The Pennsylvania State University

The Graduate School

College of Engineering

FUEL CYCLE PERFORMANCE OF THERMAL AND

FAST SPECTRUM SMALL MODULAR REACTORS

A Thesis in

Nuclear Engineering

by

Richard Hernandez

Submitted in Partial Fulfillment

of the Requirements

for the Degree of

Master of Science

May 2019

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The thesis of Richard Hernandez was reviewed and approved* by the following:

Nicholas R. Brown

Associate Professor of Nuclear Engineering, University of Tennessee

Thesis Advisor

William J. Walters

Assistant Professor in Nuclear Engineering

Arthur T. Motta

Professor of Nuclear Engineering and Materials Science and Engineering

Chair of Nuclear Engineering

*Signatures are on file in the Graduate School

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ABSTRACT

The work found in this thesis, consisted of the fuel cycle performance analysis of several

thermal and fast spectrum small modular reactor (SMR) designs and concepts. This analysis was

conducted following the guidelines found in the U.S. Department of Energy Office of Nuclear

Energy (DOE-NE) Evaluation and Screening (E&S) study chartered in 2011. The impacts of core

neutron leakage on the fuel cycle performance of these modular reactors, was investigated using

several of the specific performance criteria identified in the E&S study. This work was

accomplished by creating computer models from which neutronics investigations could be

performed.

The first study consisted of the construction of a VVER-1000 fuel assembly neutronics

model. The specifications were taken from an international computational benchmark paper. This

assembly model was used to conduct a fuel cycle investigation of the thermal spectrum VVER-

like SMR designs described in this thesis. The impacts of different neutron leakage rates, fuel

enrichments, and core power densities on the natural resource utilization and the highly radioactive

waste outputs at 100 and 100,000 years after fuel discharged were analyzed. The study showed

that core neutron leakage was the most impacting parameter to the fuel cycle performance of these

thermal SMRs, and that changes in fuel enrichment and core power density had minimal impacts

on both of these criteria studied. The 10% U-235 fuel enrichment case with 0% leakage, had a

natural resource utilization of 314.5 t/GWe-yr. As leakage was increased to 3% and 7%, the

resource utilization dropped to 344.1 and 394.7 t/GWe-yr respectively. Whereas, as fuel

enrichment was increased to 15% and 19.7%, using the 0% leakage case, resource utilization

dropped to 329 and 341 t/GWe-yr respectively. A 50% lower power density resulted in a slight

increase of about 2 t/GWe-yr, when compared to the full power cases. The activity levels of the

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SNF+HLW for all of the cases studied were between 1.2E+06 and 1.4E+06 Ci/GWe-yr at 100

years, and between 1.6E+03 and 1.8E+03 Ci/GWe-yr at 100,000 years.

The second study, involved the fuel cycle analysis of a neutronics computer model based

on the component specifications of the fast spectrum special purpose heat pipe reactor. This

concept was first proposed at Los Alamos National Laboratory (LANL). The LANL concept, was

chosen to represent the Westinghouse eVinciTM micro heat pipe reactor concept currently under

development. This was due to their fundamental similarities, as well as the ambiguity surrounding

the specific component details of the eVinciTM concept found in open source papers. The initial

investigation, consisted of analyzing the fuel cycle performance of a 10-year fuel cycle option

simulated using the built neutronics computer model. This fuel cycle length was chosen, because

Westinghouse indicated that the eVinciTMconcept is intended to operate for ten years without

refueling. A 5MW and 25 MW rated core thermal output was used in this initial fuel cycle

investigation. The natural resource utilization of the 5MW and 25MW were 9612 and 1933 t/GWe-

yr respectively for the most efficient thermal cycle available for this concept. The waste metrics

analyzed following criteria guidelines found in the E&S study, were around 1.00E+06 Ci/GWe-yr

for both rated power cases at 100 years after fuel discharged. At 100,000 years after fuel discharge,

the SNF+HLW activity in the 5MW and 25MW power cases were determined to be 3.37E+03 and

2.37E+03 Ci/GWe-yr respectively.

Following, an investigation into the most optimal fuel cycle length of the special purpose

reactor, with a rated thermal output of 25 MW, was conducted using a subset of the nine high level

criteria found in the E&S study. The environmental impacts, natural resource utilization, waste

management metrics and proliferation risk, as a result of this fuel cycle option, were all analyzed

in this optimization study. The natural resource utilization of the most effective thermal cycle was

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determined to be 1704 t/GWe-yr. The land and water use were calculated to be about 0.784 square

km and 25320 ML per GWe year respectively. Carbon emissions were 247 kilotons of CO2 per

GWe year.

Lastly, the impacts of core neutron leakage and parasitic absorption were quantified in a

third fuel cycle investigation, utilizing the 25 MWth model used in the optimization study. The

study concluded that the core neutron leakage was the most limiting factor to the fuel cycle

performance of these fast spectrum SMR concepts. A reduction in core leakage from the reference

model 18% rate to 9.7, resulted in an increase in natural resource utilization from 1704 to 350

t/GWe respectively.

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TABLE OF CONTENTS

LIST OF FIGURES ..................................................................................................................... VII

LIST OF TABLES ........................................................................................................................ IX

ACKNOWLEDGEMENTS .......................................................................................................... IX

1.INTRODUCTION ....................................................................................................................... 1

1.1 BACKGROUND .................................................................................................................. 1

1.2 THE NUCLEAR FUEL CYCLE EVALUATION AND SCREENING STUDY ................................ 4

1.3 SMR FUEL CYCLE PERFORMANCE STUDIES .................................................................... 6

1.4 VVER-BASED RUSSIAN FLOATING SMR CONCEPTS ....................................................... 8

1.5 FAST SPECTRUM MICRO HEAT PIPE NUCLEAR REACTOR CONCEPTS ............................. 12

2.FUEL CYCLE PERFORMANCE OF THERMAL SPECTRUM VVER-BASED RUSSIAN

SMRS ............................................................................................................................................. 16

2.1 VVER ASSEMBLY MODEL DESCRIPTION ....................................................................... 16

2.2 SERPENT ASSEMBLY MODEL VALIDATION .................................................................... 19

2. SERPENT ACTIVITY METRICS VALIDATION .................................................................... 20

2.4 FUEL CYCLE PERFORMANCE ANALYSIS OF VVER-BASED SMRS ................................. 22

3.FUEL CYCLE PERFORMANCE OF THE FAST SPECTRUM SPECIAL PURPOSE HEAT

PIPE SMR CONCEPT. ................................................................................................................. 31

3.1 SPECIAL PURPOSE REACTOR MODEL DESCRIPTION ....................................................... 31

3.2 SERPENT CODE TO CODE VALIDATION .......................................................................... 33

3.3 SERPENT METRICS VALIDATION ……………………………………………................................32

3.4 10-YEAR FUEL CYCLE PERFORMANCE OF EVINCI-LIKE REACTOR ................................. 35

3.5 ANALYSIS OF OPTIMAL FUEL CYCLE OF THE EVINCI-LIKE REACTOR ............................ 39

3.5.1 Nuclear Waste Management ..................................................................................... 41

3.5.2 Natural Resource Utilization .................................................................................... 42

3.5.3 Proliferation Risk ...................................................................................................... 42

3.5.4 Environmental Impacts ............................................................................................. 42

3.6 IMPACTS OF PARASITIC ABSORPTION ON FUEL CYCLE PERFORMANCE .......................... 43

3.7 IMPACTS OF NEUTRON LEAKAGE ON FUEL CYCLE PERFORMANCE ................................ 47

4.SUMMARY ............................................................................................................................... 52

REFERENCES ............................................................................................................................. 54

VII

LIST OF FIGURES

FIGURE 1-1. MASS FLOW CHART FOR THE EG02 REFERENCE CASE [6]. ............................................ 5

FIGURE 1-2. KLT-40S CONCEPT FLOW DIAGRAM [10]. .................................................................. 10

FIGURE 1-3. RITM-200 INSTALLATION INTO THE RUSSIAN ICEBREAKER ARKTIKA [12]. ............... 11

FIGURE 1-4. ILLUSTRATION OF THE EVINCITM REACTOR CONCEPT HIGHLIGHTING ITS MAIN

COMPONENTS. [15]. ................................................................................................................ 13

FIGURE 1-5. LANL SPECIAL PURPOSE REACTOR CONCEPT. .......................................................... 14

FIGURE 2-1. BENCHMARK UGD VVER-1000 ASSEMBLY CONFIGURATION [18]............................ 17

FIGURE 2-2. 2-D SERPENT MODEL OF VVER-1000 HEXAGONAL FUEL ASSEMBLY......................... 18

FIGURE 2-3. IMPACTS OF NEUTRON LEAKAGE ON THE NATURAL RESOURCE REQUIREMENTS FOR THE

FULL POWER CASES. ............................................................................................................... 25

FIGURE 2-4. IMPACTS OF NEUTRON LEAKAGE ON THE NATURAL RESOURCE REQUIREMENTS FOR THE

DE-RATED POWER CASES. ....................................................................................................... 26

FIGURE 2-5. IMPACTS OF FUEL ENRICHMENT AND POWER DENSITY ON THE SNF+HLW ACTIVITY AT

100 YEARS AFTER FUEL DISCHARGE. ...................................................................................... 27

FIGURE 2-6. IMPACTS OF NEUTRON LEAKAGE ON THE SNF+HLW ACTIVITY AT 100 YEARS AFTER

FUEL DISCHARGE. ................................................................................................................... 28

FIGURE 2-7. IMPACTS OF FUEL ENRICHMENT AND POWER DENSITY ON THE SNF+HLW ACTIVITY AT

100,000 YEARS AFTER FUEL DISCHARGE. ............................................................................... 29

FIGURE 2-8. NEUTRON LEAKAGE IMPACTS ON CALCULATED SNF+HLW ACTIVITY AT 100,000

YEARS AFTER FUEL DISCHARGE FOR THE VVER-1000 CASES STUDIED. ................................. 30

FIGURE 3-1. SERPENT MODEL OF THE SPECIAL PURPOSE REACTOR. ........................................... 32

FIGURE 3-2. CROSS-SECTIONAL VIEW OF 2-D SERPENT MODEL OF A MHTGR. .............................. 34

VIII

FIGURE 3-3. NORMALIZED NEUTRON FLUX PER LETHARGY AT BOL AND EOL OF THE EVINCI-

LIKE CORE. ............................................................................................................................. 36

