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: [email protected]

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 Reduction of Cask Dose Release date: 4/26/2015

Author: K. D’Souza, J. Exner, A. Muneeruddin, P. Ota, A. Patel Doc: RoCD

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.



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



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



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



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



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



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



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



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



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



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.



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.



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 = 0.6𝑚𝑅𝐸𝑀

ℎ𝑟∗ 300 ℎ𝑟 = 180 𝑚𝑅𝐸𝑀 (2)

The dose acquired calculation if the time is reduced by 20% is shown below:

𝐷𝑜𝑠𝑒 𝐴𝑐𝑞𝑢𝑖𝑟𝑒𝑑 = 0.6𝑚𝑅𝐸𝑀

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.



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) = 𝐼0𝑒−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) = 𝐼0𝑒−Σ𝑡𝑜𝑡𝑡 (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.



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



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.



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.



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.



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.



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



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



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



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.



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.



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.



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.



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.



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.



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.



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



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



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.



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



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

Do

se (

mre

m)

Cask #

Dry Cask Campaign Dose 2012ED = 3292 mrem

Actual = 2293 mrem

ED

Actual



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 = �̇� × 𝑡 (9)

Where D = radiation dose, �̇� = 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:

𝐷2𝑆22 = 𝐷1𝑆1

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:



𝐷 = 𝐷𝑜𝑒(−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.



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



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)



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 cm5 cm

30

40

50

60

70

80

90

100

0 0.2 0.4 0.6 0.8 1 1.2Per

cen

t R

edu

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amm

a D

ose

Cost ($/cm^2)

Percent Reduction of Gamma Dose vs Cost

Tungsten Bismuth Lead Iron



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

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80

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04 0.045Per

cen

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eutr

on

Do

se

Cost ($/cm^2)

Percent Reduction of Neutron Dose vs Cost

Borated Ployethylene Concrete



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



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



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



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



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.



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



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



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)



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 = �̇�𝑒−𝜇𝑡 × 𝑓 × 𝑁 × 𝜏 (19)

Where �̇� = 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

∝ × �̇� ×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

�̇� 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)



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



Figure 17. Cost Optimization Model for Various Shielding Materials

Borotron

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

Co

st

Collective Dose Commitment S

Cost Optimization 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

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4000

6000

8000

10000

12000

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16000

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Cost Optimization of T-Flex Tungsten Shielding




0

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Cost Optimization of T-Flex Bismuth Shielding




0

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Cost Optimization of T-Flex Iron




0

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0.00 0.05 0.10 0.15 0.20

Co

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Collective Dose Commitment S

Cost Optimization of Borated (5%) Polyethylene Shielding






The optimal thicknesses shows the relative effectiveness of the shielding material. To stay within

economic reasonability, the thickness of a given shielding material should not exceed the given

value in Table 17. Tungsten and bismuth T-Flex shielding, although expensive, provide

sufficient shielding when complying with optimal thickness values. Therefore, these types of

shielding must be strategically incorporated in certain places that need enhanced shielding such

as the annulus gap and other high radiation zones discussed in Section 4.6.1. With this economic

model in mind, RoCD will provide final recommendations in section 5.4.

5.4 Recommendations

RoCD’s objective is to provide recommendations for improving dose reduction during dry cask

storage campaigns. These recommendations take into account findings during the conceptual and

preliminary design processes and include new technical analysis and design.

5.4.1 Shielding

A new shielding design is proposed to help reduce dose to workers when manually welding the

vent and drain ports and performing the subsequent dye penetrant tests. This welding evolution is

performed after the annulus has been drained of water and the canister is completely dry, leading

to much higher dose rates at the annulus gap as shown in the survey data in Table 7. Since this is

the highest (unshielded) dose rate that workers are exposed to for a significant amount of time

during dry cask storage campaigns, proposing a new shield designed to mitigate dose during

these steps is the focus.

While gamma dose rates are reduced by a significant amount by the MPC lid, the lid does not

contain sufficient neutron dose. While many NPPs currently employ neutron shielding that

covers the entire MPC lid, our focus is on reducing dose in the area of work near the vent and

drain ports. RoCD’s shield contains borotron in order to reduce neutron dose.

Gamma dose is very high at the annulus gap due to a “streaming” effect: as gamma rays leave

the geometry of the MPC fuel storage cells radially and reach the now empty annulus, the

gamma rays enter a “free path” of air in which the top of the annulus is easily reached. This

streaming effect is accounted for in RoCD’s shield design: T-Flex tungsten is used to reduce

gamma dose, and a significant portion of T-Flex is inserted into the annulus. This allows for the

streaming effect to be mitigated.

Figure 18 shows the MPC lid surrounded by the HI-TRAC with one half of the RoCD shield in

place. The red area is the MPC (including vent and drain ports), the red to yellow region is the

annulus gap, the yellow to green region is the HI-TRAC, and everything in magenta is the

proposed shielding design.



Figure 18. Top-down view of MPC lid with one half of RoCD shielding in place. Note: not to scale.

Figures 19 and 20 are schematics for the RoCD shield. Note that only one half of the total

proposed shielding is shown; its mirror opposite will be placed on the other side of the vent/drain

port.

Overall shielding geometry dimensions chosen are based on MPC-32 lid design. The small 0.37”

protrusion in Figure 20 accounts for the small change in MPC lid height at the 6.25” point

inward from the annulus gap.



Figure 19. Top-down view of RoCD shielding

Figure 20. Side profile of RoCD shielding.



Technical design choices for the RoCD shield are as follows:

Since the annulus gap is ~2 inches inches in width, this portion of the significantly

expensive T-Flex is used in a more economic manner, leading to this design choice:

having the majority of gamma shielding be inserted into the annulus gap (1.5 inches of

thickness in the annulus, 0.9 inches of thickness above the MPC lid)

The majority of shielding above the MPC lid is borotron due to the need for neutron dose

rate reduction at that point vice gamma (as seen in Table 7).

T-Flex Tungsten was chosen primarily due to its ability to be easily sculpted into the

somewhat complex geometry required for the work area of concern, as well as its relative

weight for ease of installation and removal, unlike lead.

One of the two segments of RoCD shielding weights 84.9 lbs. Total weight is 169.8

pounds. It is recommended that the total shielding system be split into a total of four parts

in order to keep each individual segment < 50 lbs, a common workplace lifting limit. This

may be accomplished by splitting each proposed segment in two vertically 10 inches

from the outside edge (relative to the port position).

A high temperature silicone grease, such as DuPont Krytox, is recommended to coat the

portion of the RoCD shield that will enter the annulus in order to ensure ease of

installation and removal.

Thermal limits of the materials should withstand short term heating via conduction during

the welding process. This is assumed due to temperatures having an already small chance

of reaching 212 F at the annulus (the boiling point for water, a concern during lid

welding) due to the port weld being even further inward from the annulus than the lid

edge. A quantitative analysis has not been performed. Temperatures of concern are as

follows:

o T-Flex stability temperature limit = 550 F [33].

o Borotron melting point (crystalline peak) = 260 F (short term limit) [34].

o DuPont Krytox XHT stability temperature limit = 762 F [35].

5.4.1.1 Shielding Economic Evaluation

Using the dimensions of the RoCD shield prototype, an economic analysis is performed. These

dimensions are based in part on being within a reasonable range of the optimal thicknesses found

using the cost-benefit model, as well as the design choice reasons listed in section 5.4.1.

Shielding made of these materials will effectively reduce the on contact dose rate to about 60

mrem/hr cumulative gamma/neutron based on survey data shown in Table 5 and attenuation

information in Figures 21 and 22).



Figure 21. Gamma Attenuation for T-Flex W [33].

