Final Report submission

Nuclear, Plasma, and Radiological Engineering | University of Illinois at Urbana-Champaign
216 Talbot Laboratory, MC-234 | 104 South Wright Street | Urbana, IL 61801
217/333-2295 | fax: 217/333-2906 | email: nuclear@illinois.edu
Authors:
Kevin D’Souza Joel Exner
Arsalan Muneeruddin Peter Ota
Aditya Patel
Instructors:
Prof. J. F. Stubbins
Dr. C. Sollima
Advisors:
Dr. D. W. Miller
Christopher Gans
Reduction of Cask Dose
(RoCD)
NPRE 458 / 548: NUCLEAR DESIGN
2015

NPRE – 458: Nuclear Design Page 2 of 77
Release date: 4/26/2015
Author: K. D’Souza, J. Exner, A. Muneeruddin, P. Ota, A. Patel
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Abstract
Nuclear utilities are tasked with transitioning their spent fuel from wet to dry storage. With no
federal procedural standard in place for this transition, dose to workers during dry cask storage
campaigns varies significantly across the nation. RoCD investigates universal means to reduce
dose to workers during dry cask storage campaigns. After assessing this week-long process,
procedures that lead to the most dose are identified. Techniques are proposed in order to reduce
dose during these procedures. The potential impact of rearranging fuel assemblies based on
source strength is modeled. Other techniques that are assessed include increasing worker
distance, decreasing worker exposure time, and improving shielding. The cost of these
improvements are weighed against the benefit of dose reduction in accordance with the ALARA
principle.

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Table of Contents
Abstract........................................................................................................................................... 2
List of Symbols............................................................................................................................... 6
List of Figures................................................................................................................................. 8
List of Tables .................................................................................................................................. 9
1. Introduction............................................................................................................................ 10
1.1 Scope.............................................................................................................................. 10
1.2 Objectives....................................................................................................................... 10
1.3 Structure ......................................................................................................................... 10
2. RoCD Conceptual Design................................................................................................. 10
2.1 Background .................................................................................................................... 11
2.1.1 Dry Storage Origins ..................................................................... 11
2.1.2 Current Dry Storage Process........................................................ 11
2.2 General Dose Reduction Techniques ............................................................................. 13
2.2.1 Time....................................................................................... 14
2.2.2 Distance.................................................................................. 14
2.2.3 Shielding ................................................................................ 14
2.3 State-of-the-Art .............................................................................................................. 15
2.3.1 Robotic Technology............................................................... 15
2.3.2 Never Wet .............................................................................. 18
2.3.3 Spent Fuel Pool Demineralizer.............................................. 18
2.3.4 Fuel Assembly Cleaning........................................................ 18
2.3.5 Shielding Materials ................................................................ 19
2.4 Social Impact.................................................................................................................. 20
2.5 Environmental Impact.................................................................................................... 20
2.6 Regulations..................................................................................................................... 20
2.7 Ethical Impact ................................................................................................................ 21
3. RoCD Feasibility.................................................................................................................... 22
3.1 Decision Making Process............................................................................................... 22
3.2 Dry Cask Specifications................................................................................................. 23
3.3 Economic Analysis......................................................................................................... 23

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3.3.1 Dose Model............................................................................ 23
3.3.2 Shielding Cost........................................................................ 24
3.3.3 Remote Technologies............................................................. 25
3.4 Robotics Feasibility........................................................................................................ 25
3.5 Mock-Up Assessment .................................................................................................... 26
3.6 Crowd Control................................................................................................................ 26
3.7 Fuel Assembly Positioning ............................................................................................ 27
3.8 Work Breakdown Structure............................................................................................ 27
3.9 Product Breakdown Structure ........................................................................................ 28
4. RoCD Preliminary Design ..................................................................................................... 29
4.1 Source Calculation ......................................................................................................... 30
4.2 Shielding Selection......................................................................................................... 31
4.2.1 Radiation Energies................................................................. 31
4.2.2 Shielding Material Selection.................................................. 31
4.3 Distance Model .............................................................................................................. 32
4.4 Time Calculation............................................................................................................ 33
4.5 Dose Calculation ............................................................................................................ 35
4.6 Economic Calculation .................................................................................................... 37
4.6.1 Shielding Economics ............................................................. 38
4.6.2 Types of Shields..................... Error! Bookmark not defined.
5. RoCD Technical Analysis...................................................................................................... 41
5.1 MPC Neutron Flux Model.............................................................................................. 42
5.2 MPC Gamma Flux Simulation....................................................................................... 46
5.2.1 Gamma Model ....................................................................... 46
5.2.2 Gamma Ray Source ............................................................... 47
5.2.3 Results.................................................................................... 48
5.3 Economic Model ............................................................................................................ 48
5.3.1 Methodology.......................................................................... 48
5.3.2 Results.................................................................................... 50
5.4 Recommendations.......................................................................................................... 53
5.4.1 Shielding ................................................................................ 53

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5.4.2 Time....................................................................................... 58
5.4.3 Distance.................................................................................. 59
5.4.4 Crowd Control ....................................................................... 60
5.4.5 Fuel Rod Positioning.............................................................. 60
6. Conclusion.............................................................................................................................. 60
6.1 Achievements................................................................................................................. 60
6.2 Limitations ..................................................................................................................... 61
6.3 Future Activities............................................................................................................. 61
7. References.............................................................................................................................. 61
Appendix A: Ethics Exercise........................................................................................................ 64
Appendix B: Serpent MPC Model Code ...................................................................................... 65
Appendix C: Serpent Burnup Calculation Codes ......................................................................... 73

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List of Symbols
ALARA As Low As (is) Reasonably Achievable
AWS Automated Welding System
BWR Boiling Water Reactor
CEA French Atomic Energy Commission
COGEMA Compagnie Generale des Matieres Nucleaires (French Nuclear Company)
DSC Dry Storage Cask
EDF Électricité de France (French electric utility company)
EWS Engineering Workstations
GTAW Gas Tungsten Arc Welding
HDPE High Density Polyethylene
HI-STORM Holtec International Storage Module
HRA High Radiation Area
IAEA International Atomic Energy Agency
INPO Institute of Nuclear Power Operations
ISFSI Independent Spent Fuel Storage Installation
ISOE Information System on Occupational Exposure
MCNP Monte Carlo N-Particle Transport Code
MPC Multi-Purpose Cask
MRS Monitored Retrievable Storage
NATC North American Technical Center for ALARA
NPP Nuclear Power Plant
NRC Nuclear Regulatory Commission
PWR Pressurized Water Reactor

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QC Quad Cities Generating Station
R&D Research and Development
RP Radiation Protection
SFP Spent Fuel Pool
SNF Spent Nuclear Fuel
UIUC University of Illinois at Urbana-Champaign

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List of Figures
Figure 1. Fuel rod loaded in the spent fuel pool. .......................................................................... 12
Figure 2. Workers preparing the cask for welding. ...................................................................... 12
Figure 3. Manual welding of lid/ports. ......................................................................................... 12
Figure 4. Dry casks on Independent Spent Fuel Storage Installation (ISFSI) at D.C. Cook
Generating Station. ....................................................................................................................... 13
Figure 5. MAGRITTE - French Virtual Reality Design for Telerobotics [3] .............................. 16
Figure 6. Japanese Two-Arm Bilateral System [3]....................................................................... 17
Figure 7. Operators of the Two-Arm Bilateral System [3]........................................................... 17
Figure 8. Ishikawa Diagram.......................................................................................................... 22
Figure 9. Flow Chart of Product Breakdown Structure. ............................................................... 29
Figure 10. Intensity of 1 MeV gamma rays versus the thickness of lead and tungsten................ 32
Figure 11. Dry Cask Campaign Dose (2012)................................................................................ 35
Figure 12. Economic Analysis of Gamma Shielding Materials. .................................................. 40
Figure 13. Economic Analysis of Neutron Shielding Materials................................................... 41
Figure 14. MPC-32 Assembly Loading Patterns. ......................................................................... 45
Figure 15. Plot of gamma ray intensity over distance inside a source material, steel, concrete and
air. ................................................................................................................................................. 47
Figure 16. Optimization of Radiation Protection Cost-Benefit Model......................................... 49
Figure 17. Cost Optimization Model for Various Shielding Materials ........................................ 52
Figure 18. Top-down view of MPC lid with one half of RoCD shielding in place. Note: not to
scale............................................................................................................................................... 54
Figure 19. Top-down view of RoCD shielding ............................................................................ 55
Figure 20. Side profile of RoCD shielding................................................................................... 55
Figure 21. Gamma Attenuation for T-Flex W [33]....................................................................... 57
Figure 22. Neutron Attenuation for Borotron [34]. ...................................................................... 57

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List of Tables
Table 1. NRC 10 CFR Part 61 Dose Limits.................................................................................. 21
Table 2. Cost data from 1995 RP Calculation, Cook Nuclear Power Plant.................................. 23
Table 3. Cost data for different shielding type used in PWR/BWR dry cask............................... 25
Table 4. Work Breakdown Structure. ........................................................................................... 28
Table 5. Dose Rate at Various Locations (HI-TRAC).................................................................. 31
Table 6. Dose estimate from fuel loading campaigns................................................................... 34
Table 7. Federal Regulation 10 CFR 20 limits. ............................................................................ 35
Table 8.Assumptions and equations utilized by the economic model.......................................... 38
Table 9. Percent Reduction of Gamma Dose for Various Shielding Materials and Thicknesses. 39
Table 10. Cost Data for Gamma Shielding Materials................................................................... 39
Table 11. Thickness Required to Achieve 90% Reduction in Gamma Dose for Shielding
Materials ....................................................................................................................................... 39
Table 12. Summary of Neutron Shielding Materials.................................................................... 41
Table 13. Westinghouse Standard 17x17 Fuel Assembly Specifications..................................... 43
Table 14. Serpent MPC Neutron Model Results .......................................................................... 44
Table 15. Properties of fuel assemblies used in simulation.......................................................... 48
Table 16. Summary of Assumptions for the Cost-Benefit Analysis............................................. 50
Table 17. Optimal Thickness for Various Materials..................................................................... 51
Table 18. Overall dose with 10% time reduction. ........................................................................ 59

