Review and Evaluation of 

the Nuclear Regulatory 

Commission Safety 

Research Program 

A Report to the

U.S. Nuclear Regulatory 

Commission 

Advisory Committee on Reactor Safeguards 

NUREG-1635

Vol. 9
Review and Evaluation of

the Nuclear Regulatory 

Commission Safety 

Research Program 

A Report to the

U.S. Nuclear Regulatory

Commission 

Manuscript Completed: April 2010 

Date Published: June 2010 

Advisory Committee on Reactor Safeguards 

NUREG-1635

Vol. 9
ii
iii

ABSTRACT 

This report to the U.S. Nuclear Regulatory Commission (NRC) presents the observations and 

recommendations of the Advisory Committee on Reactor Safeguards (ACRS) concerning the NRC 

Safety Research Program being carried out by the Office of Nuclear Regulatory Research. In its 

evaluation of the NRC research activities, the ACRS considered the programmatic justification for 

the research as well as the technical approaches and progress of the work. The evaluation 

identifies research crucial to the NRC mission. This report does not address the research being 

done at NRC on issues of reactor security or the threat of sabotage. The ACRS views on current 

work in the area of security have been reported in separate documents. Two pertinent, 

interdisciplinary efforts, the State-of-the-Art Reactor Consequence Analyses (SOARCA) Project and 

the study of sump screen blockage are not addressed in this report. These projects are actively 

followed by the Committee. The ACRS has been providing interim reports on the technical 

approach and progress of these activities. 

***** 

Paperwork Reduction Act Statement 

This NUREG does not contain information collection requirements and, therefore, is not subject to 

the requirements of the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). 

Public Protection Notification 

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for 

information or an information collection requirement unless the requesting document displays a 

currently valid OMB control number. 
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v

TABLE OF CONTENTS 

ABSTRACT..................................................................................................................iii 

FIGURES...................................................................................................................vi 

ABBREVIATIONS.......................................................................................................vii 

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

2. GENERAL OBSERVATIONS AND RECOMMENDATIONS..........................................3 

3. LIFE BEYOND 60................................................................................................9 

4. ADVANCED NON-LWR DESIGNS........................................................................13 

5. DIGITAL INSTRUMENTATION AND CONTROL SYSTEMS......................................15 

6. FIRE SAFETY..................................................................................................17 

7. REACTOR FUEL..............................................................................................23 

8. HUMAN FACTORS AND HUMAN RELIABILITY......................................................29 

9. MATERIALS AND METALLURGY.........................................................................33 

10. NEUTRONICS AND CRITICALITY SAFETY .....................................................................39 

11. OPERATIONAL EXPERIENCE .........................................................................43 

12. PROBABILISTIC RISK ASSESSMENT..................................................................45 

13. RADIATION PROTECTION ...............................................................................51 

14. NUCLEAR MATERIALS AND WASTE ................................................................55 

15. SEISMIC AND STRUCTURAL ENGINEERING ...................................................... 57 

16. SEVERE ACCIDENTS AND SOURCE TERM ................................................... 59 

17. THERMAL HYDRAULICS ................................................................................. 65 

18. REFERENCES...................................................................................................................73 
vi

FIGURES 

1. Current NRC Research Activities in Fire Safety ................................................................... 21

2. Current NRC Research Activities in Reactor Fuel .........................................................27 

3. Current NRC Research Activities in Human Factors and Human Reliability...........................31 

4. PWSCC Safety Assessment Elements.......................................................................38 

5. Current NRC Research Activities in Neutronics and Criticality Safety..............................41 

6. Current NRC Research Activities in Operational Experience................................................44 

7. Current NRC Research Activities in PRA..................................................................49 

8. Current NRC Research Activities in Radiation Protection.............................................53 

9. Current NRC Research Activities in Severe Accident and Source Term...........................63 

10. Current NRC Research Activities in Thermal-Hydraulics Research ................................71 
vii

ABBREVIATIONS 

ABWR Advanced Boiling Water Reactor 

ACRS Advisory Committee on Reactor Safeguards 

ACWI Advisory Committee on Water Information 

AECL Atomic Energy of Canada Limited 

AFCI Advanced Fuel Cycle Initiative 

AIChE American Institute of Chemical Engineers 

ALARA As Low as Reasonably Achievable 

AMPs Aging Management Programs 

ANL Argonne National Laboratory 

ANS American Nuclear Society 

AOOs Anticipated Operational Occurrences 

APWR Advanced Pressurized Water Reactor 

ASME American Society of Mechanical Engineers 

ASP Accident Sequence Precursor 

ASTM American Society for Testing and Materials 

ATHEANA A Technique for Human Event Analysis 

ATWS Anticipated Transient without Scram 

BIP Behavior of Iodine Project 

BWR Boiling Water Reactor 

CAROLFIRE Cable Response to Live Fire 

CCI Core Concrete Interaction 

CDF Core Damage Frequency 

CFD Computational Fluid Dynamics 

CFR Code of Federal Regulations 

CHRISTI-FIRE Cable Heat Release, Ignition, and Spread in Tray Installations during 

FIRE 

COL Combined License 

CRDM Control Rod Drive Mechanism 

CSARP Cooperative Severe Accident Research Program 

DI&C Digital Instrumentation and Control 

DOE Department of Energy 

EAC Environmentally Assisted Cracking 

EPAct Energy Policy Act 

EPIX Equipment Performance and Information Exchange System 

EPR Evolutionary Power Reactor 

EPRI Electric Power Research Institute 

ESBWR Economic Simplified Boiling Water Reactor 

FDP Fitness for Duty Programs 

FSME Office of Federal and State Materials and Environmental Management 

Programs 

GALL Generic Aging Lessons Learned 

GDC General Design Criteria 

GEH General Electric-Hitachi 

GNEP Global Nuclear Energy Partnership 

GSI Generic Safety Issue 

GWd/t Giga Watt day per metric ton 

HAMMLAB Halden Man-Machine Laboratory 

HDPE High Density Polyethylene 
viii

ABBREVIATIONS (Cont’d) 

HERA Human Event Repository and Analyses 

HF Human Factor 

HFE Human Factors Engineering 

HRA Human Reliability Analysis 

HRP Halden Reactor Project 

HSI Human System Interface 

HTGR High Temperature Gas-cooled Reactor 

I&C Instrumentation and Control 

IAEA International Atomic Energy Agency 

IASCC Irradiation Assisted Stress Corrosion Cracking 

ICRP International Commission on Radiological Protection, 

IEEE Institute of Electrical and Electronics Engineers 

IGSCC Intergranular Stress Corrosion Cracking 

INPO Institute of Nuclear Power Operations 

ISG Interim Staff Guidance 

ISI In-Service Inspection 

JAEA Japan Atomic Energy Agency 

JNES Japan Nuclear Energy Safety Organization 

LANL Los Alamos National Laboratory 

LBB Leak before Break 

LBLOCA Large-Break Loss-of-Coolant Accident 

LER Licensee Event Report 

LERF Large Early Release Frequency 

LSTF Large Scale Test Facility 

LOCA Loss-of-Coolant Accident 

LWR Light Water Reactor 

MACCS MELCOR Accident Consequence Code System 

MASLWR Multi-Application Small Light Water Reactor 

MCCI Molten Core Concrete Interaction 

MCNP Monte Carlo N-Particle Transport Code System 

MDM Materials Degradation Matrix 

MeV Million Electron Volts 

MOU Memorandum of Understanding 

MOX Mixed Oxide 

NAS National Academy of Sciences 

NASA National Aeronautics and Space Administration 

NCRP National Council on Radiation Protection and Measurements 

NDE Non-Destructive Examination 

NEA Nuclear Energy Agency 

NEI Nuclear Energy Institute 

NFPA National Fire Protection Association 

NGNP Next Generation Nuclear Plant 

NIST National Institute of Standards and Technology 

NMSS Office of Nuclear Material Safety and Safeguards 

NOAA National Oceanic and Atmospheric Administration 

NPPs Nuclear Power Plants 

NRC Nuclear Regulatory Commission 

NRO Office of New Reactors 
ix

ABBREVIATIONS (Cont’d) 

NRR Office of Nuclear Reactor Regulation 

NSIR Office of Nuclear Security and Incident Response 

NUPEC Nuclear Power Engineering Corporation 

OECD Organization for Economic Cooperation and Development 

OSU Oregon State University 

PARCS Purdue Advanced Reactor Core Simulator 

PCI Pellet Cladding Interaction 

PCT Peak Clad Temperature 

PINC Program for the Inspection of Nickel Alloy Components 

PIRT Phenomena Identification and Ranking Table 

PMMD Proactive Management of Materials Degradation 

PNNL Pacific Northwest National Laboratory 

PRA Probabilistic Risk Assessment 

PSA Probabilistic Safety Assessment 

PSHA Probabilistic Seismic Hazard Analysis 

PTS Pressurized Thermal Shock 

PSI Paul Scherrer Institute 

PUMA Purdue University Multidimensional Integral Test Assembly 

PUREX Plutonium Uranium Extraction 

PWR Pressurized Water Reactor 

PWSCC Primary Water Stress Corrosion Cracking 

R&D Research and Development 

RBHT Rod Bundle Heat Transfer 

RCS Reactor Coolant System 

RES Office of Nuclear Regulatory Research 

RG Regulatory Guide 

ROP Reactor Oversight Process 

RPV Reactor Pressure Vessel 

RTF Radioiodine Test Facility 

SCALE Standardized Computer Analysis for Licensing Evaluation 

SDP Significance Determination Process 

SERENA Steam Explosion Resolution for Nuclear Applications 

SGTR Steam Generator Tube Rupture 

SNL Sandia National Laboratories 

SOARCA State-Of-the-Art Reactor Consequence Analyses 

SPAR Standardized Plant Analysis Risk Model 

SRM Staff Requirements Memorandum 

SRP Standard Review Plan 

SSCs Structures, Systems, and Components 

SSE Safe Shutdown Earthquake 

SSHAC Senior Seismic Hazard Analysis Committee 

SSWICS Small-scale Water Ingression and Crust Strength 

THERP Technique for Human Error Rate Prediction 

THIEF THermally Induced Electrical Failure 

TRACE TRAC-RELAP Advanced Computational Engine 

TSUNAMI Tools for Sensitivity and UNcertainty Analysis Methodology Implementation 

UREX Uranium Extraction 

U.S. United States 
x

ABBREVIATIONS (Cont’d) 

USGS United States Geological Survey 

V&V Verification & Validation 

VHTR Very-High-Temperature Reactor 

WRS Weld Residual Stress 

xLPR Extremely Low Probability of Rupture 
1

I. INTRODUCTION 

In this report, the Advisory Committee on 

Reactor Safeguards (ACRS) presents the 

results of its review and evaluation of the 

Nuclear Regulatory Commission (NRC) 

Safety Research Program. The NRC 

maintains a Safety Research Program to: 

• Ensure its regulations and regulatory 

processes have sound technical bases 

and these bases are refined as new 

knowledge develops. 

• Prepare for anticipated changes in the 

nuclear industry that could have safety 

implications. 

• Develop improved methods to carry out 

its regulatory responsibilities. 

• Maintain an infrastructure of expertise, 

facilities, analytical capabilities, and data 

to support regulatory decisions. 

The current research program, organized by 

the Office of Nuclear Regulatory Research 

(RES), is closely coupled to specific, nearterm issues to support regulatory activities 

and initiatives in the Offices of Nuclear 

Reactor Regulation (NRR), New Reactors 

(NRO), Nuclear Material Safety and 

Safeguards (NMSS), Federal and State 

Materials and Environmental Management 

Programs (FSME), and Nuclear Security and 

Incident Response (NSIR). For the purposes 

of this report, the ongoing research has been 

examined in terms of the following technical 

disciplines: 

• Advanced Non-LWR Designs 

• Digital Instrumentation and Control 

Systems 

• Fire Safety 

• Reactor Fuel 

• Human Factors and Human Reliability 

• Materials and Metallurgy 

• Neutronics and Criticality Safety 

• Operational Experience 

• Probabilistic Risk Assessment 

• Radiation Protection 

• Nuclear Materials and Waste 

• Seismic and Structural Engineering 

• Severe Accidents and Source Term 

• Thermal Hydraulics 

This report does not address the research 

being done at NRC on issues of reactor 

security or the threat of sabotage. The ACRS 

views on current work in the area of security 

have been provided in separate letter reports. 

Two pertinent, interdisciplinary efforts, the 

State-Of-the-Art Reactor Consequence 

Analyses (SOARCA) Project and the study of 

sump screen blockage are not addressed in 

this report. These projects are actively 

followed by the Committee. The ACRS has 

been providing interim reports on the 

technical approach and progress of these 

activities. 

 

Chapter 2 of this report provides a synoptic 

account of research activities in each of the 

technical disciplines and highlights some of 

the accomplishments of the work. Additional 

details on each of the research areas are 

included in Chapters 4 through 17. 

In its review of the NRC Safety Research 

Program, the ACRS has focused on the 
2

technical and regulatory justification for the 

ongoing research activities. The ACRS 

supports research that: 

• Identifies and resolves current safety and 

regulatory issues. 

• Provides technical bases for the 

resolution of foreseeable safety issues. 

• Develops the capabilities of the agency to 

independently review risk-significant 

proposals and submittals by licensees 

and applicants. 

• Supports agency initiatives, including the 

move toward a much greater use of risk 

information in the regulatory process and 

to evolve NRC safety regulations to be 

“technology neutral”. 

• Improves the efficiency and effectiveness 

of the regulatory process. 

• Maintains technical expertise within the 

agency and associated facilities in 

disciplines crucial to the agency mission 

and that are not readily available from 

other sources. 

The research to support reactor license 

renewal beyond 60 years is discussed in 

Chapter 3. In that Chapter, the ACRS notes 

that much of the current research for “life 

beyond 60” is focused on known degradation 

and analysis methods to assure processes 

currently in place to manage aging beyond 40 

years remain valid beyond 60 years. ACRS 

identifies several additional research areas 

the nuclear industry needs to pursue. ACRS 

notes that NRC will have to follow those 

industry-led research efforts and integrate 

confirmatory research with these efforts. 
3

2. GENERAL OBSERVATIONS AND RECOMMENDATIONS 

General Observations Concerning the 

Organization of the NRC Safety Research 

Program 

Activities within the NRC Safety Research 

Program are very closely tied to needs of the 

line organizations (NRR, NRO, NMSS, FSME, 

and NSIR) as identified through “User Need” 

documents. This strategy for the identification 

of research undertakings together with the 

associated process for the prioritization of 

research needs is working well. Research 

activities are yielding products that are useful 

to the line organizations and they are doing 

so on a timely basis. 

A risk that could arise within such a strategy 

for the organization of research is that longterm, more speculative research would not be 

initiated. The Office of Nuclear Regulatory 

Research does mitigate this risk. Limited 

duration studies of the feasibility and potential 

of more speculative research are funded. 

Pursuit of the research, once feasibility has 

been established, does require the 

endorsement and support by the affected line 

organization. 

A consistent feature of the current research 

program is enthusiastic support of the 

activities by the affected line organization. 

Line organizations appear to have close, 

continuing involvement in the conduct and the 

direction of the research. It has become 

common for presentations of research results 

to the ACRS to include statements of support 

from line organizations and even 

presentations on the uses to be made by line 

organizations of research results. 

Collaborations in the Conduct of Research 

Major research activities at NRC often include 

collaboration with other Federal agencies, 

industrial institutions, and international 

partners. NRC is taking good advantage of 

opportunities to leverage its resources and 

expertise on issues of common interest to 

others. The collaboration is providing high 

quality peer input to the organization and 

conduct of research. 

In the past, in-depth collaborations with 

international partners were characteristic of 

research in only a few technical areas. Such 

collaborations have become much more 

common. ACRS endorses such collaborations 

on both the conduct of research and the 

analysis of results. International 

collaborations will become ever more 

important as the nuclear industry becomes 

more international in nature. Collaborations in 

research will help in keeping safety regulation 

of nuclear power cohesive throughout the 

developed world. 

Research Planning 

There is much more emphasis on the 

detailed, documented planning of research 

now than in the past. In many areas, 

research plans are being subjected to 

extensive peer review by the broader nuclear 

community both by design and as a 

consequence of extended collaborations with 

other elements of this larger community. This 

detailed planning of research has greatly 

helped in the conduct and quality of the 

research. 

Documentation of Research Results 

RES is subjecting its research results to more 

peer review. Documentation of the final 

results is variable. In some cases, 

documentation is outstanding and the results 

are readily retrieved by interested parties 

outside the agency. In other cases, 

documentation is limited to reports readily 

identified and retrieved only within NRC. 
4

Areas Deserving Attention in the Future 

There is a growing emphasis on the use of 

numerical simulation to resolve issues of 

reactor safety. Some particular aspects of this 

trend are noted in our comments below on 

individual technical areas of research. The 

Department of Energy (DOE) is undertaking a 

major computational initiative in support of the 

existing nuclear reactors. DOE seeks to 

establish what it calls high fidelity simulation 

of the thermal hydraulics, neutronics, and 

mechanics of nuclear power plants during 

normal operations and anticipated operational 

occurrences (AOOs). The effort is centered 

at the DOE national laboratories but both 

licensees and nuclear steam suppliers are 

participating in this initiative. It is, then, likely 

that products of the initiative may well be 

utilized in submissions made to the NRC for 

regulatory purposes. NRC needs to be in a 

position to evaluate products of massively 

parallel computations. 

The growth in the use of computations to 

resolve reactor safety issues has not been 

accompanied by a similar growth in the 

development of experimental facilities 

capable of yielding data suitable for the 

validation of computational predictions. 

Indeed, experimental facilities within the U.S. 

available to the NRC for obtaining validation 

data continue to dwindle. There is some 

confidence that adequate experimental 

facilities remain in the rest of the world and 

that these facilities are available to NRC 

through formal or informal collaborations. 

NRC is taking advantage of these foreign 

capabilities in areas of research such as 

severe accidents, materials, and seismic 

research. The process is not, however, 

seamless. Greater availability of test reactors 

could speed the cost-effective resolution of 

safety issues. Lack of hot cells within this 

country and the challenges of transportation 

do handicap the post test examination of 

irradiated specimens. 

NRC is moving toward greater use of 

probabilistic assessments of risk in its 

regulatory processes. NRC relies on methods 

of probabilistic risk assessment developed 

decades in the past. There is an interest 

within the larger risk analysis community of 

developing modernized tools with superior 

capabilities for risk analysis. This 

modernization is in addition to the expansion 

of scope of risk analysis to include greater 

breadth by including seismic risk, fire risk, and 

risk during shutdown operations as well as 

greater depth to Levels II and III. Of particular 

interest is the development in modernized risk 

assessment tools of superior risk metrics that 

can be used to focus and guide NRC 

functions of monitoring and inspection. NRC 

needs to be involved in research to advance 

the development of tools for the conduct of 

risk analyses. 

Major Observations on Individual Areas of 

Research 

Major observations, conclusions, and 

recommendations concerning specific 

research activities are summarized below. 

