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

Volume 11
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Review and Evaluation of 

the Nuclear Regulatory 

Commission Safety 

Research Program

A Report to the 

U.S. Nuclear Regulatory 

Commission

Manuscript Completed: June 2014

Date Published: December 2014

Advisory Committee on Reactor Safeguards

NUREG-1635

Volume 11

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.

v

TABLE OF CONTENTS

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

FIGURES.................................................................................................................................. vii

TABLES.................................................................................................................................... vii

ABBREVIATIONS......................................................................................................................xi

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

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

3. FUKUSHIMA FORENSICS: UNDERSTANDING SEVERE ACCIDENT 

PROGRESSION............................................................................................................11

4. ADVANCED REACTOR DESIGNS ...............................................................................15

5. DIGITAL INSTRUMENTATION AND CONTROL SYSTEMS.........................................19

6. FIRE SAFETY ...............................................................................................................23

7. REACTOR FUEL...........................................................................................................31

8. HUMAN FACTORS AND HUMAN RELIABILITY...........................................................37

9. MATERIALS AND METALLURGY.................................................................................41

10. NEUTRONICS AND CRITICALITY SAFETY.................................................................51

11. OPERATIONAL EXPERIENCE .....................................................................................55

12. PROBABILISTIC RISK ASSESSMENT.........................................................................59

13. RADIATION PROTECTION...........................................................................................67

14. NUCLEAR MATERIALS AND WASTE..........................................................................73

15. SEISMIC AND STRUCTURAL ENGINEERING.............................................................75

16. SEVERE ACCIDENTS AND SOURCE TERM...............................................................79

17. THERMAL HYDRAULICS .............................................................................................85

.

18. REFERENCES..............................................................................................................91

vii

FIGURES

1. Current NRC Research Activities in Fire Safety.......................................................................29

2. Current NRC Research Activities in Reactor Fuel ...................................................................35

3. Current NRC Research Activities in Human Factors and Human Reliability ............................40

4. Current NRC Research Activities in Materials and Metallurgy .................................................49

5. Current NRC Research Activities in Operational Experience...................................................58

6. Current NRC Research Activities in PRA ................................................................................65

7. Current NRC Research Activities in Radiation Protection........................................................71

8. Current NRC Research Activities Related to Severe Accident and Source Term ....................83

9. Current NRC Research Activities in Thermal-Hydraulics Research.........................................90

TABLES

1. International Collaborative Severe Accident Experimental Research Programs with NRC 

Participation ............................................................................................................................84

ix

ABBREVIATIONS

ABWR advanced boiling-water reactor

ACRS Advisory Committee on Reactor Safeguards

ADAMS Agencywide Documents Access & Management System

AECL Atomic Energy of Canada Limited

AEOD Office for Analysis and Evaluation of Operational Data

AIChE American Institute of Chemical Engineers

ALARA as low as is 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 Program

ASTM American Society for Testing and Materials

ATWS anticipated transient without scram

BIP Behavior of Iodine Project

BWR boiling-water reactor

CAROLFIRE cable response to live fire

CCF common-cause failure

CFD computational fluid dynamics

CFR Code of Federal Regulations

CHRISTI-FIRE cable heat release, ignition, and spread in tray installations during fire

CONOPS concept of operations

CRDM control rod drive mechanism

CSARP Cooperative Severe Accident Research Program

DI&C digital instrumentation and control

DOE U.S. Department of Energy

EAC environmentally assisted cracking

EDMGs extensive damage management guidelines

EIS environmental impact statement

EMUG European MELCOR user group

EOPs emergency operating procedures

EPAct Energy Policy Act

EPIX Equipment Performance and Information Exchange System

EPMDA expanded proactive materials degradation assessment

EPR evolutionary power reactor

EPRI Electric Power Research Institute

ESBWR economic simplified boiling water reactor

ESP early site permit

FCI fuel-coolant interaction

FMEA failure modes and effects analysis

FSME Office of Federal and State Materials and Environmental Management 

Programs

GDC general design criteria

GSI generic safety issue

GUI graphical user interface
x

ABBREVIATIONS (Cont’d)