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1. Introduction
This project explores the possibility of reducing the overall acquired dose during dry storage
campaigns. Currently NPPs report an average of up to 300-500 mrem per dry cask. However,
some NPP’s have reported a dose of 150 mrem for a single dry cask. The industry goal for the
overall dose received is less than 100 mrem per dry cask.
1.1 Scope
Dry cask storage is an essential step between spent fuel pool storage and permanent storage. This
project focuses on dose optimization during this transition.
RoCD incorporates general dose reduction techniques as well as in-depth analysis of fuel rod
positioning to reduce worker dose. These techniques are applicable to dry cask storage
campaigns throughout the world.
1.2 Objective
The objective of this project is to reduce worker dose to more desirable levels during dry cask
storage campaign based on the ALARA principle. Methodology for accomplishing this objective
includes the following: assessing current work practices, time-motion studies of work, shielding
assessment and design, and strategic fuel assembly positioning.
1.3 Structure of the Report
The structure of this report is organized into broad chapters with specific sub-sections:
1) Introduction
2) Conceptual Design – Development and motivation of RoCD.
3) Feasibility – Initial analysis of the ease of dose reduction technique implementation.
4) Preliminary Design – Considers information from Chapters 2 & 3 and develops inputs for
technical analysis.
5) Technical Analysis – Perform analysis on radiation and economic models.
6) Conclusion – Achievements, Limitations and future activities of RoCD.
2. RoCD Conceptual Design
The main purpose behind RoCD’s mission is to uphold the ALARA principle as is required by
the NRC. RoCD seeks to reduce the amount of radiation dose that is acquired from transferring
SNF from wet storage to dry storage. This senior design project is an idea proposed by Quad
Cities Senior Radiation Protection Technical Specialist Christopher Gans and NATC Regional
Director David W. Miller. The goal is to further minimize the dose received by the employees
working on the transferring process by looking into relevant dose reduction techniques. The idea

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proposes the challenge of making the wet to dry storage transition safer. This chapter contains
background on the current loading and welding process, dose reduction techniques, economic,
ethical and social impacts.
2.1 Background
This section summarizes the reason for dry storage canisters. It also discusses the current dry
storage canister loading process.
2.1.1 Dry Storage Origins
As spent fuel pools (SFPs) at nuclear power plants (NPPs) near their capacities, more and more
nuclear utilities in the US are tasked with transitioning SNF from wet storage to dry storage.
Without a long-term spent nuclear fuel (SNF) storage option in sight, and with the NRCs clearer
commitment to dry storage after the events of Fukashima, dry storage is the sole option for SNF
storage beyond the SFP in the US.
2.1.2 Current Dry Cask Storage Process
The current dry cask storage process is generally standard between NPPs, so a broad step-by-step
procedure of this process follows. Data on current work practices referenced in this report have
been provided by D.C. Cook and Quad Cities Generating Stations.
Quad Cities and D.C. Cook use storage and transport canisters supplied by Holtec (HI-TRAC,
HI-STORM 100 and MPC-32/68). The outer steel envelope is made of Alloy 22, a Ni-Cr-Mo
alloy. This steel layer is generally 3-5” thick and is designed for corrosion resistance and
welding. Borated shielding (polyethylene) is about 6” thick. The storage canister is nearly 20’
tall and has a diameter of about 11’. It weighs over 180 tons. Due to these massive dimensions, it
is necessary to use a crane to transport the cask around the facility. Prior to fuel loading, the
MPC is inserted into the HI-TRAC. There is a small space between the MPC and HI-TRAC.
This space is called the annulus gap. The annulus gap is filled with demineralized water prior to
insertion in the spent fuel pool to provide shielding during later steps.
Fuel Loading Process

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The empty storage container (HI-TRAC and MPC) is lifted by a
crane and placed above the spent fuel pool. After it is lowered
into the pool and is fully submerged and filled with water, the fuel
assemblies are loaded into the MPC individually (Figure 1). The
loading pattern varies based on fuel age and thermal limits. The
loading crane is controlled by a worker that is at a safe distance
above the SNF pool. Once all the fuel rods are loaded into the
storage canister, a MPC lid (with drain and vent ports) is lowered
and placed on top of the cask. The cask is then pulled from the
pool, beginning the next stage of the storage process.
Drying, Preparation, and Decontamination
Once the cask is pulled from the pool, it is dried, prepped, and decontaminated. The annulus gap
filled with water provides shielding during this step. Figure 2 shows the workers performing
these preparation procedures.
Welding
Once the cask is dried, the lid welding process begins. The MPC lid is welded remotely or by a
worker if remote welding isn’t available. After the lid is welded, an inert gas (usually helium) is
pumped into the MPC lid ports to remove any residual moisture. The annulus gap that was filled
Figure 2. Workers preparing the cask for welding.
Figure 1. Fuel rod loaded in the
spent fuel pool.
Figure 3. Manual welding of lid/ports.

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with water is drained and dried. Vent and drain port covers are installed by a few workers and
manually welded shut (Figure 3). Next, a closure ring around the annulus gap is installed and
welded remotely or by a worker if remote welding isn’t available.
Transport to Storage Area
After the welding process is completed, the DSC is transported to the storage area. Before it can
be transported, the DSC is placed into the HI-STORM cask. After both the HI-TRAC and MPC
is placed inside the HI-STORM, the HI-TRAC is removed and prepped for reuse. The HI-
STORM and MPC is transported to the storage bay, where it will remain until final storage.
Figure 4. Dry casks on Independent Spent Fuel Storage Installation (ISFSI) at D.C. Cook Generating Station.
Dose reduction methods that can be applied during these steps and an optimized procedure
focusing on dose reduction will be discussed in later sections.
2.2 General Dose Reduction Techniques
There are three general dose reduction techniques: reducing time near the source, increasing
distance from the source and increasing shielding between the worker and source. An important
goal of any work procedure that involves radiation exposure is ALARA: “As Low As
Reasonably Achievable.” With ALARA being a federal mandate, all dose reduction techniques
must be considered. Economic considerations are based on the term “reasonable” and can vary
from plant to plant.
This section explains these fundamental concepts. The concepts will be applied in Section 3.

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2.2.1 Time
The general dose reduction method of time reduction will be discussed in the following section.
The equation for dose acquired is shown below:
𝐷𝑜𝑠𝑒 𝐴𝑐𝑞𝑢𝑖𝑟𝑒𝑑 = 𝐷𝑜𝑠𝑒 𝑅𝑎𝑡𝑒 ∗ 𝑇𝑖𝑚𝑒 (1)
The dose acquired is directly proportional to the time spent in the radiation zone. An example of
a dose calculation based on time is shown below:
DC Cook’s projected dose rate during the cask welding/processing step is 0.6 mREM/hr. The
total man hours spent on this step is estimated at 300.
𝐷𝑜𝑠𝑒 𝐴𝑐𝑞𝑢𝑖𝑟𝑒𝑑 = 0.6
𝑚𝑅𝐸𝑀
ℎ𝑟
∗ 300 ℎ𝑟 = 180 𝑚𝑅𝐸𝑀 (2)
The dose acquired calculation if the time is reduced by 20% is shown below:
𝐷𝑜𝑠𝑒 𝐴𝑐𝑞𝑢𝑖𝑟𝑒𝑑 = 0.6
𝑚𝑅𝐸𝑀
ℎ𝑟
∗ (300 ℎ𝑟 ∗ 0.8) = 0.6
𝑚𝑅𝐸𝑀
ℎ𝑟
∗ 240 ℎ𝑟 = 144 𝑚𝑅𝐸𝑀 (3)
Equations 2 and 3 show the effect of time reduction on total acquired dose. Reducing time is an
effective dose reduction method and the feasibility will be analyzed in later sections.
2.2.2 Distance
Increasing the distance between the worker and the radiation source reduces the worker’s dose.
Ways that distance will be utilized in reducing dose are as follows:
 Investigating robotic technologies for remote operation near the source
 Strategically arranging fuel assemblies in the MPC (multipurpose canister) in order to
distance workers from the assemblies with the highest source.
 Ensuring unnecessary personnel are not present in radiation areas, particularly during
work associated with higher dose rates.
2.2.3 Shielding
Ionizing radiation consists of alpha particles, beta particles, neutron radiation, gamma rays, x-
rays and ultraviolet radiation. Alpha particles and beta particles are stopped by very thin solid
materials and can be ignored due to the thickness of the dry shielded canister. High energy UV
radiation is also stopped by thin solid materials and can be ignored. X-rays do not have enough
energy to penetrate through the DSC. Gamma rays and neutrons are the only forms of radiation
that RoCD is concerned with while sealing the dry canister.