FIGURE 3-4. CALCULATED ACTIVITY OF SNF+HLW AT 100 YEARS AFTER FUEL DISCHARGE FOR

ALL CASES. ............................................................................................................................. 38

FIGURE 3-5. CALCULATED ACTIVITY OF SNF+HLW AT 100,000 YEARS AFTER FUEL DISCHARGE

FOR ALL CASES. ...................................................................................................................... 39

FIGURE 3-6. K-EFFECTIVE MULTIPLICATION FACTOR FOR MONOLITH MODIFICATIONS DONE ON THE

EVINCI-LIKE CONCEPT. ....................................................................................................... 44

FIGURE 3-7. MIGRATION AREA FOR MONOLITH MODIFICATIONS ON THE EVINCI-LIKE CONCEPT. 45

FIGURE 3-8. SNF+HLW ACTIVITY AT 100 YEARS AFTER FUEL DISCHARGE FOR MONOLITH

ANALYSIS. .............................................................................................................................. 46

FIGURE 3-9. SNF+HLW ACTIVITY AT 100,00 YEARS AFTER FUEL DISCHARGE FOR THE MONOLITH

INVESTIGATION. ..................................................................................................................... 47

FIGURE 3-10. K-EFFECTIVE MULTIPLICATION FACTOR FOR LEAKAGE MODIFICATIONS DONE ON THE

EVINCI-LIKE CONCEPT. ....................................................................................................... 48

FIGURE 3-11. MIGRATION AREA FOR LEAKAGE MODIFICATIONS ON THE EVINCI-LIKE CONCEPT. 49

FIGURE 3-12. SNF+HLW ACTIVITY AT 100 YEARS AFTER FUEL DISCHARGE FOR THE NEUTRON

LEAKAGE INVESTIGATION. ..................................................................................................... 50

FIGURE 3-13. SNF+HLW ACTIVITY AT 100,000 YEARS AFTER FUEL DISCHARGE FOR THE NEUTRON

LEAKAGE INVESTIGATION. ..................................................................................................... 51

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LIST OF TABLES TABLE 2-1. SPECIFICATIONS OF SERPENT MODEL OF VVER-1000 FUEL ASSEMBLY. ..................... 18

TABLE 2-2. SERPENT MODEL VERIFICATION WITH BENCHMARK RESULTS. ..................................... 19

TABLE 2-3. SERPENT MHTGR NEUTRONICS MODEL COMPARISON WITH THE EG02 E&S STUDY

REFERENCE CASE.................................................................................................................... 21

TABLE 2-4. IMPACTS OF FUEL ENRICHMENT AND POWER DENSITY ON THE FUEL CYCLE

PERFORMANCE PARAMETERS. ................................................................................................ 24

TABLE 3-1 SERPENT ACTIVITY METRICS VALIDATION WITH EG02 REFERENCE CASE. ................... 34

TABLE 3-2. FUEL CYCLE PERFORMANCE PARAMETERS FOR THE 5MW AND 25MW CASES,

INCLUDING ADJUSTED ENRICHMENTS FOR THE 10-YEAR OPERATING TIME. ............................ 37

TABLE 3-3. SUMMARY OF CALCULATED FUEL CYCLE PARAMETERS OF THE EVINCI-LIKE

CONCEPT, THE EG01 LIGHT WATER REACTOR (LWR) REFERENCE CASE, AND THE EG02

MHTGR CASE. ....................................................................................................................... 40

TABLE 3-4. NATURAL RESOURCE UTILIZATION FOR THE PARASITIC ABSORPTION ANALYSIS. ......... 46

TABLE 3-5. NATURAL RESOURCE UTILIZATION FOR THE NEUTRON LEAKAGE INVESTIGATION. ...... 50

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ACKNOWLEDGEMENTS

I would like to thank my graduate research advisor Dr. Nicholas R. Brown. His support,

technical knowledge and continued guidance, were fundamental towards the accomplishment of

the work conducted in this thesis. He has invested numerous advising hours towards the personal

development of my capabilities as a nuclear researcher.

I would also like to extend special gratitude to Dr. Michael Todosow from Brookhaven

National Laboratory. He provided the opportunity to work on several DOE involved projects. His

technical knowledge also contributed towards the completion of the work found in this thesis.

Additionally, I would like to thank Dr. Arthur Motta and Dr. William Walters for being a part of

my Master’s thesis committee. Their constructive comments on my work have allowed me to

further better the contents of this thesis.

I want to also extend a sincere thank you to my friends and family. They proved to be my

number one fans and motivated me to continue working hard no matter the obstacles. Lastly, I

would like to extend a great thank you to my girlfriend Brittany who has stuck with me throughout

my graduate study journey.

Disclaimer: Any opinions, findings, and conclusions presented in this thesis are only those of the

author and do not necessarily reflect the views of the funding agencies.

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1. Introduction

1.1 Background

The substantial energy demands needed to support the growing human population is ever

increasing. In light of recent climate changes, such as rising temperatures and sea levels, the U.S

cannot continue to rely solely on the burning of fossil fuels to meet energy needs. Nuclear energy

provides a valuable alternative due to its ability to produce vast amounts of energy with virtually

zero carbon emissions into the environment. Presently, the 98 operational commercial nuclear light

water reactors (LWRs) in the U.S were responsible for producing about 20% of the total electricity

output in 2017 [1]. The majority of these reactors were built in the 1970s and are currently under

a 60-year lifetime operational limit. This means that the nuclear industry is approaching a

retirement curve of its commercial power plants. With the continuous increasing energy demands

of the country, the nuclear industry must revolutionize its approach away from the Gen-II LWR

designs and develop more inherently safe, deployment flexible, reliable and economical concepts.

One approach to revolutionize the nuclear industry into the modern world, has been the

ongoing research into advanced small modular reactor (SMR) concepts. The Department of Energy

has acknowledged the capability of SMRs to provided national energy security as a result of

several advantages not achievable by large conventional LWRs [2]. Due to their small sizes, SMRs

require fewer initial investments, operational costs and experience flexible deployment

availabilities not feasible with large reactors [2]. Further, these advanced SMR designs have

inherent safety features such as lower core power densities, than those found in larger cores, as

well as negative temperature feedback coefficients, that would innately mitigate an unexpected

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surge in core power due to a reactivity insertion. Several of these concepts also possess non-

proliferation characteristics such as core encapsulation and minimal plutonium contents in the

spent fuel.

Power scalability is another major advantage of SMRs. Higher power outputs can be achieved

simply by the addition of more modular reactors. Lower power outputs can be attained by

decreasing the size of the SMR concept used. Currently, there are about 50 SMR concepts being

developed by member states all around the world, and they can be designed to produce net electric

power outputs between a few megawatts to a few hundred megawatts [2,3]. The advantages

mentioned, allow these SMR concepts to be suitable for a broader range of applications such as

offshore deployments, decentralized environments, desalination processes and localized power

needs.

There is however, from a technical point of view, one disadvantage to SMRs when

compared to large reactors. In nuclear reactors, neutrons are produced in the core through fission

reactions. These neutrons produced are then removed through absorption reactions, with fuel and

non-fuel elements, and leakage out of the core. Leakage results in a loss of neutrons available for

fission and therefore decreases the reactivity of the core. When compared to the larger LWR cores

currently used in the commercial U.S. fuel cycle, these SMR experience higher radial and axial

neutron leakage rates out of the core [4]. This is due to the relatively smaller core sizes of these

concepts. The higher leakage rates result in negative impacts on the fuel cycle performance of

these concepts.

The work performed in this thesis investigated the impacts of high neutron leakage rates

on the fuel cycle performance of both thermal and fast spectrum SMRs. In the first study, found

in Chapter 2, a fuel cycle analysis of Russian VVER-based thermal spectrum SMR was conducted.

3

These modular reactor designs, described in section 1.4, were developed for offshore applications.

Several advantages are manifested with the use of an offshore floating nuclear plant. Among these,

the threat of natural disasters such as tsunamis, earthquakes and tornadoes are eliminated.

Additionally, an expansion of the capabilities of the nuclear industry can be achieved by opening

the door to the possibility of new nuclear applications such as off-grid and desalination processes

[5]. The VVER-based SMR study investigated the impacts of fuel enrichment, power density and

neutron leakage on two fuel cycle parameters; the natural resource utilization and the spent nuclear

fuel (SNF) and high-level waste (HLW) metrics. These two parameters were analyzed following

the guidelines of the evaluation and screening (E&S) study described in section The Nuclear

Fuel Cycle Evaluation and Screening Study [6].

The second study included in this thesis consisted of a fuel cycle analysis of the

eVinciTMmicro heat pipe reactor developed by the Westinghouse Electric Company. Due to its

similarity to eVinciTM concept, the Los Alamos National Laboratory (LANL) fast spectrum special

purpose reactor concept, described in section 1.5, was chosen for the scope of this study. An

analysis of the fuel cycle performance of the special purpose reactor was conducted in this thesis,

found in Chapter 3, following a subset of the nine high-level criteria of the E&S study [6]. Natural

resource utilization, waste management metrics and environmental impacts due to several fuel

cycle options were all investigated. Additionally, the final scope of the study included an analysis

of the impacts of parasitic absorption and core neutron leakage on the fuel cycle performance of

the special purpose reactor concept.