GWd/t gigawatt-days per metric ton

HAMMLAB Halden Man-Machine Laboratory

HDPE high-density polyethylene 

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

ICG-EAC International Cooperative Group on Environmentally Assisted Cracking

ICRP International Commission on Radiological Protection

ICSP international collaborative standard problem

IDEAS Integrated Decision-tree Human Event Analysis System

IEEE Institute of Electrical and Electronics Engineers

IFRAM International Forum for Aging Management

IGSCC intergranular stress-corrosion cracking

INPO Institute of Nuclear Power Operations

IPEEE individual plant examination of external events

iPWRs integral pressurized-water reactors

IRSN Institute De Radioprotection Et De Surete Nucleaire

ISG-TP International Steam Generator Tube Integrity Program

ISI inservice inspection

JAEA Japan Atomic Energy Agency

JNES Japan Nuclear Energy Safety Organization

KM knowledge management

LANL Los Alamos National Laboratory

LER licensee event report

LERF large early release frequency

LPSD low power and shutdown

LRGDs license renewal guidance documents

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

MCAP MELCOR Cooperative Assessment Program

MCCI molten core concrete interaction

MCNP Monte Carlo N-Particle Transport Code System

MDEP Multinational Design Evaluation Program

MeV million electron volts 

MORs monthly operating reports

MOX mixed oxide
xi

ABBREVIATIONS (Cont’d)

MSPI Mitigating System Performance Index

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

NMSS Office of Nuclear Material Safety and Safeguards

NPPs nuclear power plants

NRC U.S. Nuclear Regulatory Commission

NRO Office of New Reactors

NRR Office of Nuclear Reactor Regulation

NSIR Office of Nuclear Security and Incident Response

NTTF Near-Term Task Force

NUPEC Nuclear Power Engineering Corporation

OECD Organization for Economic Cooperation and Development

ORNL Oak Ridge National Laboratory

PARCS Purdue Advanced Reactor Core Simulator

PCI pellet cladding interaction

PCMI pellet cladding mechanical interaction

PCT peak clad temperature

PIRT phenomena identification and ranking table

PNNL Pacific Northwest National Laboratory

PRA probabilistic risk assessment

PSHA probabilistic seismic hazard analysis

PSI Paul Scherrer Institute

PSU Pennsylvania State University

PUMA Purdue University Multidimensional Integral Test Assembly 

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

RPV reactor pressure vessel

RTF radioiodine test facility

SAMGs severe accident management guidelines

SCALE standardized computer analysis for licensing evaluation

SERENA steam explosion resolution for nuclear applications

SDP significance determination process

SGTR steam generator tube rupture

SMR small modular reactor

SNAP Symbolic Nuclear Analysis Package
xii

ABBREVIATIONS (Cont’d)

SNL Sandia National Laboratories

SOARCA State-of-the-Art Reactor Consequence Analyses

SPAR standardized plant analysis risk

SRM staff requirements memorandum

SSCs structures, systems, and components

SSE safe-shutdown earthquake

SSHAC Senior Seismic Hazard Analysis Committee

SSWICS small-scale water ingression and crust strength 

TRACE TRAC-RELAP advanced computational engine

U.S. United States

USGS United States Geological Survey

V&V verification & validation

VHTR very-high-temperature reactor

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 

U.S. 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, 

near-term issues to support regulatory 

activities and initiatives in the Offices of 

Nuclear Reactor Regulation (NRR), New 

Reactors (NRO), Nuclear Material Safety 

and Safeguards (NMSS), Nuclear Security 

and Incident Response (NSIR), and Federal 

and State Materials and Environmental 

Management Programs (FSME). 

RES has succeeded over the last few years 

in its effort to tie research activities it 

undertakes to near-term issues confronted 

by the NRC line organizations (NRO, NRR, 

NMSS, NSIR, and FSME). 