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2.3.3.1 Gamma Shielding
Equation 4 is used to find the number of transmitted gamma rays through a medium [2]. I0 is the
number of transmitted photons without any shielding. The thickness is t and μ is the linear
attenuation coefficient. The larger the attenuation coefficient, the more gamma rays the material
absorbed. I is the number of photons at distance t through the shielding. A thicker piece of
material will absorb more high energy electromagnetic radiation.
𝐼(𝑡) = 𝐼0 𝑒−𝜇𝑡
(4)
2.3.3.2 Neutron Shielding
Equation 5 is used for neutron interactions with matter [2]. It is the same as the gamma ray
transmission equation except the linear attenuation is replaced by the total macroscopic cross
section of the material. A larger cross section absorbs or scatters more neutrons. I(t) is the
intensity of uncollided neutrons. This equation does not account for scattered neutrons.
𝐼(𝑡) = 𝐼0 𝑒−Σ 𝑡𝑜𝑡 𝑡
(5)
To account for scattered neutrons, a Serpent code is used. Serpent is a three-dimensional
continuous-energy Monte Carlo reactor physics burnup calculation code. Shielding parameters
are an input for the code. More details on Serpent are in Section 5.1.
2.3 State-of-the-Art
Dose-reducing innovations are already being implemented. Methods include implementing a
variety of robotic technologies, innovative types of shielding, and analyzing how the effect of
cleaning and repositioning the fuel affects the amount of dose received. RoCD expects to present
these ideas and assess how each of these innovations can be beneficial to the primary objective.
2.3.1 Robotic Technology
Existing robotic technology can be classified into two different categories. The two types of
robotic implementations that can be utilized are remote technology and automated technology.
The automated system has an employee supervising from a different room that is either far away
from the source or shielded so that no harmful radiation is experienced. This system is more like
an assembly line, much like the automotive industry. Everything functions by itself through
coding and robotic interactions. There will be a few individuals enforcing quality control through
either a live video feed or through specific shielded areas. The next system that can be
implemented is the remote technology system. Remote technology systems require an operator to
control the machinery directly so the human element is still present. The employee is attempting
to work in 3D space through a monitor. Gaging depth in this scenario is difficult and additional
training must be provided in order to perform.

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2.3.1.1 Remote Technology
Considering the concern of the amount of dosage received by laborers, nuclear companies and
associations around the world began developing remote handling systems. Many are prototypes
that still need to be tested and optimized, while others are used currently in nuclear facilities. The
following information depicts the current progress of leading innovators in countries around the
world.
France has increased its attention on developing robotic solutions to handle radioactive materials.
The CEA is leading the cause along with its associates (COGEMA, EDF, and Framatome). Some
examples of projects they are working on are of the following: the Dexterous Arm for
Teleoperation (BD250), and the Virtual Reality for Telerobotics (MAGRITTE). The Dexterous
Arm for Teleoperation provides an impressive reduction of cost and includes a force-feedback
teleoperation control system. This helps with increasing the execution processes of the system. It
was marketed by SICN Company, and was vetted in 1998 at COGEMA La Hague plant.
MAGRITTE is a virtual system that trains future operators on how to handle remote control
systems. It was established mainly as a safety technique, which allows operators to practice
before testing their skills on an actual radioactive system. This allows for a safer environment
and provides a confidence boost to the operating personnel. It also prevents mistakes from
happening. [3]
Figure 5. MAGRITTE - French Virtual Reality Design for Telerobotics [3]
Japan is another leader in the robotic industry. They currently have many machines being
prototyped to be used in the nuclear industry. It is often compared to the robots that are operating
in the automobile industries. They are investigating many possibilities of how robots can serve
their dose reducing agenda. They are currently researching robotic inspection, maintenance, and
handling of SNF. An example of robotic involvement with handling nuclear waste is the two-arm
bilateral servo manipulator system. This system entails of a robotic system that has a cameras on
both sides of the machine and allows it to be controlled remotely by another individual. This

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robotic system has countless applications once further testing and optimization is done in the
future. [3]
Figure 6. Japanese Two-Arm Bilateral System [3]
Figure 7. Operators of the Two-Arm Bilateral System [3]
Incorporation of remote technologies is beneficial because it can be a feasible addition to a NPP.
It increases the distance between the source and the employee resulting in a negligible amount of
dose experienced. Further feasibility analysis is done in Section 3.4.

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2.3.1.2 Automated Technology
Automated technologies have been a solution many countries have looked at in reducing the
amount of exposure that employees receive. Some of those technologies are mentioned in the
Remote Technologies section above.
The United States was also in the process of creating a mechanical system for handling SNF at
Yucca Mountain, until the geological repository was closed down through the loss of federal
funding and political reasons in 2011. The waste handling building had five different systems
that were incorporated. Those five systems included of securing, raising, unloading, holding,
reloading, and encasing. The main mechanical systems at this building would be automated in an
assembly line fashion. The mechanical process that would be in place had the potential to be
efficient and safe at the same time. It solved the problem of personnel dose rate by having
majority of the work being conducted by assembly line machines. Direct employee contact with
the system could be modified to video loop feeds and select safe locations where temporary
shielding would be present. [3]
A completely automated process presents many solutions to the dose problem. It completely
takes the employee out of contact with the source. This is beneficial in that it allows a complete
reduction of the dose experienced by employees. Further feasibility analysis is done in Section
3.4.
2.3.2 Never Wet
NeverWet is a silicon based liquid that is sprayed on the surface of a material. When said
material comes in contact with water, a small air pocket is formed between the water and the
coating, allowing for any moisture to quickly be shed from the surface. This is because a surface
that has high micro-roughness and low surface energy will exhibit hydrophobic properties. The
use of NeverWet technology at the Robinson plant reduced the decontamination time and saved
an estimated 30 mrem/cask [4].
2.3.3 Spent Fuel Pool Demineralizer
A demineralizer intended to remove radioactive ions in spent fuel pools via ion exchange may
assist in lowering worker dose.
2.3.4 Fuel Assembly Cleaning
At some nuclear power plants, fuel assemblies are cleaned between fuel cycles in order to lower
Co-60 deposit amounts. Fuel assemblies can be cleaned either ultrasonically or with resin. This
cleaning leads to less Co-60 circulating through the plant, therefore lowering the dose to workers
in NPPs.

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It is not current practice to clean fuel assemblies prior to placing the fuel assemblies into wet
storage as SNF. However, removing Co-60 deposits (CRUD) from the fuel assemblies could
lower gamma dose to workers placing these fuel assemblies into dry storage.
2.3.5 Shielding Materials
The materials used for shielding are broken up into two categories, gamma and neutron
shielding.
Gamma radiation is best absorbed by materials with heavy nuclei. Lead is commonly used due to
its high atomic weight and density. Tungsten is also considered due to its density and reduced
toxicity. Other dense metals are rarer than tungsten and lead, therefore, are not considered
feasible for economic reasons.
Heavy, thick shielding is necessary for gamma radiation protection. The amount of shielding
may vary for different canisters based on their purpose. It may not be economical to have 12
inches of heavy lead shielding for a transportation cask but the shielding must be adequate for
NRC spent fuel transportation policies. Since the cask design will stay the same, additional
shielding may be added while workers are present. Lead blankets and plates are considered for
additional gamma shielding.
Tungsten can be mixed with plastic to form plastic pellets. Information provided by Exelon
indicates that the tungsten pellets are 1.16 times as effective as a solid sheet of lead of the same
thickness.
Materials with high neutron scattering cross sections and low atomic weights are good for
slowing neutrons. The most efficient way to remove energy from a neutron is to moderate it with
a particle with similar mass. The mass of hydrogen is nearly identical to the mass of a neutron.
Therefore, a large amount of hydrogen is good for reducing the energies of neutrons.
High density polyethylene ([C2H4]n) has a high density of hydrogen. Borated high-density
polyethylene (HDPE) is common in neutron shielding in medical vaults, nuclear reactors and
other radiation applications [5]. It combines a neutron absorber with a material that has a high
density of hydrogen. Borated HDPE is lightweight, cost-effective and durable. Sheets of HDPE
are ideal for neutron protection while having high mobility. These sheets are recommended for
additional neutron protection around a DSC while it is being loaded and sealed.

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2.4 Social Impact
The social impacts corresponding to this objective is in majority all positive. Minimizing the
amount of radiation experienced by employees is seen as a socially acceptable enterprise. A
standardized procedure to move SNF from wet storage to dry storage creates a community
amongst nuclear power plants that can further influence better techniques for dose reduction.
These innovative techniques can then be shared from power plant to power plant so that an
increased reduction can be seen on a collective front.
2.5 Environmental Impact
The overarching environmental issues that need to be taken into account when approaching this
problem all revolve around the effect of the radiation that is being emitted from the dry casks.
The entire process of taking the SNF out of the pool and putting it into the dry cask is taken care
of in the nuclear facility. Nuclear power plants are designed so that there is not much leakage
outside to the environment. Therefore, there is not much environmental impact within the scope
of this project. After the casks are transported out, they need to be stored in a secure location
away from any type of water supplies or residential areas. Mixed waste, a byproduct of using
lead shielding, can be hard on the environment and thus increasing the amount of nuclear waste
that is produced each year. The incorporation of different types of shielding materials will help
reduce the levels of mixed waste products. This will lead to a more beneficial environmental
impact.
2.6 Regulations
Federal and local regulations need to be assessed thoroughly at the beginning of our process. An
action towards reducing worker dose cannot be performed if it will violate an existing policy.
Federal occupational dose limits for adults are listed in Table 1, which are established by the
NRC in 10 CFR Part 61 [6].
Notably, local control levels for maximum occupational exposure at QC have not yet been
received. Additionally, Exelon sets an upper limit of 2 rem/yr for all employees. These levels are
more limiting than those of Table 1. However, even these local control limits will be less limiting
than our objective’s constraint: to lower total worker dose to under 0.100 man-rem for the
operation of concern.