Figure 22. Neutron Attenuation for Borotron [34].



To find the acquired dose during welding when utilizing RoCD shielding, it is assumed that it

takes 20 minutes for 2 workers to complete the weld. Calculations for the total amount saved are

shown below. The dose rate of 193 mrem/hr is determined by finding the difference between

RoCD shielding and typical lead blanket shielding. The ALARA factor used is based on D.C.

Cook ($32,632).

193𝑚𝑟𝑒𝑚

ℎ𝑟∗ 2 𝑤𝑜𝑟𝑘𝑒𝑟𝑠 ∗ 20 𝑚𝑖𝑛𝑢𝑡𝑒𝑠 = .129 𝑚𝑎𝑛 𝑟𝑒𝑚

$32632.00 𝑠𝑎𝑣𝑒𝑑

𝑚𝑎𝑛 𝑟𝑒𝑚∗ .129 𝑚𝑎𝑛 𝑟𝑒𝑚 ≈ $4200.00 𝑠𝑎𝑣𝑒𝑑

It is determined that about $4200.00 is saved per cask by utilizing RoCD shielding. The cost of

RoCD shielding is determined to be $8600.00 based on material pricing. The number of casks

until the “break-even” point is found:

$8600.00

𝑠ℎ𝑖𝑒𝑙𝑑 ÷

$4200 𝑠𝑎𝑣𝑒𝑑

𝑐𝑎𝑠𝑘≈ 3 𝑐𝑎𝑠𝑘𝑠 𝑢𝑛𝑡𝑖𝑙 𝑏𝑟𝑒𝑎𝑘 𝑒𝑣𝑒𝑛 𝑝𝑜𝑖𝑛𝑡

The break-even point is defined as how many casks must be loaded before a recommended dose

reduction technique is worthwhile. A break-even point of 3 cask loadings can be accomplished in

one season of dry cask storage campaigns. It can be concluded that RoCD prototype shielding is

a reasonable addition in order to reduce dose to workers.

5.4.2 Time

As discussed in Section 4.4, reducing the overall time of the dry cask loading process is not a

realistic dose reduction method. Based on analysis of the current storage procedure, tasks are

completed at a reasonable rate and there is little room for improvement. It is worth mentioning

that the first cask loading of a campaign takes considerably longer than the rest of the loadings.

There is possible improvement to be made for this but there is little impact on overall dose. The

overall dose reduction with a 10% decrease in overall time is shown below. It is based off of the

last task from Table 6 which had a total dose of 277.4.



Table 18. Overall dose with 10% time reduction.

Regardless of likelihood, reducing the overall time of the storage process by 10% does not

reduce the overall acquired dose considerably. Based on this information, time reduction is not a

recommended method.

5.4.3 Distance

There is no cost effective way to reduce distance during the dry cask campaign besides crowd

control (discussed in the section below). Workers are needed to weld and perform other

operations that are close to the MPC. Remote operations, such as robotics and automation, were

considered too expensive. They should be examined in the future if these solutions become

economical.

5.4.3.1 Distance Economics

To justify the distance recommendation, the following economic analysis is performed:

Assuming all acquired welding dose is eliminated due to this addition, a breakeven analysis is

performed.

$32632 𝑠𝑎𝑣𝑒𝑑

𝑚𝑎𝑛 𝑟𝑒𝑚 ∗ 0.15 𝑟𝑒𝑚 ∗ 2 𝑤𝑜𝑟𝑘𝑒𝑟𝑠 ≈ $10,000 𝑠𝑎𝑣𝑒𝑑

To find the breakeven point, it is assumed that a robotic welding system cost about $1.5 million

to incorporate.

$1.5 𝑚𝑖𝑙𝑙𝑖𝑜𝑛 ÷$10,000

𝑐𝑎𝑠𝑘= 150 𝑐𝑎𝑠𝑘𝑠 𝑢𝑛𝑡𝑖𝑙 𝑏𝑟𝑒𝑎𝑘𝑒𝑣𝑒𝑛 𝑝𝑜𝑖𝑛𝑡

Based on this breakeven point, RoCD does not recommend the incorporation of remote

technologies for welding processes. This analysis does not account for power plants that have

already purchased a robotic welding machine.

Task (Create Schedule) Number of WorkersDuration Person Hrs Eff. Dose Rate Est. mrem

Mobilize Equip./Prep Cask 10 10.8 108 0.02 2.16


Decon/Movement 10 3.6 36 1 36


Transfer MPC to HI STORM 10 9 90 0.75 67.5

Place on ISFSI Pad 5 4.5 22.5 0.4 9

TOTAL 249.66



5.4.4 Crowd Control

Section 3.6 shows the potential for improvements in terms of crowd control. The final

recommendations based on the aforementioned information requires a very organized procedure

that stresses the number of workers required to perform each task. Workers that are not required

for certain tasks need to be clear of the radiation zone until they are needed. This will prevent

unnecessary dose acquisition and reduce overall dose during a storage campaign. Other benefits

of detailed crowd control include clearing up the work area which allows for a faster process.

5.4.5 Fuel Rod Positioning

It has been demonstrated that there is a lower gamma flux outside of the MPC from a radially

decreasing assembly region source term than a uniform assembly region source term, even when

these two differently modeled sources have the same overall source strength.

Similarly, the Serpent MPC model shows that neutron flux reduction is achievable by

strategically arranging fuel assemblies based on burnup.

It is recommended that efforts be made to place fuel assemblies inside of MPCs in a manner that

equates to a radially decreasing source strength when possible. These efforts must be made in

conjunction with the burnup and thermal limitations of the DSC manufacturer that NPPs already

take into account. Further work is required to combine gamma and neutron flux into a single

model in order to make a cohesive fuel assembly loading arrangement decision.

6. Conclusion

Following the technical analysis, a summary of RoCD’s achievements, limitations, and future

activities are listed below.

6.1 Achievements

• Proof of concept for fuel rod repositioning based on gamma and neutron models

• Impact assessed for strategic fuel assembly arrangement based on dose

• Code in place for use: MPC-32 geometry modeled, self-shielding assessed for

neutrons, uniform fuel varied by single parameters via Serpent burnup

calculations.

• Optimal shielding material thicknesses found using optimization cost-benefit

analysis

• Designed prototype neutron and gamma shielding for annulus gap and

surrounding area



• Evaluated the effectiveness of time reduction and remote technologies

6.2 Limitations

• RoCD does not combine neutron and gamma results into cohesive fuel assembly

loading arrangement recommendations

• MPC thermal and burnup limits assumed to be met and were not taken into

account during gamma and neutron flux modeling

• Optimal Serpent neutron population not determined (run time vs accuracy)

• Gamma model does not account for self-shielding or MPC fuel storage cell

materials

• Time constraints prevented a more accurate gamma model

• Economic model has several assumptions based on available data

6.3 Future Activities

• Potential incorporation of recommended dose reduction techniques at NPP

• Consider burnup/enrichment/cooling time/quantity of fuel assemblies collectively

to achieve lowest cumulative neutron and gamma flux outside of MPC

7. References

[1] Nuclear Regulatory Commission. N.p., 24 Mar. 2015. Web. 26 Apr. 2015.

<http://www.nrc.gov/waste/spent-fuel-storage/dry-cask-storage.html>.

[2] Knoll, Glenn F. 2010. Radiation Detection and Measurement Fourth Edition. Hoboken: John

Wiley & Sons, Inc.

[3] International Atomic Energy Agency. Remote Technology Applications in Spent Fuel

Management. International Atomic Energy Agency. Vienna: IAEA, 2005. Web. 19 Feb.

2015.

