Handbook of nuclear chemistry

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Handbook of nuclear chemistry

  1. 1. 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 63.2.13.1 Nondestructive Assay ......................................................... 2903 63.2.13.2 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 63.3.3.1 Gamma Ray Detectors ........................................................ 2909 63.3.3.2 Low-Resolution Gamma Spectroscopy (LRGS) .............................. 2911 63.3.3.3 High-Resolution Gamma Spectroscopy (HRGS) ............................ 2913 63.3.4 Neutron Counting Techniques ................................................ 2915 63.3.4.1 Neutron Detectors ............................................................ 2915 63.3.4.2 Gross Neutron Counting ..................................................... 2917 63.3.4.3 Neutron Coincidence Counting .............................................. 2918 63.3.4.4 Multiplicity Coincidence Counting ........................................... 2921 63.3.4.5 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
  2. 2. 63.3.5 Spent Fuel Measurement ...................................................... 2923 63.3.5.1 Gamma Methods .............................................................. 2924 63.3.5.2 Neutron Methods ............................................................. 2925 63.3.5.3 Combined Gamma/Neutron Methods ....................................... 2926 63.3.5.4 Optical Methods .............................................................. 2927 63.3.6 Unattended NDA Systems .................................................... 2928 63.3.6.1 Unattended Gamma-Based NDA Systems .................................... 2930 63.3.6.2 Unattended Neutron-Based NDA Systems ................................... 2931 63.3.6.3 Other Unattended NDA Systems ............................................. 2934 63.3.7 Other NDA Techniques ....................................................... 2935 63.3.7.1 Physical Property Measurement .............................................. 2935 63.3.7.2 Calorimetric Techniques ...................................................... 2937 63.3.7.3 X-Ray Measurements ......................................................... 2938 63.3.7.4 Analytical NDA Techniques at Laboratories .................................. 2940 63.3.8 New and Novel Technologies ................................................. 2941 63.3.8.1 New Technologies ............................................................. 2941 63.3.8.2 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 63.4.2.1 Spent Fuel Solutions .......................................................... 2951 63.4.2.2 Uranium Hexafluoride in Pressurized Cylinders ............................. 2952 63.4.2.3 Plutonium Oxide Powders .................................................... 2954 63.4.2.4 Uranium Dirty Scrap Materials ............................................... 2956 63.4.3 Safeguards Analytical Laboratories ........................................... 2957 63.4.3.1 Off-Site Laboratories .......................................................... 2957 63.4.3.2 On-Site Laboratories .......................................................... 2958 63.4.4 Isotopic Analysis .............................................................. 2960 63.4.4.1 Isotopic Analysis by Mass Spectrometry ..................................... 2960 63.4.4.2 238 Pu Abundance by Alpha Spectrometry ..................................... 2967 63.4.4.3 Gamma Spectrometry of Nuclear Material Samples ......................... 2970 63.4.5 Elemental Assay ............................................................... 2971 63.4.5.1 Ignition Gravimetry of U, Pu, Th ............................................ 2971 63.4.5.2 Uranium Titration ............................................................ 2972 63.4.5.3 Plutonium Titration .......................................................... 2973 63.4.5.4 Controlled Potential Coulometry of Plutonium ............................. 2975 63.4.5.5 Isotope Dilution Assays ....................................................... 2978 63.4.5.6 Spectrophotometric Determination of Hexavalent Plutonium .............. 2982 63.4.5.7 X-ray Absorption and Fluorescence Spectrometry ........................... 2983 63.4.5.8 Assay of Alternative Nuclear Materials ....................................... 2984 2894 63 Nuclear Safeguards Verification Measurement Techniques
  3. 3. 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 63.5.2.1 Cotton Swipe and Other Swipe Materials .................................... 2988 63.5.2.2 Air Filters ...................................................................... 2989 63.5.2.3 Water, Soil, Vegetation, and Biota Samples .................................. 2990 63.5.2.4 Process and Structural Materials ............................................. 2991 63.5.3 Safeguards Analytical Laboratories ........................................... 2991 63.5.3.1 Clean Laboratory for Safeguards ............................................. 2991 63.5.3.2 The Network of Analytical Laboratories (NWAL) of the IAEA ............. 2994 63.5.4 Sample Screening Methods ................................................... 2994 63.5.4.1 Gamma Spectrometry ......................................................... 2995 63.5.4.2 X-Ray Fluorescence Spectrometry ............................................ 2995 63.5.5 Bulk Sample Analysis ......................................................... 2995 63.5.5.1 Tracers ......................................................................... 2996 63.5.5.2 Sample Preparations and Separations ........................................ 2997 63.5.5.3 Thermal Ionization Mass Spectrometry ...................................... 2998 63.5.5.4 Inductively Coupled Plasma Mass Spectrometry ............................ 2998 63.5.6 Particle Analysis ............................................................... 2999 63.5.6.1 Sample Preparation ........................................................... 3000 63.5.6.2 Thermal Ionization Mass Spectrometry ...................................... 3000 63.5.6.3 Secondary Ion Mass Spectrometry ........................................... 3001 63.5.6.4 Scanning Electron Microscopy with X-Ray Spectrometry ................... 3002 Nuclear Safeguards Verification Measurement Techniques 63 2895
  4. 4. 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
  5. 5. 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
  6. 6. 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
  7. 7. 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
  8. 8. 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
  9. 9. 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
  10. 10. 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
  11. 11. 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. 63.2.13.1 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
  12. 12. 63.2.13.2 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
  13. 13. 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
  14. 14. 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
  15. 15. 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
  16. 16. 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. 63.3.4.3). 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
  17. 17. 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. 63.3.3.1 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
  18. 18. 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
  19. 19. 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. 63.3.3.2 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
  20. 20. 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
  21. 21. 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. 63.3.3.3 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
  22. 22. 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
  23. 23. 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. 63.3.4.1 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
  24. 24. 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
  25. 25. 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. 63.3.4.2 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
  26. 26. 63.3.4.3 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
  27. 27. (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
  28. 28. 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
  29. 29. 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. 63.3.4.4 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
  30. 30. 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). 63.3.4.5 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

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