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Table 1. NRC 10 CFR Part 61 Dose Limits
Dose Limit rem/year
Total effective dose
equivalent
5
Sum of deep-dose
equivalent and committed
dose equivalent to any
individual organ other than
those listed
50
Skin tissue 50
Lens of the eye 15
Reference [7] covers federal regulations for the control of exposure in high radiation areas
(HRA). The following entries from [7] could be of use:
1. “In place of the controls required by paragraph (a)* of this section for a high radiation
area, the licensee may substitute continuous direct or electronic surveillance that is
capable of preventing authorized entry.”
2. “The licensee may apply to the Commission for approval of alternative methods for
controlling access to high radiation areas”
*paragraph (a) outlines requirements for locking and controlling entry for areas >100 mrem/hr.
These regulations will allow for the use of a remote, electronic HRA guard system. Additionally,
the NRC could be receptive to new, inventive means of controlling HRAs.
2.7 Ethical Impact
Basing off of the ALARA principle, the nuclear facility is responsible for lowering the amount of
radiation experienced by workers. It is ethically binding that the nuclear facility in question
makes a reasonable effort to keep the dose experienced below the limits described by
regulations. This way, they are in accordance with ethical and legal constraints. The entire scope
of this project revolves around lowering employee dose. The impact will thereby be positive.
Appendix A outlines an exercise that concluded with RoCD’s ethics statement: “RoCD is
committed to researching ways in which nuclear power plant utilities can reduce radiation
exposure to their employees in accordance with the ALARA principle. Helping NPPs safely and

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efficiently reduce health risks to their workers is our goal, and is in keeping with our shared core
values: integrity, safety, and accountability.”
3. RoCD Feasibility
This section contains the decision making process, DSC specifications, economic analysis,
robotic feasibility, mock-up assessment, crowd control, fuel assembly positioning and the work
breakdown structure.
3.1 Decision Making Process
The primary objective of RoCD is lowering dose to the workers loading dry storage casks. There
are three general techniques to lower dose, as discussed earlier: minimizing the time that the
worker spends in the area, maximizing the distance of the worker from the source, and increasing
shielding between the source and the worker.
In addition, RoCD is interested in optimizing fuel assembly placement in the DSC based on a
variety of factors, including distancing workers from the radioactivity of the source (the SNF). A
computer model is used to assess optimization techniques.
For each dose-lowering technique, costs vs. man-rem saved is assessed in order to determine the
economic viability of the technique.
Figure 8. Ishikawa Diagram

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3.2 Dry Cask Specifications
The Holtec MPC-32 is used as the basis for calculations and design during later portions of this
report. Since it is proprietary information, specifications obtained are not made explicit during
this report
3.3 Economic Analysis
No new design for optimizing fuel transfer to dry cask can be implemented until it is proven that
it is economically viable compared to existing technology. The ALARA principle guides the
economic viability of the dose-lowering techniques explored by RoCD. Different NPPs each
have their own definition of “reasonable,” those definitions are based on costs. While each man-
rem saved is worth different amounts of money to different plant operators, our client is QC, and
it serves as the basis for our economic calculations. The amount of worker dose that we remove
via a particular method will be calculated into the money saved per man-rem according to QC,
which will then be compared to the cost of the dose-reducing method used. If money saved via
man-rem saved is greater than the costs of the dose-reducing method, the method will be
considered recommendable.
3.3.1 Dose Model
In this section, a preliminary cost-benefit calculation is performed from the data received from
1995 RP Calculation report received from Cook Nuclear Power Plant. The variables are defined
in the calculation below and multipliers are applied based on the changes that have occurred
since the 1989 calculation [8]. The equation used in the 1989 calculation is as follows:
𝐶𝑜𝑠𝑡 𝑝𝑒𝑟 𝑝𝑒𝑟𝑠𝑜𝑛 𝑟𝑒𝑚 𝑠𝑎𝑣𝑒𝑑 =
𝐴𝑣𝑒𝑟𝑎𝑔𝑒 𝐴𝑛𝑛𝑢𝑎𝑙 𝑆𝑎𝑙𝑎𝑟𝑦×𝐵𝑒𝑛𝑒𝑓𝑖𝑡 𝐹𝑎𝑐𝑡𝑜𝑟 ×𝐷𝑜𝑠𝑒 𝑊𝑜𝑟𝑡ℎ 𝐹𝑎𝑐𝑡𝑜𝑟
𝐴𝑛𝑛𝑢𝑎𝑙 𝐷𝑜𝑠𝑒 𝐿𝑖𝑚𝑖𝑡 (𝐴𝐷𝐿)
(6)
Table 2. Cost data from 1995 RP Calculation, Cook Nuclear Power Plant
Average Annual Salary $50,780
Benefit Factor 1.43
Annual Dose Limit 2 rem/year
Dose-Worth Factor 0.1 to 1.0
The Dose-Worth factor will remain at the 1989 value of 0.5. Therefore, the above Equation 6 can
be solved by using the values from Table 2, and the cost per person rem was calculated to be
$18,154. The NATC 2013 data on occupational exposure indicates that the cost per person rem
at DC Cook NPP is $32,632

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According to IAEA report, the ALARA-based credit for the reduced total population dose of 1
person-rem is $1000 [9]. Information System on Occupational Exposure (ISOE) estimates that it
costs from $10,000 to $80,000 per person rem which varies from different plants and utility
companies [10].
Each nuclear power plant has its own monetary value that can be used to evaluate dose reduction
proposals. Currently, Exelon fleet which operates 17 nuclear reactors developed a new formula
which can help determining $/person-rem saved in different sites. The formula starts off by
taking a base of $20,000 for each site and then using a multiplier factor of 1, 2, 3, or 4 based on
the current sites quartile rating for collective radiation exposure rating [10]. For Quad Cities, the
current cost per person rem saved is calculated using the following formula:
$20,000 × 𝐼𝑁𝑃𝑂 𝑅𝑃 𝑅𝑎𝑡𝑖𝑛𝑔 =
𝐷𝑜𝑙𝑙𝑎𝑟𝑠
𝑀𝑎𝑛 𝑅𝑒𝑚
(7)
The INPO RP Rating for Quad Cities falls on 3rd
quartile, and thus the cost per person rem saved
is $60,000 which is substantially higher than the national average for all US nuclear power
plants.
Therefore, the primary goal of radiation protection management at any nuclear power plant is to
economically optimize every possible way to minimize occupational dose. In particular, Quad
Cities Generating Station has higher INPO RP rating due to several factors which are still
unknown. These factors will be beneficial in providing the optimization of transferring spent fuel
from pool to the dry cask in a way that can lower the worker dose and still uphold the economic
viability.
3.3.2 Shielding Cost
For the purpose of this project, cost of different shielding can be evaluated to analyze the design
cost of reducing the worker dosage. One of the major contributions to reducing neutron dose
exposure is the implementation of shielding. Currently, around 12 sites in the United States use
high temperature lead wool shielding as a first layer of lid shielding. It corresponds to 60%
gamma attenuation of Co-60. The cost of high-temperature lead wool shielding is around $9,000
to $14,000 depending on various nuclear sites. Then, the composite shield is used which is
composed of Borated PE and Tungsten T-Flex. It is useful for the neutron and gamma shielding.
The cost of composite shield is anywhere from $75,000 to $125,000. This makes the total cost of
the lid shielding $84,000 to $139,000. This can be an important economic factor in reducing the
worker’s neutron dose exposure [11]. In addition to lid shielding, another way to reduce the dose
received by workers is to use movable shielding while welding the dry cask. The cost of 4’ by 8’
Borotron sheet 1 inch thick is over $600. When water is sucked out of the annulus gap, the
annulus lead shielding is used for gamma attenuation which costs around $50,000. All of the cost
for different shielding is shown below.

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Table 3. Cost data for different shielding type used in PWR/BWR dry cask
Shielding Type Cost
Lead wool shielding $9,000 to $14,000
Composite shielding $75,000 to $125,000
Borotron sheet (15-20
used per dry cask)
$600 per sheet * (15 to
20 sheets) = $9,000 to
$12,000
Annulus lead shielding $50,000
Total Cost for High
Dose Shielding
$143,000 to $201,000
Different types of shielding can be utilized in a way which can reduce the cost in a cost efficient
way. Waste storage costs of shielding materials is not considered as an additional cost because
the NPPs already dispose of shielding and much of the shielding is stored on site and reusable.
3.3.3 Remote Technologies
Ideally, the preliminary design of reducing the worker dose attempts to incorporate both
automated and remote systems. The steps of removing the SNF from the pools can potentially all
be done through mechanical systems, all functioning through limited human contact. The
transportation of the cask should be done remotely if possible to avoid any additional exposure.
The next problem that needs to be addressed is the factor of welding the dry cask shut. Currently,
both manual and automatic welding is used for two different welds. Incorporating another
automatic welding machine for the manual weld deems too costly and can add additional time
the employees are around the cask. Remote welding systems could be a valid solution to look
into in place of manual welding.
As of now, there are existing remote welding systems, specifically at Berkeley, that allow an
operator to successfully weld up to 100 feet away remotely[12]. The Berkeley’s welder has been
known to be the most reliable in the industry, and has been operational since 1995 [12]. The
AWS uses GTAW, which is known to provide a better quality weld.
3.4 Robotics Feasibility
There are many constraints with incorporating automated systems into nuclear facilities,
especially when dealing with high level wastes. In terms of reliability, if there is a problem with
a machine in an assembly line type of system, then there needs to be a way to repair the machine
remotely because the assembly line machine will be handling high level nuclear waste. This will
be more costly because the company would then need to investigate into remote robotic repairing
practices.