[4] Hall, Steve. ALARA Aspects of DSC Campaign at Robinson. N.p.: North American ISOE

ALARA Symposium- Regional Meeting, 12 Jan. 2015. PPT.

[5] "Borated Polyethylene | Radiation, Nuclear, Neutron Shielding | NELCO."Nelco. N.p., n.d.

Web. 20 Feb. 2015.



[6] Nuclear Regulatory Commission (2015, February 19). 10 CFR Part 20, Subpart C—

Occupational Dose Limits [Online]. Available: http://www.nrc.gov/reading-rm/doc-

collections/cfr/part020/part020-1201.html.

[7] Nuclear Regulatory Commission (2015, February 19). 10 CFR Part 20, Subpart G – Control

of Exposure from External Sources in Restricted Areas [Online]. Available:

http://www.nrc.gov/reading-rm/doc-collections/cfr/part020/part020-1601.html.

[8] Cassidy, John. Radiation Protection Documentation. Rep. Bridgman: Donald C. Cook

Nuclear Plant, 1995. Print.]

[9] Optimization Strategies for Cask Design and Container Loading in Long Term Spent Fuel

Storage. Vienna: IAEA, 2006. Web. 20 Feb. 2015.

[10] Information System on Occupational Exposure. Rep. N.p.: North American Technical

Center, 2013. Print.

[11] Dry Cask Operations. Rep. Lisle: NPO- A Brand of Eichrom Technologies, 2014. Print.

[12] Lento, Mario. "Automated Welding for the Nuclear Power Industry." Docsrush. Berkeley

Process Control Inc, 2001. Web. 19Feb. 2015.

<http://www.docsrush.net/1590147/automated-welding-article-v1-moog-semiconductor-

wafer-handling.html>.

[13] Nuclear Regulatory Commission (2015, March 10). Spent Fuel Storage FAQ [Online].

Available: http://www.nrc.gov/waste/spent-fuel-storage/faqs.html

[14] Nelson, G., and D. Reilly. "Gamma-Ray Interactions with Matter." N.p.: n.p., n.d. N. pag.

Matter and Interactions. Los Alamos National Lab. Web. 17 Mar. 2015.

<http://www.lanl.gov/orgs/n/n1/panda/00326397.pdf>.

[15] Rinard, P. "Neutron Interactions with Matter." N.p.: n.p., n.d. N. pag. Matter and

Interactions. Federation of American Scientists. Web. 17 Mar. 2015.

<https://fas.org/sgp/othergov/doe/lanl/lib-www/la-pubs/00326407.pdf>.

[16] “Subpart C--Occupational Dose Limits.” NRC: 10 CFR 20.1201 Occupational Dose Limits

for Adults. N.p., n.d. Web. 20 Mar. 2015. http://www.nrc.gov/reading-rm/doc-

collections/cfr/part020/part020-1201.html

[17] “Terminology.” Leonardo 8.1 (1975): 67-69. Web.

[18] “External and Internal Radiation Exposures.” EHS. N.p., n.d. Web. 20 Mar. 2015.

<http://www.orcbs.msu.edu/radiation/programs_guidelines/radmanual/16rm_exposure.ht

m>.



[19] “External and Internal Dose Calculation.” Center For Applied Physical Sciences Research

Institute King Fahd University of Petroleum and Minerals (n.d.): 1-34. Web.

<http://faculty.kfupm.edu.sa/PHYS/maalej/Maalej%20Web%20documents/Presentations/

External%20and%20Internal%20Dose%20Calculation.pdf>.

[20] "Alphabetical Elements List." Chemicool. N.p., n.d. Web. 26 Apr. 2015.

http://www.chemicool.com/elements

[21] Cember, Herman, Thomas E. Johnson, and Parham Alaei. “Introduction to Health Physics:

Fourth Edition.” Medical Physics 35.12 (2008): 571-74. Web. 01 May 2015.

<http://faculty.mu.edu.sa/public/uploads/1380476387.5015Health-Physics-by-Herman-

Cember.pdf>.

[22] Nuclear Regulatory Commission (2015, February 19). 10 CFR Part 72.106—Controlled

Area of an ISFSI or MRS [Online]. Available: http://www.nrc.gov/reading-rm/doc-

collections/cfr/part020/part020-1201.html.

[23] Ewing, Rodney C., and Allison Macfarlane. "Nuclear Waste Yucca Mountain." Science

AAAS. N.p., 26 Apr. 2002. Web. 19 Feb.2015.

<https://www.sciencemag.org/content/296/5568/659.full>.

[24] International Atomic Energy Agency. Remote Technology in Spent Fuel Management.

International Atomic Energy Agency. Vienna: IAEA, 1997. Web. 19 Feb. 2015

[25] Operation and Maintenance of Spent Fuel Storage and Transportation Casks/Containers.

Rep. no. IAEA-TECDOC-1532. N.p.: n.p., n.d. Operation and Maintenance of Spent

Fuel Storage and Transportation Casks/Containers. International Atomic Energy

Agency, Jan. 2007. Web.

[26] Ragheb, Magdi. 2015. Nuclear Power Engineering. February 18. Accessed February 19,

2015.

http://mragheb.com/NPRE%20402%20ME%20405%20Nuclear%20Power%20Engineeri

ng/index.htm.

[27] Singleterry, Robert C., Jr., and Sheila A. Thibeault. Materials for Low-Energy Neutron

Radiation Shielding. Tech. no. NASA/TP-2000-210281. Hampton: Langley Research

Center, 2000. Print.

[28] “US Inflation Calculator.” US Inflation Calculator. N.p., n.d. Web. 20 Feb. 2015.

<http://www.usinflationcalculator.com/inflation/consumer-price-index-and-annual-

percent-changes-from-1913-to-2008/>. “US Inflation Calculator.” US Inflation

Calculator. N.p., n.d. Web. 20 Feb. 2015



[29] Costing of Spent Nuclear Storage. Vienna: International Atomic Energy Agency, 2008. 12-

14. Print.

[30] Westinghouse Electric Company LLC. Generation of ORIGEN-ARP Cross Section

Libraries for 17x17 Fuel Assemblies at the D.C. Cook Unit 2 Nuclear Power Plant. 2011.

Print.

[31] Holtec International. MPC-32 Fuel Basket Layout and Weld Details. 2009. Print

[32] Holtec International. Dose Versus Distance from a HI-STORM 100S Version B Containing

the MPC-32. 2009. Print

[33] “Gamma Shielding Attenuation Guide.” Npo, A Brand of Eichrom Technologies.

http://www.eichrom.com/npo/productliterature. Web. Apr. 2015.

[34] “Neutron Shielding.” Npo, A Brand of Eichrom Technologies.

http://www.eichrom.com/npo/productliterature. Web. Apr. 2015.

[35] “DuPont™ Krytox® XHT-BD, XHT-BDX, and XHT-BDZ Performance Lubricants.”

DuPont. http://www2.dupont.com/Lubricants/pt_BR/assets/downloads/H-91815-

2_Krytox_XHT_BDXZ.pdf. Web. Apr. 2015.

Appendix A: Ethics Exercise

The following organizations were assessed as models for RoCD’s ethics statement: Boeing,

Exelon, the Food & Drug Administration, and the University of California. Ethics statement

excerpts from their respective websites can be seen in the following diagram, along with their

core values.



Appendix B: Serpent MPC Model Code

%-- This is one example of a MPC

Model Code input code run for RoCD.