For the purposes of this report, the ongoing 

research has been examined in terms of the 

following technical disciplines:

• advanced reactor designs

• digital instrumentation and control 

systems

• fire safety 

• reactor fuel 

• human reliability and human factors 

• 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 

The interdisciplinary research effort of the 

State-of-the-Art Reactor Consequence 

Analyses (SOARCA) project and uncertainty 

quantification is not addressed in this report. 

Throughout the project, the ACRS provided 

reports on the technical approach of this 

activity.

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.
2

In its review of the NRC Safety Research 

Program, the ACRS has focused on the 

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 risk informed and 

“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 accident at Fukushima Dai-ichi, in 

Japan, has increased interest in accident 

initiators, accident management, and 

consequences for accidents that are within 

the design basis as well as those beyond 

the design basis. Research to focus on the 

forensics of the Fukushima accident is 

discussed in Chapter 3. Such efforts offer a 

unique opportunity to better understand 

boiling-water reactor (BWR) severe accident 

progression, and thereby to develop better 

measures for mitigating beyond

design-basis events. The NRC line 

organizations will benefit from research 

initiatives that increase the technical 

understanding of severe accidents.
3

2. GENERAL OBSERVATIONS AND RECOMMENDATIONS

Highlights of the major elements of the 

ongoing research dealing with the safety of 

nuclear power plants are presented here. 

Recommendations concerning these 

activities are also discussed.

General Observations

The U.S. Nuclear Regulatory Commission 

(NRC) has succeeded over the last few 

years in its effort to tie research activities it 

undertakes to near-term issues being 

confronted by its line organizations (Office 

of New Reactors (NRO), Office of Nuclear 

Reactor Regulation (NRR), Office of 

Nuclear Material Safety and Safeguards 

(NMSS), Office of Nuclear Security and 

Incident Response (NSIR), and Office of 

Federal and State Materials and 

Environmental Management Programs 

(FSME). Over 75 percent of research 

activities support specific needs of these 

offices. The Commission directs about 10

percent of the Office of Nuclear Regulatory 

Research (RES) activities through 

agency-mandated programs (e.g., Accident 

Sequence Precursor (ASP) program) and 

the Commission tasking memoranda. A 

small portion of the research budget 

focuses on long-term research subjects 

expected to be critical in 5 to 10 years.

The strategy for the identification of 

research subjects, through “User Need” 

documents, together with the associated 

process for the prioritization of research 

needs, has worked reasonably well. 

Research activities are yielding useful 

products to the line organizations in a timely 

manner. However, in some cases research 

focused on line organization needs may be 

terminated prematurely, precluding 

appropriate and needed understanding that 

would be of use for future regulatory issues. 

Major research activities often include 

collaborations with other Federal agencies, 

industry, universities, and international 

partners. Such collaborations can provide 

timely and thoughtful peer input while giving 

the agency the ability to leverage its 

expertise and resources on key topics of 

common interest (e.g., Fukushima-related 

topics). In addition, such collaborations 

provide an opportunity to help train new 

NRC staff as they participate in these 

multi-party research efforts. 

The Advisory Committee on Reactor 

Safeguards (ACRS) continues to encourage 

such active collaborations to effectively 

share knowledge and experience that 

contribute to intermediate- and long-term 

research objectives. These collaborative 

programs are often the first to suffer in times 

of declining resources. Collaborative 

research should remain an integral part of 

agency initiatives.

Future Research Focus Areas

In its 2012 report to the Commission on the 

NRC Safety Research Program 

(NUREG-1635, Vol. 10), the ACRS noted 

that the Japan Near-Term Task Force 

(NTTF) report provides detailed near-term 

and long-term recommendations for 

ensuring nuclear reactor safety based on 

initial lessons learned from the Fukushima 

event. Many of these recommendations 

require research to support their effective 

implementation. Furthermore, in the coming 

years, as data from the Fukushima recovery 

effort are collected, new insights could point 

to the need for additional work. This is 

discussed in more detail in Chapter 3.