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Contrastingly, remote control technologies are more reliable because they allow the operator to
have more control over what is happening to the SNF. They are directly responsible for what
happens to the nuclear waste. Thus, making safety another key factor that needs to be addressed.
Remote technologies are a lot safer because of the reduction of dose that is experienced by the
employee. Mechanical systems are also safer because they allow for minimal direct human
contact. When considering the economic portion of the design process, it will cost a lot of money
in implementing the system of additional automated technologies, than the ones already in place.
Remote technology also requires a substantial cost. The training of the operators is a lengthy
process. In the long run, the company saves money because fewer employees will be needed to
run the process from pool to dry cask. Thus, the company spends less on the value per
person/rem. Based off this feasibility analysis, switching to more automated and remote
technologies to save workers from extra dose is achievable but not probable.
3.5 Mock-Up Assessment
Key steps in the procedure from transferring SNF from wet storage to dry storage are prepared
for via a mock-up. The purpose of the mock-up is to ensure worker familiarity with the
procedure in order to accomplish the work safely and efficiently.
Observing this mock-up would be instrumental in assessing ways in which worker dose could be
lowered. Potential sites for observing a mock-up include Quad Cities Power Station and D.C.
Cook Nuclear Generating Station.
3.6 Crowd Control
Crowd control is an important method of dose reduction that will be explored in RoCD. There
are tasks during the pool to cask storage procedure that require many workers to be in high
radiation areas. As a result, steps that do not require many workers are often overcrowded with
unnecessary personnel. A recent campaign to move SNF from wet to dry storage at Robinson
NPP concluded that one of the greatest threats to dose limits is crowd control. Simple crowd
control techniques can be utilized to significantly reduce worker dose at a negligible economic
risk.
Many plants utilize a “safe zone.” Workers should wait in the safe zone (low radiation) until a
remotely operated swing gate is opened when the task is ready to be performed. This has been
found to be highly effective in terms of limiting time spent in a high radiation area. This remote
operated swing gate is an inexpensive addition and does not require any adjustments to the plant
design.

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3.7 Fuel Assembly Positioning
The arrangement of the fuel assemblies inside of the DSC should factor into the dose rate outside
of the DSC. Fuel assemblies in the center of the cask use the outer assemblies’ materials
essentially as shielding. This “self-shielding” dynamic is assessed in a computer model in order
to investigate the possible dose rate outside of the DSC.
Serpent and MATLAB is used to model dose rate outside of the DSC. Input parameters include
DSC dimensions, materials, thicknesses, fuel assembly amount, fuel type, fuel assembly type,
fuel burnup, and years of wet storage (cooling time). Calculation packages for HI-STORM
transport casks from Holtec International may provide additional insight into this model.
While employees primarily work radially adjacent to the DSC, the welders operate above the
DSC. Therefore, the dose rate is calculated in the vertical and radial direction. This type of
modeling is not possible for the MATLAB simulation. A more rigorous calculation with three
dimensional code is recommended for precise dose rates.
3.8 Work Breakdown Structure
The work breakdown structure table is a table that depicts the actions that are taken to carry out
RoCD’s objective. The table helps illustrate the amount of time taken for each of the steps and
states who led the task. The chart also presents the order of activities that are taken. It is split up
into three main phases, conceptual design, preliminary design and technical design.

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Table 4. Work Breakdown Structure.
3.9 Product Breakdown Structure
The product breakdown structure chart is used to visualize the thought process for making the
decision on which dose reduction techniques should be used in our final protocol. It shows a
rough procedure on the very left of the chart and on the right side there are different decision
parameters that are followed in order to find a feasible solution to the dose problem. It must be
emphasized that the solution should be help lower the total worker dose to below 100 mrem and
the dose saved should outweigh the cost of the dose minimizing technique.

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Figure 9. Flow Chart of Product Breakdown Structure.
4. RoCD Preliminary Design
The objective of this section is to consider calculations involved in the preceding sections and
develop an input for the technical analysis. This section establishes all the relevant equations
which are used to calculate neutron and gamma calculations as well as final recommendations to
optimize the worker dose. The primary design is divided into six different analysis that include:
source calculations, shielding selection, distance calculation, dose calculation and economic
calculation. Specifically, the source calculation, distance calculation and dose calculation are
used to help determine the impact of rearrangement of spent fuel in the dry cask to lower worker
dose. The importance of shielding selection and economic calculation is to provide a guideline
for developing an optimal dose saving process that is economically viable.

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The equations in this section set a fundamental base for the neutron and gamma simulations
performed in Section 5. Most of the calculations in this section are performed from the data
obtained from DC Cook nuclear power plant.
The six analyses in this section are interconnected with each other. The source calculation is
calculated using the fuel assembly data obtained from DC Cook nuclear power plant. The
shielding selection, distance calculation and dose calculation are based from source calculations’
results that provided how much neutron and gamma flux will be radiated. The economic
calculation is performed using the shielding calculation and dose calculation and it generates a
process to optimize worker dose in the most cost effective way.
4.1 Source Calculation
Dose to the workers is dependent on the source to which the workers are exposed. RoCD
considers the fuel assemblies placed inside of the MPC during the dry storage cask campaign as
the source. Each fuel assembly has a number of parameters that may affect the overall source
strength: cooling time, burnup, enrichment, and fuel assembly type. Co-60 depositions will not
be considered during technical analysis.
The source term for a given DSC can be calculated by using the above fuel assembly parameters
as inputs. Gamma and neutron flux will be modeled in Matlab and in Serpent (a MCNP burnup
calculation code), respectively.
In order to calculate the source term, a burnup calculation will be made using Serpent.
Currently, RoCD is interested in how varying the fuel assembly cooling time, burnup, and the
fuel assembly layout in the MPC can beneficially lower dose to workers during the dry cask
storage campaign.
RoCD is assessing the MPC used at DC Cook Power Station: the Holtec MPC-32, intended for
use with PWR fuel assemblies. Although the focus is on only one MPC type, the findings on
percent-reduction of dose based on radial loading will apply to all MPCs.
While the NRC only requires 5 years of fuel assembly cooling time and has authorized fuel
assembly movement to dry storage after as few as 3 years, the industry standard is 10 years of
cooling [13]. However, there is no upper limitation for cooling time according to federal
regulations. With this in mind, RoCD assesses how dose rate could be lowered utilizing a
combination of fuel assemblies that have been cooled for 10, 20, and 30 years. The longer a fuel
assembly has been allowed to stay in the SFP, the higher the amount of radionuclide decay inside
of the fuel assemblies.

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4.2 Shielding Selection
As listed in the conceptual design, gamma and neutron radiation are the only types of radiation
that penetrate through the DSC lid and walls. This section discusses the energies of the gamma
rays and neutrons and general requirements for shielding materials.
4.2.1 Radiation Energies
Some gamma ray energy data is supplied by Holtec in report HI-2084188. Gamma ray creation
rates are tabulated for three SNF assemblies. These rates are divided into seven energy groups all
between 0.3 and 3.0 MeV [32]. The gamma ray data is not listed here because the report is
proprietary. This data is used in the gamma ray simulation in Section 5.2 Gamma Simulation.
Neutron data from the same assemblies is also tabulated. The range of energies for neutrons is
0.1 and 20 MeV [32]. More details on the neutron simulation are in Section 5.1 Neutron Model.
To properly assess which steps to include supplementary shielding, it is important to understand
the dose rates from different locations around the storage canister. Survey data taken from a
particular DCSC at D.C. Cook show gamma and neutron data from the top of the MPC, the side
of the HI-TRAC, and the annulus gap (after water is drained) can be seen in Table 7.
Table 5. Dose Rate at Various Locations (HI-TRAC)
Location
Gamma
(mrem/hr)
Neutron
(mrem/hr)
Top of MPC 12-15
19-25 (at edge)
65-70 (at
center)
Side of HI-
TRAC
20-25 15-20
Annulus Gap 700-800 20-30
Gamma dose rate at the annulus gap and neutron dose rate at the top of the MPC are the highest
dose rates. Since welders work in these areas, these dose rates will be the focal point for the
RoCD shielding design.
4.2.2 Shielding Material Selection
High atom density material is desired for all types of shielding. The gamma shielding will either
be lead or tungsten. Pure tungsten is slightly better than pure lead when it comes to shielding.
Figure 10 is a direct comparison of lead and tungsten attenuation of 1 MeV gamma rays. The

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attenuation coefficients for lead and tungsten at 1 MeV are .77 cm-1
and 1.08 cm-1
, respectively
[14].
Borated high density polyethylene is industry standard for neutron shielding. Water was
previously (and can still be) used, but borated PE is easier to move. PE is easily purchased in 1”
thick sheets and is lightweight. It slows and captures some neutrons as they pass through it. The
total macroscopic cross section for PE is 0.449 cm-1
for 1 MeV neutrons [15].
Exact pricing of borated HDPE is available from multiple vendors. Shielding is reusable, so
additional shielding will be one-time costs for the life of the shielding. More detail on pricing
and economics will be calculated in Section 4.6 Economic Calculations.
Figure 10. Intensity of 1 MeV gamma rays versus the thickness of lead and tungsten.
4.3 Distance Model
Increasing distance is an important component of reducing employee dose. There are many
different approaches to achieve this solution. The most basic and simplest form of increasing
distance that can be integrated into the current procedure is to make sure every employee is

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accounted for. There must be announcements to clear specific areas, and warnings for other
unaware employees preventing them from entering a possible radiation site zone. Furthermore, a
greater separation from the source can be accomplished by physically situating an employee
away from the source. If the person is not working on the dry cask, they should not be near the
dry cask perimeter. Robotic instrumentation is also a feasible solution to reducing the amount of
dose that is attributed to each worker. The economic calculation for incorporating automated and
remote robotics into nuclear facilities is seen below in the Economic Calculation Section.
4.4 Time Calculation
Time reduction will be considered in this section. A recent assessment at Cook developed a
person rem estimate using information gathered from benchmarking other sites that performed
fuel loading campaigns using Holtec systems. Table 6 shows the acquired dose estimate for a DC
Cook’s 2012 Dry Cask campaign.