Additional details on the research activities in 

the various technical disciplines are provided 

in Chapters 4 through 17. 

Advanced Non-LWR Designs

The first research priority in the area of 

advanced non-LWR designs should be to 

support the advanced thermal-spectrum, 

high-temperature, gas-cooled reactor (HTGR) 

consistent with the 2005 Energy Policy Act 

(EPAct 2005). Plans now being developed by 

the NRC staff for this research seem to be 

well founded. A low priority should be given 

to monitor the DOE activities with respect to 

the Advanced Fuel Cycle Initiative (AFCI). 

Digital Instrumentation and Control Systems

The staff has issued a new plan for its 

research on safety issues associated with the 

transition to software-based digital 

instrumentation and control for safety 

systems. The Steering Panel formed for this 

work should assist in keeping research 
5

focused on products that will be of use to the 

line organizations. 

Fire Safety

The NRC's fire safety research program is at 

the forefront of a rapidly evolving 

understanding of fire events and the 

assessment of risks from fire damage in 

nuclear power plants. Structured 

collaboration with industry has provided costeffective solutions and technical insights that 

surpass independent efforts. The NRC is a 

recognized leader in national and 

international fire safety research. 

New reactor designs rely heavily on 

integrated digital instrumentation, protection, 

and control systems. Extensions of operating 

licenses and obsolescence of analog 

equipment will also likely lead to replacement 

of many currently installed instrumentation 

and control (I&C) systems with digital 

platforms. Limited information is available 

regarding the effects from fire, heat, and 

smoke on digital equipment. Research 

should address such issues as the effects of 

fire and heat on fiber optic cables, the effects 

of heat on digital equipment, and smoke 

damage to digital signal processing and 

computation modules. 

Reactor Fuel 

There is ample capability to assess risks of 

failure of current fuels due to centerline 

melting or departure from nucleate boiling 

events. The NRC does not have empirical or 

analytical capability to quantitatively assess 

the risks of Pellet Cladding Interaction (PCI) 

fuel failures during AOOs on current or future 

fuel designs. 

Advanced fuel designs are in development 

and it is imperative that NRC maintain a core 

capability to analyze the safety of new fuel 

designs. Of particular concern is the source of 

experimental data on advanced fuel and 

cladding behavior under conditions of design 

basis and beyond design basis accidents. 

Human Factors and Human Reliability

The ACRS finds the NRC’s Human Factors 

(HF)/Human Reliability Analysis (HRA) 

research program and activities to be based 

on sound rationale and to be focused on both 

near- and long-term needs of the agency. The 

research program is yielding products of use 

to the regulatory process. 

Materials and Metallurgy

The ACRS enthusiastically supported the 

initial vision of the Proactive Materials 

Degradation Assessment Program and has 

recommended its continuation. However, the 

program seems to have lost its momentum 

and focus. Current research appears to 

concentrate on conventional reactive 

management instead of anticipating and 

preventing materials degradation surprises. 

The work, published in Expert Panel Report 

on Proactive Materials Degradation 

Assessment (NUREG/CR-6923), identified 

target items of low knowledge and high-tomedium susceptibility to degradation. These 

target items were essentially predictions of 

possible materials degradation problems 

which had yet to occur. They represented 

areas of greatest uncertainty and concern to 

the experts involved in the identification 

process. The next step in the proactive 

process should be the validation of one or 

more of these predictions by experimental 

means or by focused inspections at operating 

plants. In the absence of a quantitative 

confirmation of the predictions made in 

NUREG/CR-6923, it is doubtful that the 

program can attain its primary goal. If such 

confirmations are not possible, continuation of 

the program should be reconsidered. 

The effectiveness of the various materials, 

fabrication, and water chemistry changes 

introduced to mitigate environmentally 

assisted cracking in Boiling Water Reactors 

(BWRs) and Pressurized Water Reactors 

(PWRs) should be confirmed by long-term 

and aggressive confirmatory testing. Some 

mitigation methods may lose effectiveness 
6

over time, while others (individually or in 

combination) may demonstrate the ability to 

protect critical components from stress 

corrosion cracking for the 60 year life of the 

uprated plants. 

With the closure of the Steam Generator 

Action Plan, the vast amount of knowledge 

gained during this long term research 

program should be thoroughly documented. 

Neutronics and Criticality Safety

NRC needs to establish a long term strategy 

to upgrade its neutronics analysis capabilities 

to take advantage of modern computer 

architecture in a way that facilitates the 

analysis of low symmetry reactor core 

designs including quantitative analysis of 

parametric uncertainty. 

A move to reactors based on technologies 

other than light water is not imminent. There 

are, however, DOE initiatives in both gascooled reactors and sodium-cooled reactors 

that will challenge NRC neutronics analyses 

capabilities. Some of the challenges are 

associated with data bases. NRC can 

encourage the DOE and foreign counterparts 

to cooperate in the development of data 

bases adequate for safety analyses of 

advanced reactor concepts and reactors 

using higher fuel enrichments and new fuel 

compositions such as mixed oxide (MOX) 

fuels enriched in other actinides. 

Operational Experience

The Operational Experience Research 

program is being managed and executed in a 

manner consistent with the needs of the NRC. 

It provides data and tools necessary for 

regulatory decision making and for the 

assessment of regulatory effectiveness. 

Probabilistic Risk Assessment

It has been nearly twenty years since NRC 

sponsored the NUREG-1150 Study, “Severe 

Accident Risks: An Assessment for Five U.S. 

Nuclear Power Plants.” It is time to begin 

planning for a new NRC Level 3 probabilistic 

risk assessment (PRA) to demonstrate 

advances in PRA methods and suitable 

approaches for new reactors. 

There is a need for the NRC to develop a 

common approach for the elicitation of expert 

opinions in areas of regulatory interest where 

knowledge is incomplete. Common 

approaches for the treatment of model and 

parametric uncertainty are also needed. 

PRAs contain a wealth of information 

regarding the ways undesirable plant states, 

such as core damage and large release of 

radioactivity, can occur. This information is 

not utilized in formulating security 

requirements to evaluate their benefit and 

their impact on safety. The ACRS 

recommends that RES establish a research 

project to explore the possibility of riskinforming security requirements and building 

on PRAs to create a unified framework for the 

evaluation of both safety and security. 

Radiation Protection

The staff has developed an appropriate and 

robust research program in the areas of 

radiation protection. This program includes 

radiation protection of workers and 

radiological assessments related to 

decommissioning and waste management. 

Nuclear Materials and Waste

The NRC has been successful in leveraging 

its nuclear materials and waste research 

efforts with those of other federal agencies 

and other organizations. These collaborations 

are productive and provide significant synergy 

with NRC research activities. 

Seismic and Structural Engineering 

Efforts by the NRC staff to revitalize the 

seismic safety research in support of 

regulatory activities have been exemplary. 

There is a well developed research plan that 
7

has been broadly reviewed for both technical 

quality and programmatic impact. Credentials 

and expertise of those pursuing the research 

are quite impressive. The research is having 

impact on the regulatory processes and is 

expected to have even greater impact in the 

future. The active research involving 

collaborations with the larger seismic 

community is assuring the agency has at 

hand expertise to help resolve novel issues 

that arise as new plant designs are submitted 

for certification and new license applications 

are submitted for staff review. 

Severe Accidents and Source Term

The ACRS supports the overall strategy that 

the NRC staff has developed to support 

regulatory decisions for severe accidents 

through computer code and associated model 

development such as the MELCOR code and 

targeted experimental data analysis and 

evaluation. This approach can successfully 

maintain and update NRC modeling 

capabilities for severe accident safety 

analyses for light water reactors and can set 

the stage for advanced reactor accident 

analysis. The NRC approach to leverage 

resources by international experimental 

collaborations is notable and should be 

continued. The planned program extensions 

and experimental collaborations are well 

worth the investment. 

Thermal Hydraulics

The RES staff is to be commended for the 

progress that has been made in developing 

and moving forward with incorporation of 

TRACE into the regulatory process. Much 

work remains to be done to enable its reliable 

use for the analysis of the new LWR designs, 

an urgent matter which should be conducted 

with high priority. The priorities for further 

development of TRACE require careful 

evaluation.

The international collaborative efforts are also 

to be commended, as they take advantage of 

facilities that are of a scale and capability that 

do not currently exist in the U.S. 

Furthermore, they draw on the expertise of 

international partners, who have continued to 

maintain a very high level of capability in the 

thermal-hydraulics field. However, 

complementary development of national 

facilities to address safety related thermal 

hydraulics issues should be seriously 

considered. Such facilities would enable the 

retention of U. S. expertise and provide the 

capability to conduct experiments for 

supporting confirmatory thermal-hydraulic 

analyses of new reactor designs. 

NRC currently has only a modest effort in the 

area of computational fluid dynamics (CFD) 

and it is limited to some use of commercial 

CFD codes. Although commercial CFD 

codes are used in the process industry for 

qualitative indications of phenomena, they are 

validated to a much less rigorous standard 

than codes for nuclear use, and the source 

codes are not available. They appear to 

include a number of ad hoc fixes to improve 

stability and robustness which may affect their 

predictive capability for situations that cannot 

be studied experimentally. 

It is inevitable that the licensees will 

increasingly capitalize on the extraordinary 

advances in computing power and 

computational science to resolve problems 

which the current generations of thermalhydraulic codes such as TRACE are unable 

to do. A recent LWR Sustainability Research 

and Development (R&D) Program, prepared 

by DOE in close collaboration with industry 

R&D programs, considers features such as a 

next-generation system analysis code as 

essential components to improve 

understanding and utilization of safety 

margins. NRC thermal-hydraulic research has 

to position the agency to address the 

research coming from the LWR Sustainability 

Program. 
8
9

3. LIFE BEYOND 60 

The NRC position on license renewal is 

based on: 

(1) a policy decision that if plants are safe 

enough to operate currently, maintaining 

that licensing basis is sufficient for future 

operation and 

(2) a technical position that current 

regulations (especially 10CFR 50.65, the 

maintenance rule) and operating 

procedures are sufficient to manage 

degradation in active systems and 

components and that license renewal 

should focus on the aging management of 

passive, long-lived, safety-related 

components. 

There are no indications at present that the 

NRC licensing position will or should change. 

From this perspective, the critical research 

required for license renewal beyond 60 years 

relates to aging management of passive 

components and structures, namely the 

reactor coolant system, the reactor internals, 

concrete, cables, and support system 

components such as buried piping. 

In accordance with Title 10 of the Code of 

Federal Regulations (10 CFR) Part 54, 

“Requirements for Renewal of Operating 

Licenses for Nuclear Power Plants,” those 

plants that enter into a period of extended 

operation are required to have in place aging 

management programs (AMPs) that mitigate 

aging degradation effects during the period of 

continued operations. As a result of the 

engineering margins incorporated into the 

design of US nuclear reactors and AMPs, the 

plants’ structures, systems, and components 

(SSCs) inherently afford a measure of 

protection against the deleterious effects of 

aging. For those plants that may elect to 

pursue subsequent license renewal periods 

beyond the 60 year operating period, the staff 

is investigating whether additional research 

into aging effects is needed. 

The potential for new modes of degradations 

and gaps in our understanding of currently 

recognized modes of degradation has been 

recognized. In 2004, the Electric Power 

Research Institute (EPRI), through an expert 

opinion elicitation process, developed a 

Materials Degradation Matrix (MDM), which 

identified these potential mechanisms and 

provided the status of current research 

progress and needs. In addition, the NRC 

has developed the Proactive Management of 

Materials Degradation (PMMD) program to 

address the regulatory issues involved with 

previously unidentified materials aging 

processes, and the Generic Aging Lessons 

Learned (GALL) report covering aging issues 

associated with Part 54 reviews. However, 

the scope of the assessments performed for 

the PMMD program were more limited in 

scope than is required for an assessment in 

terms of 10 CFR Part 54 and did not include 

for example degradation of cables, concrete, 

and buried piping. 

In order to obtain a broader range of inputs, 

the NRC and DOE jointly sponsored a public 

workshop on February 19–21, 2008, in 

Bethesda, MD, to discuss potential research 

and development issues related to ensuring 

that, if licensees elect to pursue subsequent 

license renewal periods, continued long-term 

operation can be conducted safely. Panel 

and public discussions were held on: systems 

structures and components, materials 

degradation issues, diagnostics and 

prognostic technologies, and the future needs 

(beyond the scope of 10 CFR Part 54) of the 

nuclear industry to continue long-term 

operation. Participants included 

representatives from the NRC, DOE, industry, 
10

national laboratories, academia, the 

International Atomic Energy Agency, the 

public, and international organizations. 

The workshop participants recognized 

radiation embrittlement of reactor pressure 

vessels as a fundamental issue for life 

extension beyond 60 years. Based on our 

current understanding of embrittlement 

mechanisms, recent studies by the NRC on 

pressurized thermal shock (PTS) indicate that 

margins are considerably larger than 

expected and most vessels, being operated 

as they currently are, should probably be able 

to operate past 60 years. It also appears that 

there is sufficient margin on upper shelf 

toughness to justify extended operation. 

To assess the need for additional work it is 

important not only to consider the level of our 

scientific understanding of the nature of 

possible aging mechanisms, but also the 

adequacy of current engineering approaches 

to the management of these aging issues. 

Thus, the critical issue for reactor vessels is 

whether our current correlations for 

embrittlement can be extrapolated to 80 years 

of operation. These correlations are largely 

empirical and caution must be used in 

extrapolating them to conditions outside the 

database on which they are based. For the 

current renewal period, the reactor pressure 

vessel materials surveillance program 

provides a means to detect unexpected 

degradation. The surveillance program was 

not planned to extend for the duration of 

operations now being discussed, and a 

research program to determine whether there 

are new mechanisms of embrittlement that 

could become significant and thus invalidate 

our current understanding is warranted. The 

potential need for modifications to the 

surveillance programs to address extended 

life was recognized in the ACRS report on 

PTS, and there appear to be industry efforts 

under way to address this need. NRC staff 

needs to be aware of these efforts. 

The limited availability of surveillance 

materials suggests that a more fundamental 

understanding of radiation damage will be 

needed to assure integrity of reactor vessels 

at the higher fluences associated with lives 

beyond 60 years. 

Issues with core internals such as irradiation 

assisted stress corrosion cracking are not 

fully resolved even for the case of license 

extension to 60 years. Licensees have 

typically committed to participate in the 

industry programs for investigating and 

managing aging effects on reactor internals; 

to evaluate and implement the results of the 

industry programs; and to submit an 

inspection plan reactor internals based on 

these efforts to the NRC for review and 

approval. Improved materials choices are 

available that have been demonstrated to at 

least substantially mitigate many of the 

materials problems that have been 

experienced in current LWRs, but the degree 

of improvement of performance that can be 

achieved for core internal materials is unclear. 

New reactors may be able to design around 

the problem, e.g., by ensuring that all 

internals are replaceable. For existing 

reactors demonstration of acceptable 

performance for life beyond 60 years may 

require both improved understanding and 

improved inspection techniques. 

In addition to the research associated with the 

passive, long-lived, safety-related 

components associated with 10 CFR Part 54, 

life beyond 60 efforts will seek to make 

extended operation economically and 

operationally justifiable. For example, plants 

seeking extended operation will very likely 

need to convert to digital instrumentation and 

control systems. While the NRC needs to 

address the safety issues associated with 

such changes, the research needed to 

develop such systems is clearly the 

responsibility of industry and DOE. 
11

Similarly in addressing issues such as the 

cracking of nickel alloys used in the reactor 

coolant system and steam generators, NRC 

effort will focus on the potential for cracking in 

the materials in use such as Alloy 600TT and 

Alloy 690. Industry and DOE efforts could 

consider the development of less expensive 

and easier to weld alloys. 

It was recognized at the workshop that cable 

aging and buried piping can be issues both in 

terms of 10 CFR Part 54 and economical 

operation. Only a small fraction of cable and 

buried piping are typically involved in 

evaluations related to 10 CFR Part 54. 

However, managing the aging of much larger 

fractions of cable and buried piping will be 

important to economic operation of plants 

during an extended period of operation. NRC 

programs directed towards extended 

operation will need to consider confirmatory 

research on degradation mechanisms that 

can affect safety, but recognize that industry 

has the primary responsibility to address 

aging problems, especially those that 

primarily affect the ability to operate 

economically. 

The staff has had focused discussions with 

DOE, the domestic industry (e.g., the Nuclear 

Energy Institute and the Electric Power 

Research Institute), and potential international 

partners to begin development of an 

integrated research plan in order to better 

leverage resources and prevent unnecessary 

duplication of efforts. In addition, public 

outreach continues to ensure appropriate 

stakeholder participation. In the longer term 

(fiscal year 2010 and beyond), the NRC, in 

collaboration with DOE, the industry, and 

international partners, will begin work on the 

priority areas identified in the integrated “Life 

Beyond 60” research plan to ensure that 

adequate technical information is available for 

regulatory decisions if licensees pursue 

subsequent license renewal terms. 

The Office of Nuclear Regulatory Research 

needs to determine how its research in this 

area will be integrated with that being done by 

DOE and the industry. 

The work to date in this area has focused on 

the additional characterization of recognized 

degradation processes and what is necessary 

to assure that current aging mitigation 

strategies remain valid for lifetimes greater 

than 60 years. These current efforts should 

be augmented by further efforts on the Proactive Materials Degradation Assessment 

Program to identify degradation processes 

not now manifest or anticipated. 
12
4. ADVANCED NON-LWR DESIGNS 

Background 

13

 

The DOE Generation IV program for 

advanced reactors was first developed in 

2000-2002 with the publication of the 

Generation IV Technology Roadmap (DOE, 

2002). From 2002 to 2005, the primary goal 

of the Generation IV program in the US was 

to focus on the use of high-temperature 

(850oC to 1000oC) process heat and 

innovative approaches to yield energy 

products, such as hydrogen, that might 

benefit the transportation and the chemical 

industries. To that end, the DOE began to 

interact with reactor designers, potential 

process heat users, and industrial and 

international organizations to support the Next 

Generation Nuclear Plant (NGNP) design and 

development needs for the thermal-spectrum, 

gas-cooled, graphite-moderated reactor 

concept; the so-called Very High Temperature 

Reactor (VHTR). The DOE effort was 

reinforced by the passage of the Energy 

Policy Act of 2005 (EPAct 2005),1

 which 

authorized appropriation of funds for research 

and construction activities for the NGNP 

project. DOE selected the VHTR as the lead 

design concept for the NGNP. Specifically, 

the VHTR program is focused on this type of 

reactor design and the EPAct 2005 authorizes 

the NRC to collaborate with the DOE in safety 

research related to licensing issues as the 

project proceeds through licensing to 

construction and operation. 

As part of this joint effort on safety research 

for NGNP licensing, the NRC has completed 

initial activities to provide the background 

information for the VHTR reactor. The 

activities included: 

• development of information sources 

for VHTR technology; and 

1

 See Subtitle C: Next Generation Nuclear Plant 

Project. 