This correlates with MPC Loading

Pattern MPC 5 seen in Figure 16 of

this report --%

%-- MPC-32, fuel assemblies varied

by burnup, 2-D infinite cyl --%

%-- Westinghouse Standard 17x17

assembly used as input --%

%-- Pin #'s in the 1s. Surface #'s

in the 1000s. Universe #'s in the

100s. Lattice #'s in the 110s. Cell

#'s in the 100s (and coincide w/

universe #'s). gcu (universe 0)

will be highest universe (i.e.

overall MPC geometry) --%

%-- high fuel pin --%

pin 1

high 0.4096

void 0.418

zircaloy 0.475

air

%-- guide tube --%

pin 2

void 0.572

zircaloy 0.612

air

%-- low fuel pin --%

pin 3

low 0.4096

void 0.418

zircaloy 0.475

air

%-- mid fuel pin --%

pin 4

mid 0.4096

void 0.418

zircaloy 0.475

air

%-- 17 x 17 fresh fuel assembly

lattice: universe 110 --%

lat 110 1 0 0 17 17 1.26

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 2 1 1 2 1 1 2 1 1 1 1 1

1 1 1 2 1 1 1 1 1 1 1 1 1 2 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 2 1 1 2 1 1 2 1 1 2 1 1 2 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 2 1 1 2 1 1 2 1 1 2 1 1 2 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 2 1 1 2 1 1 2 1 1 2 1 1 2 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 2 1 1 1 1 1 1 1 1 1 2 1 1 1