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Table 6. Dose estimate from fuel loading campaigns.
Figure 11 shows the estimated dose and the actual dose obtained during the 12 cask loadings
done in the dry cask storage campaign of Cook in 2012. The dose is reduced to nearly 100 mrem
during this campaign. The techniques used in Cook will heavily influence the optimal procedure
that is developed in Section 5.4. Time based recommendations are listed in Section 5.4
Task (MPC 1) Number of Workers Duration Person Hrs Eff. Dose Rate Est. mrem
Mobilize Equip./Prep Cask 10 100 1000 0.02 20
Load Fuel in MPC 3 20 60 0.5 30
Decon/Movement 10 6 60 1 60
Welding/Processing 6 50 300 0.6 180
Transfer MPC to HI STORM 10 10 100 1 100
Place on ISFSI Pad 5 5 25 0.4 10
TOTAL 400
Task (MPC 2-3) Number of Workers Duration Person Hrs Eff. Dose Rate Est. mrem
Mobilize Equip./Prep Cask 10 12 120 0.02 2.4
Transfer MPC to HI STORM 10 10 100 0.9 90
TOTAL 347.4
TOTAL 327.4
Decon/Movement 10 4.5 45 1 45
TOTAL 302.4
TOTAL 277.4

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Figure 11. Dry Cask Campaign Dose (2012)
4.5 Dose Calculation
External dose calculation can be described as a function of time, distance and shielding. In order
to calculate dose rate, the first step is to layout dose limits set by the NRC as listed in Table 7
[16]. These dose limits are in unit of rem/yr. The unit rem, often referred to as roentgen
equivalent man, is commonly used for equivalent absorbed dose of radiation. It takes into
account the relative biological effectiveness of different forms of ionizing radiation. It also takes
into account for the varying ways in which the radiation transfers its energy to human tissue. The
unit rem can be defined using the following equation:
𝑟𝑒𝑚 = 𝑟𝑎𝑑 × 𝑄 (8)
Q is described to be as the quality factor.
Table 7. Federal Regulation 10 CFR 20 limits.
0
100
200
300
400
500
600
700
800
1 2 3 4 5 6 7 8 9 10 11 12
Dose(mrem)
Cask #
Dry Cask Campaign Dose 2012
ED = 3292 mrem
Actual = 2293 mrem
ED
Actual

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Body Part Dose Limit (rem/yr)
Whole body (TEDE) 5
Lens of eye 15
Skin 50
Extremities 50
Internal organs 50
The quality factor converts absorbed dose (rad) to dose equivalent (rem) [17]. The quality factor
is equivalent to one for gamma and beta source. Therefore, the rem equals rad for gamma and
beta sources.
Dose exposure is classified into external and internal radiation exposures. External radiation
exposure is caused from a source external to the body which penetrates the skin and leads to a
dose of ionizing radiation. These exposures occur from neutrons, beta particles or gamma rays.
Internal radiation exposure is caused from radioactive materials present in the body through
different uptake which includes: skin contact, ingestion and inhalation. Primary internal exposure
occurs when radioactive material is inhaled and deposited in the respiratory system [18].
Dose as function of time is a linear function. It can be calculated using the following equation:
𝐷 = 𝐷̇ × 𝑡 (9)
Where D = radiation dose, 𝐷̇ = radiation dose rate and t = exposure time [19].
Dose as a function of distance is directly correlated to the area of the source [19]. The
relationship of dose as a function of distance is shown below:
𝐷(𝑟) ∝
𝐴
4π𝑟2
(10)
Furthermore, the radiation dose decreases with increase in distance. This is proved by using the
following equation:
𝐷2 𝑆2
2
= 𝐷1 𝑆1
2
(11)
Where D1 is dose rate at distance 1, S2 is distance 1, D2 is dose rate at distance 2 and S2 is
distance 2 [19].
The above Equation 11 indicates that radiation dose decreases as the square of the distance from
the radiation source emitting radiation. This equation is used to create a general relationship
between distance and dose rate for neutron and gamma model.
Finally, the dose as function of shielding is calculated using the following equation:

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𝐷 = 𝐷𝑜 𝑒(−𝜇𝑥)
(12)
Where D is the dose rate after passing through a thickness x
Do is the original radiation dose rate
µ is the linear attenuation coefficient (cm-1
)
x is the thickness of attenuating material (cm)
The Equation 12 is used to determine the relationship between dose and shielding. This equation
is a simple shielding calculation. It can be used to assume how much additional shielding is
required to help mitigate dose experienced by workers. This equation is also used to find the
percent dose reduction for each shielding material that can help create the final shielding design.
Overall, dose variation can be related to time, shielding and distance by the following
correlation:
𝐷 ∝
𝐴 𝑡 𝑒(−𝜇𝑥)
𝑑2
(13)
Where D is the dose rate, A is the activity, t is time, 𝜇 is the linear attenuation coefficient and
𝑥 represents the thickness of the attenuating material and d2
represents distance.
The dose rates due to the neutron source, gamma source and the overall dose is listed in Figure
11 for all of the employees working during the first dry cask campaign [19]. The average neutron
dose rate at the top surface of the DSC is modeled using Serpent software.
4.6 Economic Calculation
The goal of RoCD is to reasonably lessen dose for workers during the wet to dry storage
transition process. The term reasonably is rather vague but stems from the ALARA model (As
Low as Reasonably Achievable). “Reasonable” takes into account the economics and ease of
implementation. This section delves into the economic analysis and what steps must be taken to
determine whether a dose reduction method implementation is reasonable.
The various methods of saving dose that are explored in Section 4.4 must be analyzed in terms of
cost and dose reduction effectiveness. A tabulated Excel program is utilized to help with this
analysis. This program takes into account plant quartile and allows for the user to select a type of
dose reduction method. Each method has a known dose saved and cost. Using these parameters,
the program performs a check for how many dry cask processes that must take place before the
“break-even” point. Knowing the number of casks that are loaded and stored per year, the break-
even point can be determined in number of years. This is valuable information for a NPP because
it provides a tangible worth for a dose saving method. Using this economic model, an optimal
procedure can be formulated.

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Table 8.Assumptions and equations utilized by the economic model.
Known Variable Description
$
𝑚𝑎𝑛 𝑅𝑒𝑚 𝑠𝑎𝑣𝑒𝑑
a
This value is known based on the IPO factor
multiplied by the base price of $20,000. This is the
value that a plant attributes to 1 saved man rem.
𝑚𝑎𝑛 𝑅𝑒𝑚 𝑠𝑎𝑣𝑒𝑑
𝑐𝑎𝑠𝑘
b
A method of dose reduction can be evaluated and
the amount of dose that is saved per cask is
reported.
𝑐𝑜𝑠𝑡 𝑜𝑓 𝑖𝑚𝑝𝑙𝑒𝑚𝑒𝑛𝑡𝑎𝑡𝑖𝑜𝑛 ($) c
Implementation costs of dose reduction methods
are determined and reported.
𝑎 ∗ 𝑏 =
$
𝑐𝑎𝑠𝑘
(14)
The Equation 14 represents the dollar value a plant accredits to a dose reduction method
𝑐
𝑎∗𝑏
=
$
$
𝑐𝑎𝑠𝑘⁄
(15)
Using the Equations 14 and 15, an economic model is created to determine the number of casks
it requires until a ‘break-even’ point is reached. The reason the economic model is developed is
to determine the steps of an optimal wet to dry storage procedure. Also, this economic model
helps to provide a guideline for developing a dose reduction process that is fiscally feasible for
the power plant.
4.6.1 Shielding Economics
To achieve the collective dose goal of under 100 mrem, it is necessary to make some adjustments
to the current storage procedure. Based on the Feasibility Analysis, the most viable option for
dose reduction is the incorporation of additional shielding. Although reducing time and
increasing distance from a radiation source are effective in lowering acquired dose, these
methods are already optimized and incorporated in the current procedure. NPPs place a large
emphasis on limiting the time spent in high radiation zones. The work on and around the storage
canister cannot be done at an increased distance. Therefore, shielding will be the focus of
RoCD’s final recommendations. Additionally, worker positioning during certain tasks will be
explored.
Due to the ease of transporting and repositioning, blanket and sheet shielding are the best types
of shielding to use on the top of the MPC/HI-TRAC. Moveable shielding is optimal for use on
the side of the MPC/HI-TRAC because it can be set up on the ground and placed around the