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1.2 The Nuclear Fuel Cycle Evaluation and Screening Study

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Evaluation and

Screening (E&S) study was begun in 2011, with the goal of categorizing all potentially

promising fuel cycle options for comparison using specified performance metrics. Forty

“Evaluation Groups” (EG) were classified based on fundamental fuel cycle characteristics such

as once-through, limited or continuous recycle fuel options, type of nuclear fuel used, isotopic

enrichment of the fuel, neutron spectra of the core and its irradiation system [6]. Each EG had

an “Analysis Example” (AE) that represented the specific type of fuel, neutron spectra and

irradiation environment of the group. The results of the E&S study were released in 2014 as

part of the DOE-NE Fuel Cycle Options (FCO) campaign to provide information to serve as

the basis for future nuclear Research and Development (R&D) [7]. The most promising EGs

with the highest possible improvements to the current U.S nuclear fuel cycle were all

continuous recycle options in the E&S study [6].

The performance metrics of the E&S study, consisted of nine evaluation criteria

and twenty-five specific metrics that provided information on the natural resource utilization,

radioactivity and quantities of the outputted waste and environmental impacts that would result

because of the implementation of a specific fuel cycle option. Further, the E&S also focused

on the proliferation and security risk due to materials found within the AEs studied, as well as

the developmental, deployment, institutional and financial challenges associated with a fuel

cycle. The AEs were then used to form the basis of performance bins. These ranged from “A”

(the “best”) to up to “F” (the “worst”) [6].

5

A subset of the guidelines found in these evaluation metrics was the framework used in the

work present in this thesis. All the concepts studied were categorized under the EG02 in the

E&S study. The EG02 group is categorized by a once-through uranium fuel cycle, with U-235

fuel enrichments greater than 5% but less than 20%. The AE for this evaluation group is a

once-through, thermal spectrum, helium-cooled modular High-Temperature Gas Cooled

Reactor (mHTGR) with a rated thermal output of 350 MW, shown in Figure 1-1. The prismatic

core of the reactor system utilizes Tristructural-isotropic (TRISO) Low-Enriched Uranium

(LEU) fuel with a U-235 enrichment of 15.5 wt.%. The discharge burnup of the fuel is 120

GWd/tU with a fuel residence time of 4.9 Effective Full Power Years (EFPY). The net thermal

efficiency is rated at 50%. The spent fuel, depleted uranium and low-level waste is sent for

disposal at the end of the fuel cycle [6].

Figure 1-1. Mass flow chart for the EG02 reference case [6].

6

1.3 SMR Fuel Cycle Performance Studies

Previous studies have looked into the possible impacts of the design characteristics of

SMRs on nuclear fuel cycle performance. A study conducted in collaboration with the Brookhaven

and Oak Ridge National Laboratories, investigated the potential negative fuel cycle performance

impacts due to the high neutron leakage rates of SMRs [4]. The study centered around the effects

of core leakage on natural resource utilization, waste outputs, as well as the environmental impacts

of SMRs. The analysis in this reference was performed following the guidelines found in the

DOE-NE Fuel Cycle E&S study described in Section 1.2 [6]. Specifically, the results in this

reference were compared to those of the EG01 reference case in the E&S study. This AE consisted

of a once-through, thermal spectrum, LEU Pressurized Water Reactor (PWR), with a fuel

enrichment of just under 5% [4].

The work done in the thermal spectrum SMR fuel cycle referenced study was conducted

using a theoretical “cartridge type’ light water modular reactor with a U-235 fuel enrichment of

just under 5%, using uranium oxide fuel [4]. The approach consisted of constructing a one-batch

representative SMR model using a typical 17x17 PWR assembly. This is similar to the study

approach performed in Chapter 2 of this thesis. In addition to their smaller core size, the study

indicated that due to a lack of boron in the coolant for beginning of life (BOL) reactivity control,

and the design of the SMR’s smaller fuel assemblies, the core would experience higher neutron

leakage rates relative to large conventional PWRs. This was due to higher variations in the axial

and radial power profiles of these concepts [4].

The analysis in the paper included a full power and 50% de-rated power case to represent

large PWR and light water SMR systems respectively. The fuel discharge burnup of the full power

7

case was found to be 30.5 GWd/tU. The de-rated case had a discharge burnup of 32.4 GWd/tU

and double the fuel residence time of the full power case [4]. When compared to the EG01

reference case in the E&S, with a fuel discharge of 50 GWd/tU, the reduction in fuel utilization

due to increased neutron leakage in the SMR concepts is quantified. The hypothetical SMR in this

study had a BOL neutron leakage of between 6-7% whereas the EG01 reference case experienced

about 3% neutron leakage, which is typical of that found in large conventional PWR cores [4].

Also investigated in this study, were the impacts of the reduced fuel discharge burnup on

the natural resource utilization and waste metrics of this SMR fuel cycle option. The natural

uranium required to sustain both SMR power cases were between about 345 and 366 t/GWe-year.

This was much higher than then 190 t/GWe-year needed to sustain the fuel cycle of the EG01

reference case. Additionally, the normalized mass output per energy year of the spent fuel and

highly radioactive waste for the full and de-rated power cases was 36.28 t/GWe-yr and 34.16

t/GWe-yr respectively. This was considerably higher than the 22.13 t/GWe-year outputted by the

EG01 analysis example. Both of these negative resulting impacts of higher neutron leakage rates

resulted in increased environmental impacts due to the operation of this SMR fuel cycle [4].

Lastly, the study in reference also looked at an infinitely reflective core case with 0%

neutron leakage. This part of the investigation concluded that this case had a fuel discharge burnup

of 44 GWd/tU and a natural resource utilization of 254 t/GWe-year. This case performed

significantly better than the higher leakage cases but still not equal to the discharge burnup of the

EG01 reference case. The cause of the differences was mentioned to be the use of a one-batch fuel

system instead of the three-batch system used in larger commercial PWRs; as it is known that

multi-batch configurations increase fuel utilization [4].

8

1.4 VVER-based Russian Floating SMR Concepts The current goal of the Offshore Floating Nuclear Plant (OFNP) concept is to develop

revolutionary reactor designs that are inherently safe, reliable and economically attractive [5]. The

offshore modular reactor systems studied in this work are of Russian origin and offer deployment

flexibilities not achievable by large conventional reactors. Russia has a track record of over 300

reactor years of experience using marine reactors to power artic icebreaker and military ships [8,9].

Thus, this knowledge served as the technological basis used to develop the concepts studied in this

thesis.

The three floating SMR concepts considered in this study were the ABV-6M, KLT-40s

and the RITM-200s. The ABV-6M and KLT-40s are currently under development, whereas the

RITM-200s has already been deployed and in service. The cores of these floating SMR concepts,

all have triangular pitch lattices with light water coolant designs that are based on VVER-like fuel

assembly technology. Russian VVERs are the equivalent to American pressurized water reactors

(PWRs). Additionally, all these concepts utilize a once through fuel cycle with U-235 enrichments

in excess of 5% but not surpassing the LEU enrichment limit of 20%.

The ABV-6M is an OFNP concept designed for both land and offshore floating power plant

operations [5]. The original design was completed in 1996 by the developer OKBM Afrikantov

based in Nizhny Novgorod, Russia [10]. The ABV is the smallest of the floating SMR concepts

studied, with a rated thermal output of 38 MW and an electric power output of 8.6 MW [5]. It

utilizes uranium dioxide type fuel in both pellet and silumin matrix form with a fuel enrichment

between 16.5% and 19.7% U-235 [5,10]. The average fuel discharge burnup in this reference is

indicated to be 94.5 GWd/tIHM, with an operation period between refueling of 8 years [10]. The

9

concept has a service life of 50 years [11]. The design pressure of the ABV-6M is 15.7 MPa and

the operating temperature is 300 °C; this is similar to that found in a typical VVER system [5].

The ABV design features 100% natural circulation in the primary circuit eliminating the risks of

pump failures [5].

The Russian KLT-40s is an icebreaker-type pressurized SMR system used in

piloted floating Nuclear Power Plants (NPP) [12]. The principal designer of this concept is

Afrikantov OKB Mechanical Engineering (OKBM) [8]. The design is currently under construction

in Severodvinsk, Russian in order to demonstrate the possible advantages achievable by this

reactor’s technology [10]. The features of the KLT-40 system are presented in the flow diagram

shown in Figure 1-2. This floating SMR has a rated thermal output of 150 MW and an electric

power output of 35 MW [13]. The core of the KLT-40s uses a U-235 fuel enrichment of less than

20%, enhancing the non-proliferation capabilities of this system [8]. Both uranium dioxide pellet

or cermet type fuel can be used in this concept, with an average fuel burnup of 45.4 GWd/tU and

a claimed cycle length of operation between refueling of about 3-4 years [8]. The spent fuel can

be store within the floating power unit housing the reactor [8]. The design pressure of the primary

coolant is 12.7 MPa, with operating inlet and outlet temperatures of 280 °C and 317 °C

respectively. Several modular reactors can be connected in series, for use in icebreakers, sea

vessels, as well as floating nuclear power plants. [13]. Several safety features of the KLT-40s

include control rod emergency shutdown systems, primary and secondary auxiliary heat removal

systems and passive containment cooling systems designed to make this concept inherently safe

under accident scenarios [8].

10

Figure 1-2. KLT-40s concept flow diagram [10].

The RITM-200 is a SMR concept designed to replace the aging Russian nuclear

icebreaker power plant fleet. The technology instrumented into the design of the RITM-200 has

been developed and validated with over 175,000 operational hours, and throughout a service

lifetime of 34 years [14]. Recently, a completed RITM-200 reactor system was installed in the

Russian icebreaker Arktika, shown in Figure 1-3, and it is currently in active service. The volume

of the core of this concept is several times larger than that used in nuclear submarines, and it is

capable of maintaining operational integrity throughout over 600,000 service lifetime power

changes [14]. The RITM-200 has a rated thermal power output of 175 MW and an electrical power

output of 50 MW [13]. The core of the concept can utilize both uranium dioxide pellets and cermet

11

type fuel, with a fuel enrichment of less than 20% in order to enhance the non-proliferation

capabilities of the concept [1,6]. The operating period between refueling is claimed to be about 7

years [14]. The operating pressure of the RITM-200 is 15.7 MPa with a primary coolant

temperature of 295 °C [13]. The circulation of the primary coolant loop is mainly forced, with

about 30% of it being natural circulation [14]. Overall, the design of the RITM-200 is very similar

to features found in VVER reactors.