In its review and evaluation of the NRC 

Safety Research Program, the ACRS has 

identified several topical areas for research 

that will support NRC line organizations in 

implementing various lessons-learned from 

Fukushima to the U.S. nuclear reactor fleet.

These include:
4

• protection from external hazards

• protection from severe accidents

• emergency response and severe 

accident management capabilities

• accident tolerant instrumentation

• improved understanding of severe 

accident phenomena

The ACRS continues to recommend that the 

Commission develop an integrated plan for 

providing the necessary technical basis for 

implementation of the lessons learned from 

events that occurred at the Fukushima 

Dai-ichi nuclear plant site. 

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 Reactor Designs

The U.S. Department of Energy (DOE) has 

decreased funding in advanced reactors 

that are not based on light-water reactor 

technologies. Therefore, NRC research is 

properly being refocused on generic issues 

for advanced reactors.

Safety research efforts on small modular 

reactor (SMR) designs are in their early 

stages. The NRC has identified important 

issues for licensing of proposed designs 

(e.g., multi-module operation and 

performance of unique safety systems). 

These efforts provide significant input to the 

development of safety research necessary 

for future reactor licensing activities. Any 

future research efforts are being 

appropriately tied to the expected design 

certification issues identified for the mPower 

and NuScale concepts.

Digital Instrumentation and Control Systems

Digital systems safety assurance continues 

to be a challenge for the NRC. The NRC 

and the Electric Power Research Institute 

(EPRI) have independently evaluated and 

identified the limitations of failure (fault) 

modes and effects analysis (FMEA) for 

regulatory assurance of complex logic in 

digital systems – particularly in the presence 

of interactions and feedback paths. Both 

organizations have been researching a 

broader range of hazard analysis (HA) 

techniques for evaluating safety assurance.

The ACRS encourages close collaboration 

on digital system research under the 

memorandum of understanding (MOU) on 

Cooperative Nuclear Safety Research 

between the two organizations.

Quantifying the reliability of digital systems 

in probabilistic risk assessments (PRAs) is a 

major challenge. The ACRS has noted the 

progress that has been achieved in analysis 

of digital system failure modes. The use of 

these results, however, is not coordinated 

with digital system PRA research.

The ACRS continues to see a lack of 

integration of control of access, safety, and 

cybersecurity in the design stage and 

licensing as an impediment to ensuring 

secure digital instrumentation and control 

(DI&C) safety systems.

Fire Safety

The NRC’s fire safety research program 

continues to make progress developing an 

understanding of fire events and an 

assessment of risk from fire damage in 

nuclear power plants. Structured 

collaboration with industry provides 

cost-effective 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 

user-identified needs. This process has 
5

been responsive to immediate and 

near-term technical issues. It should remain 

an important part of integrated planning. 

Where identification of need is clear and 

beneficial, research priorities and programs 

should more actively anticipate additional 

emerging applications and potential 

intermediate- to long-term requirements that 

are not fully dictated by current user needs.

Guidance is currently being developed for 

the assessment of risk during low-power 

and shutdown operating modes. Continued 

research is needed to evaluate issues such 

as fire initiation frequencies, human-caused 

fires, effectiveness of detection and 

suppression capabilities, and propagation of 

heat and smoke through compromised fire 

barriers that are uniquely associated with 

personnel activities and system 

configurations during plant shutdown.

Research projects should address early 

detection of incipient fire, the effects from 

fire damage and heat on fiber optic cables, 

the effects of heat on digital equipment, and 

the effects of smoke damage to digital 

signal processing and computation 

modules.

Reactor Fuel

Current research addressing fuel 

fragmentation, axial relocation, and 

dispersal is very well structured, has made 

excellent progress, and is well on its way to 

generating experimental and analytical 

results to support regulatory decisions. 

The FRAPCON and FRAPTRAN fuel 

performance codes meet most near-term 

agency needs. There is ample capability to 

assess risks of fuel failure due to fuel center 

melting, excessive mechanical strain, and 

departure from nucleate boiling events. In 

the future, light-water reactor (LWR)

operations may include more load following 

that may increase the likelihood of pellet 

clad interaction. The NRC is developing 

capability to quantitatively assess pellet 

cladding interaction (PCI) fuel failures 

during anticipated operational occurrences 

(AOOs) for current or future fuel designs. 