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perimeter of the canister. Finally, the annulus gap is best shielded by ring shielding. Ring
shielding is a thin, snake-like ring that sits around the top of the annulus gap.
All of the different forms of shielding can be made of various materials, and each type of
material must be analyzed for effectiveness (in terms of dose reduction) and cost.
Table 9 shows the percent reduction of 1 MeV gamma rays for viable shielding materials [14].
The shielding thickness ranges from 1 cm to 5 cm. The materials are listed in order of shielding
effectiveness.
Table 9. Percent Reduction of Gamma Dose for Various Shielding Materials and Thicknesses
Material
Density
(g/cm^3)
Micro. x-
section
(cm^-1)
Percent Reduction with Respect to
Thickness
1 cm 2 cm 3 cm 4 cm 5 cm
Tungsten 19.25 1.27 72.03 92.18 97.81 99.39 99.83
Lead 11.34 0.805 55.31 80.03 91.07 96.01 98.22
Bismuth 9.87 0.712 50.93 75.93 88.19 94.20 97.16
Iron 7.874 0.472 37.63 61.10 75.74 84.87 90.56
It can be observed that tungsten is the most effective gamma shielding material. Lead and
bismuth have relatively similar shielding effectiveness, while iron is the least effective. In order
to determine which material to use, the cost of each must be determined. Table 10 summarizes
the cost data for the various gamma shielding materials [20]. The materials are listed in order of
cost by volume (high to low).
Table 10. Cost Data for Gamma Shielding Materials
Material
Density
(g/cm^3)
Cost by weight
($/g)
Cost by volume
($/cm^3)
Tungsten 19.25 0.0295 0.5679
Bismuth 9.87 0.0283 0.2793
Lead 11.34 0.0002 0.0023
Iron 7.874 0.0002 0.0016
For the sake of analysis, the benchmark for percent reduction will be set at 90%. The thickness
required to achieve this percent reduction for each material is shown in Table 11.
Table 11. Thickness Required to Achieve 90% Reduction in Gamma Dose for Shielding Materials
Material
Thickness required
to achieve 90%
dose reduction (cm)

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Tungsten 1.81
Lead 2.86
Bismuth 3.23
Iron 4.88
It is observed that tungsten requires much less thickness for the indicated gamma dose reduction
benchmark than any other materials.
By combining information from Table 9-Table 11, it is possible to illustrate the thickness, cost
and effectiveness in the same graphic (Figure 12).
Figure 12. Economic Analysis of Gamma Shielding Materials.
Figure 12 shows the percent reduction of gamma dose as a function of the cost per area of
shielding. The thickness of the shielding material used is shown in the data labels. The cost
effectiveness and the relative thickness can be compared between materials. For iron and lead,
there is not much added cost for increasing thickness due to the low cost of these materials.
Tungsten and bismuth, however, have a significant added cost for increasing their thicknesses.
Using thickness data labels along with the conclusions from the economic analysis (Figure 12),
the optimal material for the different types of shielding (blanket, moveable, and ring) can be
determined.
Next, neutron shielding effectiveness is explored. The same methodology that is seen in the
analysis of gamma shielding is used for neutron shielding. The neutron shielding materials are
1 cm
2 cm
1 cm
2 cm
3 cm
1 cm
2 cm
3 cm
1 cm
2 cm
3 cm
4 cm
5 cm
30
40
50
60
70
80
90
100
0 0.2 0.4 0.6 0.8 1 1.2
PercentReductionofGammaDose
Cost ($/cm^2)
Percent Reduction of Gamma Dose vs Cost
Tungsten Bismuth Lead Iron

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summarized in Table 12 [15, 20]. The thickness of material and the percent reduction of neutron
dose is tabulated as well.
Table 12. Summary of Neutron Shielding Materials
Material
Cost
($/cm^3)
Micro. X-
Section
(cm^-1)
Percent Reduction with Respect to Thickness
1 cm 2 cm 3 cm 4 cm 5 cm
Borated
Poly.
0.008 0.1907 17.36 31.71 43.57 53.36 61.46
Concrete 1.18E-04 0.2572 22.68 40.21 53.77 64.26 72.36
Water -- 0.56 42.88 67.37 81.36 89.35 93.92
Again, it is possible to illustrate the cost effectiveness and dose reduction on one diagram.
Figure 13. Economic Analysis of Neutron Shielding Materials.
Figure 13 shows the percent reduction of neutron dose as a function of cost. The thickness
required is included in the data labels.
5. RoCD Technical Analysis
The primary objective of this section is to perform technical analysis of the neutron model,
gamma model and final recommendations to optimize worker dose using economic analysis. The
inputs are taken from the preliminary design to model them using serpent and matlab for neutron
and gamma model respectively. The final recommendations optimize the worker dose includes
using extra shielding keeping it cost effective.
1 cm
2 cm
3 cm
4 cm
5 cm
1 cm
2 cm
3 cm
4 cm
5 cm
0
10
20
30
40
50
60
70
80
0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04 0.045
PercentReductionofNeutronDose
Cost ($/cm^2)
Percent Reduction of Neutron Dose vs Cost
Borated Ployethylene Concrete

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The neutron model and gamma model is performed using several assumptions from the data
obtain from DC Cook nuclear power plant. Both the models supports that the rearrangement of
spent fuel rods have correlation to the worker dose.
5.1 MPC Neutron Flux Model
Serpent, a Monte Carlo burnup calculation code, is used to model the effectiveness of
strategically arranging fuel assemblies of higher source values towards the center of the MPC.
The MPC is modeled as a 2-D infinite cylinder containing 32 fuel assemblies, based on the
Holtec MPC-32 used for storing spent PWR fuel assemblies. Geometry infinite in the axial
direction is applied in order for Serpent’s output of neutron leakage to only apply to the radial
direction, which better gauges the effectiveness of self-shielding in a given MPC fuel assembly
arrangement. Self-shielding refers to a fuel assembly’s material ability to attenuate its own
neutrons or a neighboring assembly’s. Additionally, the MPC-32 has neutron-absorbing
materials between each of its fuel storage cells. Each of these materials are included in the
Serpent model of neutron leakage outside of the geometry of the MPC. Neutron leakage is
measured by one of the reaction mode counters in the Serpent output file, which records each
instance of a neutron leaking outside of the model’s geometry.
MPC fuel assembly arrangements are tested for neutron shielding effectiveness by varying one
of two properties: burnup or cooling time. The total number of a given type of fuel assembly
inside of the MPC will not vary from one arrangement to the next as long as said arrangements
are being compared to each other.
The Serpent MPC model includes the following assumptions:
 Thermal and burnup limits are assumed to be met
 Fuel assemblies are 17x17 Westinghouse Standard
 All fuel is 3.6 weight % U-235 initial enrichment
 The final boundary of the MPC is the outer diameter of the outermost fuel storage cells
 Each fuel assembly is in the exact center of its respective MPC fuel storage cell
 MPC sides are equidistant (no protrusions)
 Neutron absorbing material (boral) is on each fuel storage cell wall, including the
outermost fuel storage cell walls (note: boral thickness was decreased by half in order to
help compensate for this assumption)
 For code symmetry purposes, the outermost fuel storage cell walls are assumed to have
half of actual steel thickness
 There are no gaps between boral and the corners of the fuel storage cell walls
 Assume any void space is filled with air (vice pressurized Helium)
 Each fuel assembly contains 264 identical fuel pins, plus 25 guide tubes

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 Due to cross section libraries limitations, materials are assumed to be at either 600 K or
300 K (for fuel and non-fuel, respectively)
Table 13 shows the dimensions of the fuel assemblies used in the MPC Serpent model [30].
MPC-32 dimensions are based on a Holtec schematic [31].
Table 13. Westinghouse Standard 17x17 Fuel Assembly Specifications
Fuel Cell Pitch 1.26 cm
Fuel Pellet OD 0.8192 cm
Cladding ID 0.83566 cm
Cladding OD 0.950 cm
Guide Tube ID 1.143 cm
Guide Tube OD 1.224 cm
An example of one run of the Serpent MPC model code can be seen in Appendix B.
In order to provide varying inputs into the Serpent MPC model, a burnup calculation is run in
Serpent. Burnup is calculated for a single fuel pin of 3.6% enrichment as previously discussed.
This fuel pin is subjected to burnup of 10, 20, and 40 MWd/kgU. 214 decay and fission yield
products are tracked through each of these burnup amounts. Not all of these nuclides contribute
to neutron flux: key contributors to gamma flux are also tracked for other potential purposes.
These three burnup calculations can be seen in Appendix C.
For each of these three burnup amounts, fuel composition is obtained for the fuel pin after
cooling times of 5, 10, 20, and 30 years from the completion of burnup.
In total, twelve sets of depleted fuel pin data were collected. The fuel composition of these
depleted fuel pins are applied to the fuel assembly MPC model in Serpent. The burnup and
cooling times chosen for analysis reflect typical values in NPP SFP inventory.
The Serpent burnup calculation includes the following assumptions:
 Each burnup step contains 500 source neutrons, 1000 active cycles, and 20 inactive
cycles
 The gap between fuel pellet and cladding may be ignored
 Burnup occurs in light, non-borated water
 In order to simulate adjacency to other fuel pins, Serpent boundary condition 3 is applied
(i.e. neutrons do not escape during the burnup process, they simply reenter at the opposite
end of the fuel pin)
 During this calculation, fuel is at 900 K and non-fuel materials are at 600 K

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With the MPC-32 geometry modeled and fuel composition calculated for the twelve
aforementioned types, Serpent is then used to model different arrangements of fuel assemblies
inside of the MPC in order to assess neutron flux outside of the MPC.
Figure 14 shows the six MPC loading patterns that have been modeled. Referring to Figure 14:
MPC 1 was compared to MPC 2, MPC 3 was compared to MPC 4, and MPC 5 was compared to
MPC 6. Note that for each set of compared MPC loading patterns, the total number of fuel
assemblies of any given type between the two patterns stays the same (e.g. MPC 5 and MPC 6
both contain four assemblies of 40 MWd/kgU burnup, 12 assemblies of 20 MWd/kgU burnup,
and 16 assemblies of 10 MWd/kgU burnup).
Calculation type: criticality does not refer to modeling the MPC as a critical system; this
modeled system’s keff was expectedly much less than 1. “Criticality” refers to the fact that this
was modeled using actual nuclear fuel data vice an “external source” calculation where neutron
population is determined by the user at every step.
Table 14. Serpent MPC Neutron Model Results shows noteworthy Serpent MPC model inputs
and results from these six MPC loading patterns.
Table 14. Serpent MPC Neutron Model Results