Figure 1-3. RITM-200 installation into the Russian icebreaker Arktika [12].

12

1.5 Fast Spectrum Micro Heat Pipe Nuclear Reactor Concepts The innovative technology found in micro heat pipe reactors, first developed at LANL in

the 1960s, offers a revolutionary design capable of meeting the present demands of decentralized

energy markets not feasible with large conventional reactors. Military applications, disaster relief

operations, remote communities and off-grid operations, such as artic mining, can benefit from the

advantages of these concepts [16]. Heat pipe fission reactors, like the SAFE-400 and the HOMER-

15, are also being investigate for deep space exploration applications due to their technological

readiness, reliability, inherent safety and low production cost [17,18].

Westinghouse Electric Company has selected the eVinciTM micro heat pipe reactor to meet

the demands of decentralized energy applications [19]. Its simple design can operate similar to a

nuclear battery without the need of a centralized power grid. The eVinciTM

concept, shown in

Figure 1-4, aims to meet the decentralized market’s energy needs with a simple design that features

inherent safety, reliability, affordability and proliferation resistance. The heat pipe concept has a

high technology readiness level (TRL) [19]. By 2019, Westinghouse is aiming to developed a full-

scale electrical model of the concept to demonstrate its feasibility, in preparation for an expected

commercial deployment to be achieved by the year 2024 [19].

The eVinciTM

reactor concept is a liquid metal cooled fast spectrum reactor that utilizes a

once-through LEU fuel source. Its solid core design, is intended for autonomous operations of

about 10 years in length without the need of refueling [19]. The core is fully encased in a type-316

stainless steel monolith that further enhances its non-proliferation capabilities. Drilled within the

solid core are fuel pellet and heat pipe channels in a one-to-two ratio. Each fuel channel is

surrounded by three adjacent heat pipe channels, filled with liquid potassium, that passively

remove heat from the core in a lower than atmospheric pressure environment. The heat removed

13

is transported to an ex-core condenser without the need of mechanical valves, pumps or pipes used

in conventional Gen- II reactor coolant loop circuits. Thus, eliminating the risk of loss-of-coolant

accidents (LOCA), due to rupturing pipes, as well as loss of circulation accidents due to pump

failures. A central emergency core shutdown system, in addition to a strong negative temperature

coefficient, further enhances the inherent safety of the concept in the event of a reactivity-initiated

accident [19].

Figure 1-4. Illustration of the eVinciTM

reactor concept highlighting its main components. [15].

Due to the ambiguity of specific details about the main components of the eVinciTM

, the

analysis in Chapter 3 of this thesis was done on a similar fast spectrum heat pipe reactor concept

from an assessment report carried out by the Idaho National Laboratory (INL) [20]. The special

purpose reactor, shown in Figure 1-5, is a 5 MWth (2 MW𝑒) concept being developed at LANL for

use in remote sites and off-grid applications. The horizontally oriented hexagonal core, is a Type

316 stainless steel monolith consisting of six 60 ° sector wedges with drilled fuel and heat pipe

14

channels. The 2,112 fuel channels contain solid cylindrical uranium dioxide pellets with a U-235

enrichment of 19.75 wt.% and a total core mass of 5.22 metric tons. The 1,224 heat pipe channels

each contained 100 grams of liquid potassium at maximum operating temperature of 675 °C. The

core is encased in an alumina (Al2O3) reflector material and the top and bottom sections of the fuel

channels contain beryllium oxide (BeO); this was done to increase the reactivity of the core. 12

rotating boron carbide (B4C) control drums, embedded in the alumina reflector, serve as the

primary form of reactivity control in the concept. Lastly, two emergency B4C control rods inside

a central hexagonal volume can be used to quickly shutdown the core [20].

Figure 1-5. LANL Special Purpose Reactor concept.

The intended cycle length of the special purpose reactor is 5 years. Two thermal cycles can

be used with this reactor system; a simple and heat recuperated Brayton cycle with thermal

efficiencies of 29.4% and 40.3% respectively. A depletion calculation in the INL paper in reference

3, concluded that after 5 years of operation the discharge burnup of the fuel in the heat pipe concept

is <2,000 MWd/MTU. The neutronics analysis also determined that the U-238 doppler broadening

15

and thermal expansion of the fuel, as well as heat swelling of the steel were the primary means of

negative reactivity feedback in the tightly coupled core [20].

Several reflector materials were studied to see how they impacted the beginning of life

(BOL) reactivity in the core. BeO, beryllium metal, stainless steel and alumina materials were all

studied with the latter showing the most positive results [20]. Further, the core is highly sensitivity

to a fuel radius increase/decrease because of changes in the parasitic absorption rates. As a result,

no cladding was incorporated around the fuel/heat pipe channels in order to achieve core criticality

[20]. An analysis of the radioactivity levels of the metal coolant, confirmed the in-core activation

of the liquid potassium throughout the cycle [20]. This is common in liquid metal cooled reactors,

and the core has been designed to be well contained so that it poses no health risk to working

personnel [21].

16

2. Fuel Cycle Performance of Thermal Spectrum VVER-based

Russian SMRs

Parts of the work presented in this chapter was previously published as a first author in an

American Nuclear Society (ANS) conference paper for the 2018 winter meeting & expo occurring

between November 11-15 in Orlando, Florida [22].

2.1 VVER Assembly Model Description The model used in this VVER-based thermal SMR study, was constructed according to the

specifications found in a VVER-1000 LEU fuel assembly computational benchmark from the

Nuclear Energy Agency (NEA) office of Organization for Economic Co-operation and

Development (OECD) [23]. The benchmark fuel assembly, shown in Figure 2-1 , is typical of a

Russian VVER-1000 reactor core. The uniform LEU fuel assembly consist of 331 hexagonal cells

of four different types including 12 Gadolinia (Gd2O3) enriched fuel pins, a central tube and 18

guide tubes through which moderator fluid flows. The Gadolinia fuels cells have a U-235 fuel

enrichment of 3.6% w/o and a gadolinium oxide content of 4.0% w/o. The remainder fuel cells

have an enrichment of 3.7% w/o U-235. The hexagonal cells all have an equal outer dimension of

1.275 cm. Additionally, the cladding material used for all the cells was Zirconium. The moderator

fluid used in the fuel assembly in the benchmark was light water with an initial boron concentration

of 600 parts per million (ppm) [23].

17

Figure 2-1. Benchmark Uranium-Gadolinia VVER-1000 assembly configuration [18].

The fuel assembly was modeled using the Serpent Monte Carlo reactor physics code

initially developed at the VTT Technical Research Centre of Finland [24]. The 2-D Serpent model,

shown in Figure 2-2, was constructed according to the benchmark specifications. The broad

capabilities of Serpent allowed for the construction of a 10 x 10 hexagonal fuel assembly lattice

with a pin pitch of 1.275 cm. The fuel pins in the model had an outer fuel radius of 3.86 mm and

an outer cladding radius of 4.58 mm. The guide tubes were modeled as pins with an inner radius,

through which moderator fluid flows, surrounded by zirconia cladding. The central guide tube had

an outer cladding radius 5.62 mm and the remaining 18 guide tubes, scattered around the assembly,

had an outer cladding radius of 6.32 mm.

18

Table 2-1. Specifications of Serpent model of VVER-1000 fuel assembly.

Assembly Lattice 10 x 10 Hexagonal

Pin Pitch 1.275 cm

Fuel Pin Outer Radius 3.86 mm

Fuel Cladding Outer Radius 4.58 mm

Central Guide Tube Outer Radius 5.62 mm

Outer Guide Tube Outer Radius 6.32 mm

Cladding Material Zirconia

Figure 2-2. 2-D Serpent model of VVER-1000 hexagonal fuel assembly.

The temperature of the fuel elements was set at 1027 °K and the remaining non-fuel

elements were set at a temperature of 575 °K in accordance to the benchmark specifications.

Further, the linear power of the assembly was set at 3.971E+04 W/cm and was modeled from an

initial operating non-poisoned state. Thirty uniform depletion steps were used in the burnup

calculations. To better model the burnup calculations in the Gadolinia rods, additional radial rings

19

were included inside these fuel pins. Periodic boundary conditions were used in the model to

represent an infinite lattice. Thus, all neutrons entering a periodic boundary were diverted to the

opposite boundary surface [24]. A neutron population of 50,000 was used and the number of active

and inactive cycles was set at 500 and 50 respectively to allow for the convergence of the fission

source, resulting in an uncertainty in k-effective values of +/- 20 pcm.

2.2 Serpent Assembly Model Validation

In order to provide validation of the Serpent model constructed using the specifications,

burnup calculations were conducted and results for several parameters were compared with those

found in the OECD-NEA benchmark reference. This is because the burnup calculations included

in the benchmark were carried out using different neutronics codes and cross section libraries than

the ones used in the Serpent calculations performed in this Chapter. Validation was carried out by

comparing the K-infinity multiplication factor and the isotopic compositions of Pu-241, in both

the gadolinia and non-gadolinia fuel rods, at the 20 and 40 GWd/t burnup steps with the benchmark

data.

Table 2-2. Serpent model verification with benchmark results.

Burnup

Step

(GWd/t)

K-infinity

Diff.

(pcm)

Pu-241

Concentration

U rods

(Diff %)

Pu-241

Concentration

Uranium-

Gadolinia rods

(Diff %)

20 96 +/- 20 2.9 8.1

40 120 +/-

40 2.9 4.7

The values shown in Table 2-2 above are the relative differences between the benchmark

and assembly model built in for this study at the two burnup steps. The comparative difference in

20

values is more pronounced in the gadolinia fuel rods. The disagreement is most likely due to the

different cross sections used. The benchmark case utilizes the JEF 2.2 library, whereas the

ENDF/B VII.0 library was used for the Serpent model.