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 burnup beyond the current 

regulatory limit of 62 gigawatt-days per 

metric ton (GWd/t) and the amount of 

experimental data provided to support these 

proposed changes to regulatory limits. 

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.

Human Factors and Human Reliability

The human factors (HF) and human 

reliability analysis (HRA) research program 

has evolved into a carefully coordinated 

series of projects that are extending 

knowledge and providing improved methods 

to be used in regulatory activities. 

The Integrated Decision-tree Human Event 

Analysis System (IDHEAS) HRA 

methodology program has integrated 

experience from the International and U.S. 

HRA Empirical Studies to identify the 

strengths and weaknesses of existing HRA 

methods and develop a hybrid approach. 

The methodology needs to be tested on a 

large-scale application to ensure that it is 

practical. The guidance must be improved, 

especially to strengthen the qualitative 

analysis.

The ACRS has identified three areas of HF

and HRA research that could benefit from 

increased emphasis in the future: (1) the 

work on medical applications should receive 

more emphasis, (2) the development of 

HRA and HF approaches to support NSIR 

applications needs to be explored, and 

(3) the work on the effects of degraded I&C 
6

systems on human performance should be 

expanded. 

Materials and Metallurgy

The ACRS supported the initial vision of the 

proactive materials degradation assessment 

(PMDA) and the later expansion of its 

original scope to include concrete structures 

and electrical cable insulation. The 

expanded materials degradation 

assessment (EMDA) program, under the 

joint sponsorship of the NRC and DOE, has 

identified key areas of concern for 

subsequent license renewal operation (time 

frames of 60 to 80 years). The current 

research programs are adequate for all 

areas of concern except one. Recent 

research indicates that the exposure of 

stainless steels and nickel-base alloys to 

high-temperature water in the absence of 

radiation damage can result in a significant 

decrease in tearing resistance and fracture 

toughness. If confirmed, this effect is a 

potential safety issue and NRC research on 

this topic should be pursued.

The NRC research program is conducting 

effective work in: the xLPR (extremely low 

probability of rupture) program; the 

improvement of nondestructive evaluation 

(NDE); and the confirmation of industry 

steam generator inspection and integrity 

models.

The ACRS continues to support the 

agency’s active participation in international 

efforts relating to materials degradation, 

such as the International Cooperative Group 

on Environmentally Assisted Cracking 

(ICG-EAC) and the recent cooperative 

agreement between the NRC and the 

French Institute De Radioprotection et de 

Surete Nucleaire (IRSN ) to develop 

techniques for inspection of coarse-grained 

materials in dissimilar metal welds. 

Neutronics and Criticality Safety

The current NRC research programs on 

neutronics and criticality safety are 

appropriately prioritized and are well

supported by user needs.

Expanded application capability, improved 

accuracy, and enhanced computational 

efficiency demands for neutronics and 

criticality analyses should be achieved as a 

result of both near term and long term 

research programs. Licensees of operating 

reactors continue to optimize core designs 

using improved analytical techniques, some 

of which require full core, coupled 

approaches to safety analyses. In addition,

similar capabilities must be applied to 

assure appropriate safety analyses for 

integral small modular and large advanced 

reactors.

Operational Experience

The operational experience research 

programs continue to provide data and tools 

necessary for regulatory decisionmaking.

The primary sources of operating 

experience data are the Institute of Nuclear 

Power Operations (INPO) Equipment 

Performance and Information Exchange 

System (EPIX), Licensee Event Reports

(LERs), and inspection reports. These 

sources provide information about 

operational failures. When these data are 

used to support risk analysis, they must be 

augmented with estimates of the successes 

(i.e., the number of equipment demands or 

operating times). The successes are highly 

plant-specific, involving operating practices 

and details of surveillance and testing 

procedures.