1 1 1 1 1 2 1 1 2 1 1 2 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

%-- 17 x 17 next fuel assembly


lat 120 1 0 0 17 17 1.26

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

3 3 3 3 3 2 3 3 2 3 3 2 3 3 3 3 3

3 3 3 2 3 3 3 3 3 3 3 3 3 2 3 3 3

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

3 3 2 3 3 2 3 3 2 3 3 2 3 3 2 3 3

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

3 3 2 3 3 2 3 3 2 3 3 2 3 3 2 3 3

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

3 3 2 3 3 2 3 3 2 3 3 2 3 3 2 3 3

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

3 3 3 2 3 3 3 3 3 3 3 3 3 2 3 3 3

3 3 3 3 3 2 3 3 2 3 3 2 3 3 3 3 3

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3

%-- 17 x 17 fresh fuel assembly


lat 130 1 0 0 17 17 1.26

4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4

4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4

4 4 4 4 4 2 4 4 2 4 4 2 4 4 4 4 4

4 4 4 2 4 4 4 4 4 4 4 4 4 2 4 4 4

4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4

4 4 2 4 4 2 4 4 2 4 4 2 4 4 2 4 4

4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4

4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4

4 4 2 4 4 2 4 4 2 4 4 2 4 4 2 4 4

4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4

4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4

4 4 2 4 4 2 4 4 2 4 4 2 4 4 2 4 4

4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4

4 4 4 2 4 4 4 4 4 4 4 4 4 2 4 4 4

4 4 4 4 4 2 4 4 2 4 4 2 4 4 4 4 4



4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4

4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4

% -- high fuel assy: universe 100 -

-%

surf 1000 sqc 0 0 10.71 %--- assy -

-%

surf 1001 sqc 0 0 11.1633 %-- gap

between assy and MPC slot --%

surf 1002 sqc 0 0 11.3538 %-- boral

--%

surf 1003 sqc 0 0 11.7277 %--

stainless steel --%

cell 100 100 fill 110 -1000

cell 101 100 air 1000 -1001

cell 102 100 boral 1001 -1002

cell 103 100 steel 1002 -1003

cell 104 100 steel 1003

% -- low fuel assy: universe 300 --

%

surf 3000 sqc 0 0 10.71 %--- assy -

-%

surf 3001 sqc 0 0 11.1633 %-- gap


surf 3002 sqc 0 0 11.3538 %-- boral

--%

surf 3003 sqc 0 0 11.7277 %--

stainless steel --%

cell 300 300 fill 120 -3000

cell 301 300 air 3000 -3001

cell 302 300 boral 3001 -3002

cell 303 300 steel 3002 -3003

cell 304 300 steel 3003

% -- mid fuel assy: universe 400 --

%

surf 4000 sqc 0 0 10.71 %--- assy -

-%

surf 4001 sqc 0 0 11.1633 %-- gap


surf 4002 sqc 0 0 11.3538 %-- boral

--%

surf 4003 sqc 0 0 11.7277 %--

stainless steel --%

cell 400 400 fill 130 -3000

cell 401 400 air 3000 -3001

cell 402 400 boral 3001 -3002

cell 403 400 steel 3002 -3003

cell 404 400 steel 3003

%-- "empty" corner space for MPC-32


surf 2000 sqc 0 0 11.7277

cell 200 200 air -2000

cell 201 200 air 2000

%-- MPC-32 geometry: global

universe 0 --%

surf 9000 sqc 0 0 70.3662

cell 900 0 fill 111 -9000

cell 901 0 outside 9000

%-- MPC-32 lattice: universe 111--%

lat 111 1 0 0 6 6 23.4554

200 300 300 300 300 200

300 400 400 400 400 300

300 400 100 100 400 300

300 400 100 100 400 300

300 400 400 400 400 300

200 300 300 300 300 200

%-- external source calculation --%

%-- fuel, 3.6% initial enrichment,

40.0 MWd/kgU burnup, 10 year

cooling time --%

mat high -1.09677E+01

8016.09c -1.18518E-01

92235.09c -2.33706E-03

92238.09c -8.33555E-01

7014.09c -1.13479E-16

7015.09c -2.17570E-09

8017.09c -9.81491E-08

90234.09c -3.09726E-57

92234.09c -5.85142E-06

93235.09c -6.91480E-13

92236.09c -4.33731E-03

93236.09c -2.32047E-10

94236.09c -4.92966E-13

92237.09c -9.79511E-12

93237.09c -2.16354E-04

94237.09c -3.18276E-35

93238.09c -1.59438E-14

94238.09c -6.25759E-05

93239.09c -4.50722E-11

94239.09c -1.95106E-03

94240.09c -1.89970E-03

96240.09c -2.81294E-15

92241.09c -2.03650E-11

94241.09c -3.23183E-04

95241.09c -2.07996E-04

96241.09c -8.21618E-47

94242.09c -5.06055E-04



95242.09c -1.13081E-12

95342.09c -8.76382E-08

96242.09c -2.30283E-10

31069.09c -4.63965E-15

32070.09c -1.09135E-16

31071.09c -4.09411E-12

32072.09c -6.31706E-09

32073.09c -1.67677E-08

32074.09c -4.94514E-08

33074.09c -5.18991E-14

33075.09c -1.31118E-07

32076.09c -3.80764E-07

34076.09c -2.48638E-09

34077.09c -8.47965E-07

34078.09c -2.56102E-06

34079.09c -4.67709E-06

35079.09c -1.13139E-09

34080.09c -1.46337E-05

36080.09c -5.74221E-11

35081.09c -2.28389E-05

34082.09c -3.66383E-05

36082.09c -3.79467E-07

36083.09c -3.48330E-05

36084.09c -1.22332E-04

38084.09c -3.87368E-12

36085.09c -1.01882E-05

37085.09c -8.15097E-05

36086.09c -1.88171E-04

37086.09c -1.30755E-67

38086.09c -1.80000E-07

37087.09c -2.73649E-04

38087.09c -1.83265E-09

38088.09c -3.83678E-04

38089.09c -4.88999E-27

39089.09c -5.17985E-04

38090.09c -4.91035E-04

39090.09c -1.27588E-07

40090.09c -1.62092E-04

39091.09c -7.14643E-24

40091.09c -6.73164E-04

40092.09c -7.15298E-04

40093.09c -7.69818E-04

41093.09c -8.57847E-10

40094.09c -8.53064E-04

41094.09c -5.48367E-10

42094.09c -1.41968E-12

40095.09c -4.80332E-22

41095.09c -5.78213E-22

42095.09c -8.38111E-04

40096.09c -8.71281E-04

42096.09c -3.44949E-05

42097.09c -8.57533E-04

42098.09c -8.69160E-04

44098.09c -8.72053E-15

43099.09c -8.60876E-04

44099.09c -6.17344E-08

42100.09c -9.85112E-04

44100.09c -7.66326E-05

44101.09c -8.28744E-04

44102.09c -7.86925E-04

44103.09c -4.89907E-33

45103.09c -4.43829E-04

44104.09c -5.15409E-04

46104.09c -2.28828E-04

46105.09c -3.38798E-04

44106.09c -1.68353E-07

46106.09c -3.22014E-04

46107.09c -1.87320E-04

47107.09c -2.17312E-10

46108.09c -1.20302E-04

48108.09c -1.66743E-10

47109.09c -6.62641E-05

46110.09c -3.97394E-05

47310.09c -4.46017E-10

48110.09c -1.96544E-05

47111.09c -1.58699E-154

48111.09c -1.96352E-05

48112.09c -9.86539E-06

48113.09c -3.41326E-08

49113.09c -4.53451E-08

48114.09c -1.14520E-05

50114.09c -2.16650E-10

48315.09c -3.32444E-33

49115.09c -1.76837E-06

50115.09c -1.73321E-07

48116.09c -4.32891E-06

50116.09c -2.09786E-06

50117.09c -4.01016E-06

50118.09c -3.43470E-06

50119.09c -3.57076E-06

50120.09c -3.55109E-06

52120.09c -1.08932E-16

51121.09c -3.58648E-06

50122.09c -4.61456E-06

52122.09c -1.29674E-07

50123.09c -5.96680E-16

51123.09c -4.62190E-06

52123.09c -7.32680E-10

50124.09c -7.85915E-06

51124.09c -1.20922E-26

52124.09c -1.02099E-07

50125.09c -9.03990E-122

51125.09c -6.27182E-07

52125.09c -9.89706E-06

50126.09c -1.79067E-05

51126.09c -8.51108E-13

52126.09c -5.31144E-07

54126.09c -7.51210E-11

52327.09c -1.38336E-16

53127.09c -4.65086E-05



52128.09c -8.74363E-05

54128.09c -1.37307E-06

52329.09c -3.29247E-39

53129.09c -1.43924E-04

54129.09c -1.15521E-08

52130.09c -3.88039E-04

54130.09c -9.90812E-06

53131.09c -1.65408E-142

54131.09c -4.64204E-04

54132.09c -1.10633E-03

56132.09c -4.45849E-16

54133.09c -3.17343E-215

55133.09c -1.21986E-03

56133.09c -2.34749E-11

54134.09c -1.61059E-03

55134.09c -3.59394E-06

56134.09c -1.29104E-04

55135.09c -2.21064E-04

56135.09c -9.23090E-08

54136.09c -2.54956E-03

55136.09c -1.66157E-90

56136.09c -1.25144E-05