• conduct of a PhenomenaIdentification and Ranking Table 

(PIRT) process for the VHTR in 

collaboration with the DOE and its 

contractors to identify the safety 

phenomena that require additional 

research and development. These 

important phenomena are associated 

with tools, standards, data, etc. 

required for the design and licensing 

review of such reactor technology. 

Current Research Activities 

The staff is focused on the development of 

appropriate evaluation models/methods/ 

guidance using information from past 

prismatic, gas-cooled, reactor designs and 

pebble-bed designs as well as the current 

conceptual VHTR designs from the industrial 

teams working with the DOE and its national 

laboratory contractors. 
14

The staff has developed a plan for advanced 

reactor research. The plan is well-written and 

comprehensive. Because the final design for 

the VHTR has yet to be determined, the plan 

is focused on research applicable to the 

range of possible designs. 

In fiscal year 2006, DOE shifted the emphasis 

of the Advanced Fuel Cycle Initiative program 

(AFCI) and the Generation IV fast-reactor 

development program. At that time, spent 

fuel management efforts began to focus on 

using a closed fuel cycle with reprocessing 

and recycling to a fast reactor as a main 

objective of the DOE spent fuel management 

and nuclear energy programs. The Global 

Nuclear Energy Partnership (GNEP) program 

was announced in early 2006. This new 

priority was short-lived and since the start of 

fiscal year 2009 the DOE has established the 

AFCI research and development program to 

address fuel cycle challenges, such as 

advanced fuel development. 

Assessment and Recommendations

Despite rather extensive changes made by 

DOE in its programs and plans for 

development of advanced reactors, the NRC 

staff has been able to develop a focused plan 

on research needed to support the licensing 

of a very high temperature gas-cooled 

reactor. This plan addresses: 

• the reactor fuel, 

• nuclear graphite and structural 

materials performance, 

• human factors and instrumentation, 

• Thermal-fluid analysis and 

containment analysis, 

• structural and seismic analysis, and 

• accident and source term analysis 
5. DIGITAL INSTRUMENTATION AND CONTROL SYSTEMS 

The challenges that arise in the transition 

from the use of analog to digital, softwarebased, technology for instrumentation and 

control systems in nuclear power plants are 

widely appreciated. These challenges are 

being met as software-based digital 

technology is applied in upgrading nuclear 

plant control systems. There have been, 

however, no applications of such digital 

systems to reactor safety systems which are 

the systems of primary interest to the NRC. In 

part, this has been due to the lack of 

standards and guidance from the regulators 

on the critical issues involved in the 

application of digital technology to reactor 

safety systems. Among the challenges that 

have been identified are: 

15

• testing methods for software based 

systems, 

• software quality and associated 

development and test processes, 

• the representation of software based 

systems in PRAs, and 

• the maintenance of safety division 

independence in a software-based 

communications environment. 

Safety systems involve simple actuations and 

do not involve complicated feedback and 

control functions. But, software-based digital 

systems offer the designer much greater 

flexibility and functionality than analog 

systems. Taking advantage of the enhanced 

flexibility and functionality carries with it 

increased circuit complexity, a new mode of 

common cause failure in software 

programming, and the potential for loss of 

safety division independence through interdivision data communication. 

Candidate Generation III Nuclear Power Plant 

Control Room 

Recognizing the difficulties faced by industry 

and the NRC, the staff, in 2007, established a 

new research plan. The staff also established 

a Steering Committee to identify technical and 

regulatory areas involving digital 

instrumentation and control requiring 

guidance and developed and oversaw a plan 

of action in concert with industry. Seven key 

issues were identified: 

• Cyber security 

• Diversity and Defense In Depth 

• Risk-informing Digital Instrumentation 

and Control 

• Highly Integrated Control Room - 

 Communications 

• Highly Integrated Control Room - 

 Human Factors 

• Licensing Process Issues 

• Digital instrumentation and control 

systems in nuclear fuel facilities. 

Task Working Groups were formed to 

address each issue. Results to date include 

the issuance of five Interim Staff Guidance 

(ISG) documents with the ISGs for Licensing 

Process Issues (ISG-6) and Digital I&C 

Systems in Safety Applications at Fuel Cycle 

Facilities (ISG-7) completing development. A 

pilot program for ISG-6 is under 

consideration. ACRS considers that these 
16

ISGs have provided much needed and 

valuable guidance for industry and for staff 

reviews. 

Another noteworthy research project is the 

study on Diversity Strategies for Nuclear 

Power Plant Instrumentation and Control 

Systems. The focus of the study is to develop 

a metric for system diversity and to address 

the issue of how much diversity is sufficient 

for safety systems. 
6. FIRE SAFETY 

Background 

The current focus of the fire safety research 

program is to support the NRC's regulatory 

needs during licensee transitions to riskinformed, performance-based fire protection 

programs that meet the requirements of 10 

CFR 50.48(c) and the referenced 2001 edition 

of the National Fire Protection Association 

(NFPA) Standard NFPA-805. Continuing 

research activities also develop improved 

information to support deterministic fire 

protection programs for licensees that do not 

adopt a risk-informed approach. In addition to 

direct support for currently operating reactors, 

the fire safety research programs also provide 

input to new reactor licensing reviews and 

assessment of the risk from fires in new 

reactor designs. 

Current Research Activities 

The NRC research activities in fire safety are 

depicted in Figure 1. Research activities 

primarily support programs in the Office of 

Nuclear Reactor Regulation (NRR) and, to a 

lesser extent, the Office of Nuclear Material 

Safety and Safeguards (NMSS). These 

activities are grouped into four technical 

areas: 

• fire modeling, 

• fire testing, 

• fire & electrical systems analysis, and 

• fire risk assessment. 

There are also efforts underway on fire 

research knowledge management. The 

following sections briefly discuss the major 

projects in each area. 

Fire Modeling

Fire models provide a phenomenological 

basis for the evaluation of fire growth, 

detection, and suppression, and the analysis 

of potentially risk-significant fire scenarios. 

CHRISTI-FIRE 

The objective of the CHRISTI-FIRE (Cable Heat 

Release, Ignition, and Spread in Tray Installations 

during FIRE) experimental program is to perform 

fire tests on grouped electrical cables to better 

understand the fire hazard characteristics, 

including heat release rate and flame speed. This 

type of quantitative information will be used to 

develop more realistic models of cable fires for 

use in fire probabilistic risk assessment (PRA) 

analyses. 

 

In 2007, NRC and EPRI completed a 

collaborative project for verification and 

validation (V&V) of five commonly used fire 

modeling codes (NUREG-1824). That effort 

identified strengths and weaknesses in each 

code with respect to specific fire modeling 

issues. In 2008, a Phenomena Identification 

and Ranking Table (PIRT) exercise was 

completed to assess the predictive 

capabilities of these fire models for a number 

of postulated fire scenarios (NUREG/CR6978). A collaborative project between RES 

and EPRI is using the results of the 

verification and validation of fire models along 

with the PIRT evaluations to develop an 

integrated "Fire Model Users' Guide" for the 

approved models. The Guide will describe 

the capabilities of each code to evaluate 

specific phenomena during realistic fire 

scenarios, including limitations, precautions, 

17
18

and lessons learned from practical 

applications. The Users' Guide will benefit 

risk analysts who use the fire models to 

evaluate potentially important fire scenarios, 

and it will support inspection efforts that use 

the models as input to the significance 

determination process (SDP). 

Future research projects are planned to 

extend the current fire modeling capabilities to 

address specific limitations documented in the 

V&V reports and the PIRT evaluations, and to 

improve predictions of phenomena observed 

during fire experiments. 

Fire Testing

This element of the research program 

includes two major projects. 

The Cable Heat Release, Ignition, and 

Spread in Tray Installations during Fire 

(CHRISTI-FIRE) program is performing fullscale fire tests to measure the heat release 

rates and flame spread characteristics from 

fires in bundles of electrical cables. The tests 

are conducted with realistic cable tray 

configurations and loadings, and they include 

cables with thermoplastic and thermoset 

insulations. The quantitative data collected 

by this project will be used to develop more 

realistic cable fire models for PRAs and to 

enhance the predictive capabilities of fire 

modeling codes. 

Small-scale fire tests will be performed to 

evaluate the performance of spent nuclear 

fuel shipping cask seals during beyonddesign-basis fires that exceed the 

manufacturers' rated temperatures. This 

information will be used by NMSS to further 

develop risk insights related to the 

transportation of spent fuel. 

Fire & Electrical Systems Analysis

The experience from actual fire events 

confirms that damage to electrical cables may 

disable equipment and cause unexpected 

responses from instrumentation and control 

(I&C) signals. Realistic evaluation of fireinduced circuit damage, particularly involving 

spurious signals caused by "hot shorts", is a 

very significant effort in fire risk assessments 

and in pilot applications of the methods 

described in NFPA-805 and NUREG/CR6850. 

The Direct Current Electrical Shorting In 

Response to Exposure-Fire (DESIREE-FIRE) 

project will extend the Cable Response to 

Live Fire (CAROLFIRE) test program 

(NUREG/CR-6931) to examine fire damage to 

DC circuits. The tests are being performed in 

collaboration with EPRI. They will include 

realistic cable tray configurations and loading, 

and they will evaluate the effects from fire 

damage to circuits for DC-operated 

components and typical I&C applications. 

A planned future project will compile the 

results from industry-sponsored cable tests, 

CAROLFIRE, and DESIREE-FIRE. A 

structured expert elicitation process will 

examine the available data and develop 

conditional probabilities and uncertainties for 

various AC and DC circuit failure modes that 

may be caused by fire-induced cable 

damage. 

Fire Risk Assessment

Integration of fire risk into a full-scope PRA 

framework is a very important research 

activity. The requirements of NFPA-805 and 

the guidance in NUREG/CR-6850 provide 

cornerstones for the development of riskinformed, performance-based fire protection 

programs and a comprehensive assessment 

of fire risk. 

A collaborative project with EPRI is evaluating 

elements of human reliability analysis (HRA) 

methods that apply to post-fire mitigation 

actions. The goal of this project is to 

recommend specific methods and guidance 

for the modeling and quantification of human 

errors during fire scenarios. 
19

An NMSS-sponsored quantitative risk 

assessment is being performed to evaluate 

the risk from potential fires and explosions of 

"red oil" at the Mixed-Oxide Fuel Fabrication 

Facility. This effort included a pioneering 

attempt to use the methods of probabilistic 

risk assessment to estimate risks of red oil 

fires in fuel reprocessing facilities. The 

feasibility of using probabilistic methods 

rather than the usual integrated safety 

assessment methods for nuclear facilities is 

noteworthy. The results from this probabilistic 

study will be used to support design and 

licensing reviews. 

Pilot applications of the methods described in 

NFPA-805 and NUREG/CR-6850 are 

currently in progress. A revision to 

NUREG/CR-6850 is planned to incorporate 

lessons learned from these applications and 

improvements to fire modeling and fireinduced damage assessments from on-going 

research activities. 

Fire Research Knowledge Management

Fire research continues to be a rapidlyevolving element of the RES mission. 

Compilation and dissemination of the 

information gained from this research, 

including new results and insights, is vital for 

understanding the current state of knowledge 

and planned near-term advancements. Three 

excellent resources prepared by RES include: 

• "The Browns Ferry Nuclear Plant Fire of 

1975 and the History of NRC Fire 

Regulations," NUREG/BR-0361, January 

2009. This DVD preserves the history of 

the Browns Ferry fire and documents its 

influence on the development of 

enhanced fire protection regulations. 

• "Fire Protection and Fire Research 

Knowledge Management Digest," 

NUREG/BR-0465, Revision 1, February 

2009. This CD contains a compilation of 

fire-related reference materials that are 

useful for NRC inspectors, reviewers, 

licensees, and other stakeholders. It 

includes 10 CFR Part 50, guidelines for 

fire protection, fire inspection manuals 

and procedures, generic letters, bulletins, 

information notices, regulatory guides, 

and fire-related NUREG reports. 

• "A Short History of Fire Safety Research 

Sponsored by the U.S. Nuclear 

Regulatory Commission, 1975 – 2008," 

NUREG/BR-0364, June 2009. This report 

provides an historical perspective on 

NRC-sponsored fire safety research, 

summaries of current research activities, 

and planned near-term research 

programs. 

These references are updated annually. 

Assessment and Recommendations 

The NRC's fire safety research program is at 

the forefront of a rapidly evolving 

understanding of fire events and the 

assessment of risks from fire damage in 

nuclear power plants. Structured 

collaboration with industry provides costeffective solutions and technical insights that 

surpass independent efforts. The NRC is a 

leader in national and international fire safety 

research. 

The focus and priorities for current research 

projects are determined primarily by useridentified needs. This process is responsive 

to immediate and near-term technical issues, 

and it should remain an important part of 

integrated planning. However, research 

priorities and programs should more actively 

anticipate emerging applications and 

intermediate- to long-term requirements that 

are not fully dictated by current user needs. 

Data and insights from the CAROLFIRE tests 

have considerably advanced the 

understanding of fire-induced cable damage. 

It is expected that the results from current 

projects such as CHRISTI-FIRE and 

DESIREE-FIRE will also provide valuable 

knowledge. The NRC should continue to 

encourage and support additional testing and 
20

fire experiments through collaborative efforts 

with US industry and international 

organizations. 

Other current research projects are also 

evaluating HRA methodologies. The goal of 

those activities is to recommend or develop 

an integrated methodology that captures the 

best practices and reduces the variability of 

assessments from the current assortment of 

HRA methods. The resulting integrated HRA 

methodology should apply equally to 

scenarios that are initiated by internal events, 

internal hazards (e.g., fires and floods), and 

external events (e.g., earthquakes) during any 

plant operating mode (e.g., full power, low 

power, or shutdown). The fire research 

described above is focused primarily on only 

one facet of this problem (i.e., human 

performance after fires that occur during full 

power operation). This work should be more 

closely coordinated with other RES activities 

to avoid potential duplication or divergence of 

NRC-sponsored HRA research and 

conclusions. 

Guidance is currently being developed for the 

assessment of risk during low power and 

shutdown operating modes. Many of the 

current fire research programs and results are 

directly applicable to the evaluation of fire 

damage during these operating conditions. 

However, additional information is needed to 

evaluate such issues as fire initiation 

frequencies, human-caused fires, 

effectiveness of detection and suppression, 

and propagation of heat and smoke through 

compromised fire barriers that are uniquely 

associated with personnel activities and SSC 

configurations during plant shutdown. 

New reactor designs rely heavily on 

integrated digital instrumentation, protection, 

and control systems. Extensions of operating 

licenses and obsolescence of analog 

equipment will also likely lead to replacement 

of many currently installed I&C systems with 

digital platforms. Limited information is 

available regarding the effects from fire, heat, 

and smoke on digital equipment. Research 

projects should address such issues as the 

effects from fire damage and heat on fiber 

optic cables, the effects of heat on digital 

equipment, and smoke damage to digital 

signal processing and computation modules. 
Independent Evaluation of Licensees’ 

Submittals in the Area of Fire Safety

Including Evaluating Licensees’ Transition to the Riskinformed, Performance-based Fire Protection Programs 

that Meet the Requirements of 10 CFR 50.48(c) 

Fire Safety Research 

Fire Modeling Fire Testing Fire PRA 

Fire Research 

Knowledge Management 

Fire & Electrical Systems/ 

Circuit Analysis 

• Development 

of Fire Model 

Users Guide 

• Application 

of the PIRT 

process to 

electrical 

circuits’ fireinducedfailures in 

NPPs 

• CHRISTI-FIRE 

Program at 

NIST 

• Spent Fuel 

Shipping Cask 

Fire Testing 

• International 

PRISM 

Program 

(OECD/IRSN) 

Figure 1. Current NRC Research Activities in Fire Safety

• DESIREE-FIRE 

Direct Current 

Circuit Testing 

• Circuit Fire 

Testing Data 

Analysis 

• Support for 

Development of Fire 

PRA Standard 

• RES/EPRI 

Collaboration in 

evaluating elements of 

HRA methods that 

apply to post-fire 

mitigation actions 

• Risk Assessment of 

potential “Red Oil” 

Hazards 

21
22
7. REACTOR FUEL 

Background 

Fuel integrity is an important element of 

nuclear safety technology. The oxide fuel 

pellets and their zirconium alloy cladding are 

the initial physical barriers against inadvertent 

release of radionuclides to the environment in 

the event of an accident. The ability of 

nuclear fuel to operate without failure in the 

demanding thermal, mechanical, radiation, 

and chemical environments of BWR and 

PWR cores has improved dramatically over 

the past three decades. The various 

mechanisms capable of causing fuel failure 

have been identified and addressed 

effectively by materials, design, and 

operational improvements. The industry’s 

current goal of achieving zero fuel failures 

during normal operation by 2010 is within 

reach. 

During this same period, performance 

differences among fuel designs provided by 

various manufacturers have become relatively 

small, and many utilities have developed the 

capability to operate their cores with fuel 

assemblies provided by two or more 

suppliers. Fuel manufacturers have focused 

on evolutionary design changes with the 

occasional introduction of new materials such 

as M5 and Zirlo fuel cladding. Not only have 

these compositional changes improved the 

corrosion and hydriding resistance of fuel 

cladding during normal operation, they have 

also improved resistance to embrittlement 

during loss of coolant accidents. 

23

 

The fuel industry is now evaluating more 

significant design changes. Lead use 

assemblies are in operation in which chemical 

additives have been used to improve fission 

gas retention and mechanical properties of 

the UO2 fuel pellets. Both new and 

evolutionary fuel cladding materials are in the 

Halden Reactor Project (HRP) 

Halden tests on high-burnup fuel under loss-ofcoolant accident conditions supported NRC 

research on cladding embrittlement. 

The Norwegian Institute for Energy Technology 

(Institutt for Energiteknikk or IFE) manages the 

HRP for the Organization for Economic 

Cooperation and Development/Nuclear Energy 

Agency. The HRP is based at IFE’s facility in 

Halden, Norway. This facility includes the Halden 

Boiling Water Reactor (HBWR), which currently 

operates at 18 to 20 megawatts. The HBWR is 

fully dedicated to instrumented in-reactor testing of 

fuel and reactor materials. The NRC has used 

numerous research products from this 

internationally funded cooperative effort. 
development and regulatory pipeline. In 

BWRs, changes in fuel bundle design from 

the current 10x10 arrays to 11x11 are under 

evaluation as are changes in the composition 

and thermal processing of fuel channels. 

The NRC must maintain the capability to 

evaluate the adequacy of the testing and 

analyses supporting these new developments 

and the impact of these changes on nuclear 

safety. 

Current Research Activities 

The overall NRC fuel behavior research 

program is summarized in Figure 2. As 

shown, the research is categorized from left 

to right by sources of experimental and 

analytical data, regulatory issues of interest, 

analytical models in which new information 

will be incorporated , and finally by users and 

user needs. 