Many plant-specific PRAs have documented 

thorough compilation of success data. 

Lacking accurate plant-specific success 

data, it must be recognized that there 

is substantially greater uncertainty in the 

count of successes compared to the count 

of failures. Likewise, any generic 

industry-wide failure rate data should 

include substantial uncertainty caused by

plant-to-plant variability in both the number 

of failures and successes.
7

Some years ago, the NRC’s former Office 

for Analysis and Evaluation of Operational 

Data (AEOD) published a number of reports 

that investigated this issue and provided 

industry-wide failure rate estimates for a 

variety of equipment. RES should examine 

the impact of the uncertainty in success 

data in their applications and analyses. 

Expanding the previous work of AEOD with 

research to fully incorporate these 

uncertainties may be appropriate. 

Probabilistic Risk Assessment

A substantial fraction of probabilistic risk 

assessment (PRA) research resources is 

allocated to immediate user needs and 

short-term requirements to support existing 

PRA model infrastructure for the Reactor 

Oversight Process. The ACRS continues to 

encourage increased sharing of the 

Standardized Plant Analysis Risk (SPAR)

model maintenance and support activities 

among the regional offices, headquarters 

staff, and contractors, allowing RES to focus 

more effectively on advancement of 

state-of-the-art risk assessment methods 

and practices.

The ACRS continues its strong support for 

the Level 3 PRA project. It will improve the 

understanding of integrated site-level risk 

and provide a context to define needs for 

additional intermediate- and long-term 

research. The ACRS recommends that 

RES should use evolving knowledge and 

insights from the Level 3 PRA project to 

risk-inform initiatives for specific research 

that may be proposed as a consequence of 

NRC and industry responses to the 

Fukushima Dai-ichi accidents.

Proposed new small modular reactor 

designs introduce regulatory challenges that 

must be addressed to understand and 

manage the risk from potential single-unit, 

multiple-unit, and site-level accidents. The 

ACRS recommends that RES should 

pursue a pilot study to apply the 

risk-informed licensing framework proposed 

in NUREG-1860, “Feasibility Study for a 

Risk-Informed and Performance-Based 

Regulatory Structure for Future Plant 

Licensing, Volumes 1 and 2,” for a specific 

small modular reactor design. The study 

should address design, operations, and 

site-level issues that affect accident 

progression, mitigation options, emergency 

planning, and offsite consequences.

Substantial efforts have been made to 

improve security against potential physical 

and cyber attacks on our operating nuclear 

power plants. New plants will further 

integrate security protections from the initial 

stages of their designs. Despite these 

efforts, concerns remain that the applied 

security controls may not be allocated 

optimally to cope with the full spectrum of 

potential threats. The ACRS recommends 

that RES should initiate a research project

and pilot applications to examine the use of 

quantitative risk assessment methods to 

inform security programs and practices.

During 2013, decisions were made to 

permanently shut down four US reactors. 

Others may follow over the next few years. 

These actions raise questions about 

whether the NRC has fully addressed the 

risks from decommissioning activities, 

including disposal of used nuclear fuel and 

high-level wastes. The ACRS recommends 

that RES should initiate a project to 

examine how quantitative risk information 

can be used to inform regulatory decisions 

in these interrelated disciplines. 

RES has completed an important update to 

NUREG-1855, “Guidance on the Treatment 

of Uncertainties Associated with PRAs in 

Risk-Informed Decision Making,” which 

contains practical guidance on methods to 

systematically identify, document, and 

quantify sources of uncertainty. However, a 

disciplined assessment of uncertainty is not 

yet applied consistently throughout the 

agency. The ACRS recommends that RES 

should initiate efforts to ensure that an 

appropriate characterization of uncertainty 

is performed in all agency analyses.
8

The ACRS encourages active collaboration 

among RES, other Government agencies, 

industry research organizations such as 

EPRI, and universities to effectively share 

knowledge and experience that contribute to 

intermediate- and long-term research 

objectives. These collaborative programs 

are often the first to suffer in times of 

declining resources. Collaborative research 

should remain an integral part of RES 

initiatives.