55137.09c -1.04416E-03

56137.09c -3.13314E-04

56138.09c -1.39696E-03

57138.09c -6.68626E-09

58138.09c -4.54800E-13

57139.09c -1.30479E-03

58139.09c -2.84676E-12

56140.09c -1.41913E-91

58140.09c -1.31646E-03

58141.09c -8.09720E-39

59141.09c -1.18583E-03

58142.09c -1.20704E-03

60142.09c -3.40172E-05

59143.09c -2.31432E-86

60143.09c -5.81532E-04

58144.09c -5.45271E-08

60144.09c -1.72657E-03

60145.09c -7.11695E-04

60146.09c -7.62804E-04

60147.09c -5.80611E-106

61147.09c -1.46447E-05

62147.09c -2.77373E-04

60148.09c -3.94968E-04

61148.09c -1.66344E-142

61348.09c -1.52606E-33

62148.09c -1.04670E-04

62149.09c -2.09410E-06

60150.09c -1.77249E-04

62150.09c -2.84340E-04

62151.09c -4.79889E-06

63151.09c -3.85939E-07

62152.09c -1.45752E-04

63152.09c -1.01892E-09

64152.09c -2.10657E-10

63153.09c -1.06975E-04

64153.09c -4.72281E-17

62154.09c -3.29978E-05

63154.09c -4.68637E-06



cooling time --%

mat low -1.09693E+01

8016.09c -1.18507E-01

92235.09c -2.18693E-02

92238.09c -8.47171E-01

7014.09c -2.56678E-18

7015.09c -3.54950E-10

8017.09c -1.40142E-08

90234.09c -3.14611E-57

92234.09c -3.92384E-07

93235.09c -1.42962E-14

92236.09c -1.54100E-03

93236.09c -1.91169E-11

94236.09c -2.03831E-15

92237.09c -8.55355E-13

93237.09c -2.48596E-05

94237.09c -6.02205E-37

93238.09c -4.39170E-16

94238.09c -1.24086E-06

93239.09c -5.14575E-14

94239.09c -1.46451E-03

94240.09c -2.83777E-04

92241.09c -1.61670E-12

94241.09c -2.82218E-05

95241.09c -1.77239E-05

96241.09c -1.59050E-49

94242.09c -3.75284E-06

95242.09c -3.11481E-14

95342.09c -2.41398E-09

96242.09c -6.31363E-12

31069.09c -3.27488E-16

32070.09c -5.52644E-18

31071.09c -2.91729E-13

32072.09c -1.05931E-09

32073.09c -3.55971E-09

32074.09c -1.09561E-08

33074.09c -4.03825E-15

33075.09c -3.30555E-08

32076.09c -1.00154E-07

34076.09c -1.07421E-10

34077.09c -2.48031E-07

34078.09c -6.53586E-07

34079.09c -1.37438E-06

35079.09c -2.98980E-10

34080.09c -3.90269E-06

36080.09c -6.78695E-12

35081.09c -6.17883E-06



34082.09c -1.01052E-05

36082.09c -1.80071E-08

36083.09c -1.56493E-05

36084.09c -2.91252E-05

38084.09c -1.58149E-13

36085.09c -3.19381E-06

37085.09c -2.34708E-05

36086.09c -5.68227E-05

37086.09c -1.50871E-68

38086.09c -7.33597E-09

37087.09c -8.10147E-05

38087.09c -1.11570E-10

38088.09c -1.14346E-04

38089.09c -7.86020E-27

39089.09c -1.55244E-04

38090.09c -1.50980E-04

39090.09c -3.92299E-08

40090.09c -4.37782E-05

39091.09c -1.04010E-23

40091.09c -1.97580E-04

40092.09c -2.05084E-04

40093.09c -2.17709E-04

41093.09c -1.97635E-10

40094.09c -2.33616E-04

41094.09c -4.75297E-11

42094.09c -1.18111E-13

40095.09c -5.55031E-22

41095.09c -6.68133E-22

42095.09c -2.35639E-04

40096.09c -2.33383E-04

42096.09c -5.99458E-07

42097.09c -2.24556E-04

42098.09c -2.19924E-04

44098.09c -3.86385E-16

43099.09c -2.34130E-04

44099.09c -1.56185E-08

42100.09c -2.46725E-04

44100.09c -2.89379E-06

44101.09c -2.05884E-04

44102.09c -1.78420E-04

44103.09c -3.25942E-33

45103.09c -1.29444E-04

44104.09c -9.09927E-05

46104.09c -6.13746E-06

46105.09c -5.31990E-05

44106.09c -2.72414E-08

46106.09c -3.28994E-05

46107.09c -1.66658E-05

47107.09c -1.82588E-11

46108.09c -9.05086E-06

48108.09c -2.23120E-12

47109.09c -5.64059E-06

46110.09c -3.11443E-06

47310.09c -8.29270E-12

48110.09c -2.64669E-07

47111.09c -3.39652E-155

48111.09c -1.67568E-06

48112.09c -1.00453E-06

48113.09c -3.23564E-08

49113.09c -5.77480E-09

48114.09c -1.57628E-06

50114.09c -1.60923E-12

48315.09c -1.70892E-33

49115.09c -5.90875E-07

50115.09c -3.35755E-08

48116.09c -7.64324E-07

50116.09c -8.46142E-08

50117.09c -7.33808E-07

50118.09c -6.49136E-07

50119.09c -7.12865E-07

50120.09c -7.03065E-07

52120.09c -1.36060E-18

51121.09c -7.07466E-07

50122.09c -8.88737E-07

52122.09c -4.11542E-09

50123.09c -1.74786E-16

51123.09c -9.07918E-07

52123.09c -1.01314E-11

50124.09c -1.53024E-06

51124.09c -9.46308E-28

52124.09c -3.93487E-09

50125.09c -4.44831E-122

51125.09c -1.44240E-07

52125.09c -1.82739E-06

50126.09c -3.26767E-06

51126.09c -1.55313E-13

52126.09c -1.16928E-07

54126.09c -2.07227E-12

52327.09c -7.06335E-17

53127.09c -9.06093E-06

52128.09c -1.87801E-05

54128.09c -3.76585E-08

52329.09c -2.26410E-39

53129.09c -3.07722E-05

54129.09c -4.92550E-11

52130.09c -9.36240E-05

54130.09c -2.76023E-07

53131.09c -1.60302E-142

54131.09c -1.45399E-04

54132.09c -2.33191E-04

56132.09c -1.52616E-17

54133.09c -3.47642E-215

55133.09c -3.41044E-04

56133.09c -4.50640E-14

54134.09c -4.09290E-04

55134.09c -1.80177E-07

56134.09c -5.38757E-06

55135.09c -7.15632E-05

56135.09c -4.16833E-10

54136.09c -6.05133E-04



55136.09c -3.41009E-91

56136.09c -8.99591E-07

55137.09c -2.67416E-04

56137.09c -7.21546E-05

56138.09c -3.60432E-04

57138.09c -1.71563E-09

58138.09c -5.06457E-14

57139.09c -3.45540E-04

58139.09c -3.64402E-13

56140.09c -1.66997E-91

58140.09c -3.36110E-04

58141.09c -9.23127E-39

59141.09c -3.16918E-04

58142.09c -3.17503E-04

60142.09c -8.47937E-07

59143.09c -2.96223E-86

60143.09c -2.98050E-04

58144.09c -3.11957E-08

60144.09c -3.26457E-04

60145.09c -2.14392E-04

60146.09c -1.73164E-04

60147.09c -6.73083E-106

61147.09c -7.69227E-06

62147.09c -1.08894E-04

60148.09c -9.87779E-05

61148.09c -8.64205E-143

61348.09c -7.92834E-34

62148.09c -5.63950E-06

62149.09c -2.09424E-06

60150.09c -4.02907E-05

62150.09c -6.44229E-05

62151.09c -4.90912E-06

63151.09c -3.98539E-07

62152.09c -3.63987E-05

63152.09c -2.79377E-09

64152.09c -5.50662E-10

63153.09c -1.35773E-05

64153.09c -3.40941E-17

62154.09c -5.47011E-06

63154.09c -3.61842E-07

64154.09c -4.63471E-07

63155.09c -2.40506E-07

64155.09c -8.20962E-07

63156.09c -2.20360E-79

64156.09c -3.59692E-06

64157.09c -1.70403E-08

64158.09c -1.21133E-06

66158.09c -3.73738E-15

65159.09c -1.63986E-07

64160.09c -6.74845E-08

65160.09c -8.58857E-25

66160.09c -2.39005E-09

66161.09c -2.53947E-08

66162.09c -1.37115E-08

66163.09c -5.69393E-09

66164.09c -1.39489E-09

67165.09c -1.63219E-09

67366.09c -2.07702E-13

68166.09c -1.25201E-10

68167.09c -5.17485E-13

68168.09c -1.53431E-13

68170.09c -1.32620E-13

95243.09c -5.98039E-08

96243.09c -6.01823E-11

94244.09c -5.11444E-12

96244.09c -9.66720E-10

96245.09c -6.28974E-12

96246.09c -1.27735E-13

96247.09c -8.02435E-17



cooling time --%

mat mid -1.09686E+01

8016.09c -1.18512E-01

92235.09c -1.34474E-02

92238.09c -8.43965E-01

7014.09c -1.44582E-17

7015.09c -7.91358E-10

8017.09c -3.23163E-08

90234.09c -3.13643E-57

92234.09c -1.17373E-06

93235.09c -8.62158E-14

92236.09c -2.81691E-03

93236.09c -7.58057E-11

94236.09c -2.54082E-14

92237.09c -3.89901E-12

93237.09c -7.29995E-05

94237.09c -3.21267E-36

93238.09c -4.91871E-15

94238.09c -8.32502E-06

93239.09c -1.54944E-12

94239.09c -1.94044E-03

94240.09c -8.32198E-04

96240.09c -5.75447E-17

92241.09c -4.12652E-12

94241.09c -1.28645E-04

95241.09c -8.22535E-05

96241.09c -4.61792E-48

94242.09c -4.70435E-05

95242.09c -3.48860E-13

95342.09c -2.70366E-08

96242.09c -7.07630E-11

31069.09c -1.19378E-15