2.3 Serpent Activity Metrics Validation

The purpose of this validation was to verify that the code we are using, will accurately

calculate the activity metrics of the burnup calculations carried out in this chapter. This verification

was done because the fuel cycle metrics performed in the E&S study use the ORIGEN neutronics

code, whereas we are using the Serpent code to conduct this analysis. To commence the metrics

validation, a 2-D core model of a 350 MWth mHTGR was constructed and several parameters

were compared with the EG02 reference case found in the Appendix B of the E&S study [6]. As

mentioned before, this reference case comparison was used because these VVER-based Russian

SMR concepts would be categorized under this evaluation group. Both models contained LEU fuel

enriched to 15.5% U-235, and had a discharge burnup of 120GWd/t.

In order to convert the activity metrics outputted from Serpent, in Becquerels (Bq) per cm,

to the proper units used in the E&S study, Curies (Ci) per GWe-year, the following equation was

used:

𝐴𝑐𝑡𝑖𝑣𝑖𝑡𝑦 (𝐶𝑖

𝐺𝑊𝑒−𝑦𝑟) =

𝐴𝑐𝑡𝑖𝑣𝑖𝑡𝑦(𝐵𝑞

𝑐𝑚)∗(

1 𝐶𝑖

3.7𝐸10 𝐵𝑞)∗𝑀𝑎𝑠𝑠 𝐹𝑙𝑜𝑤 (

𝑡

𝐺𝑊𝑒−𝑦𝑟)

𝐹𝑖𝑠𝑠𝑖𝑙𝑒 𝑀𝑎𝑠𝑠 (𝑡

𝑐𝑚)

(2.1)

The fissile mass, in t/cm, is another parameter that is outputted by the Serpent code. The

mass flow in equation 2.1, is given by the following equation:

21

𝑀𝑎𝑠𝑠 𝐹𝑙𝑜𝑤 (𝑡

𝐺𝑊𝑒−𝑦𝑟) =

𝑃𝑜𝑤𝑒𝑟𝑡ℎ𝑒𝑟𝑚𝑎𝑙(𝑀𝑊)

𝐵𝑢𝑟𝑛𝑢𝑝(𝐺𝑤𝑑

𝑡)∗𝑃𝑜𝑤𝑒𝑟𝑒𝑙𝑒𝑐𝑡𝑟𝑖𝑐(𝑀𝑊)

∗ 365𝑑𝑎𝑦𝑠

𝑦𝑟 (2.2)

Where the ratio of the electric and thermal powers is the net thermal efficiency. As in the

E&S study, the net thermal efficiency was normalized to 33% to allow for a direct comparison

between fuel cycle options [6].

The mass flow is basically the tonnage of uranium fuel needed to produce 1 GWe per

energy year in a respective fuel cycle. The fissile mass of the mHTGR Serpent model was

calculated to be about 3.54E-02 tons, and the mass flow for both cases was calculated to be about

6.1 t/GWe-yr. Using the outputted Serpent model radioactivity data, the spent nuclear fuel (SNF)

and high-level waste (HLW) activity metrics were compared with the E&S study EG02 reference

case at 100 and 100,000 years after fuel discharge. The results are shown below in Table 2-3.

Table 2-3. Serpent mHTGR neutronics model comparison with the EG02 E&S study reference

case.

Reactor type Enrichment

(U-235 %)

Discharge

BU

(GWd/t)

Charge

Mass Flow.

(t/Gwe-yr)

Calculated

Activity

(Ci/Gwe-yr)

100 years

Calculated

Activity

(Ci/Gwe-yr)

100,000 years

HTGR

(EG02

reference

case)

15.5 120 6.1 1.43E+06 2.05E+03

HTGR

(Serpent

model)

15.5 120 6.1

1.39E+06

(-2.9%

Diff.)

2.03E+03

(-0.98% Diff.)

22

2.4 Fuel Cycle Performance Analysis of VVER-based SMRs

Following the validation of both the VVER assembly model and activity calculation

capabilities of the Serpent code, the assembly model was loaded with three different U-235 fuel

enrichments: 10%, 15% and 19.7%. These three cases were chosen to represent the fuel

enrichments of the different VVER-based concepts. For each enrichment case, a full power and

50% de-rated power case was included to represent the power density found in large reactor cores

and SMR cores respectively. The full power density case was the same as that used in section 2.1,

and 50% de-rated power case had a power density of 1.985E+04 W/cm. Additionally, several

neutron leakage assumptions were made in the analysis. A typical large pressurized water reactor,

such as a VVER, has about a 3% neutron leakage rate. Further, due to their small size, these SMR

concepts will experience higher axial and radial neutron leakage than large conventional reactors

[4]. Increasing leakage rates will result in a decrease in the discharge burnup of the fuel. For all

cases, the fuel cycle performance of three leakage cases: 0%, 3% and 7%, were compared.

The linear reactivity model (LRM) can be used to predict the cycle length of a reactor. In

PWRs, such as the VVER-based SMRs, the variation of the initial reactivity is approximately

linear [26]. Using this knowledge, the linear reactivity model was used to calculate the cycle length

for the three leakage cases using the following linear equation:

ρ =Kinfinity−1+(

percent leakage

100)

Kinfinity (2.3)

Where ρ is the reactivity of the core and K-infinity is the multiplication factor of each case

at a respective burnup case. In the case of an infinite lattice with no leakage, the fuel’s discharged

burnup is simply when the K-infinity reaches a value of 1.0 and the reactor is only critical because

23

of the delayed neutron source. For cases with leakage involved, equation 2.3 can be then used to

predict when the discharge of the fuel will occur.

The impacts of fuel enrichment, rated power density and neutron leakage on two E&S

study criteria was then investigated. The criteria investigated were the natural resource utilization

(NRU) and the SNF+HLW radioactivity at 100 and 100,000 years of the discharged fuel.

The NRU is calculated using the following equation:

NRU (t

GWe−yr) = Mass Flow (

t

GWe−yr) ∗ (

enruLEU−enrutails

enruNU−enrutails) (2.4)

In equation 2.4, 𝑒nruLEU is the enrichment of the fuel, enruNU is 0.711 for the natural

uranium enrichment and enrutails is 0.25 for the tails lost in the production of the fuel [27]. The

mass flow is calculated using equation 2.2. The impacts of fuel enrichment and power density on

the NRU required to maintain the fuel cycle being investigated is shown in Table 2-4 below. The

net thermal efficiency for all these cases was assumed to be 33% and a 0.2% fabrication loss was

included as in the E&S report [6]. Higher fuel enrichment results in an increase in the initial excess

reactivity in the core. This allows the reactor to achieve a longer fuel cycle length. Additionally,

the results show that there are no significant NRU differences between the full and de-rated power

cases. The de-rated cases had a slightly higher fuel burnup. Overall the performance for all these

cases was similar as they are in the lowest metric bin (D) in the E&S study for this criterion.

24

Table 2-4. Impacts of fuel enrichment and power density on the fuel cycle performance

parameters.

Reactor Type Enrichment

(%)

Discharge

Burnup

(GWd/t)

Fuel Residence

Lifetime

(EFPY)

Natural

Resource

Utilization

(t/Gwe-yr)

VVER-1000 10.0 74.2 6.8 314.5

VVER-1000 15.0 107.9 9.9 328.9

VVER-1000 19.7 137.2 12.5 341.1

VVER-1000

(De-rated) 10.0 75.0 13.7 312.4

VVER-1000

(De-rated) 15.0 108.3 19.8 325.7

VVER-1000

(De-rated) 19.7 137.2 25.1 341.0

HTGR

(EG02 AE) 15.5 120.0 4.9 305.7

Table 2-4 also shows that the mHTGR reference case for the EG02 group, is marginally

more efficient at utilizing natural resources than the VVER-based light water concepts. The

discharge burnups of these concepts, are exceptionally high compared to NRC standards for PWR

fuel. From a fuel cycle analysis perspective, due to a higher U-235, the discharge burnup will be

undoubtedly higher because the fuel would be able to maintain core criticality for much longer

than normal PWR fuel cycle lengths. These results were also modeled assuming an unrealistic

infinite lattice, with no neutron leakage. This further increase the discharge burnup of the fuel.

25

Since the work in this thesis was done under a fuel cycle analysis umbrella, more work will

be needed to determine how the integrity of these fuel materials will hold under these extreme

discharge burnups. With these higher burnups, the risk of fuel failure, during normal and accident

conditions, is increased. Although, a safety feature of these SMRs is a reduction in the core power

density. Therefore, under normal operating conditions, these fuels will most likely never reach the

high temperatures that could be experienced in a large PWR core.

Following, the impacts of neutron leakage on the NRU requirements of the above

enrichment and power density cases was investigated. Increasing leakage in a reactor core results

in a decreased in the discharge burnup, and thus an increase in the charge mass flow (see equation

2.2). This results in an increase in the natural resource required to produce the same amount of

energy. This can be seen from the results shown in Figure 2-3 and Figure 2-4.

Figure 2-3. Impacts of neutron leakage on the natural resource requirements for the full power

cases.

26

Figure 2-4. Impacts of neutron leakage on the natural resource requirements for the de-rated

power cases.

The de-rated power cases in this investigation of the impacts of leakage on the NRU, have

slightly higher discharge burnup values than the full power cases. This results in slightly lower

natural resource required to sustain the fuel cycle. Following, the impacts of fuel enrichment and

power density on the SNF+HLW activity at 100 and 100,000 years was investigated.

At 100 years after fuel discharge, the radioactivity found in the spent fuel is primarily due

to fission products. Therefore, variations in activities at this stage will be primarily due to the

fission products mass quantities and their specific activities; the main contributors being Cs-137

and Sr-90 each with a half-life of about 30 years [6]. The results of the impacts of fuel enrichment

and power density on the SNF+HLW activity at 100 years is shown in Figure 2-5.

27

Figure 2-5. Impacts of fuel enrichment and power density on the SNF+HLW activity at 100

years after fuel discharge.