Radiation Protection

The staff has developed an appropriate and 

robust research program in the area of 

radiation protection. This program includes 

radiation protection of workers and the 

public and radiological assessments related 

to radiation exposure and health risks 

around NRC-licensed nuclear facilities.

The ACRS is supportive of RES’s 

participation in national and international 

efforts relating to radiation protection. 

These collaborations provide excellent 

opportunities for the NRC staff to benefit 

from the work of other organizations in the 

United States and around the world.

Nuclear Materials and Waste

Research activities in the general areas of 

nuclear materials and waste reflect the 

current hiatus in the resolution of policy 

issues. The current activities address 

issues encountered or anticipated in the 

inspection and monitoring activities.

Electrochemical separation has been 

suggested as a means to facilitate the safe 

disposal of waste. RES has a research 

program to better understand the possible 

separation processes and hazards that 

could arise in the possible use of 

electrochemical preprocessing of waste for 

disposal. This is of necessity a very 

exploratory effort. Further detailed work in 

this area should await more definitive 

proposals from potential licensees.

Seismic and Structural Engineering 

The NRC seismic and structural research 

program in support of regulatory activities is 

being conducted under a well-developed 

research plan that has been broadly 

reviewed for both technical quality and 

programmatic elements. The program 

funds state-of the-art work via contracts to 

several renowned organizations in the field, 

including national laboratories and 

universities, as well as international 

cooperative programs and collaborative 

research with other governmental agencies 

such as DOE. This program will adequately 

support the NRC staff’s capabilities to 

evaluate potential seismic and flooding risks 

to U.S. nuclear plants.

Severe Accidents and Source Term

The NRC makes use of its severe accident 

expertise and analysis capabilities to 

support regulatory decisions for operating 

nuclear power plants and for certifying new 

and advanced reactor designs. Severe 

accident analysis tools also help in the 

transition to a more risk-informed regulatory 

framework. The agency’s long-term severe 

accident and containment response 

evaluation development plan focuses on 

two areas: (1) maintenance and 

development of its severe accident 

computer codes and (2) continued 

collaboration in international experimental 

research programs.

The ACRS agrees with the approach that 

the NRC staff has developed to support 

regulatory decisions for severe accidents 

via computer code development validated 

by experimental data analysis. This 

approach has successfully allowed the NRC 

to maintain and update its modeling 

capabilities for severe accident analyses.

Ongoing Fukushima assessments may 

identify some deficiencies in BWR-specific 

and ex-vessel modeling capabilities. The 

NRC should participate in efforts to obtain 

data on accident progression in the 
9

damaged Fukushima reactors. These data 

can be used to refine severe accident 

models for boiling-water reactors.

Thermal Hydraulics

Excellent progress has been made in 

developing and incorporating the NRC’s 

systems thermal-hydraulics code, TRACE, 

into the regulatory process. Further 

development should focus on 

implementation of the four-field 

thermal-hydraulics model, as recommended 

by the TRACE peer-review group. The 

research program on interfacial area 

transport should be phased out as the 

results are of limited value for TRACE 

applications.

The ACRS is supportive of the agency’s 

active participation in international 

collaborative efforts, as they take advantage

of facilities that are of a scale and capability 

that do not currently exist in the United 

States. They also draw on the expertise of 

international partners, who have continued 

to maintain a high level of capability in 

thermal-hydraulics. Complementary 

development of unique U.S. facilities to 

support confirmatory accident analyses for 

new fuel and reactor designs should be 

seriously considered. Integral tests would 

enhance confidence in regulatory analysis 

of new reactor designs. Such integral tests 

cannot be carried out in any existing 

facilities.

NRC currently has modest, but productive, 

efforts in the area of computational fluid 

dynamics (CFD) through the use of 

commercial CFD codes, which have played 

a role in improving the technical bases for 

certain licensing decisions. In this direction, 

licensees are and will continue to capitalize 

on the extraordinary advances in computing 

power and computational science to 

address many critical safety issues, e.g., 

prediction of the behavior of full-scale 

components, such as advanced 

accumulators, based on small-scale 

experiments. The agency should maintain 

independent confirmatory capabilities that 

keep pace with such developments in 

industry. The ACRS has suggested 

strategies to establish and maintain such

capability. 

