter containing 1010 total U atoms would be expected to yield approximately 107 ions of the major isotope (238 U) and approximately 105 ions of the minor isotope (235 U) under ideal conditions. Practical consid- erations such as pre-sputtering and duty cycle would reduce these values by as much as a factor of 10. The result is that the enrichment or ratio of 235 U/238 U can only be measured with 1–5% uncertainty for a pure U particle that is 1 mm in size. Realistic particles encountered in environmental samples are frequently smaller or less pure, and certain interference effects can also reduce the quality of such data. An obvious improvement is offered by a large- geometry SIMS instrument like the Cameca IMS-1280 that has significantly higher sensitivity and resolution to reduce the effects of molecular ion interferences (Ranebo et al. 2009). Scanning Electron Microscopy with X-Ray Spectrometry Another method to locate particles containing elements of interest is the scanning electron microscope combined with energy-dispersive X-ray spectrometry (SEM-XRS). The SEM instrument used in the clean laboratory consists of a JEOL 6490 SEM with Oxford INCA X-ray spectrometer. The software supplied with this instrument allows an automated search capability in which a specified area of the sample planchet is sequentially scanned in fields of approximately 100 Â 100 mm. The backscattered electron detector is used to detect particles containing high-Z (atomic number) elements. When such a particle is detected, the energy- dispersive X-ray spectrum of the particle is accumulated for a specified time such as 10 s, the resulting spectrum is stored and the search continues. At the end of an analysis, which may take many hours, the data can be sorted according to the element of interest and particles revisited for further analysis, either by energy-dispersive XRS or wavelength-dispersive XRS, which is capable of measuring elemental composition with reasonable accuracy down to element ratios of 1 part per thousand (0.1%). Particles identified in this way can be manipulated inside the vacuum system of the SEM or offline with an optical microscope and further analysis can be carried out by bulk, SIMS, or other mass spectrometric methods to obtain a complete elemen- tal and isotopic analysis of a single interesting particle. References Abhold ME, Baker MC, Bourret S, Polk P, Vo DT, Ishikawa M, Sao Y, Uxchikoshi S, Yokota Y (2001) In: 42nd INMM annual meeting, Indian Wells Agboraw E, Johnson S, Creusot C, Poirier S, Saukkonen H, Chesnay B, Sequeira V (2006) In: IAEA symposium on international safeguards, IAEA-CN-148/200, Vienna, Austria Aggarwal SK, Alamelu D (2005) Novel approach for determining 238 Pu by TIMS using IEC method. Int J Mass Spectrom 241:83 Aggarwal SK, Chitambar SA, Kavimandan VD, Almaula AI, Shah PM, Parab AR, Sant VL, Jain HC, Ramaniah MV (1980) Precision and accuracy in the determination of 238 Pu/(239 Pu + 240 Pu) alpha activity ratio by alpha spectrometry. Radiochim Acta 27(1):1–5 Aggarwal SK, Choursiya G, Duggal RK, Singh CP, Rawat AS, Jain HC (1985) A comparative study of different methods of preparation of sources for alpha spectrometry of plutonium. Nucl Instrum Methods Phys Res A 238(2/3):463–468 3002 63 Nuclear Safeguards Verification Measurement Techniques
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