Reactors with different neutron spectra will produce different quantities of fission products

and will result in dissimilar radioactivity levels at 100 years. Therefore, since the same thermal

system is being used for all the fuel enrichment and power density cases, the variations in

radioactivity at 100 years will be due to changes in the time the fuel spends in the reactor core

[28]. As enrichment increases, so does the fuel residence time. This results in fission products

having more time to decay in the core and thus a lower radioactivity at 100 years. Likewise, the

de-rated power cases have fuel residence times that are about twice as long as the full power cases.

The outcome is a lower SNF+HLW activity at 100 years in the de-rated cases. When comparing

the EG02 case, the cause of the increased activity at 100 years is not so simple. The mHTGR

analysis example used in this evaluation group, is graphite-cooled and therefore has a slightly

harder neutron spectrum than the light water SMR concepts. This could be a result of the different

neutron spectra between the two thermal systems as different mass quantities of fission products

are being formed in these two reactors. Overall the performance for all these cases at 100 years

was similar as they are in the same metric bin (C) in the E&S study.

28

Increasing neutron leakage also impacts the activity metrics at 100 years due to a change

in the fuel residence time. As neutron leakage increases, the fuel residence time decreases and

the SNF+HLW activity at 100 years of the discharged fuel will increase as a result. The effects

of leakage in the SNF+HLW activity at 100 years is shown in Figure 2-6.

Figure 2-6. Impacts of neutron leakage on the SNF+HLW activity at 100 years after fuel

discharge.

At 100,000 years after fuel discharge, the activity in the spent fuel is several orders of

magnitude lower than at 100 years. At this stage, the fission products are no longer the primary

contributors to the radioactivity. The dominant contributors to the activity are the amounts of

actinides in the SNF+HLW. Actinides are transuranic atoms which are heavier than the uranium

atoms found in the fuel of these reactor systems. They are formed by the capture of a neutron, by

a uranium atom, that does not lead to a fission event. In a thermal spectrum uranium fueled reactor,

the primary actinide contributor of activity at 100,000 years is Pu-239 with a half-life of about

24,100 years [6]. Pu-239 is formed by the capture of a neutron by a U-238 atom. Other plutonium

29

isotopes (e.g. Pu-240, Pu-241) are also contributors to the activity at 100,000 years. The effects of

fuel enrichment and power density on the SNF+HLW at 100,000 years is shown in Figure 2-7.

Figure 2-7. Impacts of fuel enrichment and power density on the SNF+HLW activity at 100,000

years after fuel discharge.

When analyzing the effects of power density on the activity at 100,000 years, the de-rated

cases have slightly higher total activity levels in the SNF+HLW than found in the full power cases.

This is due to the actinides in the de-rated cases having slightly higher fission rates and therefore

higher burnups. There are no significant impacts on the activity at 100,000 years as a result of

changes in the fuel enrichment. One important observation is the considerable increase in

radioactivity in the EG02 reference case. Upon inspection of the activities of the plutonium

isotopes outputted from Serpent, it was found that the activity of these isotopes was higher in the

mHTGR model than in the light water VVER-based SMRs. This could be a result of the harder

graphite-moderated mHTGR causing an increase in actinide formation in the spent fuel. Overall

the performance of the SNF+HLW activity for all these cases at 100,000 years was similar as they

are in the same metric bin (C) in the E&S study.

30

Increasing neutron leakage also impacts the activity metrics of the spent fuel at 100,000

years. This is because as neutron leakage increase, the discharge burnup of the fuel decreases. As

a result, there is less actinide build-up in fuel with increasing core leakage. The trend shown in

Figure 2-8 below supports this conclusion.

Figure 2-8. Neutron leakage impacts on calculated SNF+HLW activity at 100,000 years after

fuel discharge for the VVER-1000 cases studied.

31

3. Fuel Cycle Performance of the Fast Spectrum Special Purpose

Heat Pipe SMR concept.

Parts of the work presented in this chapter was previously published as a first author in a journal

article in the Annals of Nuclear Energy that was accepted on November 28, 2018 [29].

3.1 Special Purpose Reactor Model Description

The model was constructed according to the specifications from an INL reference paper on

the 5MW special purpose heat pipe nuclear reactor [20]. As mentioned before in Section 1.5, this

was done as a result of the lack of specific details about the components found in the Westinghouse

eVinciTM concept. The model was built using the Serpent Monte Carlo reactor physics code [23].

The 2-D Serpent model consists of an active core, with a height of 1.5 meters and a length of about

1.0 meters across, and it is made up of six identical 60° sector block wedges each thermally

separated by a void spacing. The core structure in the Serpent model is made up of a type-316

stainless steel solid monolith represented by a 45 x 45 hexagonal lattice and a pin pitch of 2.7731

cm (see Figure 3-1). It consists of 2,112 fuel channels and 1,224 heat pipe channels in about a 1-

to-2 overall core ratio. The fuel channels have an outer diameter of 1.412 cm surrounded by a

0.0065 cm gap, filled with Helium gas, and no cladding. The fuel used in the Serpent model is

uranium dioxide with an enrichment of 19.75% U-235 and a total mass, including U-235 and U-

238, of 4.60 metric tons. The heat pipe channels have an outer diameter of 1.575 cm and were

filled with Potassium material. The fuel and heat pipe channels were inserted into the monolith in

Serpent, using 3-D hexagonal prism cells and 9 x 9 hexagonal lattices with a pin pitch of 1.60 cm.

The fuel to fuel as well as the fuel to heat pipe web thicknesses, were modeled according to the

specifications found in the INL special purpose reactor paper [20]. A central 3-D hexagonal void

32

was incorporated in the model to represent the special purpose reactor’s Emergency Core Cooling

System (ECCS). The core was encased in a cylinder made up of alumina (Al2O3) material, shown

in brown in Figure 3-1.

Figure 3-1. SERPENT model of the Special Purpose Reactor.

For the purpose of this study, the boron carbide control drums (which are used instead of

control rods) were not included in the Serpent model. The temperature of the fuel materials was

set to 950 °K and the potassium coolant, steel monolith, helium gaps and alumina reflector were

set to 900 °K. A black boundary condition, in which the neutrons escaping out of the core are

eliminated, was applied to all directions in the model. The cross-section data used in this study

was taken from the ENDF/B-VII.0 library. For all cases studied in this chapter, the following

neutron parameter settings were employed. The neutron population was set at 50,000 per batch

generation. To allow the fission source to converge, 50 inactive neutron generation cycles were

implemented. Lastly, the number of active generation cycles was set to 500 to allow the

33

convergence of the k-effective multiplication factor with a statistical error of less than 20 pcm for

all cases.

3.2 Serpent Model Code to Code Validation

A code to code validation of the heat pipe reactor model constructed in Serpent was carried

out by comparing fuel burnup calculations, with data from the INL special purpose reactor paper.

A neutronics analysis of the concept conducted by INL indicated that, using a power rating of 5

MWth, after 5 years of operation the fuel had a discharge burnup of <2.0 GWd/tU [20]. The power

density in the model was then set to 0.00109 KW/g to achieve a power rating of 5 MWth. The

resulting burnup of the fuel in the Serpent model after 5 years of operation is 1.98 GWd/tU. This

provided some sort of confirmation that the special purpose reactor and the model reactor systems

utilize fuel at a similar rate.

3.3 Serpent Metrics Validation

As in chapter 2 of this thesis, a verification of the activity metrics capabilities of Serpent

was carried out by comparing several parameters from a 350 MWth mHTGR model, shown in

Figure 3-2, with that from the EG02 reference case in the DoE’s E&S study. EG02 was chosen as

the reference case for this study because the heat pipe concept in this analysis, fall under that

evaluation group in the E&S study [6]. Both mHTGR models had a fuel enrichment of 15.5% U-

235 and a fuel discharge of 120 GWd/tU. The activity metrics were normalized to 33% net thermal

efficiency as in the E&S [6]. The activity metrics outputted by Serpent were converted into the

proper units using equations 2.1 and 2.2. The results, shown in Table 3-1, indicated that Serpent

and the activity metrics from the E&S study assented with one another.

34

Figure 3-2. Cross-sectional view of 2-D Serpent model of a mHTGR.

Table 3-1 Serpent Activity metrics validation with EG02 reference case.

Reactor type

Activity

(Ci/GWe-yr)

100 years

Activity

(Ci/GWe-yr)

100,000 years

mHTGR

(EG02 reference) 1.43E+06 2.05E+03

mHTGR

(Serpent model)

1.39E+06

(2.9% Diff.)

2.03E+03

(0.98% Diff.)

35

3.4 10-Year Fuel Cycle performance of eVinciTM-like Reactor

The eVinciTM micro heat pipe reactor is designed to operate for about 10 years without

refueling throughout a range of different thermal power levels. For this matter, the initial scope of

this study consisted of burnup calculations of a 5 MWth and 25 MWth power case operated to 10

years. As specified earlier, net thermal efficiencies of 29.4% and 40.3%, for both the Simple

Brayton Cycle (SBC) and the heat recuperated Brayton Cycle (HRBC) respectively, were included

due to the possibility of implementing both of these thermal cycles in the special purpose reactor

system. Additionally, calculations were conducted on adjusted enrichment cases to determine the

fuel cycle which resulted in a final critical k-effective of 1.0 +/- 100 pcm, for both thermal power

cases, after 10 years of operation. This was done because the initial analyses, using 19.75%

enriched U-235 fuel, had excess reactivity leftover. The adjusted enrichment cases, 18.6% and

19.6% for the 5 MWth and 25 MWth respectively, resulted in no remaining excess reactivity after

10 years.

In order to characterize the bulk of the occurring fission reactions found in the eVinciTM-

like core, the spectra of the normalized neutron flux per unit lethargy at the beginning of life (BOL)

and end of life (EOL) of the 10-year fuel cycle are shown in Figure 3-3. The majority of the fissions

are happening within the range of 1.0E-02 MeV and 1.0E+00 MeV. This is consistent with the

characteristics of a fast spectrum reactor core.

36

Figure 3-3. Normalized neutron flux per lethargy at BOL and EOL of the eVinci-like core.