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Figure 14. MPC-32 Assembly Loading Patterns.
The results of the RoCD MPC neutron model shown in Table 14 suggest three key points:
 Spent fuel pool cooling time has very little effect on the neutron source term. MPC
loading patterns 1 and 2
 While the Serpent model has a limited selection of material temperatures to choose from,
the neutron model is not affected notably by introducing either a much higher (900 K) or
much lower (300 K) temperature. While the difference in neutron leakage is 0.64136%
between the same MPC loading when material temperature changes from 600 K to 300
K, in reality, the actual fuel assembly temperature is not be as low as 300 K, so this error
is exaggerated in this case. The reason in the error is due to scattering cross sections
varying for a given material as temperature changes.
 Strategically arranging fuel assemblies based on burnup can reduce neutron flux outside
of the MPC. Neutron leakage decreased by 10.8% from MPC 3 to MPC 4 by swapping

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the higher source term (assemblies with 10 MWd/kgU burnup) at the perimeter of the
MPC with the lower source term (assemblies with 40 MWd/kgU burnup) at the center of
the MPC. The same effect can be seen when only four fuel assemblies are switched in a
similar manner, i.e. from MPC 5 to MPC 6, where neutron leakage was reduced by
5.32% (note that statistical error data is not available for this Serpent output data).
5.2 MPC Gamma Flux Simulation
Along with a neutron model to compare radial and uniform sources, a gamma model is
necessary. MATLAB was used for this model.
5.2.1 Gamma Model
Using MATLAB, a gamma source region could not be created. Instead, the source region is
created by making line sources throughout the region. There is one line source per centimeter of
source region. An accurate radial dependence could not be created using a line source method for
the source region, therefore, the simulation uses a 1-D slab geometry. Figure 15 shows a plot of
two gamma ray intensities coming from the source region. One relies on a decreasing source
with a maximum at the center of the slab, the other is uniform throughout the slab. The axis of
this simulation centers the slab at 70 cm. The steel and concrete regions have thicknesses of 3 cm
and 32 cm, respectively. The thicknesses of the chosen to simulate a MPC.

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Figure 15. Plot of gamma ray intensity over distance inside a source material, steel, concrete and air.
5.2.2 Gamma Ray Source
Both sources have the same amount of gammas created per second over the entire source region.
Since this is true, the intensity numbers listed on the y-axis are arbitrary. The ratio between the
gamma creation rates for the decreasing and uniform source will not change for a given distance
from the center if the intensity changes. This ratio is determined by the minimum (140 cm) and
maximum (70 cm) gamma production rates for the decreasing source.
5.2.2.1 Source Strength
The center and edge of the decreasing source uses the data from two assemblies with different
burnup and enrichments. This is shown in Table 15. All assemblies in the data used have a
cooling time of 7 years. The assemblies with the lowest and highest radioactivity are used.
Further detail on how many gamma rays produced in these assemblies are protected by Holtec
Industries and are not be listed in this report. The rate of gamma ray production is given for these
assemblies and is used for the simulation. The actual simulation uses the ratio of total gammas
created by the hotter fuel and the cooler fuel and not the actual gammas created. This way the

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decreasing source can be scaled down to the magnitude of the uniform source. The rate of
production was given in seven different energy ranges, 0.3-0.45, 0.45-0.7, 0.7-1.0, 1.0-1.5, 1.5-
2.0, 2.0-2.5 and 2.5-3.0 MeV. Linear attenuation coefficients are found for each energy range so
the varying energy of the gammas was accounted for.
Table 15. Properties of fuel assemblies used in simulation.
Burnup (MWD/MTU) Enrichment (%) Cooling Time
(Years)
Center Assembly 60000 3.4 7
Edge Assembly 35000 1 7
5.2.3 Results
The gamma simulation depicts a decrease in gamma rays of 15% between the decreasing model
and the uniform model. This provides a proof of concept that putting hotter assemblies in the
center may be beneficial to reduce the amount of radiation that escapes a DSC. However, there
are a few assumptions in this model that could produce errant results. First of all, the model is
not accounting for cylindrical geometry. This may change the magnitude of the decrease in
gammas that escape the cask. Secondly, the source region is generalized into one region. A better
model can be made that labels each assembly as its own region in two or three dimensions.
Third, the linear transfer coefficient is used in these models. The energy transfer coefficient
should be used. An ideal simulation would use a more advanced code to account for these errors.
Lastly, the current simulation does not use the actual magnitude of gamma rays. Without it, dose
cannot be calculated. Although not a complete simulation, this model suggests an investigation
in fuel assembly arrangement may reduce the dose to workers which benefits companies that
load DSC’s.
5.3 Economic Model
ALARA is defined by a cost-benefit model. This model is used by IAEA and the International
Commission on Radiation Protection. The methodology and results of a cost-benefit analysis will
be discussed. This model will help determine the optimal use of shielding materials and can be
incorporated in material selection and shielding design.
5.3.1 Methodology
In order to model an economic analysis, it is very important to implement ALARA concept
adopted by IAEA and ICRP (International Commission on Radiological Protection). The ICRP
recommended that ALARA principle can be implemented on the basis of optimization of
radiation protection efforts [21]. The optimization model is defined as the attainment of a
balance between the radiation safety benefits obtained from the resources committed to radiation

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safety and benefits obtained by committing these resources to other avenues [21]. The sample
plot of method of cost-benefit is shown in Figure 16.
Figure 16. Optimization of Radiation Protection Cost-Benefit Model
The above Figure 16 shows the cost of the sum of the detrimental effects of radiation, the
detriment, that are assumed to be directly proportional to the collective dose to the population
being protected plus the cost of radiation protection as a function of the collective dose [21]. The
amount of radiation protection leading to the minimum in the curve is considered the optimum
degree of radiation protection.
The net benefit B of optimization of radiation protection cost-benefit model is determined by the
following equation [21]:
𝐵 = 𝑉 − 𝑃 − 𝑋 − 𝑌 (16)
Where V = gross value of procedure, P = cost of the procedure, exclusive of the cost of
protection, X = protection cost and Y = detriment cost
The cost of detriment (Y) can be calculated using the following equation [21]:
𝑌 = 𝛼𝑆 (17)
The cost of the shield when shield thickness is optimized is [21]:
𝑋 = 𝐶𝐴𝑡 (18)

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Where C = cost per unit volume of the shield, A = area of the shield, t = thickness of the shield
The decreased collective dose due to the additional shielding is given by [21]:
𝑆 = 𝐻̇ 𝑒−𝜇𝑡
× 𝑓 × 𝑁 × 𝜏 (19)
Where 𝐻̇ = maximum dose-equivalent rate in the shielded area
f = ratio of the average to maximum dose rates in the shielded area
N = number of people in the shielded area
𝜏 = lifetime of the shielded installation
µ = effective attenuation coefficient of shielding material
t = thickness of the additional shielding.
The following equation may be used in order to solve for optimum shield thickness [21]:
𝑡 = ln (
𝐶𝐴
∝ × 𝐻̇ ×𝑓 ×𝑁 × 𝜏 ×𝜇
) × (
1
−𝜇
) (20)
Based on this methodology, various shielding options are analyzed and the results are presented.
5.3.2 Results
An analysis on various materials often used in shielding is performed using the methodology in
Section 5.3.1. Using equations 17-19, the radiation protection cost benefit of these materials is
optimized. Assumptions for these equations are summarized in Table 16.
Table 16. Summary of Assumptions for the Cost-Benefit Analysis
Variable Value Used Assumption
𝐻̇ 500 mREM
Standard value used for each
shielding material.
f 1
The dose is assumed to be
relatively constant throughout
the work area.
N 10
Based on current work
procedure.
τ 20 years (Section 4.2.2)

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The results are shown in Figure 17. The optimal thickness for shielding materials based on
Equation 20 is tabulated below:
Table 17. Optimal Thickness for Various Materials
Material
Optimal
Thickness (m)
Borotron 0.175306
T-Flex (Tungsten) 0.029848
T-Flex (Bismuth) 0.036028
T-Flex (Iron) 0.151047
Borated (5%) Poly. 0.168321

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Figure 17. Cost Optimization Model for Various Shielding Materials
0
500
1000
1500
2000
2500
3000
3500
4000
4500
5000
0.00 0.02 0.04 0.06 0.08 0.10 0.12 0.14
Cost
CollectiveDose Commitment S
CostOptimization of Borotron Shielding
X (Cost of Shield in $)
Y (Monetary Value of Health Detriment)
X+Y (Cost of rad. Protection + Cost of rad. Health detriment)
0
2000
4000
6000
8000
10000
12000
14000
16000
0.00 0.10 0.20 0.30 0.40 0.50
Cost
CollectiveDose Commitment
CostOptimization of T-Flex Tungsten Shielding
0
2000
4000
6000
8000
10000
12000
14000
16000
0.00 0.10 0.20 0.30 0.40 0.50
Cost
CostOptimization of T-Flex Bismuth Shielding
0
1000
2000
3000
4000
5000
6000
7000
8000
0.00 0.05 0.10 0.15 0.20 0.25
Cost
CostOptimization of T-Flex Iron
0
1000
2000
3000
4000
5000
6000
0.00 0.05 0.10 0.15 0.20
Cost
CollectiveDose Commitment S
Cost Optimization of Borated (5%) Polyethylene
Shielding

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