32070.09c -2.64819E-17

31071.09c -1.03208E-12

32072.09c -2.44712E-09

32073.09c -7.58193E-09

32074.09c -2.28797E-08

33074.09c -1.42195E-14



33075.09c -6.64549E-08

32076.09c -1.98418E-07

34076.09c -4.67266E-10

34077.09c -4.81072E-07

34078.09c -1.30057E-06

34079.09c -2.66566E-06

35079.09c -6.00330E-10

34080.09c -7.67625E-06

36080.09c -1.81116E-11

35081.09c -1.21581E-05

34082.09c -1.97199E-05

36082.09c -7.35978E-08

36083.09c -2.76781E-05

36084.09c -5.88670E-05

38084.09c -7.08098E-13

36085.09c -5.99544E-06

37085.09c -4.52517E-05

36086.09c -1.08112E-04

37086.09c -3.68879E-68

38086.09c -3.22303E-08

37087.09c -1.54976E-04

38087.09c -3.60104E-10

38088.09c -2.18372E-04

38089.09c -7.22929E-27

39089.09c -2.96116E-04

38090.09c -2.85608E-04

39090.09c -7.42111E-08

40090.09c -8.62463E-05

39091.09c -9.90448E-24

40091.09c -3.78824E-04

40092.09c -3.95555E-04

40093.09c -4.21785E-04

41093.09c -4.09706E-10

40094.09c -4.56074E-04

41094.09c -1.56439E-10

42094.09c -3.93295E-13

40095.09c -5.60215E-22

41095.09c -6.74373E-22

42095.09c -4.58476E-04

40096.09c -4.58417E-04

42096.09c -4.38573E-06

42097.09c -4.44167E-04

42098.09c -4.38945E-04

44098.09c -1.63206E-15

43099.09c -4.59968E-04

44099.09c -3.13531E-08

42100.09c -4.93986E-04

44100.09c -1.31802E-05

44101.09c -4.13258E-04

44102.09c -3.67059E-04

44103.09c -3.63559E-33

45103.09c -2.55254E-04

44104.09c -2.03603E-04

46104.09c -3.51276E-05

46105.09c -1.26761E-04

44106.09c -6.27649E-08

46106.09c -9.03818E-05

46107.09c -5.03092E-05

47107.09c -5.64198E-11

46108.09c -2.97310E-05

48108.09c -1.60283E-11

47109.09c -1.82449E-05

46110.09c -9.85593E-06

47310.09c -5.16247E-11

48110.09c -1.90244E-06

47111.09c -5.59537E-155

48111.09c -4.98719E-06

48112.09c -2.73906E-06

48113.09c -3.43443E-08

49113.09c -1.45096E-08

48114.09c -3.86875E-06

50114.09c -1.49141E-11

48315.09c -2.10503E-33

49115.09c -1.12874E-06

50115.09c -7.38103E-08

48116.09c -1.72505E-06

50116.09c -3.89037E-07

50117.09c -1.63508E-06

50118.09c -1.42134E-06

50119.09c -1.53430E-06

50120.09c -1.51323E-06

52120.09c -2.09468E-17

51121.09c -1.53410E-06

50122.09c -1.94069E-06

52122.09c -1.98314E-08

50123.09c -3.10034E-16

51123.09c -1.97128E-06

52123.09c -6.90518E-11

50124.09c -3.34074E-06

51124.09c -2.71769E-27

52124.09c -1.64883E-08

50125.09c -5.49226E-122

51125.09c -2.96211E-07

52125.09c -4.07080E-06

50126.09c -7.31325E-06

51126.09c -3.47600E-13

52126.09c -2.43086E-07

54126.09c -1.09093E-11

52327.09c -9.78574E-17

53127.09c -1.98728E-05

52128.09c -3.96069E-05

54128.09c -1.95550E-07

52329.09c -2.50598E-39

53129.09c -6.54261E-05

54129.09c -5.31076E-10

52130.09c -1.89761E-04

54130.09c -1.37434E-06

53131.09c -1.54793E-142

54131.09c -2.78310E-04

54132.09c -4.88822E-04



56132.09c -6.67042E-17

54133.09c -3.23897E-215

55133.09c -6.66228E-04

56133.09c -8.47279E-13

54134.09c -8.15388E-04

55134.09c -7.68033E-07

56134.09c -2.46267E-05

55135.09c -1.29790E-04

56135.09c -3.63090E-09

54136.09c -1.23242E-03

55136.09c -6.11503E-91

56136.09c -2.99071E-06

55137.09c -5.31038E-04

56137.09c -1.48657E-04

56138.09c -7.14846E-04

57138.09c -3.48775E-09

58138.09c -1.42948E-13

57139.09c -6.81323E-04

58139.09c -1.16577E-12

56140.09c -1.54016E-91

58140.09c -6.67420E-04

58141.09c -8.79270E-39

59141.09c -6.24173E-04

58142.09c -6.26815E-04

60142.09c -4.81673E-06

59143.09c -2.68628E-86

60143.09c -5.13268E-04

58144.09c -4.66404E-08

60144.09c -7.09541E-04

60145.09c -4.10526E-04

60146.09c -3.53168E-04

60147.09c -6.21282E-106

61147.09c -1.25579E-05

62147.09c -1.95557E-04

60148.09c -1.97483E-04

61148.09c -1.47288E-142

61348.09c -1.35124E-33

62148.09c -2.36149E-05

62149.09c -2.02052E-06

60150.09c -8.28311E-05

62150.09c -1.36423E-04

62151.09c -4.80422E-06

63151.09c -3.88943E-07

62152.09c -7.60341E-05

63152.09c -2.41201E-09

64152.09c -5.00275E-10

63153.09c -3.63984E-05

64153.09c -9.58255E-17

62154.09c -1.24310E-05

63154.09c -1.34426E-06

64154.09c -1.77676E-06

63155.09c -5.28540E-07

64155.09c -1.80209E-06

63156.09c -5.07933E-79

64156.09c -1.30494E-05

64157.09c -2.46289E-08

64158.09c -3.61861E-06

66158.09c -1.38283E-14

65159.09c -5.01720E-07

64160.09c -2.17222E-07

65160.09c -3.86087E-24

66160.09c -1.48507E-08

66161.09c -7.44445E-08

66162.09c -5.51607E-08

66163.09c -2.39530E-08

66164.09c -3.78207E-09

68164.09c -1.86267E-17

67165.09c -8.10008E-09

67366.09c -3.20037E-13

68166.09c -7.30381E-10

68167.09c -5.39372E-12

68168.09c -2.85929E-12

68170.09c -4.94002E-13

95243.09c -1.80076E-06

96243.09c -2.92039E-09

94244.09c -2.14044E-10

96244.09c -6.89913E-08

96245.09c -8.55599E-10

%-- cladding --%

mat zircaloy -6.55

8016.06c -0.00125

24000.06c -0.001

26000.06c -0.00135

28000.06c -0.00055

40000.06c -0.98135

50000.06c -0.0145

% -- helium. this weight density

could use updating to reflect ~44.7

psi & a higher temp, but cross

section is likely ~irrelevant for

He -- %

mat helium -.00001

2004.06c -1

%-- air --%

mat air -0.00173

7014.06c -0.78

8016.06c -0.22

%-- Boral --%

mat boral -2.644

5010.06c -0.0044226

5011.06c -0.201474

13027.06c -0.6861

6000.06c -0.0682

%-- stainless steel --%



mat steel -7.92

24000.06c -0.19

25055.06c -0.02

26000.06c -0.695

28000.06c -0.095

%-- water, moderator. May be cut if

not used--%

mat water -0.74 moder lwtr 1001

1001.06c -0.666666

8016.06c -0.333334

%-- bc 1 = neutron escapes --%

set bc 1

%-- miscellaneous variables --%

set title "MPC-32 with 17x17 fuel

assemblies"

set seed 416416430

therm lwtr lwe6.12t

set egrid 5E-6 1E-9 15.0

set gcu 0

set pop 5000 250 100

set acelib

"/home/codes/serpent/xsdata/endfb68

/sss_endfb68.xsdata"

set ures 1 [3]

Appendix C: Serpent Burnup Calculation Codes

10 MWd/kgU

calculation :

%-- burnup

calculation --%

set title "3.6%

initial enrichment

Westinghouse Standard

17x17 Pin-cell 10.0

MWd/kgU burnup

calculation + 30

years cooling time"

pin 1 %-- gap ignored

for burnup calc

fuel 0.4096

zircaloy 0.475

water

%-- geometry (based

on fuel pitch) --%

surf 1000 sqc 0 0

0.63

cell 1 0 fill 1 -1000


%-- 4.5% initial

enrichment fuel --%

mat fuel -10.97 burn

1

8016.09c -0.1185

92235.09c -0.03173

92238.09c -0.84977

%-- cladding --%

mat zircaloy -6.56

8016.06c -0.00125

24000.06c -0.001

26000.06c -0.00135

28000.06c -0.00055

40000.06c -0.98135

50000.06c -0.0145

%-- water, moderator

--%

mat water -0.74 moder

lwtr 1001

1001.06c -0.666666

8016.06c -0.333334

%-- thermal

scattering library --

%

therm lwtr lwe6.12t

%-- set xs library

set acelib

"/home/codes/serpent/

xsdata/endfb68/sss_en

dfb68.xsdata"

%-- boundary

condition to simulate

adjacency to other

fuel pins

set bc 3

%-- global universe

setting

set gcu 0

%-- set symmetry of

fuel pin IAW Serpent

def'n

set sym 12

%-- 2 group structure

(group boundary at

0.625 eV)

set nfg 2 0.625E-6

%-- source neutron

population, active

cycles, inactive

cycles

set pop 500 1000 20

%-- set decay and

fission yield

libraries

set nfylib



dfb68.nfy"

set declib



dfb68.dec"