The first criterion from the E&S study evaluated in this initial study of the 10-year cycle

of the eVinciTM-like core, is the resource utilization requirements. The NRU standard evaluates the

amount of natural uranium, in tons, needed to maintain the power generation of the fuel cycle. This

quantity is normalized to a GWe-year for comparison with other fuel cycles. It also describes the

mass flow of materials from its initial mining, milling and enrichment processes required for

nuclear fuel production [6]. Ultimately, the NRU criterion is also a measure of the amount of

environmental impacts of these processes due to the sustainment of the fuel cycle. The results of

the 10-year fuel cycle cases are shown in Table 3-2, for all the enrichment cases, including both

the SBC and the HRBC thermal cycles. These were computed using equations 2.2 and 2.4 from

Chapter 2. The calculated natural resource utilization values include a 0.2% fabrication loss, as in

the E&S study [6].

37

Table 3-2. Fuel cycle performance parameters for the 5MW and 25MW cases, including adjusted

enrichments for the 10-year operating time.

Reactor Type Enrichment

(%) Efficiency

Fuel

Burnup

(GWd/t)

Fuel

Residence

Lifetime

(EFPY)

Charge

Mass Flow

(t/Gwe-yr)

Natural

Resource

Utilization

(t/Gwe-yr)

eVinci (5MW)

SBC 19.8 29.4 3.98 10.0 312.2 13179.4

eVinci (5MW)

HRBC 19.8 40.3 3.98 10.0 227.7 9612.3

eVinci (5MW)

SBC 18.6 29.4 3.98 10.0 312.2 12402.2

eVinci (5MW)

HRBC 18.6 40.3 3.98 10.0 227.7 9045.4

eVinci

(25MW)

SBC

19.8 29.4 19.8 10.0 62.7 2646.86

eVinci

(25MW)

HRBC

19.8 40.3 19.8 10.0 45.8 1933.44

eVinci

(25MW)

SBC

19.6 29.4 19.8 10.0 62.7 2626.50

eVinci

(25MW)

HRBC

19.6 40.3 19.8 10.0 45.8 1918.56

The activity metrics were also explored following the guidelines found in the E&S study

for this criterion. The SNF+HLW is particularly important because it determines the shielding and

long-term storage requirements needed to handle these highly radioactive waste materials.

As discussed in chapter 2, the activity at 100 years is primarily as a result of the specific

activities and mass quantities of the fission products present in the spent nuclear fuel. The primary

isotopic contributors being Cs-137 and Sr-90. The variations in the activity at this stage is a result

of the different thermal efficiency and fuel residence times of reactor systems. Since all these cases

have the same fuel residence time of 10 years the disparity in radioactivity at 100 years, shown in

38

Figure 3-4, will be due to the difference in the net thermal efficiencies of the simple Brayton and

heat recuperated Brayton thermal cycles.

Figure 3-4. Calculated activity of SNF+HLW at 100 years after fuel discharge for all cases.

At 100,000 years the activity is primarily due to the quantities of actinides accumulated in

the SNF+HLW. In a uranium reactor, the primary contributor is the Pu-239 isotope, with a half-

life of about 24,100 years, formed when a U-238 atom captures a neutron. The fast spectrum heat

pipe reactor produces a minimum amount of Pu-239 throughout its 10-year cycle [20]. But when

the activity is normalized to a GWe-year for comparison with other analysis examples in the E&S,

due to the low discharge burnup resulting in a high charge mass flow (Eq. 2.2), this results in

elevated activity levels as shown in Figure 3-5. Further, the thermal more thermal efficient HRBC

results in lower radioactivity in both power cases due to a lower charge mass flow.

39

Figure 3-5. Calculated activity of SNF+HLW at 100,000 years after fuel discharge for all cases.

3.5 Analysis of Optimal Fuel Cycle of the eVinciTM-like Reactor

To analyze its optimal fuel cycle, the Serpent model of the eVinciTM-like special purpose

reactor is used in a burnup calculation until it has a k-effective 1.0 +/- 100 pcm. The configuration

used in this part of the thesis is described in section 3.1, and utilized a fuel enrichment of 19.75%.

With a power output of 25 MWth, the fuel had a discharge burnup of 22.54 GWd/tU and a

residence time of 11.34 years. The analysis in this section followed the specified guidelines found

in the E&S report [6]. Part of the nine high-level “performance metrics” and 25 complementary

specific metrics were evaluated in this study; including the uranium utilization and activity at 100

and 100,000 years of the SNF+HLW [6]. A summary of the performance of the calculate

parameters of the eVinci-like model is shown in Table 3-3. Results for both the SBC and HRBC

are included in this analysis along with the bin classification of each parameter in the respective

criterion evaluated.

40

Table 3-3. Summary of calculated fuel cycle parameters of the eVinci-like concept, the EG01

light water reactor (LWR) reference case, and the EG02 mHTGR case.

1 Note: the water use metric calculation is consistent with the E&S, but based on values generated

for large reactors.

Criterion Metrics eVinci-like eVinci-like EG01 EG02

SBC HRBC LWR mHTGR

Nucl

ear

Was

te M

anag

emen

t

Mass of SNF+HLW disposed, t/GWe-yr 55.1/F 40.2/F 21.9/E 9.22/D

Activity of SNF+HLW (@100 years),

MCi/GWe-yr

1.35/C 0.98/B 1.34/C 1.43/C

Activity of SNF+HLW (@100,000 years),

10-4 MCi/GWe-Yr

31.9/D 23.2/D 16.5/C 20.5/C

Mass of DU+RU+RTh disposed, t/GWe-yr 2276/F 1660/F 167/E 296/F

Volume of LLW, m3/GWe-yr 1308/D 1029/D 399/C 414/C

Res

ourc

e U

tili

zati

on Natural Uranium required per energy

generated, t/GWe-yr

2336/D 1704/D 189/D 306/D

Natural Thorium required per energy

generated, t/GWe-yr

0/A 0/A 0/A 0/A

Envir

onm

enta

l Im

pac

t

Land use per energy generated, km2/GWe-

yr

1.05/E 0.784/D 0.175/B 0.210/C

Water use per energy generated, ML/GWe-

yr1

25920/B 25320/B 23890/B 23990/B

Carbon emission - CO2 released per energy

generated, kt CO2/GWe-yr

336.5/E 247.0/E 44.1/B 54.9/B

41

3.5.1 Nuclear Waste Management

The nuclear waste management metric in the E&S is focused on categorizing the type,

quantity, and characteristics of the wastes produced during a fuel cycle. Three different types of

waste groups are identified in the E&S for and were included in the analysis in this study:

SNF+HLW, depleted+recovered Uranium and Thorium (DU)+(RU)+(RTh) and low-level waste

(LLW). A more in-depth discussion of these metrics can be found in Appendix C of the E&S study

[6].

The SNF+HLW results for both thermal cycles are shown in Table 3-3. They are within

the lowest performance bin for this metric due to the low discharge burnup of the fuel, resulting in

high mass flow demands needed to produce an GWe per energy year. The HRBC results in fewer

quantities of SNF+HLW due to its improved net thermal efficiency. The DU+RU+RTh is the

remaining waste mass after the quantities of the SNF+HLW have been deducted from the NRU

input. Again, the performance of both thermal cycles for this metric is place in the lowest bin due

to the low discharge burnup of the fuel.

Compared to the EG01 and EG02 reference cases, the amounts of LLW produced in both

thermal cycles underperformed for this metric. Due to the high enrichment of the fuel used in

eVinci-like reactor system, high input of separative work (SW) is needed. This enrichment

requirement is the primary contributor to the high amounts of LLW being produced during the fuel

cycle. The highest amount of greater than class C (GTCC) production is produced during normal

reactor operations. For this metric, the concept was assumed to be similar to a sodium fast-cooled

reactor (SFR) system due to the lack of a category including heat pipe reactors in the E&S study.

More on specifications of SFR systems can be found in this reference [30].

42

3.5.2 Natural Resource Utilization

The results of the resource utilization requirements needed to sustain both thermal cycles

can be found in Table 3-3. They are both categorized under the lowest performing bin in the E&S

study, due to the low discharge burnup of the fuel. When compared to the analysis examples, the

natural resource requirements of the heat pipe reactor are about 10 times higher than both that

needed for both the EG01 and EG02 reference cases. This is due to the low fuel utilization of the

eVinci-like core configuration. The resulting values include a 0.2% fabrication loss, as in the

E&S study [6].

3.5.3 Proliferation Risk

For the purpose of this study, only the physics-based technical aspects were evaluated in

the analysis of the heat pipe reactor. The eVinci uses LEU fuel which poses minimal

proliferation risk. Further, the core components in the heat pipe reactor are encapsulated by a steel

monolith. This may further enhance the proliferation resistance of the concept.

3.5.4 Environmental Impacts

The analysis of this performance metric is focused on the environmental impacts and

emissions caused by the input of raw sources required to produce fuel, the normal operations of

the reactor, as well as the outputted waste materials of the fuel cycle. The results for land use and

water use requirements as well as carbon dioxide emissions, all normalized to a GWe per energy

year, are shown in Table 3-3. These environmental impacts were assumed to be similar to those of

the SFR system on a per energy basis. Due to the large amounts of natural uranium required and

the enrichment of the fuel used, the front end of fuel cycle (FEFC) demands, such as mining and

43

milling, contribute to the poor performance of this heat pipe reactor for this metric. Water use per

energy year was average compared to the other EGs, with most of the water demands coming from

the operation of the reactor. Since this metric was assumed to be similar than a large SFR, it is

possible that the actual water usage is lower in the much smaller eVinci-like reactor. Lastly, the

carbon emissions of the heat pipe concept place it in the lowest performing bin for this metric. The

major contributor is the 19.75% U-235 enrichment demands of the fuel, resulting in higher mining

and milling demands for this fuel cycle.