The suitability of new cladding materials is 

determined by extensive vendor testing in 

laboratories and test reactors, and by 

performance evaluations of lead-use 

assemblies in operating power plants. Fuel 

suppliers typically perform laboratory tests to 

evaluate cladding embrittlement during loss of 

coolant accidents. However, fuel suppliers 

rarely validate predictions of fuel performance 

during design basis accident conditions with 

experimental data. In the past, the industry 

relied on NRC to provide information on fuel 

behavior under accident conditions. Because 

of the cost, complexity and requirements for 

specialized test facilities such data are 

currently obtained from internationally funded 

programs such as those at CABRI or Halden. 

Experimental data on the performance of fuel 

and cladding systems under severe accident 

conditions are not provided by licensees. 

Severe accidents, of course, pose the bulk of 

the risk to the public health and safety. 

Experimental data obtained in research 

sponsored by the NRC show that cladding 

interactions with the fuel play a central role in 

the rate and extent of core degradation and 

fission product release under severe accident 

conditions. For example, recent single rod 

loss-of-coolant accident (LOCA) experiments 

at the Halden reactor project have 

demonstrated conditions under which very 

high burnup fuel can fragment, relocate into 

the ballooned volume and subsequently be 

ejected from the cladding. These data will be 

invaluable in determining future high burnup 

limits. 

The NRC is in the process of completing 

experimental studies of high-burnup fuel and 

cladding behavior under design basis 

accident conditions. The NRC has sponsored 

in-pile tests of fuel behavior during reactivity 

insertion events and out-of-pile tests of fuel 

behavior under design basis loss of coolant 

conditions. These experimental 

investigations have been conducted using an 

impressive combination of national and 

international collaborations. Results of the 

research have led to well considered 

proposals for changes in the regulations. The 

proposed changes would make the 

regulations more realistic and could decrease 

burden on both the staff and licensees, 

especially as new fuel and cladding systems 

are proposed. Implementation of these 

proposed changes to the regulations has 

been slow. Acceptance of the changes to the 

regulations now awaits results of additional 

testing that could be regarded as confirmatory 

in nature. 

The staff has also completed revisions of its 

fuel performance computer codes FRAPCON 

and FRAPTRAN. These computer codes are 

used to independently confirm analyses done 

by the vendors and other licensees. The 

modifications allow the computer codes to be 

used to evaluate fuels taken to burnups of up 

to 62 GWd/t and to evaluate the performance 

of mixed oxide (MOX) fuels. MOX fuels have 

been irradiated in the Catawba reactor as part 

of a DOE program to dispose of excess 

weapons-grade plutonium. The NRC is also 

examining the severe accident behavior of 

high-burnup and MOX fuels in its severe 

accident research program. 

24
Assessment and Recommendations 

The upgraded fuel performance models 

appear to meet most near-term agency 

needs. There is ample capability to assess 

risks of fuel failure due to fuel center melting 

or departure from nucleate boiling events. 

However, the NRC still does not have 

empirical or analytical capability to 

quantitatively assess the more likely risk of 

Pellet Cladding Interaction (PCI) fuel failures 

during anticipated operational occurrences 

(AOOs) on current or future fuel designs. The 

stress corrosion fuel failure mechanism that is 

active during some AOOs may compromise 

fuel integrity as licensees continue to uprate 

power, increase fuel burnup, and introduce 

new fuel and cladding materials. Despite 

these needs, NRC is not developing any 

quantitative capabilities to assess the fuel 

vulnerability to PCI during operational 

transients. There is a wealth of experimental 

data on the PCI phenomenon available to the 

NRC to develop and validate a practical PCI 

failure model for use in regulatory decision 

making. 

Advances in computing power and 

computational simulation are making it 

possible to examine fuel performance in 

vastly more detail than is done with either 

FRAPCON or FRAPTRAN. Whether such 

detail is needed will depend critically on what 

efforts are made by licensees to extend fuel 

burnups beyond the current regulatory limit of 

62 GWd/t and the amount of experimental 

data provided to support these proposed 

changes to regulatory limits. There is 

widespread expectation that the nuclear 

industry may push for extending burnup limits 

to 85 GWd/t. Certainly, a recent research 

strategy document prepared for DOE and 

nuclear energy industry proposes that fuel 

burnups be extended to 85 GWd/t. There 

appears to be some confidence within the 

nuclear industry that such extensions of fuel 

burnup can be done by extrapolating the 

currently available bases of fuel performance 

data and models. Emergence of new 

physics complicated such extrapolations of 

fuel performance for burnups beyond 40 

GWd/t. This necessitated the experimental 

research on fuel behavior under accident 

conditions that the NRC is currently 

completing. It is not evident that no new 

phenomena will arise in connection with the 

extrapolation of fuel burnup to 85 GWd/t. 

Consequently, there will be a continuing need 

for the agency to independently evaluate the 

safety of proposed changes in the nature and 

burnup limits of reactor fuels. 

Lead-test assemblies of MOX fuels have 

emerged from their second cycle of irradiation 

in the Catawba reactor. These MOX fuels are 

being tested as part of a DOE program to 

dispose of excess weapons-grade plutonium 

by using it as reactor fuel. It appears that the 

NRC does not have plans for any research 

examinations of these novel fuels either after 

the first cycle of irradiation or after 

subsequent cycles of irradiation. In light of 

the limited NRC experience with MOX fuels, 

the ACRS recommends that there be a 

research program to follow closely the postirradiation examination of the lead-test 

assemblies planned by DOE. 

NRC must maintain the expertise in the area 

of reactor fuel to respond to future design 

changes. Because of license extensions, 

power uprates, and the prospect of additional 

new reactors, it is anticipated that the vendors 

will introduce new fuel and cladding systems. 

The challenge in maintaining what amounts to 

an essential core competency of the agency 

arises because of the limited availability of 

expertise outside of the agency that is 

independent of licensees. Current manpower 

working for the agency in this field either 

directly or by contract is experienced and 

there is a need to groom newer generations in 

the field. A major difficulty in doing so is the 

decline in the United States of in-pile test 

facilities and hot cells for examinations of 

irradiated fuels and cladding. Long-term 

collaboration with international partners 

having the facilities and staff capable of 

undertaking pertinent in-reactor studies may 

well be essential for NRC to maintain an 

adequate level of expertise in reactor fuels. 

One option available to the NRC for the 

25
26

development of fuel staff expertise is the 

stationing of promising staff members at 

appropriate international projects such as the 

Halden project for one to two years. This 

opportunity has been widely used by industry 

and should be considered by the NRC. 

Reprocessing of nuclear fuels may be an 

emerging technology that merits research 

attention by NRC. Aqueous reprocessing of 

irradiated fuels has been done for many years 

within the nuclear weapons community using 

the PUREX process. Though familiar, this 

process has not been trouble free and there 

are many known hazards. NRC has gained 

some exposure to the associated safety 

issues of “red oil,” hydroxylamine nitrate, 

ammonium nitrate and the like through its 

review of the construction authorization 

application for the DOE’s MOX Fuel 

Fabrication Facility. Modifications of the 

PUREX process – the so-called UREX 

processes – were being considered by DOE 

in its now abandoned Global Nuclear Energy 

Partnership initiative. Undoubtedly, safety 

issues similar to those of the PUREX process 

and new safety issues will arise if these new 

processes are pursued. 

Even more challenging will be the 

pyrometallurgical reprocessing of spent 

nuclear fuel which was also being considered 

by DOE as a longer-term development. In the 

current political environment, it is unlikely that 

the NRC will face the challenge of regulating 

technologies other than modest changes to 

the PUREX process. 
Figure 2. Current NRC Research Activities in Reactor Fuel

27
28
8. HUMAN FACTORS AND HUMAN RELIABILITY 

Human performance continues to be a 

significant element of the safe operation of 

nuclear facilities. The challenge NRC 

research faces is how to respond using the 

current state of the practice to short-term 

needs emanating from program offices while 

at the same time maintaining a research 

program that is long-term and aims at 

improving our understanding and modeling 

human performance. To achieve an 

appropriate balance between these two 

competing objectives the staff has organized 

its activities around four goals: 

• ensuring that Human Factors 

(HF)/Human Reliability Analysis 

(HRA) methods and guidance have 

sound, up-to-date technical bases; 

• expanding the HF/HRA infrastructure 

for anticipated changes made by the 

nuclear industry; 

• improving HF/HRA methods to reduce 

uncertainty and promote the state of 

the art; and 

• maintaining the infrastructure of 

expertise, facilities, capabilities, and 

data. 

These goals of the staff’s research activities 

are well directed towards meeting the agency 

needs (see also Figure 3). The staff’s 

participation in domestic and international 

collaborations strengthens these activities 

and should be encouraged. These 

collaborations include various OECD groups, 

EPRI, NASA, and professional organizations 

such as ANS, IEEE, and AIChE. 

Human Factors Research 

10 CFR Part 26, “Fitness for Duty Programs” 

(FFD), was updated and published in the 

Federal Register on March 31. 2008. To 

ensure consistency among various offices, 

the staff is developing training materials for 

personnel involved in the implementation of 

the updated Part 26.

Recognizing that the methods for assuring 

personnel ’fitness for duty” are evolving, the 

staff is exploring the use of new methods in a 

project titled “Methods for Assessment of 

FFD.” Both of these activities are important to 

the regulation of existing nuclear power plants 

(NPPs) and should continue. 

The new generation of nuclear power plants 

will differ from the current plants in three 

important ways: 

• the plant design, 

• the extensive use of digital 

instrumentation and control (DI&C) 

systems, 

• and human-system interface (HSI). 

Consequently, Human Factors Engineering 

(HFE) must adapt to this new reality. 

Recognizing this need, the staff convened a 

group of experts from research organizations, 

vendors, and utilities to prioritize research 

needs. NUREG/CR-6947, “Human Factors 

Considerations with Respect to Emerging 

Technology in Nuclear Power Plants,” 

contains the results of this process, i.e., the 

prioritization of the research areas and the 

rationale for the ranking. These results are 

used by the staff to develop a long-term 

research plan. The ACRS commends the 

staff for taking such a systematic approach to 

the prioritization of research topics. 

29
Of the twenty topics identified as having high 

priority, the staff is currently pursuing two. 

The project “Automation and Complexity” 

aims at developing guidance for reviewing the 

operators’ interface with automation displays. 

The project “Human Factors under Degraded 

I&C” investigates how degraded or failed I&C 

systems may affect operator situation 

awareness and performance. Both of these 

projects will contribute to the expansion of the 

human factors infrastructure that will enable 

the agency to deal with anticipated changes 

in the industry. This research is supported by 

experiments performed at the Halden ManMachine Laboratory (HAMMLAB) in Norway. 

HAMMLAB uses a simulator control room that 

can be configured as a prototype advanced 

ontrol room with an integrated surveillance 

ility 

enchmark test (discussed below) will be 

US plant. 

 models and to recommend 

lasses of problems for which a model is 

ncy to use or guidance on 

hich model(s) should be used in specific 

ese models on the basis of a comparison of 

essful, 

ese HRA projects will offer a significant 

arch activities include 

lear 

product Materials’” 

,” 

• “GEH Fuel Cycle Facility Review,” and 

sound rationale and to be focused on both 

near- and long-term needs of the agency. 

c

and control system. 

A testing capability similar to that at Halden’s 

HAMMLAB does not exist in the US. Some 

have advocated that such a capability be 

developed in this country. It is not evident that 

such a capability would be cost effective. It is 

noteworthy that the next human reliab

b

conducted at a simulator for a 

Human Reliability Analysis 

Human Reliability Analysis (HRA), as an 

accident sequence evolves, is an area where 

several models exist. For example, the EPRI 

software tool “HRA Calculator” recommends 

the Cause-Based Decision Tree Model and 

THERP as the default methods for all postinitiating-event human failure events. The 

NRC itself is using ATHEANA and SPAR-H. 

There is clearly a need to investigate and 

evaluate the validity of the assumptions 

behind these

c

appropriate. 

This need has been recognized by the 

Commission. The Commission has asked the 

ACRS to “work with the staff and external 

stakeholders to evaluate the different HRA 

models in an effort to propose either a single 

model for the age

w

circumstances”. 

In response, the staff has initiated two 

research projects. The project “HRA Method 

Benchmarking” is utilizing HAMMLAB to 

simulate accident sequences. It is cosponsored by Halden, the Swiss Federal 

Nuclear Inspectorate, and EPRI. The 

objective is to collect and analyze crew 

performance data, to apply HRA models to 

predict crew performance, and to evaluate 

th

the simulator data with the model predictions. 

The project “HRA Model Differences” will 

build on the conclusions from the 

benchmarking project mentioned above and 

will recommend appropriate models as the 

SRM requires. Insights gained from operating 

experience will play an important role in the 

formulation of this recommendation. The staff 

is continuing the development of the Human 

Event Repository and Analysis (HERA) 

system. HERA is developing taxonomy for 

collecting human performance data, collecting 

real and simulated events, and developing 

tools enabling the use of HERA to inform 

human reliability analyses. If succ

th

advance in the evolution of methods. 

In addition to nuclear reactor applications, the 

staff’s HF/HRA rese

several worthwhile projects aimed at nuc

facilities including: 

• “HRA for By

• “Qualitative HRA for Spent Fuel 

Handling’” 

• “AREVA Fuel Cycle Facility Review

• “MOX Fuel Cycle Facility Review.” 

Overall, the ACRS finds the HF/HRA research 

program and activities to be well based on 

30
Human Factors and 

Human Reliability Research 

Providing Safety Perspectives on the Impacts of 

Human Performance on Regulated Activities

Maintaining 

Current 

Infrastructure 

• Halden Reactor 

Project 

(OECD/NEA) 

• Human Event 

Repository and 

Analysis (HERA) 

• HRA Data 

Repository 

• Technical Basis 

Support for Personal 

Security 

Rulemaking 

• Methods for 

Assessment of 

Fitness for Duty 

(FFD) 

• HRA Model 

Differences 

Improving 

HF/HRA Methods 

Expanding HF/HRA 

Infrastructure for New 

Applications 

• Human Factors 

under degraded I&C 

• HRA for Byproduct 

Materials 

• HRA Methodology 

for Fire Analysis 

• Qualitative HRA for 

Spent Fuel Handling 

Developing Technical 

Bases and Guidance

Documents

• Model Based 

HRA 

 

• HRA Method 

Benchmarking 

Using Simulator 

Data 

Figure 3. Current NRC Research Activities in Human Factors and Human Reliability

31
32
9. MATERIALS AND METALLURGY 

Background 

Materials and metallurgy continues to be the 

most active area of research within the NRC. 

This is appropriate in light of the efforts 

required by the agency to address known and 

emerging materials degradation phenomena 

in aging LWRs, and to monitor the 

effectiveness of licensees’ aging 

management programs. As plants age, known 

degradation mechanisms will continue to 

affect components important to safety, and 

new degradation mechanisms may develop. 

The agency must develop the technical 

capabilities to minimize the risk of material 

degradation surprises and to assess the 

effectiveness of the industry initiatives to deal 

with materials degradation. 

Sectioning of reactor vessel head 

penetrations from Washington Nuclear 

Power, Unit 1, a canceled plant 

Current Research Activities 

assisted stress corrosion cracking and other 

environmental degradation mechanisms. 

RES is making excellent use of domestic and 

international cooperative programs to 

accelerate progress, reduce costs, and 

resolve key issues related to the detection, 

understanding, and mitigation of materials 

degradation phenomena. These include the 

Program for the Inspection of Nickel Alloy 

Components (PINC), the Stress Corrosion 

Cracking and Cable Aging Program, the 

OECD Pipe Failure Data Exchange, the 

Halden Reactor Project, the Cooperative 

Irradiation-assisted Stress Corrosion Cracking 

(IASCC) Program (CIR II), and the Zorita 

Internals Research project. 

Current materials and metallurgy research 

activities are conducted by the Component 

Integrity Branch and the Corrosion and 

Metallurgy Branch within the Division of 

Engineering of RES. The former directs 

research focused on fracture mechanics, nondestructive examination (NDE), and safety 

assessments, and the latter on corrosion, 

metallurgy, and advanced reactors. Topical 

areas of research are grouped into five broad 

categories: 

• Component Integrity Assessments 

• Non-Destructive Evaluation 

• Environmentally Assisted Cracking 

• Proactive Management of Materials 

 Degradation 

• Steam Generator Tube Integrity Component Integrity Assessments

 

The research activities in the five areas are 

appropriate and address the important 

materials issues. The results of these 

research activities will improve the agency’s 

ability to independently evaluate licensees’ 

efforts to prevent or mitigate environmentally 

This area of research addresses the integrity 

of dissimilar metal welds in piping, control rod 

drive mechanisms, and pressure vessels; 

develops probabilistic and deterministic 

fracture mechanics tools; evaluates 

effectiveness of primary water stress 

corrosion cracking (PWSCC) mitigation 

33
34

methods; and validates methods used to 

determine weld residual stresses (WRS) in 

dissimilar metal welds. In addition to 

research on metallic components, work is 

being done to confirm the acceptability of high 

density polyethylene (HDPE) piping for long 

term safety related applications. 

 A new research project has been initiated to 

develop and validate a probabilistic fracture 

mechanics tool to assess extremely low 

probability of rupture (xLPR) in piping 

systems susceptible to degradation 

mechanisms such as PWSCC. These 

systems have rupture probabilities of < 10-6 

per reactor year. As shown in Figure 4, this 

code will incorporate flaw distributions, 

fracture mechanics, fluid mechanics, 

corrosion, and metallurgical variables in a 

manner similar to that used in development of 

the new pressurized thermal shock rule. 

The PWSCC mitigation program will evaluate 

the effectiveness of industry developed 

mitigation strategies such as full structural 

weld overlay and mechanical stress 

improvement. The objective of this work is to 

ensure that the probability of rupture of 

current leak before break (LBB) systems 

remains extremely low in PWR components 

subject to PWSCC. An important input to this 

research is the validation of estimated weld 

residual stresses in dissimilar metal welds. 

This research will include determinations of 

weld residual stresses in simple plates and 

cylinders and in complex mockups of PWR 

components. Weld residual stresses will be 

measured using x-ray diffraction, neutron 

diffraction, and incremental deep hole drilling 

techniques. 

High density polyethylene (HDPE) materials 

have attractive properties for service water 

and buried piping applications, but they are 

subject to aging and degradation mechanisms 

that differ from those of metal piping and 

other types of piping conventionally used. A 

growing number of licensees propose to use 

HDPE in safety-related applications. Failure 

mechanisms and service life as affected by 

temperature, time, and flaws need to be 

researched. The research will include 

fracture-mechanics-based flaw tolerance 

evaluations of base materials as well as 

fusion joints. 

The integrity of the reactor pressure vessels 

has been studied for decades. Licensee 

obligations to ensure the structural integrity of 

the reactor pressure vessel during both 

routine operations and postulated upset 

conditions, are codified in three general 

design criteria (GDC 14, GDC 30, and GDC 

31) as well as in 10 CFR 50.61 and the 

Appendices G and H to 10 CFR Part 50. The 

technical bases for these requirements were 

largely established in the1980s through NRC 

sponsored research. 