%-- limit the energy

parameters



set egrid 5E-5 1E-9

15.0

%-- cutoff values

set fpcut 1E-9

set stabcut 1E-12

set ttacut 1E-18

set xsfcut 1E-6

%-- extra burnup

calculation options

set bumode 1 %

Uses Transmutation

Trajectory Analysis

to solve Bateman

eqn's

set pcc 1 %

Predictor-corrector

enabled

set xscalc 2 %

Cross sections from

spectrum

%-- power, cooling

time history

set powdens 40.0E-3

dep butot %-- total

cumulative burnup =

10.0 MWd/kg

0.10000

0.20000

0.50000

1.00000

2.00000

3.00000

4.00000

5.00000

6.00000

7.00000

8.00000

9.00000

10.0000

dep decstep %-- 30

years total cooling

time

365

365

365

365

365 % 5 year step

1825 % 10 year step

3650 % 20 year step

3650 % 30 year step

%-- list of isotope

inventory desired in

output

set inventory

922340

922350

922360

922370

922380

922390

932360

932370

932380

932390

942360

942380

942390

942400

942410

942420

942430

952410

952420

952430

952440

952421

962420

962430

962440

962450

962460

962470

962480

962490

972490

972500

982490

982500

982510

982520

360830

360850

380900

400930

430990

441060

451030

451050

471090

511250

501260

531350

541310

541350

551330

551340

551350

551370

561400

571400

601430

601450

611470

611480

611490

611481

621470

621490

621500

621510

621520

631530

631540

631550

631560

641520

641540

641550

641560

641570

641600

%-- gives access to

fuel comp after each

burnup step

set printm 1

40 MWd/kgU

calculation :

%-- burnup

calculation --%

set title "3.6%

initial enrichment


17x17 Pin-cell 40.0

MWd/kgU burnup

calculation + 30

years cooling time"


for burnup calc

fuel 0.4096

zircaloy 0.475

water

%-- geometry (based

on fuel pitch) --%



surf 1000 sqc 0 0

0.63

cell 1 0 fill 1 -1000


%-- 4.5% initial

enrichment fuel --%


1

8016.09c -0.1185

92235.09c -0.03173

92238.09c -0.84977

%-- cladding --%

mat zircaloy -6.56

8016.06c -0.00125

24000.06c -0.001

26000.06c -0.00135

28000.06c -0.00055

40000.06c -0.98135

50000.06c -0.0145


--%


lwtr 1001

1001.06c -0.666666

8016.06c -0.333334

%-- thermal


%

therm lwtr lwe6.12t

%-- set xs library

set acelib



dfb68.xsdata"

%-- boundary


adjacency to other

fuel pins

set bc 3

%-- global universe

setting

set gcu 0

%-- set symmetry of


def'n

set sym 12


(group boundary at

0.625 eV)

set nfg 2 0.625E-6

%-- source neutron

population, active

cycles, inactive

cycles

set pop 500 1000 20

%-- set decay and

fission yield

libraries

set nfylib



dfb68.nfy"

set declib



dfb68.dec"


parameters

set egrid 5E-5 1E-9

15.0

%-- cutoff values

set fpcut 1E-9

set stabcut 1E-12

set ttacut 1E-18

set xsfcut 1E-6

%-- extra burnup

calculation options

set bumode 1 %

Uses Transmutation

Trajectory Analysis

to solve Bateman

eqn's

set pcc 1 %

Predictor-corrector

enabled

set xscalc 2 %

Cross sections from

spectrum

%-- power, cooling

time history

set powdens 40.0E-3

dep butot %-- total

cumulative burnup =

40.0 MWd/kg

0.10000

0.20000

0.50000

1.00000

2.00000

3.00000

4.00000

5.00000

6.00000

7.00000

8.00000

9.00000

10.0000

12.0000

15.0000

20.0000

25.0000

30.0000

35.0000

40.0000

dep decstep %-- 30

years total cooling

time

365

365

365

365

365 % 5 year step

1825 % 10 year step

3650 % 20 year step

3650 % 30 year step

%-- list of isotope


output

set inventory

922340

922350

922360

922370

922380

922390

932360

932370

932380

932390

942360

942380

942390

942400

942410



942420

942430

952410

952420

952430

952440

952421

962420

962430

962440

962450

962460

962470

962480

962490

972490

972500

982490

982500

982510

982520

360830

360850

380900

400930

430990

441060

451030

451050

471090

511250

501260

531350

541310

541350

551330

551340

551350

551370

561400

571400

601430

601450

611470

611480

611490

611481

621470

621490

621500

621510

621520

631530

631540

631550

631560

641520

641540

641550

641560

641570

641600

%-- gives access to


burnup step

set printm 1

20 MWd/kgU

calculation :

%-- burnup

calculation --%

set title "3.6%

initial enrichment


17x17 Pin-cell 20.0

MWd/kgU burnup

calculation + 30

years cooling time"


for burnup calc

fuel 0.4096

zircaloy 0.475

water

%-- geometry (based

on fuel pitch) --%

surf 1000 sqc 0 0

0.63

cell 1 0 fill 1 -1000


%-- 4.5% initial

enrichment fuel --%


1

8016.09c -0.1185

92235.09c -0.03173

92238.09c -0.84977

%-- cladding --%

mat zircaloy -6.56

8016.06c -0.00125

24000.06c -0.001

26000.06c -0.00135

28000.06c -0.00055

40000.06c -0.98135

50000.06c -0.0145


--%


lwtr 1001

1001.06c -0.666666

8016.06c -0.333334

%-- thermal


%

therm lwtr lwe6.12t

%-- set xs library

set acelib



dfb68.xsdata"

%-- boundary


adjacency to other

fuel pins

set bc 3

%-- global universe

setting

set gcu 0

%-- set symmetry of


def'n

set sym 12


(group boundary at

0.625 eV)

set nfg 2 0.625E-6

%-- source neutron

population, active

cycles, inactive

cycles

set pop 500 1000 20

%-- set decay and

fission yield

libraries

set nfylib



dfb68.nfy"

set declib





dfb68.dec"


parameters

set egrid 5E-5 1E-9

15.0

%-- cutoff values

set fpcut 1E-9

set stabcut 1E-12

set ttacut 1E-18

set xsfcut 1E-6

%-- extra burnup

calculation options

set bumode 1 %

Uses Transmutation

Trajectory Analysis

to solve Bateman

eqn's

set pcc 1 %

Predictor-corrector

enabled

set xscalc 2 %

Cross sections from

spectrum

%-- power, cooling

time history

set powdens 40.0E-3

dep butot %-- total

cumulative burnup =

20.0 MWd/kg

0.10000

0.20000

0.50000

1.00000

2.00000

3.00000

4.00000

5.00000

6.00000

7.00000

8.00000

9.00000

10.0000

12.0000

15.0000

20.0000

dep decstep %-- 30

years total cooling

time

365

365

365

365

365 % 5 year step

1825 % 10 year step

3650 % 20 year step

3650 % 30 year step

%-- list of isotope


output

set inventory

922340

922350

922360

922370

922380

922390

932360

932370

932380

932390

942360

942380

942390

942400

942410

942420

942430

952410

952420

952430

952440

952421

962420

962430

962440

962450

962460

962470

962480

962490

972490

972500

982490

982500

982510

982520

360830

360850

380900

400930

430990

441060

451030

451050

471090

511250

501260

531350

541310

541350

551330

551340

551350

551370

561400

571400

601430

601450

611470

611480

611490

611481

621470

621490

621500

621510

621520

631530

631540

631550

631560

641520

641540

641550

641560

641570

641600

%-- gives access to


burnup step

set printm 1

Documents

Final Report submission