3.6 Impacts of Parasitic Absorption on Fuel Cycle Performance

Parasitic absorption takes place when a neutron is absorbed by another material component,

instead of the fuel components, resulting in an event that does not cause a fission reaction. This

type of absorption negatively impacts the fuel cycle performance of the reactor because it reduces

the quantity of available neutrons for fission. As a consequence, the reactivity of the core is reduced

and the fuel cycle length is shortened.

To quantify the impacts of parasitic absorption on the fuel cycle performance of the heat

pipe reactor concept, the original steel monolith in the core was replace with a Silicon Carbide

(SiC) monolith. From a neutronics point of view the SiC material experiences a low parasitic

absorption rate, resulting in BOL excess reactivity and a lengthening of the fuel cycle, as shown

in Figure 3-6. It also possesses material properties that are favorable at high temperatures.

Ultimately, from the perspective of the thermal and material performance in a reactor a SiC

monolith may or may not be viable. But, this material was chosen to investigate the effects of

parasitic absorption.

44

Figure 3-6. K-effective multiplication factor for monolith modifications done on the eVinci-

like concept.

The migration area (M2) is a measure of the average length between a neutron’s birth point,

as a fast neutron, and its absorption as a thermal neutron. In this part of the study, the migration

area was investigated for both monolith material options (e.g. SS-316 and SiC). Migration area,

given by equation (3.1), is a function of the critical buckling of a reactor core and the neutron

leakage experience by it [31].

M2 =

1

𝑃𝑁𝐿−1

𝐵2 (3.1)

Where 𝑃𝑁𝐿 is the non-leakage probability and B2 is the critical buckling. The results show that a

decrease in parasitic absorption, using the SiC monolith, results in a slight increase in the leakage

of neutrons out of the core. The neutron leakage rates found in the steel and SiC monolith

modifications were 18% and 20.5% respectively. This resulted in a higher migration area in the

SiC monolith at the EOL of the fuel cycle as shown in Figure 3-7.

45

Figure 3-7. Migration area for monolith modifications on the eVinci-like concept.

The impacts of parasitic absorption on the natural resource utilization and activity of the

SNF+HLW at 100 and 100,000 years was also investigated. The decrease in parasitic absorption

in the SiC monolith modification results in higher BOL reactivity in the core. This modification

results in a fuel discharge burnup of about 53.8 GWd/tU; more than double that of the steel

monolith case. As a result of the increased fuel utilization, a reduction in the natural resource

utilization requirements, shown in Table 3-4, and the activity levels of the SNF+HLW at both the

100- and 100,000-year cases, are shown in Figure 3-8 and Figure 3-9, can be achieved for this fuel

cycle.

46

Table 3-4. Natural resource utilization for the parasitic absorption analysis.

Evinci

(25MW)

Reactor

Modification

s

Enrichment

(%)

Fuel

Burnup

(GWd/t)

SBC

Charge

Mass

Flow

(t/Gwe-yr)

HRBC

Charge

Mass Flow

(t/Gwe-yr)

SBC

Natural

Resource

Utilization

(t/Gwe-yr)

HRBC

Natural

Resource

Utilization

(t/Gwe-yr)

eVinci-like

Base Model

(Steel Mon.)

19.8 22.5 55.11 40.20 2335.7 1704.0

eVinci-like

SiC monolith 19.8 53.8 23.1 16.9 979.5 714.6

Figure 3-8. SNF+HLW activity at 100 years after fuel discharge for monolith analysis.

47

Figure 3-9. SNF+HLW activity at 100,000 years after fuel discharge for the monolith

investigation.

3.7 Impacts of Neutron Leakage on Fuel Cycle Performance

The leakage of neutrons out of the core, negatively impacts the fuel cycle performance of

a reactor system. This is because the core experiences a reduction in the quantity of free neutrons

available for fission. Neutron leakage effects were quantified in this study by varying the size of

the eVinci-like reactor core using a 25MWth power case. The number of outer periphery HP

“rings” was varied, in increments of 5, from a reference case of 20 to 30 resulting in leakage values

of 18%, 12.8% and 9.6%, respectively. The amount of fissile material in the cores of all these

configurations were kept the same, and therefore the fuel density in the core decreased as the

number of outer periphery HPs was increased. These decreased fuel densities would result in lower

initial excess reactivities and higher leakage rates, for the 25 and 30 outer periphery HP

configurations, than what is actually expected had the density been kept constant. This is because

a lower fuel density increases the chances of neutron’s leakage out of the core, as a result of a

48

decreased capture cross-section value of the fuel material. However, the goal of this section was

to determine how different leakage rates impact fuel cycle performance.

As core leakage decreases, there is an increase in the BOL excess reactivity resulting in a

longer fuel cycle length, as shown in Figure 3-10. This is because the negative effects of the

geometrical buckling decrease, as the leakage of a core is reduced, allowing the EOL critical

buckling to reach a smaller value and discharge fuel burnup to increase. Lastly, as shown in Figure

3-11, the migration area has a proportional relationship with leakage as indicated by equation 3.1.

The similar EOL migration area is a result of the same material ratios, regardless of size, in the

core for all cases.

Figure 3-10. K-effective multiplication factor for leakage modifications done on the eVinci-

like concept.

49

Figure 3-11. Migration area for leakage modifications on the eVinci-like concept.

The impact of neutron leakage on the natural resource utilization, shown in Table 3-5, and

the SNF+HLW radioactivity at 100 and 100,000 years, shown in Figure 3-12 and Figure 3-13, was

also investigated. As neutron leakage decreases the fuel discharge burnup increases. This results

in a decrease in the natural resource utilization and the activity levels for both cases due to a lower

mass flow require to sustain the heat pipe’s fuel cycle.

50

Table 3-5. Natural resource utilization for the neutron leakage investigation.

Evinci

(25MW)

Reactor

Modification

s

Leakage

(%)

Fuel

Burnup

(GWd/t)

SBC

Charge

Mass

Flow

(t/Gwe-yr)

HRBC

Charge

Mass Flow

(t/Gwe-yr)

SBC

Natural

Resource

Utilization

(t/Gwe-yr)

HRBC

Natural

Resource

Utilization

(t/Gwe-yr)

eVinci-like

Base Model

20HP

18.0 22.5 55.1 40.2 2335.7 1704.0

eVinci-like

25HP model 12.8 76.1 16.3 11.9 691.6 504.5

eVinci-like

30HP model 9.7 110.0 11.3 8.2 478.5 349.1

Figure 3-12. SNF+HLW activity at 100 years after fuel discharge for the neutron leakage

investigation.

51

Figure 3-13. SNF+HLW activity at 100,000 years after fuel discharge for the neutron leakage

investigation.

52

4. Summary

Small modular reactors have the potential to revolutionize the nuclear industry, and allow

it to continue being a good energy option capable of meeting the nation’s energy demands into the

future. The fuel cycle analysis of the VVER-based SMRs showed that core leakage negatively

impacts the natural resource requirements as well as the SNF+HLW activity at 100 years

normalized to a GWe per energy year basis. Further, the U-235 enrichment of the fuel and the

rated core power density were concluded to have minimal fuel cycle impacts on these thermal

spectrum SMRs. The only mentionable impact being the slightly higher discharge burnup of the

de-rated power cases resulting in somewhat different activity levels of the SNF+HLW at 100 and

100,000 years after fuel discharge. All the enrichment and power density cases resulted in

decreased radioactivity at 100,000 years relative to the EG02 reference case from the E&S study.

The results of the special purpose reactor’s fuel cycle analysis supported Westinghouse’s

claim, that the eVinciTM

heat pipe reactor concept is designed to operate similar to a nuclear battery

and not a large-scale power source. When normalized to a GWe per energy year basis, the

eVinciTM

-like reactor performed poorly compared to other fuel cycle options, as shown with the

resulting natural resource utilization of 9650 t/GWe-yr for the 5MW HRBC case. This also resulted

in SNF+HLW radioactivity levels of 1.00E+06 and 3.37E+03 at 100 and 100,000 years

respectively, for this same HRBC thermal cycle and power level. The use of LEU, parasitic

absorption effects and core neutron leakage, resulted in a very low fuel discharge burnup and

subsequently a high charge mass flow requirement needed to sustain the eVinciTM

-like reactor’s

fuel cycle. This resulted in a high natural resource utilization normalized to a GWe-year. An

analysis of the waste metrics of the fast spectrum reactor concluded that high amounts of

53

radioactive waste will be produced as a result of its fuel cycle being scaled to energy outputs

comparable of the current U.S commercial fuel cycle. All cases exceed a SNF+HLW radioactivity

level of 1.00E+06 and 2.5E+03 Ci/GWe-yr at 100 years and 100,000 years after fuel discharge

respectively.

Parasitic absorption was investigated by replacing the steel monolith with a much more

“neutron transparent” SiC monolith material. The SiC monolith configuration resulted in a

resource utilization of about 715 t/GWe-yr. The results showed a significant decrease in the natural

resource utilization due to a longer achievable fuel cycle length. Additionally, a reduction in the

SNF+HLW activity at 100 and 100,000 years, 8.30E+06 and 2.14E+03 Ci/GWe-yr for the most

effective thermal cycle, was observed due to the use of a SiC monolith. Neutron leakage was

investigated by varying the number of outer periphery heat pipes, and thus the size of the core. A

significant reduction in the natural resource utilization and waste outputs of the fuel cycle, can be

achieved by decreasing the neutron leakage of the core from 18.0% to about 9.7%. This reduction

in neutron leakage resulted in a natural resource utilization of 349 t/Gwe-yr. As a result of this

reduction in input and output fuel cycle requirements, less environmental impacts can be expected.

In both studies of these thermal and fast spectrum SMRs, the importance of core neutron leakage

on fuel cycle performance was highlighted. It was determined, that neutron leakage was the most

negatively impacting factor to the performance of these SMRs. Improvements of the negative

effects as a result of neutrons leaking out of the core of these SMRs, will result in a decrease in

fuel cycle cost. Additionally, the environmental impacts as a result of the operation of these

modular reactors will be reduced. Thus, making SMRs a favorable option capable of taking the

nuclear industry into the future.

54

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