Major progress has been made in completing 

and closing several research projects on 

reactor pressure vessel integrity and 

incorporating remaining research in the 

component integrity assessments program. 

This research has led to updates in several 

NRC regulatory documents as well as ASME 

and ASTM codes and standards. Revisions in 

the Pressurized Thermal Shock (PTS) 

screening criterion in the newly issued PTS 

rule and the associated regulatory guides and 

Appendices G and H to 10 CFR Part 50 

provide great benefit by quantifying excessive 

conservatisms in prior regulations and 

allowing longer service life of reactor pressure 

vessels. 

The staff is now focusing pressure vessel 

research under two programs. The “Pressure 

Boundary Materials” program will develop and 

validate predictive material property models 

aimed at the refinement and generalization of 

structural integrity assessments. The 

“Probabilistic Pressure Boundary Safety 

Assessment” program will develop more 

realistic flaw evaluation tools and fracture 

mechanics models to replace deterministic 

failure criteria with probabilistic criteria. 
35

Non-Destructive Evaluation 

Various NDE methods are relied upon to 

monitor the integrity of reactor coolant 

systems. These include ultrasonic testing, 

eddy current testing, penetrant testing, 

radiographic testing, and visual testing. The 

reliability and effectiveness of these methods 

can vary considerably depending on 

component geometry, materials, and types of 

defects. There are two ongoing projects 

addressing NDE technology. These have 

been expanded to address the effectiveness 

of NDE methods applied to high density 

polyethylene piping materials and joints.

“Reliability of NDE for Nuclear Power Plant InService-Inspection” is a project focused on 

quantifying the reliability of a broad range of 

NDE techniques used in power plant inservice inspection (ISI) programs. This task 

was initiated in 2007 to evaluate the accuracy 

and reliability of the NDE methods and to 

provide recommendations to the staff to 

improve the effectiveness and adequacy of 

ISI programs. This research, which is 

expected to be completed in 2012, covers the 

entire spectrum of NDE techniques used in 

reactor construction, in-service inspection, 

and repairs. Specific evaluations of the 

effectiveness of ISI techniques will be 

provided for: 

• PWSCC in Alloy 600, 82, 182 

dissimilar metal welds and J-groove 

penetrations, 

• IGSCC in austenitic stainless steel 

welds, 

• degradation in cast stainless steel and 

weldments, 

• repairs including overlays, cladding 

and in-lays, 

• reactor internals examinations, 

• vessel penetrations, and 

• HDPE piping 

The project “Assess Emerging NDE for 

Dissimilar Metal Welds” is intended to provide 

data and correlations necessary for NRC staff 

to independently evaluate licensee programs 

for assessing the integrity of dissimilar metal 

welds. In addition, the program will continue 

to evaluate emerging NDE techniques such 

as phased array UT and synthetic aperture 

focusing techniques for the inspection of 

coarse grained materials such as cast 

stainless steels. 

Environmentally Assisted Cracking 

Environmentally assisted cracking (EAC) is a 

generic term for the various stress corrosion 

cracking mechanisms that can be active in 

BWRs and PWRs. These complex 

phenomena are influenced by applied and 

residual stresses, water chemistry, radiation 

exposure, temperature, material composition, 

microstructure, and fabrication history. In 

BWRs, these mechanisms are known as 

intergranular stress corrosion cracking 

(IGSCC) and irradiation assisted stress 

corrosion cracking (IASCC). The intergranular 

stress corrosion cracking occurs on stainless 

steel welds outside the core regions and the 

irradiation assisted stress corrosion cracking 

occurs in radiation hardened stainless steel 

core internals. In PWRs the dominant 

mechanisms are primary water stress 

corrosion cracking (PWSCC) and IASCC. 

These phenomena occur on susceptible 

nickel base alloys and dissimilar metal welds.. 

In recent years, EAC has occurred in 

components internal to the vessels in BWRs 

and PWRs, and in reactor vessel penetrations 

and dissimilar metal welds. In addition to 

primary system leakage, EAC can lead to 

serious secondary damage. At Davis-Besse 

PWSCC cracks in control rod drive 

mechanism (CRDM) nozzles led to leakage of 

borated coolant and accelerated corrosion of 

the pressure vessel head. 

Licensees have implemented improved 

inspection methods, resistant materials and 

repair procedures, and improved water 

chemistries to prevent EAC. These actions 

have been effective in reducing the frequency 

of EAC events, but they have not eliminated 

the problem. Environmentally assisted 
36

cracking continues to occur in nuclear power 

plants as metal components age and 

radiation exposure increases. 

The NRC staff must maintain capabilities to 

evaluate licensees’ analyses of the active 

degradation phenomena in their plants, and 

the effectiveness of implemented or proposed 

mitigation methods. The research projects 

now under way are designed to ensure that 

the NRC has the necessary technical 

understanding of the root causes of the 

various environmental degradation 

phenomena, their underlying mechanisms, 

and the long-term reliability of mitigation 

methods. 

The effectiveness of the various materials, 

fabrication and water chemistry changes 

introduced to mitigate EAC in BWRs and 

PWRs should be confirmed by long-term and 

aggressive confirmatory testing. Some 

mitigation methods may lose effectiveness 

over time, while others (individually or in 

combination) may demonstrate the ability to 

protect critical components from stress 

corrosion cracking for the 60 year life of the 

uprated plants 

 

The project entitled “CIR-Il Cooperative 

Agreement” involves NRC collaboration with 

the international community to develop a 

mechanistic understanding and a predictive 

model of IASCC. This understanding is 

required to ensure that current mitigation 

methods will remain effective as plants age 

and, possibly, to identify more effective 

countermeasures. 

The project on “Environmentally Assisted 

Cracking of LWRs” includes tasks addressing 

IASCC and PWSCC of nickel based alloys. 

Researchers at the Halden Reactor are 

measuring IASCC initiation and growth of 

relevant materials in both BWR and PWR 

environments. This effort includes tests of 

neutron-irradiated specimens to improve the 

understanding of IASCC initiation and the 

influence of stress relaxation on crack growth 

and arrest. It also provides data on the 

performance of electrochemical potential 

probes and monitoring techniques in radiation 

environments. This work is essential and 

should be continued. 

The project on “Investigation of Stress 

Corrosion Cracking in Selected Materials” is 

intended to develop a better understanding of 

the PWSCC mechanism affecting PWRs. 

Understanding the root cause and underlying 

mechanisms of this phenomenon is essential 

for effective long-term mitigation. 

Research addressing environmental effects 

on fatigue of steels used in LWRs has been 

completed. NUREG/CR-6909 Rev.1 and 

Regulatory Guide 1.207 have been issued 

providing designers and regulators with 

quantitative adjustment factors to account for 

the effects of environment on fatigue life of 

reactor materials. 

Proactive Materials Degradation 

Assessment 

The nuclear industry and the NRC have often 

been surprised by unexpected material 

degradation events. The project “Proactive 

Material Degradation Assessment” is an NRC 

initiative to identify materials and systems in 

LWRs where degradation can reasonably be 

expected to occur in the future. With such 

knowledge, current inspection and monitoring 

programs at plants could be reviewed and 

modified as needed to provide early 

identification of incipient degradation before it 

affects plant safety. 

The staff completed Phase 1 of the project in 

2008. A comprehensive assessment of the 

likelihood and safety significance of possible 

environmental degradation mechanisms was 

completed for approximately 1900 BWR and 

PWR components, and NUREG/CR-6923 

documenting this work was issued. The 

objective of Phase 2 of the project was to 

establish agreements with industry and 

international organizations to define research 
37

tasks addressing the identified issues of 

greatest concern. The Zorita Internals 

Research Program was such an activity in 

which high fluence core internal components 

which had operated for over 30 years would 

be examined. Today the program has evolved 

to: 

• establishing a refined plan 

• establishing international 

collaboration 

• targeting items of low knowledge 

and high/medium susceptibility to 

degradation, and 

• creating an information tool. 

The ACRS enthusiastically supported the 

initial vision of the Proactive Materials 

Degradation Assessment Program and has 

recommended its continuation. However, the 

program seems to have lost its momentum 

and focus. Current research appears to 

concentrate on conventional reactive 

management instead of anticipating and 

preventing materials degradation surprises. 

The work published in NUREG/CR-6923 

identified target items of low knowledge and 

high-to-medium susceptibility to degradation. 

These target items were essentially 

predictions of possible materials degradation 

problems which had yet to occur. They 

represented areas of greatest uncertainty and 

concern to the experts involved in the 

identification process. It would seem that the 

next step in the proactive process should be 

the validation of one or more of these 

predictions by experimental means or by 

focused inspections at operating plants. In the 

absence of a quantitative confirmation of the 

predictions made in NUREG/CR-6923, it is 

doubtful that the program can attain its 

primary goal. If such confirmations are not 

possible, continuation of the program should 

be reconsidered. 

Steam Generator Tube Integrity 

Rupture of steam generator tubes in PWRs 

can lead to accidents that allow radioactive 

materials released from the core to bypass 

the reactor containment and enter directly into 

the environment. Severe accidents involving 

containment bypass can be risk dominant in 

some PWRs. Degradation has been observed 

on both primary and secondary sides of 

steam generator tubes, and many different 

phenomena have been observed including: 

general corrosion and wastage, denting, 

crevice corrosion, pitting, intergranular attack, 

and stress corrosion cracking. 

Research and development by the industry 

and by the NRC under the Steam Generator 

Action Plan has been very effective in 

reducing the number and severity of steam 

generator materials degradation events This 

has been achieved by improved water 

chemistry control, development of reliable 

NDE methods, and the introduction of 

improved tube support designs and tubing 

materials such as Alloy 690 in replacement 

steam generators. 

NRC research has concentrated on the 

assessment and improvement of nondestructive crack detection methods, the 

development of analytical models to predict 

initiation and growth of stress corrosion 

cracks, and the development of analytical 

models to predict leak rates and rupture of 

degraded steam generator tubes. 

With the closure of the Steam Generator 

Action Plan, the vast amount of knowledge 

gained during this long term research 

program is to be thoroughly documented. 

Research will continue, to assure that that the 

safety of steam generators will persist over 

the life of these systems. A variety of 

advanced NDE and signal analysis 

techniques are being evaluated for inspecting 

original or repaired steam generator tubes. 

An analytical model based on improved 

understanding of the mechanisms of initiation 

and propagation of stress corrosion cracks in 

nickel based alloys, and the influence of 

crevice conditions will be developed and 

validated. 
An analytical model to predict leak rates or 

rupture of degraded steam generator tubes 

under normal or postulated accident 

conditions will be developed and validated 

experimentally. 

 

Figure 4. PWSCC Safety Assessment Elements 

38
10. NEUTRONICS AND CRITICALITY SAFETY 

Background 

Neutronics analysis and criticality safety are 

among the core competencies essential to the 

pursuit of the NRC mission. Neutronics and 

criticality issues are unique to the nuclear 

sciences. The licensees and, indeed, the 

public expect that the agency to have 

capabilities in these areas that are at or very 

near the state-of-the-art. There is limited 

availability of independent expertise from the 

academic community or the private sector in 

these areas. It is, then, difficult for the agency 

to rely on external contractors to provide the 

needed expertise to independently analyze 

neutronics and criticality issues. NRC must, 

then, have a strong, in-house, capability in 

these areas to make technically sound 

regulatory decisions without imposing 

conservatisms that are actually unnecessary 

when the state-of-the-art is examined. The 

in-house capability certainly must include the 

ability to execute analyses with the available 

computational tools. It must also include a 

respectable understanding of core physics 

that computer codes attempt to estimate. 

39

Currently, the NRC relies on the SCALE 

(Standardized Computer Analysis for 

Licensing Evaluation) suite of computer codes 

to analyze neutronic issues associated with 

fuel loading in nuclear reactors. This suite of 

codes is maintained by Oak Ridge National 

Laboratory. The SCALE computer codes 

were developed in the era of batch computer 

codes and do not take advantage of modern, 

multiprocessor computational architecture 

with large memory machines. The SCALE 

code suite, then, provides an adequate 

capability for analysis of the current fleet of 

operating reactors. The cores for the current 

reactors are simple and have significant 

symmetry. Such cores do not challenge the 

computational capabilities of the SCALE code 

suite. 

SCALE/TSUNAMI 

SCALE Tools for Sensitivity and UNcertainty 

Analysis Methodology Implementation 

(SCALE/TSUNAMI) utilizes first-order-linear 

perturbation theory to produce the sensitivities of a 

computed keff value to constituent cross-section 

data. The energy-dependent sensitivity data for 

each reaction of each nuclide in a system model 

can be quickly computed using 

SCALE/TSUNAMI's 1-D and 3-D analysis tools. 

These sensitivity data can be coupled with crosssection-covariance data to produce an uncertainty 

in keff due to uncertainties in the evaluated nuclear 

data. 

In addition to the SCALE code suite, NRC has 

the PARCS (Purdue Advanced Reactor Core 

Simulator) neutronics analysis code used in 

conjunction with the TRACE thermal-hydraulic 

code. PARCS does derive some of its input 

from the SCALE analyses. 

There are economic driving forces that push 

reactor designers to wring from reactor cores 

as much profitable power generation as 

possible. As a result, nuclear cores for 

advanced light water reactor designs being 

submitted for certification are not so 

symmetrical or simple to analyze as the cores 

of existing light water reactors. Local 

asymmetry, great variability in fuel 

enrichment, partial length rods, and burnable 

poisons complicate the neutronics analysis of 
40

these cores. Core analysis is becoming more 

time consuming. Complexity of analysis 

coupled with greater demands for multiple 

analyses to quantify uncertainties in 

computed results are taxing the capabilities of 

NRC’s resources for neutronics analysis. The 

labor involved in individual analyses limits the 

opportunities for exploration of uncertainties 

and low probability events that may have high 

consequences. 

Challenges to the current capabilities for 

neutronics analysis will grow as more 

advanced reactor designs based on 

technologies other than conventional light 

water are submitted for certification. The 

“pebble bed” reactor core for gas-cooled 

reactors certainly appears to pose challenges 

for neutronics analysis that will not be easily 

met with current computational resources. 

Any move to fuel enrichments greater than 

5% 235U will pose challenges to the currently 

available data bases supporting 

computations. 

Current Research Activities 

The current NRC research activities in 

neutronics analysis and criticality safety are 

depicted in Figure 5. Research activities are 

focused on maintenance and application of 

current capabilities. There are also modest 

efforts to apply these current capabilities to 

issues that the agency will have to face in the 

future. 

Assessment and Recommendations 

NRC has established an adequate neutronics 

analysis capability. It has done well 

maintaining this activity and is trying to find 

ways to apply the capability developed in a 

previous era to regulatory issues that arise 

today. There is not a crisis of need now, but 

the application of existing capabilities will 

grow more challenging. Labor intensive work 

to analyze the core of the ESBWR design is 

indicative of the challenges that will grow 

more common as the light water reactor 

technology advances. The demand for more 

extensive, quantitative analysis of 

uncertainties (in the main, parametric 

uncertainties) for neutronics analysis can be 

met by Monte Carlo methods, but these 

methods require at least hundreds of 

calculations for each core configuration. 

Such extensive calculations are not practical 

with existing computational capabilities when 

applied to complicated core designs. 

NRC needs to establish a long term strategy 

to upgrade its neutronics analysis capabilities 

to take advantage of modern computer 

architecture in a way that facilitates the 

analysis of low symmetry reactor core 

including quantitative analysis of parametric 

uncertainty. 

A move to reactors based on technologies 

other than light water is not imminent. There 

are, however, DOE initiatives in both gascooled reactors and sodium-cooled reactors 

that will challenge NRC neutronics analyses 

capabilities. Some of the challenges are 

associated with data bases. NRC can 

encourage the DOE and foreign counterparts 

to cooperate in the development of data 

bases adequate for safety analyses of 

advanced reactor concepts and reactors 

using higher fuel enrichments and new fuel 

compositions such as mixed oxide (MOX) 

fuels enriched in other actinides. 
Neutronics and Criticality 

Safety Research 

• Nuclear Analysis 

for non-LWRs 

• Development of 

Parallel SCALE 

Calculational 

Capability

• Neutronics for 

Greater than 5 

Weight Percent 

Enrichment

Independent Evaluation of Licensees’ Submittals in 

the Area of Neutronics and Criticality Safety

(A core competency essential to the agency mission) 

Maintenance of Current 

Code Capabilities 

Application of Current 

Capabilities 

Issues that the Agency will 

have to Face in the Future 

• SCALE Code 

Infrastructure

• PARCS 

Development, 

Maintenance and 

Assessment 

• Technical 

Support for 

Criticality Safety 

Analysis

• Neutronics for 

Reactor/Spent 

Fuel Analysis 

Figure 5. Current NRC Research Activities in Neutronics and Criticality Safety

41
42
11. OPERATIONAL EXPERIENCE 

Background 

When properly documented and analyzed, 

operational data provide an invaluable source 

of information that can be used to refine the 

regulatory process and improve its 

effectiveness. With the increased use of risk 

information in the regulatory process, 

operational data become even more 

important to assure the risk information has 

an adequate basis for use in decision making. 

In the operational experience research area, 

there are several programs aimed at 

capturing appropriate data, analyzing the 

data, and providing the tools necessary to use 

the data in the regulatory process. In 

addition, the operational data is used as a 

measure of the regulatory effectiveness and 

inputs to the annual report to Congress on 

significant operating events. 

RES Data Collection, Analysis, and 

Trending Program 

The operational experience research program 

is an integrated system of data collection, 

data coding, and tools for use of the data. 

Although the programs do not fall within a 

single branch within RES, the coordination 

between the various groups within RES and 

NRR appears to be good and results in 

products that improve the regulatory process. 

These efforts yield improved NRC PRA 

models for evaluating existing plants as well 

as new reactor designs submitted for 

certification. The operational research 

activities are providing quantitative measures 

for the effectiveness of the regulatory 

process. 

Current Research Activities 

The RES Data Collection, Analysis, and 

Trending Programs and their relationship to 

each other are graphically shown in Figure 6. 

For example, activity N6448, “Access to 

INPO’s EPIX System,” is providing a 

mechanism whereby NRC staff has access to 

some of the Institute of Nuclear Power 

Operations (INPO) operational data through 

EPIX so it can be used by the NRC. The data 

is then coded and integrated with other 

sources of information under research activity 

N6632, “Reactor Operating Experience Data.” 

This integrated data can then be used as 

inputs to PRA models, special studies under 

activity N6890, “Special Reliability Studies,” 

and risk-based operating experience analyses 

under activity N6631, “Computational Support 

for Risk Applications.” All of these ultimately 

provide input to programs and processes 

such as the Significance Determination 

Process, Industry Trends Program, Inspection 

Program, and the Accident Sequence 

Precursor Program. 

Assessment and Recommendations 

The Operational Experience Research 

program is being managed and executed in a 

manner consistent with the needs of the NRC. 

It provides data and tools necessary for 

regulatory decision making and for the 

assessment of regulatory effectiveness. 

There appears to be good coordination 

between the research staff and the user 

organizations. The Operational Experience 

Research program should be continued. 

43
44

Figure 6. Current NRC Research Activities in Operational Experience

Operational Experience 

Research 

• Risk-based 

Operating 

Experience 

Analyses 

 - CCF Parameters 

 - Initiating Events 

 - System Studies 

Data Collection Development of Tools 

and Databases 

Industry Trends Support 

• Access to EPIX 

Data 

• Integrated Data 

Collection and 

Coding System 

 - EPIX 

 - MSPI 

 Unavailabilities 

 - MORS 

 - LERs 

 - Fire Events 

 

• CCF Database 

• LER Search 

• Fire Event 

Database Support 

• SPAR Models 

Industry Trends 

Program 

Reactor Oversight 

Process 

ASP 

Program 

Operating 

Experience 

Clearinghouse 
12. PROBABILISTIC RISK ASSESSMENT 

Background 

Probabilistic risk assessment (PRA) is an 

essential technology for NRC as the 

regulatory system evolves to make greater 

use of risk information. Therefore, NRC must 

have state-of-the-art probabilistic risk 

assessment capabilities. While the agency 

has been responsible for the development of 

methods in widespread use today, much of its 

recent research work has been in the area of 

applications. Development of methods has 

not been a priority. The extensive use of risk 

information by both the industry and the staff 

in regulatory decision-making, the reviews of 

new reactor designs, and articles published in 

the literature have identified areas where 

methodological advances would be useful. 

This report focuses on three aspects of the 

NRC research program on (PRA)—

organization of the research program, 

technical content including recent 

advancements and continuing/future needs, 

and independent capability. 

Evolving Organization: Management of the 

PRA research program is evolving from a 

reactive mode to an organized goal-driven 

process. A reasonable first step has been 

accomplished—mapping the process as it 

currently exists. The PRA research program 

managers describe the following four steps in 

the current process: 

• Understanding, cataloging and taking 

advantage of the previously divergent 

drivers of PRA research (SRMs, User 

Needs, Commission mandates, ACRS 

recommendations, emergent needs, etc.) 

• Formally stating the goals (issue 

resolution, application support, mission 

support) 

• Detailing the alternative mechanisms for 

funding research (staff work, contracts, 

agreements and MOUs, and 

collaboration) 

• Identifying and tracking anticipated 

outcomes 

Activity is underway to move toward a more 

formally organized and managed process 

focused on meeting both long and short-term 

goals. One especially encouraging aspect of 

the process is close cooperation between 

Division of Risk Analysis of RES and its 

counterparts in other NRC line organizations 

(NRO, NRR). Through regular meetings, they 

have begun to structure current User Needs 

to advance long-term agency goals. 

The ACRS looks forward to a future goaldriven process. More work is needed to 

identify, organize, and structure the long-term 

PRA research goals. With a clear idea of what 

is needed and the time frames in which those 

needs ought to be realized, RES can be 

better prepared to take advantage of new 

opportunities as they present themselves. 

45
46

Recent Advancements: In its 2006 report to 

the Commission on the NRC Safety Research 

Program (NUREG-1635, Vol. 7), the ACRS 

recommended that “the staff needs to review 

the literature concerning Binary Decision 

Diagrams and evaluate the need to adopt this 

technology.” The Committee followed up in 

its 2008 report (Vol. 8) explaining that “[new] 

calculation methods, such as those based on 

Binary Decision Diagrams and EPRI’s Direct 

Probability Calculation (DPMTM), are exploring 

the possibility of eliminating the 

approximations in current methods... The 

ACRS continues to believe that the staff 

should undertake such an investigation and 

participate actively in the international 

activities related to the development of the 

next generation of PRA software. A shortterm product could be an evaluation of the 

potential impact of the new software on 

regulatory decisionmaking. The ACRS was 

informed that there are efforts under way to 

develop a standard PRA representation 

format which would promote independence 

between the logic models and the individual 

codes. Such a standard would facilitate 

significantly the review of the PRA logic, a 

benefit of great value to the agency.” RES 

now is following this work and is trying to 

understand how such a standard can support 

improvements in NRC PRA tools and review 

capabilities. 

In its 2008 report to the Commission on the 

NRC Safety Research Program (NUREG1635, Vol. 8), the ACRS also pointed out that 

the “quantification of uncertainties is an 

essential element of risk-informed 

decisionmaking. The uncertainties in the 

values of input parameters to the PRA are 

usually handled well and there are many 

software packages that propagate these 

uncertainties to the output quantities. 

Uncertainties in the models themselves due 

to questionable or plausible alternative 

assumptions are still not included routinely in 

PRAs. RG 1.174, Revision 1, contains a 

discussion of these uncertainties but does not 

provide acceptable methods for handling 

them. There is a need for specific guidance 

on what we mean by model uncertainties, 

how to identify them, and how to quantify or 

manage them in decisionmaking.” 

With publication of NUREG-1885, “Guidance 

on the Treatment of Uncertainties Associated 

with PRAs in Risk-Informed Decision Making,” 

the staff has taken a first step forward in 

providing needed guidance on the nature and 

treatment of uncertainty within the context of 

Level 1 PRA. There is still a need for a 

common approach for dealing with 

uncertainty throughout the agency. 

RES has been actively attacking a number of 

short-term projects to resolve issues in PRA 

and to provide PRA support to other offices. 

The more significant ones include substantial 

efforts in the PRA standards and 

development of related guidance; digital I&C 

PRA, SPAR models, SAPHIRE 8 

development, uncertainty analysis methods 

and guidance, and advanced reactor PRA 

scoping studies. 

Continuing Gaps: We have been urging 

NRC to develop agency-wide common 

approaches for the treatment of uncertainties 

and elicitation of expert opinions, for many 

years. Individual research projects and 

regulatory applications continue to be 

developed and presented to the ACRS that 

invent “new” approaches for dealing with 

uncertainties or, worse, ignore them 

altogether. 

The ACRS has discussed the need for an 

agency-wide common approach for the 

elicitation of expert opinions in the past. 

Development of risk information for regulatory 

decision-making requires often the elicitation 

of expert opinions. Important examples 

include the evaluation of the frequencies of 

LOCAs of various sizes in the context of the 

efforts to risk-inform 10 CFR 50.46 (NUREG1829) and the study (jointly done with DOE 

and EPRI and reviewed by a National 

Research Council committee) focused on 

expert opinion elicitation in the assessment of 

seismic risk (NUREG/CR-6372). Of course, 
47

the first studies to formalize the use of expert 

opinions in nuclear plant risk assessments 

were conducted as apart of the NUREG-1150 

Study. Numerous other examples involving 

expert opinion utilization can be found within 

the agency including examples in 

performance assessments for waste 

repositories and the ATHEANA model for 

human reliability assessment. Few of these 

studies build on previous NRC-sponsored 

approaches. We continue to urge that the 

staff combine the best attributes of these 

studies and develop a systematic 

methodology for the elicitation and processing 

of expert opinions that should be used 

agency-wide 

Future Needs. The staff has identified a 

number of longer-term PRA research needs 

including the integrated treatment of 

phenomena, enhanced data, advanced level 

2/3 PRA methods development including 

methods to analyze digital I&C systems 

(especially software failures) and passive 

systems. It has been nearly twenty years 

since NRC sponsored the NUREG-1150 

PRAs. We think it is time for an NRC Level 3 

PRA to demonstrate advances in PRA 

methods and suitable approaches for new 

reactors. Such an undertaking would require 

a great deal of advanced planning beginning 

with its scope. Would it include both internal 

and external accident initiators? Would it 

include shutdown operations and spent fuel 

considerations? Would it be confined to 

existing plants or would advanced light water 

reactors be included? Would mixed oxide fuel 

be considered? 

Since September 11, 2001, the security of 

nuclear power plants has been a major area 

of activity of the agency. Many new 

requirements have been imposed on 

licensees regarding both physical and cyber 

security. While there is no doubt that security 

has been improved, there is a growing 

suspicion that many of these requirements 

are not improving the security of facilities that 

are already secure enough. This, of course, 

means that they may constitute unnecessary 

regulatory burden. 

The Commission’s PRA Policy Statement of 

1995 stated that PRA use should be 

increased in regulatory matters, where 

practical within the bounds of the state-of-the 

–art, to reduce unnecessary conservatisms 

associated with the current regulatory 

requirements, regulatory guides, license 

commitments and staff practices. The ACRS 

believes that it is time the agency applied 

these principles to the security arena. 

PRAs contain a wealth of information 

regarding the ways undesirable plant states, 

such as core damage and large release of 

radioactivity, can occur. This information is 

not utilized in formulating security 

requirements to evaluate their benefit and 

their impact on safety. The ACRS 

recommends that RES establishes a research 

project to explore the possibility of riskinforming security requirements and building 

on PRAs to create a unified framework for the 

evaluation of both safety and security. It is 

likely that security would be improved, with a 

reduction in added regulatory burden. 

Independent Capability: Staff describes its 

primary role as conducting independent 

confirmatory research. The ACRS agrees 

and suggests that the following key elements 

of the goals identified by the staff and 

depicted in Figure 7 are required in order to 

develop and maintain cutting edge capability 

to conduct independent confirmatory research 

to ensure that the regulator’s perspective on 

safety is adequately supported in the PRA 

arena: 

• Supporting the reactor oversight and 

operating experience programs 

• Facilitating the implementation of riskinformed regulation 

• Expanding PRA infrastructure to 

encompass new reactor designs 
48

• Supporting continuous advancement in 

PRA state-of-the-art and state-of-practice 

The structure shown in Figure 7 could provide 

a basis for developing long-term research 

goals that can provide a constant vision of 

where the PRA research program should be 

headed. Continued and expanded 

international agreements and active domestic 

MOUs can support the long-term research 

program and leverage NRC’s research 

budget. The ACRS supports the growing 

participation in these activities. 

PRA issues related to digital I&C and HRA 

are discussed in Chapters 5 and 8, 

respectively.
49

Probabilistic Risk Assessment 

Research 

Independent Capability to Provide Quantitative Assessment of 

Risk for Regulatory Activities 

(An essential technology for NRC as it evolves the regulatory system 

to make greater use of risk information) 

Supporting the Reactor 

Oversight and Operating 

Experience Programs 

• Support for 

RASP Program 

• Support for ASP 

Programs 

• SPAR Model 

Development and 

Maintenance 

• Implementation of 

Phased Approach to 

PRA Quality 

• Support for 

Development of PRA 

Standards 

• Support for WGRisk 

Digital Risk Task 

• Software Reliability 

Quantification 

Expanding PRA 

Infrastructure to 

Encompass New 

Reactor Designs 

Supporting Continuous 

Advancement in PRA 

State-of-the-art and Stateof-practice 

• Advanced Level 2/3 

PRA Modeling 

• Dynamic PRA 

(Collaborative 

Agreement with Univ. 

of Maryland) 

• Sensitivity / 

Uncertainty 

(Collaborative 

Agreement with 

MIT) 

Facilitating the 

Implementation of Riskinformed Regulation 

• Advanced Reactor 

PRA Scoping Study 

• Development of 

NEW Reactor 

SPAR Models 

• Development of 

Regulatory 

Guidance for 

Establishing a Riskinformed Licensing 

Basis 

Figure 7. Current NRC Research Activities in PRA
50
13. RADIATION PROTECTION 

Background 

The NRC has maintained a program of 

research related to radiation protection in the 

areas of: 

• risks from radiation, 

• the sciences of internal and external 

dosimetry, and 

• the fate and transport of radioactive 

materials in the human body and in the 

environment. 

A major thrust of this research is to collect, 

analyze, and disseminate information on 

occupational exposures reported to NRC by 

licensees. This information is used to track 

the effectiveness of licensees’ As Low As 

Reasonably Achievable (ALARA) programs 

and will form the basis for future studies to 

evaluate the health effects of this group of 

workers. Another important thrust is to 

develop and maintain tools for assessments 

related to licensing, siting, environmental 

performance, and the decontamination and 

decommissioning of licensed facilities. 

51

Current Research Activities 

The current NRC research activities in the 

area of Radiation Protection are depicted in 

Figure 8. These activities are leveraged in 

several ways. First, these activities are based 

on continued staff participation on national 

(e.g., The National Council on Radiation 

Protection and Measurements, NCRP, the 

National Academy of Sciences, NAS) and 

international committees (The International 

Commission on Radiological Protection, 

ICRP, and Committees of the International 

Atomic Energy Agency, IAEA). NRC 

research is also well leveraged by working 

with a number of collaborating agencies, 

including the U.S. Environmental Protection 

Agency. 

Radiological Toolbox Graphical User 

Interface 

The staff is engaged in a number of 

collaborative activities that support the 

assessment of the behaviors of radioactive 

material in the environment and ultimate dose 

consequence analysis. 

These are all essential activities and need to 

be sustained. 

Health Effects/Dose Calculation Tools

Health effects/dose calculation tools are used 

to model and assess the health implications 

of radioactive exposure and contamination. 

VARSKIN: The NRC sponsored the 

development of the VARSKIN code in the 

1980s to assist licensees in demonstrating 

that they have approved radiation protection 

programs that include established protocols 

for calculating and documenting the dose 

attributable to radioactive contamination of 

the skin. Since that time, the code has been 

significantly enhanced to simplify data entry 

and increase efficiency. 
52

Since the release of VARSKIN 3 in 2004, the 

NRC staff has compared its dose calculations 

for various energies and at various skin 

depths, with doses calculated by the Monte 

Carlo N-Particle Transport Code System 

(MCNP) developed by Los Alamos National 

Laboratory (LANL). That comparison 

indicated that VARSKIN 3 overestimates the 

dose with increasing photon energy. For that 

reason, the NRC is sponsoring a further 

enhancement to replace the existing photon 

dose algorithm, develop a quality assurance 

program for the beta dose model, and correct 

technical issues reported by users. 

Radiological Toolbox: The NRC developed 

the radiological toolbox as a means to quickly 

access databases needed for radiation 

protection, shielding, and dosimetry 

calculations. The toolbox is essentially an 

electronic handbook with limited 

computational capabilities beyond those of 

unit conversion. Further revisions of the 

toolbox are planned as the need for additional 

data is identified by NRC staff and other 

users. The toolbox contains radioactive decay 

data, biokinetic data, internal and external 

dose coefficients, elemental composition of a 

large number of materials, radiation 

interaction coefficients, kerma coefficients, 

and other tabular data of interest to the health 

physicist, radiological engineer, and others 

working in fields involving radiation. The 

toolbox includes a means to export the 

tabular data to an Excel worksheet for use in 

further calculations. It operates in a Windows 

environment. 

Radionuclide Transport Codes (for License 

Termination and Decommissioning)

Radionuclide transport codes provide dose 

analyses in support of license termination and 

decommissioning: 

DandD: DandD is a code for screening 

analyses for license termination and 

decommissioning. The DandD software 

automates the definition and development of 

the scenarios, exposure pathways, models, 

mathematical formulations, assumptions, and 

justifications of parameter selections 

documented in Volumes 1 and 3 of 

NUREG/CR-5512. 

Probabilistic RESRAD 6.0, RESRAD-BUILD 

3.0, and RESRAD-OFFSITE Codes: The 

existing deterministic RESRAD 6.0 and 

RESRAD-BUILD 3.0 codes for site-specific 

modeling applications were adapted by 

Argonne National Laboratory (ANL) for NRC 

regulatory applications for probabilistic dose 

analysis to demonstrate compliance with the 

NRC's license termination rule (10 CFR Part 

20, Subpart E) according to the guidance 

developed for the Standard Review Plan 

(SRP) for Decommissioning. The 

deterministic RESRAD and RESRAD-BUILD 

codes are part of the family of codes 

developed by the U.S. Department of Energy. 

The RESRAD code applies to the cleanup of 

sites and the RESRAD-BUILD code applies to 

the cleanup of buildings and structures. 

Current RESRAD work is focused on 

RESRAD-OFFSITE and improving the 

treatment of complex source terms 

Cancer Incidence and Mortality Data for 

Population Living Near NRC-Licensed 

Nuclear Power Facilities

The objective of this proposed study is to 

provide the NRC with the latest cancer 

mortality data for populations living near 

NRC-licensed nuclear power facilities and to 

examine the feasibility of studying cancer 

incidence among these populations. This 

study will provide the NRC with the most 

current scientific information available for 

responding to stakeholder concerns related to 

cancer mortality rates of populations that live 

near past, present, and proposed nuclear 

power facilities. 

The Committee supports the staff’s decision 

to establish an external review committee to 

review and comment on study method, 

protocol, and final deliverables. 
 

Assessment and Recommendations 

The Committee believes that the staff has 

developed an appropriate and robust 

research program in the areas of radiation 

protection. This program includes radiation 

protection of workers and radiological 

assessments related to decommissioning and 

waste management. 

Radiation Protection 

 Research 

• Cancer Study in 

Populations Near NPPs

• Development of 

Strategies for the 

Revision of 10 CFR 20 

Standards for 

Protection Against 

Radiation

Independent Evaluation of Licensees’ Submittals in 

the Area of Radiation Protection

(A core competency essential to the agency mission) 

Development and 

Maintenance of Health 

Effect/Dose Calculation 

Tools 

Development and 

Maintenance of

Radionuclide Transport 

Codes

Emerging Research and 

Development

• Varskin Code 

• Radiological Toolbox 

• DandD code

• RESRAD 6.0 and 

RESRAD-BUILD 3.0 

Codes 

Figure 8. Current NRC Research Activities in Radiation Protection

53
54
14. NUCLEAR MATERIALS AND WASTE 

Background 

55

The NRC continues to maintain a robust 

program of research related to nuclear 

materials and radioactive waste topics related 

to licensing, facility siting, facility 

environmental performance, and the 

decontamination and decommissioning of 

licensed facilities. 

Current Research Activities 

The staff is engaged in a number of research 

and development projects. These projects 

cover a wide range of topics from 

fundamental studies on the fate and transport 

of radioactive materials in the environment to 

modernization of computer codes (GALE) 

used in risk-informed analyses and regulatory 

decisionmaking. These activities support both 

specific licensing actions and generic issues. 

Three projects are summarized below. 

Integrated Groundwater Modeling and 

Monitoring

 

NRC research is well engaged in a 

collaborative effort with Pacific Northwest 

National Laboratory (PNNL) and the U.S. 

Geological Survey (USGS) regarding uranium 

sequestration using in situ bioremediation. 

The goal of this effort is to determine the longterm efficacy of the use of bioremediation 

technologies to capture and sequester 

uranium in situ. Research includes field and 

laboratory studies of geochemical, 

hydrogeologic, and microbial processes, and 

the performance indicators needed to confirm 

long-term sequestration. Through this 

collaboration, the NRC has been able to 

participate in research that is being used by 

Office of Federal and State Materials and 

Environmental Management Programs 

(FSME) in regulatory decision-making 

regarding pending licensing actions for 

decommissioning sites, and for evaluating 

proposed future remediation strategies for in 

situ uranium leach mining. 

Flow Chart of the Integrated Monitoring 

Strategy 

GALE Code Updates 

Research is coming to closure on two projects 

to update both the computing platform and 

parameter values that are used in the GALE 

codes. Data reflecting current plant operating 

experience have been added to the code. 

Further plans are being developed for 

additional updates to incorporate new reactor 

information as it becomes available. 

Flood Hydrology

Research is underway regarding flood 

assessments in collaboration with the U.S. 

Army Corps of Engineers, Bureau of 

Reclamation (U.S. Department of the Interior), 

and PNNL regarding analytical tools and 

guidance on hurricane storm surges, probable 

maximum precipitation estimates, and flood 

assessments, respectively. The NRC senior 

technical staff representative now chairs the 

Federal Interagency Work Group on Extreme 

Storm Events [under the Federal 

Subcommittee on Hydrology (SOH) under the 

Advisory Committee on Water Information 

(ACWI)]. Participation in the SOH ensures 

broad coordination of information among 
56

participating Federal agencies on flood 

hydrology data sources, methods, and 

guidance. Both the research efforts and 

interagency activities will provide the technical 

basis for updating Regulatory Guide 1.59, 

“Design Basis Floods for Nuclear Power 

Plants.” Both are contributing information to 

NRO licensing staff in their hydrology reviews 

for new nuclear power plants. 

Assessment and Recommendations

The staff has been particularly successful in 

leveraging its efforts with those of other 

federal agencies and other organizations. 

These collaborations are productive and 

provide significant synergy with NRC 

research activities. These collaborative 

activities should continue and be encouraged. 

Research has also been effective at 

producing results that support a range of 

topics important to regulatory decision 

making. 
15. SEISMIC AND STRUCTURAL ENGINEERING 

Seismic events come to mind commonly 

when discussions of nuclear power plant 

safety turn to the topic of accident initiation by 

events external to the plant. Nuclear power 

plants are designed, of course, to cope with 

substantial seismic events – up to at least the 

safe shutdown earthquake (SSE). 

Earthquakes of greater magnitude have the 

potential to damage the power plant and 

barriers to radionuclide release. Earthquake 

damage can be a common cause mechanism 

for the failure of methods to prevent accidents 

and methods to mitigate accidents – 

extending even to the final line of defense in 

depth of public evacuation and emergency 

planning. 

57

The seismicity of the Western US is relatively 

well known. Research attentions are properly 

placed on the seismicity of the country east of 

the Rocky Mountains where events are less 

common and most of the nation’s nuclear 

power plants are located. There are seismic 

centers in the Central and Eastern US that 

have the potential to produce earthquakes 

that could damage nuclear power plants. The 

New Madrid, Charleston, Northeast, and 

Eastern Tennessee seismic centers are 

examples. Recently, the US Geological 

Survey has revised expected “return 

frequencies” for large earthquakes at some of 

these seismic centers to values higher than 

used for seismic hazard analysis in the past. 

Events at these seismic centers can affect 

large areas since seismic attenuation is 

thought to be less in the Central and Eastern 

US than in the Western US. 

The US NRC has revitalized its program of 

research into seismic events and the 

responses of nuclear power plants to these 

events. The research program has been 

developed to support regulatory activities 

especially in the areas of certification of new 

plant designs and combined license 

applications. The seismic research program 

 

Seismicity of the USA 

Seismic centers in the Central and Eastern US are 

of particular interest because the hazard these 

centers pose are not as well understood as 

seismic centers in the West and most US nuclear 

power plants are located in areas affected by 

these centers. 

has elements broadly categorized as: 

• seismic hazard characterization,

• earthquake engineering,

• international collaborations in seismic 

research, and

• development of regulatory guides.

Within these broad categories, the NRC staff 

has identified 40 activities that need to be 

pursued to support regulatory processes. 

The staff has had to prioritize these needs to 

align with available resources. The staff has 

done much to solicit input both from within the 

agency and from the larger technical 

community involved in seismic research and 

engineering to assist in this prioritization of its 

proposed research. The staff has also been 

diligent in its efforts to establish collaborations 

with other interested agencies and institutions 

to leverage resources in pursuit of 17 priority 

research activities currently underway. 
58

Within the broad research category of 

“seismic hazard characterization,” research is 

being conducted to: 

• characterize the seismicity of the 

Central and Eastern US, generally, 

and the East Tennessee seismic zone 

in particular; 

• develop practical procedures for 

implementation of the guidelines for 

updating probabilistic seismic hazard 

assessments – the so-called “SSHAC 

guidelines” on the use of expert 

opinion in analysis of seismic issues; 

• develop improved models of seismic 

attenuation in the Central and Eastern 

US; and 

• characterize the threats posed by 

possible tsunamis to nuclear plants on 

the Atlantic and Gulf coasts of the US. 

Work being done on the implementation of 

the SSHAC guidelines, while very important 

for seismic hazard analysis, may also have 

utility for other agency activities that make 

extensive use of expert opinion to bound and 

describe physical phenomena not now 

completely understood. (See also comments 

on the issues of expert opinion elicitation 

made in connection with review of the 

Probabilistic Risk Assessment research – 

Chapter 12.) 

Work done by the staff in collaboration with 

the US Geological Survey to characterize 

tsunami threats to coastal nuclear power 

plants sites is a notable accomplishment. The 

threat is probably not negligible and comes 

primarily from tsunamis initiated by 

underwater landslides along the continental 

shelf. 

Within the broad category of earthquake 

engineering, the staff is conducting research 

on: 

• random vibration theory for use in the 

review of combined license (COL) 

applications, 

• methods for two-dimensional site 

response analysis, and 

• time-domain analysis of soil-structure 

interactions during seismic events. 

These research efforts hold the promise of 

substantial modernization of staff capabilities 

for independent review of power plant site 

applications and changes in plant licensing 

basis. 

The NRC staff is continuing its collaboration 

with Japan on the large-scale testing of 

structures, systems, and components during 

shaking that simulates earthquakes. Further 

collaborations are taking place in 

understanding the impacts of the 

Kashiwazaki-Kariwa earthquakes that 

exceeded plant design bases in some cases. 

NRC staff is also engaged with the IAEA on 

seismic and tsunami threats to nuclear power 

plants. 

Efforts by the NRC staff to revitalize the 

seismic safety research in support of 

regulatory activities have been exemplary. 

There is a well developed research plan that 

has been broadly reviewed for both technical 

quality and programmatic impact. Credentials 

and expertise of those pursuing the research 

are quite impressive. The research is having 

impact on the regulatory processes and is 

expected to have even greater impact in the 

future. The active research involving broad 

collaborations with the larger seismic 

community is assuring the agency has at 

hand expertise to help resolve novel issues 

that arise as new plant designs are submitted 

for certification and new license applications 

are submitted for staff review. 
16. SEVERE ACCIDENTS AND SOURCE TERM 

Background 

Significant research was undertaken by NRC 

following the accident at the Three Mile Island 

(TMI) Unit 2. Over the last three decades, 

major experimental and model development 

studies provided an improved understanding 

of progression of severe accidents and the 

radiological consequences of such accidents. 

As a result of these studies, the system-level 

computer code, MELCOR, has emerged as 

an integrated tool for predicting severe 

accident progression and source term 

magnitude. The NRC requires such expertise 

and analysis capabilities to help support 

regulatory decisions for operating nuclear 

power plants as well as advanced nuclear 

plant designs. The tools help the staff in its 

transition to a more risk-informed regulatory 

framework. 

Support of regulatory needs implies a level of 

understanding of severe reactor accidents to 

assure adequate protection to the health and 

safety of the public. The NRC made major 

investments in severe accident research to 

achieve the needed understanding. Once this 

immediate need was met, the NRC adjusted 

its investments in severe accident research to 

required levels for on-going analysis and riskinformed activities. 

59

Current Research Activities 

The current NRC research activities in severe 

accident and source term are depicted in 

Figure 9. Severe accident research activities 

are focused on development and usage of 

MELCOR code. Much of the research 

includes international collaborative severe 

accident experimental programs. 

Severe accident research is continuing 

internationally, with on-going programs in the 

Pacific Rim (Japan and Korea) as well as 

EPICUR Facility for the study of 

radioactive iodine behavior in reactor 

containments under accident conditions 
60

Europe (France, Switzerland, and 

Germany).The NRC has an effective strategy 

to maintain its leading position in severe 

accident analytical capabilities as well as to 

maintain its competence in understanding key 

accident phenomena, by collaborations in 

international research programs. The body of 

knowledge coming from NRC’s past 

experimental work and current ongoing 

international efforts are systematically 

incorporated into the MELCOR accident 

analysis code. The NRC Cooperative Severe 

Accident Research Program (CSARP) is an 

annual forum used for the exchange of severe 

accident findings, both experimental and 

analytical. One significant outcome of this 

effort is the adoption of the MELCOR code by 

many countries and institutions as the key 

analytical tool for severe accident analyses. 

MELCOR Code Development and Usage

The MELCOR code is a fully integrated, 

engineering-level computer code whose 

primary purpose is to model the progression 

of accidents in current nuclear power plants, 

advanced reactors and more recently nonreactor systems (e.g., spent fuel pool 

behavior). The MELCOR code was first 

developed as a PRA tool and thus, modeled 

the reactor core, coolant system and safety 

systems as well as containment systems in a 

less detailed manner than more mechanistic 

thermal-hydraulic and fuel rod models. A 

major current activity is to consolidate the 

physical models and capabilities of more 

detailed severe accident codes into 

MELCOR. This effort will provide an efficient 

state-of-the-art code for severe accident 

analyses. In addition, the MELCOR code has 

been adopted by a number of international 

nuclear safety organizations (e.g., IBRAERussia, PSI – Switzerland, IRSN – France) 

and universities (e.g., University of Michigan, 

University of Texas A&M, and University of 

Wisconsin) for use in severe accident 

analysis. 

The MELCOR code and its fission product 

source term transport and consequences 

model (MACCS) are being used as an 

integral parts of the SOARCA program. 

MACCS is a widely accepted tool used for 

consequence analysis, and its continued 

development and support is important to the 

agency for both safety and security regulatory 

applications. More recently, the NRC staff is 

developing a Graphic-User-Interface as part 

of the long-term development strategy for 

MELCOR. Such an advance will allow NRO 

and NRR staff to routinely do safety analyses 

for specific plants (e.g., source term analyses 

as in Regulatory Guide 1.183 for control room 

habitability and license amendments). Such 

an advance will allow staff to do analyses 

themselves on their desktop. The ACRS is 

supportive of these efforts. 

Collaborative Severe Accident Experimental 

Programs

Collaborative severe accident research 

programs that the NRC has joined are making 

good progress, and some of the key 

accomplishments are noted below. 

ARTIST: This program involves experimental 

work underway in Switzerland at PSI. The 

original test objective was to measure the 

aerosol removal on the secondary side of 

steam generators during accidents in a PWR, 

that involved containment bypass due to 

spontaneous or induced steam generator 

tube failure. Such bypass accidents are often 

risk dominant for PWRs. The risk ascribed to 

such accidents may stem from the 

conservatism in the aerosol decontamination 

assumed in accident models. 

ARTIST tests of aerosol transport through 

secondary sides of steam generators is 

complete and has shown that overall retention 

is low (DF ~6) but dependent on particle size. 

ARTIST-II experiments are examining the 

issues of particle bounce and breakup 

discovered in doing the ARTIST work and not 

now modeled in the MELCOR computer code. 

OECD-BIP: The OECD/NEA Behavior of 

Iodine Project (BIP) has been initiated to 
61

provide separate effects and modeling studies 

of iodine behavior in a containment following 

a severe accident. As a part of this project the 

results of three Radioiodine Test Facility 

(RTF) experiments will be provided by Atomic 

Energy of Canada Limited (AECL). These test 

data will be used for validation of improved 

models of radioactive iodine now being 

developed for the MELCOR computer code

OECD-FCI SERENA: The OECD is 

sponsoring the SERENA (Steam Explosion 

Resolution for Nuclear Applications) Phase II 

program, which evaluates the capability of 

existing codes to predict if fuel-coolant 

interactions could lead to threatening, 

dynamic pressure loads in the reactor vessel 

or containment cavity. Experiments are being 

supported and performed at CEA (KROTOS) 

and KAERI (TROI) as part of the SERENA II 

program. Steam explosion models are then 

used by other international participants to 

perform analyses of KROTOS and TROI 

experiments. One specific effort by the OECD 

deserves note; i.e., a web-accessible 

database that documents steam explosion 

experiments and associated publications in 

the public domain. 

OECD-MCCI: The OECD-MCCI program is 

an international collaborative experimental 

study being conducted at the Argonne 

National Laboratory. The original focus was 

on investigating the heat transfer to an 

overlying water layer to cool molten core 

materials interacting with concrete and its 

viability for long-term coolability in currently 

operating reactors. More recently, the 

program has been addressing the long-term 

coolability in advanced light water reactor 

designs using core retention devices; e.g., the 

Evolutionary Power Reactor (EPR). The 

MCCI project has entered a second phase of 

experimentation and consists of two 

experimental efforts: Small-scale Water 

Ingression and Crust Strength Tests 

(SSWICS) and integral Core Concrete 

Interaction Tests (CCI). The SSWICS tests 

focus on quantifying the heat loss to an 

overlying water layer by mechanisms other 

than conduction limited heat transfer, and to 

measure the strength of the crust formed 

during flooding of the melt. The CCI tests are 

focused on resolving uncertainties in axial 

versus lateral power splits and respective 

concrete ablation rates. Most recently, the 

CCI integral tests are examining the melt 

behavior with underlying cooled refractory 

basemat, similar to the core retention concept 

in the EPR. Test results will be used to 

update the capabilities of the MELCOR 

computer code. 

PHEBUS-FP: Phebus-FP has completed its 

major, in-pile experiments and is preparing 

the final report on the last test. The follow-on 

program, Phebus-ISTP, is a collection of 

separate effects projects to pursue specific 

aspects of Phebus-FP findings. The separate 

effects tests are: 

- EPICUR: chemistry of iodine in the 

containment 

- CHIP: fission product chemistry in the 

reactor coolant system 

- MOZART: air oxidation of fuel cladding 

- BECARRE: interaction of steel and 

cladding with boron carbide in steam

- VERDON: fission product release from 

MOX and high burnup fuel in steam and 

air 

The EPICUR project is making progress and 

has resolved some issues. All the technical 

issues will not be completely resolved by the 

conclusion of the effort and it is likely that 

some follow-on will be needed. The CHIP has 

encountered some technical issues and it is 

not clear that all planned experiments will be 

done by the conclusion of the project next 

year. The MOZART project is completed and 

appears to have yielded high quality kinetic 

data. The BECARRE project is entering its 

last phase with larger scale tests underway. It 

has yielded useful data for modeling the 

degradation of boron carbide control rods that 

may be used in future PWRs. The VERDON 

project is a large undertaking to redevelop the 

VERCORS facility and its capability to study 

fission product release from irradiated fuel 
62

pellets under controlled temperature and gas 

composition conditions. This facility will not be 

ready at the end of the project, but the work 

should continue and some mechanism for 

providing NRC with data is being developed. 

Assessment and Recommendations 

The ACRS supports the strategy that the 

NRC staff has developed to support 

regulatory decisions for severe accidents via 

computer code development and 

experimental data analysis and evaluation. 

This approach can successfully maintain and 

update its modeling capabilities for severe 

accident analyses. The planned program 

extensions and continuations of these 

collaborations are well worth the investment. 
Severe Accident and 

Source Term Research 

MELCOR Code 

Development and Usage 

Collaborative Severe 

Accident Experimental

Programs

• Continued 

Maintenance and 

User Support for 

the Code

• MELCOR 

Development and 

Assessment

• Modernization of 

MACCS code 

SOARCA 

Project 

• ARTIST (PSI) 

• OECD-BIP 

• OECD-MCCI 

• SERENA (OECD-FCI) 

• PHEBUS-ISTI 

Independent Evaluation of Licensees’ Submittals in 

the Area of Severe Accident and Source Term

(A core competency important to risk-informed 

regulatory decisions) 

Figure 9. Current NRC Research Activities in Severe Accident and Source Term 

63
64
17. THERMAL HYDRAULICS 

Background 

Evaluation of the effects of thermal hydraulic 

phenomena on nuclear safety has always 

been a central element in the conduct of 

NRC’s regulatory mission. Of particular 

importance has been and continues to be the 

capability to independently confirm thermalhydraulic analyses in licensees’ submittals. 

Early thermal-hydraulic analyses of nuclear 

power plants employed very conservative 

bounding assumptions, assuring large safety 

margins with regard to allowable 

temperatures and pressures over a wide 

range of accident and operating conditions. 

With time, the experimental database and the 

confidence in analytical predictions have 

grown, allowing the NRC to consider 

submittals from licensees employing “bestestimate” thermal-hydraulic analyses together 

with estimates of uncertainties. These 

analyses have grown ever more 

sophisticated. It is necessary for the NRC to 

continue development of state-of-the-art 

thermal-hydraulic computational tools and 

more sophisticated understanding of 

important thermal-hydraulic phenomena. To 

this end, the NRC maintains competence in 

the thermal-hydraulics field and capability to 

conduct confirmatory analysis through its 

research program.

65

Current Research Activities 

The current NRC research activities in 

thermal-hydraulics are depicted in Figure 10 

The research activities focus on the 

development of the TRACE computer code 

for confirmatory analyses of a wide range of 

safety-significant thermal-hydraulic 

phenomena, and supporting experiments. 

There is also a modest effort to develop 

capability in multi-dimensional computational 

fluid dynamics (CFD). 

 

Upper View of the PrimärkreislaufVersuchsanlage (primary coolant loop 

test facility) PKL Test Facility in 

Erlengen, Germany 

Under the auspices of the OECD-NEA, the 

NRC has joined 13 other international 

organizations to conduct thermal-hydraulic 

experiments in the PKL facility. PKL is an 

integral test facility simulating a typical 

western-type pressurized water reactor 

(PWR) and is used for the investigation of the 

thermal hydraulic system behavior under 

accident conditions. 

TRACE Computer Code Development and 

Validation

In the mid-1990s, a prudent decision was 

made that the several primary reactor system 

thermal-hydraulic codes that were in use at 

that time be consolidated into a single code. 

The several codes included RELAP5 (for 

LOCA), TRACP (for PWR LOCA), TRACB 

(for BWR LOCA), and RAMONA (for BWR 

stability). 

 
66

The models, correlations, and solution 

methodologies in these codes did not reflect 

the state-of-the-art and required in-depth 

modernization. It was also recognized that 

they had been designed at a time when 

computer capabilities were limited and 

included many structural features, such as 

memory management, that were no longer 

needed and increased the cost of continued 

code maintenance and development. The 

availability of graphical user interfaces and 

their wide acceptance also suggested the 

desirability of incorporating similar capability 

into the NRC codes. All these considerations 

led to extensive code consolidation, model 

improvements, and implementation efforts, 

culminating in the development and validation 

of the TRACE computer code. 

TRACE is intended to serve as the main tool 

for the confirmatory analyses of a broad 

range of thermal-hydraulic problems for 

current and new reactor designs. It has the 

potential to offer significantly enhanced 

capabilities for state-of-the-art analyses of 

thermal-hydraulic issues. Several important 

technical issues, such as core stability and 

ATWS behavior, involve coupling between 

neutronics and thermal hydraulics and require 

that TRACE be properly coupled to a 

neutronics code like PARCS, an activity that 

is completed now. The integration, validation, 

and testing of the PARCS-TRACE coupled 

code is currently under way so that it can be 

reliably used for confirmatory analyses. 

TRACE also has the capability to interface 

with the CONTAIN code for containment 

response analysis as well as with other 

computational tools, including MATHCAD. 

In response to ACRS recommendations, the 

staff commissioned and completed a detailed 

peer review of TRACE. The main objective of 

the TRACE code peer review was to identify 

the strengths and deficiencies of the code, 

and provide recommendations for code 

changes and improvements. The peer review 

found no major deficiencies that would 

introduce significant errors or preclude the 

use of TRACE for analysis of postulated 

LOCAs in current LWRs. Certain 

improvements were recommended for 

treatment of the momentum equations, which 

could be particularly important for passively 

cooled systems, though TRACE was not 

explicitly reviewed for applicability to such 

systems. 

Thermal-hydraulic codes, including TRACE, 

solve an intertwined structure of approximate 

conservation equations and empirical 

correlations. The uncertainties and biases 

inevitably introduced by such empirical 

procedures need to be properly addressed. 

Because of these uncertainties, predictions of 

such codes are adequately accurate only 

within certain ranges of parameters. The 

codes cannot be given blanket approval for all 

situations to which they might be applied. In 

practice, a code, such as TRACE, must be 

qualified by assessment against a range of 

data that cover the phenomena that dominate 

the prediction of figures of merit, such as 

peak clad temperature, important to the 

regulatory process. These dominating 

phenomena change with the reactor systems 

and accident conditions being considered. In 

view of this, thermal-hydraulic codes need to 

be assessed for analyses of a specific 

accident in a particular system. 

RES has initiated a systematic assessment of 

the applicability of TRACE to analyze new 

reactor designs. Work on a detailed 

assessment of the applicability of TRACE to 

analyze ESBWR LOCAs, focusing on the 

collapsed liquid level in the reactor pressure 

vessel as the primary figure of merit, has 

already been completed. Ongoing work on 

assessment of the applicability of TRACE for 

confirmatory analyses of safety-significant 

thermal-hydraulic phenomena in the USAdvanced Pressurized Water Reactor 

(APWR) and EPR designs should be 

completed in a timely fashion to allow 

applications in the design certification 

process. 

RES has also been participating in two 

cooperative international agreements to 
67

obtain high-quality experimental data to refine 

best-estimate calculations. The first 

cooperative agreement, which has recently 

concluded, was the NUPEC BWR Full-size 

Fine-Mesh Bundle Tests (BFBT) benchmark. 

Participation in this activity allowed the NRC 

to obtain a database of sub-channel void 

fractions, pressure drops, and critical power 

measurements from a representative BWR 

fuel assembly from the Nuclear Power 

Engineering Corporation (NUPEC) in Japan. 

The extensive database and contributions of 

the benchmark participants will be used to 

improve the predictive capabilities of TRACE. 

A second international cooperative 

agreement, centered on a PWR database, 

has been recently put into place. The 

NUPEC PWR Sub-channel and Bundle Test 

(PSBT) benchmark will use a high-quality 

database of PWR void fraction and DNB data 

provided by METI/JNES. This database will 

be used to assess TRACE code models. Like 

the BFBT benchmark, the PSBT agreement 

will involve the participation of international 

experts who will take part in workshops 

intended to facilitate open discussion of the 

database and the modeling techniques used 

in other thermal-hydraulic codes. Under this 

agreement, METI/JNES will provide the 

experimental data, OECD/NEA will be in 

charge of the logistics of the event, and 

NRC/PSU NEP will prepare the benchmark 

specifications and coordinate the workshops. 

Multi-Application Small Light Water Reactor 

(MASLWR) is a system-level test facility 

constructed by Oregon State University 

(OSU) to examine thermal hydraulic 

phenomena of importance to integral type 

reactors. MASLWR is the predecessor to the 

current NuScale integral reactor with 1:3 

length-scale, 1:254 volume scale, and a 1:1 

time scale and is operated under full pressure 

and temperature conditions of the NuScale 

design. Currently, NuScale is working with 

OSU to update this integral test facility to the 

latest NuScale module design in preparation 

for design certification testing. Through 

participation in the IAEA International 

Collaborative Standard Problem (ICSP), a 

TRACE model for MASLWR will be 

developed. Predictions of this model will be 

compared for several steady state and 

transient reactor states. Test results can also 

be used to validate new analysis models or 

components for the TRACE code. The 

research effort aims towards the future 

support of licensing reviews of the NuScale 

reactor. 

Experimental Studies of Thermal-hydraulic 

Phenomena

Thermal-hydraulic phenomena involved in 

normal and accident conditions for LWRs are 

complex, and often involve the difficult-tomodel flow of two-phase mixtures (steam and 

water). Predictions from computer codes of 

such phenomena need extensive 

experimental validation, and there are many 

effects, such as those involving 

multidimensional flows in complex 

geometries, where large-scale experiments 

are the primary means of confirming the 

validity of these predictions. In view of this, 

NRC-RES has maintained two complex 

experimental facilities: 

• PUMA facility at Purdue University for 

BWR-related problems 

• RBHT facility at Penn State 

University for PWR emergency core 

cooling problems 

The Purdue University Multidimensional 

Integral Test Assembly (PUMA) is a medium 

size reduced height scaled facility and has 

been used in the past to perform integral 

LOCA tests of interest for the ESBWR design. 

Tests are being conducted at the PUMA 

facility to obtain experimental data on the void 

fraction distribution and fluid dynamics of a 

BWR suppression pool during the blowdown 

period. The results of these tests will be used 

to support the resolution of Generic Safety 

Issue 193, “BWR ECCS Suction Concerns.” 

In addition, an exploratory research program 

to develop so-called “closure relationships” for 
68

the evolution of interfacial area in two-phase 

flows is being undertaken at Purdue 

University. It is expected that when the data 

encompass the range of flow regimes 

expected in two-phase flows, then a model of 

interfacial area evolution will be incorporated 

into the TRACE code, potentially improving its 

accuracy and reliability. Results from this 

program have been slow in coming and the 

strategy for utilizing them in TRACE still 

remains to be elucidated. 

The Rode Bundle Heat transfer (RBHT) 

facility at Penn State University was 

developed to address issues related to 

emergency core cooling, including the 

development of a better understanding of 

reflood and rewetting in realistic, bundled 

geometries. Currently, this facility is being 

used to conduct oscillating reflood tests to 

determine the effect of the inlet flow rate, 

magnitude, and frequency on peak clad 

temperature (PCT). There is also an option to 

perform steam cooling with droplet injection 

tests to determine the effect of dry spacer grid 

on droplet size and distribution. 

In parallel, the NRC is collaborating with 

international groups in undertaking 

experiments in facilities abroad as noted 

below: 

OECD/NEA ROSA-2: NRC is participating in 

the OECD/NEA ROSA-2 Project to utilize the 

Large Scale Test Facility (LSTF) of ROSA 

(Rig-of-Safety Assessment) Program of JAEA 

(Japan Atomic Energy Agency) for studying 

the integral response of the core and steam 

generator. The full-height ROSA/LSTF 

integral test facility, with 1/48 volumetric 

scaling, is designed to investigate thermal 

hydraulic phenomena of interest to PWRs. 

The ROSA-2 Project will provide both integral 

and separate-effects thermal-hydraulic data 

on intermediate break LOCA and on the 

recovery from Steam Generator Tube 

Rupture (SGTR) events. These data will be 

used to assess predictive capability of 

thermal-hydraulic analysis codes including 

TRACE. 

OECD/NEA-PKL-2: The OECD/NEA-PKL2 

Program is designed to address thermalhydraulic safety issues for current PWR and 

new PWR designs.. The PKL facility is a fullheight, 1:145 power and volume scaled 

replica of a 4-loop, 1300 MW PWR. The 

experiments will focus on steam generator 

heat transfer under shutdown conditions (e.g., 

loss of RHR during mid-loop operations), fast 

cooldown transients (such as main steam line 

breaks), accident situations for new PWR 

designs (e.g., EPR Emergency Operating 

Procedures), and boron precipitation. The 

data obtained from these experiments will be 

used for TRACE assessment, to complement 

design certification reviews, where possible, 

and in the case of boron precipitation, to 

validate licensee post-LOCA long-term 

cooling strategies. 

ISP-50, ATLAS 50% DVI Line Break: ATLAS 

(Advanced Thermal-Hydraulic Test Loop for 

Accident Simulation) is a 1/2-height and 

length, 1/288-volume, and a full pressure 

integral test facility in Korea. The ATLAS 

facility will be used to simulate a direct vessel 

injection (DVI) line break, which is important 

in several new PWR designs such as APWR 

and AP1000. Participants in the International 

Standard Problem 50 (ISP-50) will involve 

“blind” and open calculations for the test using 

various safety analysis codes including 

ATHLET, CATHARE, MARS, 

RELAP5/MOD3, as well as TRACE. 

Considering the very limited integral test data 

on DVI line breaks, this will provide the 

participants with the opportunity to 

understand the relevant thermal-hydraulic 

behavior and to assess existing safety 

analysis codes against the data. The effort 

will be coordinated by the Korea Atomic 

Energy Research Institute (KAERI). 

 

Development of Multidimensional CFD 

Capabilities

The NRC currently has only a modest effort in 

the area of CFD and this is limited to using 

commercial CFD codes from ANSYS Inc. 

(FLUENT) and CD-adapco (STAR-CCM+). 
69

Assessment and Recommendations 

The staff is to be commended for the 

progress that has been made in developing 

and moving forward with incorporation of 

TRACE into the regulatory process. Much 

work remains to be done to enable its reliable 

use for the analysis of the new LWR designs, 

an urgent matter which should be conducted 

with high priority. The priorities for further 

development of TRACE require careful 

evaluation. 

The international collaborative efforts are also 

to be commended, as they take advantage of 

facilities that are of a scale and capability that 

do not currently exist in the U.S. 

Furthermore, they draw on the expertise of 

international partners, who have continued to 

maintain a very high level of capability in the 

thermal-hydraulics field. However, 

complementary development of national 

facilities to address safety related thermalhydraulics issues should be seriously 

considered. Such facilities would enable the 

retention of U. S. expertise and provide 

capability to conduct experiments for 

supporting confirmatory thermal-hydraulic 

analyses of new reactor designs. 

NRC currently has only a modest effort in the 

area of CFD, and it is limited to some use of 

commercial CFD codes. Although 

commercial CFD codes are used in the 

process industry for qualitative indications of 

phenomena, they are validated to a much 

less rigorous standard than codes for nuclear 

use, and the source codes are not available. 

They appear to include a number of ad hoc

fixes to improve stability and robustness 

which may affect their predictive capability for 

situations that cannot be studied 

experimentally. 

It is inevitable that the licensees will 

increasingly capitalize on the extraordinary 

advances in computing power and 

computational science to resolve problems 

which the current generations of thermalhydraulic codes such as TRACE are unable 

to do. A recent LWR Sustainability Research 

and Development (R&D) Program, prepared 

by DOE in close collaboration with industry 

R&D programs, identified examples of 

advanced features anticipated in the nextgeneration system analysis code, including: 

• CFD-based Coarse-Grain Modeling for 

 turbulent and two-phase flow; 

• fully implicit, nonlinear (tightly) coupled 

 multi-physics, using Jacobian-Free 

 Newton-Krylov methods; 

• high order, accurate, computationally 

 effective and robust numerical solvers; 

• parallel, high-performance computing; 

• optimized operation for multi-processors 

 (100-1000 CPU clusters); 

• adaptive model refinement; and 

• built-in sensitivity analysis and uncertainty 

 quantification machinery, to provide 

 guidance for model refinement and further 

 research (“a quantitative PIRT process”) 

LWR Sustainability R&D Program considers 

features such as a next-generation system 

analysis code as essential components to 

improve understanding and utilization of 

safety margins. 

NRC thermal-hydraulic research has to 

position the agency to address the research 

coming from the LWR Sustainability Program. 

Several possibilities for developing such 

capabilities should be evaluated taking 

budgetary constraints into account. First 

amongst these would be participation, and 

cost-sharing, in programs with international 

partners to develop next-generation 

multidimensional CFD simulation tools aiming 

at a high level of transparency, verification, 

and validation. Second, perhaps in 

conjunction with the first, consideration should 

be given to the formation of an NRC-US 

university consortium to develop such 
70

capability capitalizing on the high level of CFD 

expertise that now exists in several 

universities. Third, and perhaps again in 

conjunction of some subset of the other 

options, the possibility of building on one of 

the excellent open-source CFD platforms 

should be seriously considered. 
Thermal-Hydraulics 

Research

• Development of 

computational 

capability with 

STAR-CD CFD code

Independent Evaluation of Licensees’ Submittals in the 

area of Thermal-Hydraulics 

(A major element of many licensing actions) 

TRACE Computer Code 

Development and 

Application 

Experimental Studies of 

Thermal-hydraulic 

Phenomena 

Development of 

Multidimensional CFD 

Capabilities 

• Development of 

TRACE ABWR 

plant models 

• Development of an 

uncertainty 

methodology for 

TRACE

Figure 10. Current NRC Research Activities in Thermal-Hydraulics Research 

• Assessment of 

TRACE

• Participation in the 

international PSBT 

benchmark 

• T-H calculations for 

EPUs and new 

reactors

• RBHT test 

program at Penn 

State

• T/H Experiments 

at PUMA facility 

• Participation in 

OECD/NEA 

PKL-2 Project 

• Participation in 

OECD/ROSA 

Program

• Participation in 

ISP-50: ATLAS 

50% DVI Line 

break 

71
72
73

18. REFERENCES 

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Program,” Advisory Committee on 

Reactor Safeguards (ACRS), NUREG1635, Vol. 7, May 2006. 

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Fiscal Year 2009,” Prepared for U.S. 

Department of Energy (DOE), Office of 

Nuclear Energy, September 2008. 

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Materials Degradation Assessment,” 

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2006, February 2007. 

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Part 50, Section 46, Acceptance Criteria 

for Emergency Core Cooling Systems for 

Light-Water Nuclear Power Reactors. 

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“Severe Accident Risks: An Assessment 

for Five U.S. Nuclear Power Plants,” 

NUREG-1150, December 1990. 

7. Public Law 109-58, "Energy Policy Act of 

2005," enacted August 8, 2005. 

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Part 50, Section 65, Requirements for 

Monitoring the Effectiveness of 

Maintenance at Nuclear Power Plants.

9. Code of Federal Regulations, Title 10, 

Part 54, Requirements for Renewal of 

Operating Licenses for Nuclear Power 

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10. U.S. Nuclear Regulatory Commission, 

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Report,” NUREG-1801, Rev. 1, 

September 2005. 

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V. Bonaca, Chairman, ACRS, To R. W. 

Borchardt, Executive Director For 

Operations, NRC, Subject: Draft Final 

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Toughness Requirements for Protection 

against Pressurized Thermal Shock 

Events.” 

13. U.S. Nuclear Regulatory Commission, 

“Verification and Validation of Selected 

Fire Models for Nuclear Power Plant 

Applications,” NUREG-1824, Vols. 1-7, 

EPRI TR1011999, May 2007. 

14. U.S. Nuclear Regulatory Commission, “A 

Phenomena Identification and Ranking 

Table (PIRT) Exercise for Nuclear Power 

Plant Fire Modeling Applications,” 

NUREG/CR-6978, November 2008. 

15. U.S. Nuclear Regulatory Commission, 

“Fire PRA Methodology for Nuclear Power 

Facilities,” NUREG/CR-6850, Vols. 1-2, 

EPRI TR 1011989, September 2005. 

16. U.S. Nuclear Regulatory Commission, 

“Cable Response to Live Fire 

(CAROLFIRE),” NUREG/CR-6931, Vols. 

1-3, April 2008. 
74

17. U.S. Nuclear Regulatory Commission, 

“A Short History of Fire Safety Research 

Sponsored by the U.S. Nuclear 

Regulatory Commission, 1975-2008,” 

NUREG/BR-0364, June 2009. 

18. Code of Federal Regulations, Title 10, 

Part 26, Fitness for Duty Programs.

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“Human Factors Considerations with 

Respect to Emerging Technology in 

Nuclear Power Plants,” NUREG/CR-6947, 

October 2008. 

20. U.S. Nuclear Regulatory Commission, 

“Effect of LWR Coolant Environments on 

the Fatigue Life of Reactor Materials,” 

NUREG/CR-6909, February 2007. 

21. U.S. Nuclear Regulatory Commission, 

“Guidance on the Treatment of 

Uncertainties Associated with PRAs in 

Risk-Informed Decision Making,” NUREG1855, Draft Report for Comment, 

November 2007. 

22. U.S. Nuclear Regulatory Commission, 

“Estimating Loss-of-Coolant Accident 

(LOCA) Frequencies Through the 

Elicitation Process,” NUREG-1829, April 

2008. 

23. U.S. Nuclear Regulatory Commission, 

“Recommendations for Probabilistic 

Seismic Hazard Analysis: Guidance on 

Uncertainty and use of Experts,” 

NUREG/CR-6372, Volumes 1 and 2, April 

1997. 

24. U.S. Nuclear Regulatory Commission. 

"Residual radioactive contamination from 

decommissioning," NUREG/CR-5512, 

Vol. 1, October 1992. 

25. U.S. Nuclear Regulatory Commission. 

“Residual radioactive contamination from 

decommissioning: User's Manual DandD 

Version 2.1,” NUREG/CR-5512, Vol. 2, 

April 2001. 

26. U.S. Nuclear Regulatory Commission. 

“Residual radioactive contamination from 

decommissioning: parameter analysis,” 

NUREG/CR-5512, vol.3 October 1999. 

27. Code of Federal Regulations, Title 10, 

Part 20, Standards for Protection against 

Radiation. 

28. U.S. Nuclear Regulatory Commission, 

Regulatory Guide 1.183, “Alternative 

Radiological Source Terms for Evaluating 

Design Basis Accidents at Nuclear Power 

Reactors,” July 2000. 

29. U.S. Nuclear Regulatory Commission, 

Regulatory Guide 1.174,” An Approach for 

Using Probabilistic Risk Assessment in 

Risk-Informed Decisions on Plant-Specific 

Changes to the Licensing Basis,” 

Revision 1, November 2002. 

30. Generic Safety Issue 193, “BWR ECCS 

Suction Concerns.” 

31. Letter Dated September 24, 2008, From 

William J. Shack, Chairman, ACRS, To 

Dale E. Klein, Chairman, NRC, Subject:

Development of The TRACE ThermalHydraulic System Analysis Code