11

3. FUKUSHIMA FORENSICS: UNDERSTANDING SEVERE ACCIDENT 

PROGRESSION

Eventually the damaged reactors at the 

Fukushima Dai-ichi site will be opened and 

dissected. This process presents an 

opportunity for the collection of data and 

observations concerning the progression of 

severe reactor accidents in boiling-water 

reactors (BWRs) that at least parallels the 

opportunity provided by the recovery of the 

Three Mile Island Unit II (TMI-2) for 

understanding severe accidents in 

pressurized-water reactors (PWRs). The 

understanding provided by TMI-2 forensics 

led to a wide range of changes in the 

modeling of severe accidents in PWRs. 

‘Separate-effects’ tests were conducted to 

put quantitative substance on qualitative 

insights gained from examinations of the 

fuel and structures, fission products, and 

their interactions throughout the reactor 

coolant system. Unfortunately, there have 

been a very limited number of BWR-specific

tests, and similar advances were not made 

in the modeling of BWR accidents. 

Forensic efforts at Fukushima, assisted by 

international collaborations for conducting 

additional analysis and companion 

experiments, offer significant potential for 

improved understanding of 

beyond-design-basis accident behavior, 

associated fission product source term 

release and transport, as well as the ability 

of severe accident analyses to inform onsite 

and offsite actions to mitigate the 

consequences of such events.

The progression of severe reactor accidents 

in BWRs may be quite different than 

accident progression in PWRs. Channel 

boxes and the cruciform control blades used 

in BWR fuel assemblies would certainly 

affect the early stages of cladding oxidation, 

fuel rod heatup, and core degradation.

Continued degradation and accident 

progression could be affected by natural 

convection to the massive steel structures 

making up steam separators and steam 

dryers above the reactor core. The BWR 

geometry with its internal structures 

represents a shorter natural convection 

transport path than the natural convection to 

the steam generators in PWRs. Based on 

results from a limited number of smaller 

scale prototype BWR experiments, the 

MELCOR code, with BWR specific models 

incorporated into it, predicts that BWR 

relocation is an incoherent phenomenon 

with small amounts of material draining 

through the lower core support plate into the 

lower reactor plenum. However, standard 

MELCOR code models and industry severe 

accident codes assume that liquefied BWR 

fuel and cladding accumulate within the 

core region similar to what was observed in 

the PWR accident at TMI-2. Forensic 

investigations into the reactor cores at 

Fukushima can certainly help resolve these 

uncertainties. 

Based on what is currently known about the 

events at Dai-ichi, it is expected that there 

remained sufficient water to quench the 

core material in the lower plenum below the 

core or in the pedestal region below the 

reactor pressure vessel.

In-vessel, massive steel structures making 

up over 200 control rod drives (CRD) in the 

lower plenum may also provide heat sinks 

to quench relocating core debris. These 

CRD structures and instrumentation 

penetrations may also serve as fins for heat 

removal. Although water inventory and 

steel structures in the BWR lower head may 

temporarily quench liquefied core debris, 

long-term debris stabilization requires 

continued water addition to the reactor 

pressure vessel with associated removal of 

decay heat to an ultimate heat sink. 

If cooling of core debris in the lower plenum 

of a BWR reactor vessel cannot be 

maintained by continued water injection (or 
12

if the relocated debris has been fragmented 

to sizes sufficiently small that it is not 

coolable), the debris mass will heat and 

remelt. Contact of this material with the 

vessel lower plenum may lead to local 

failures of the vessel lower head 

instrumentation tube or drain line 

penetrations with release of the core 

material. This would add to the incoherency 

of the accident progression and lead to 

different end-states than currently simulated 

in ex-vessel models in severe accident 

analysis computer codes. 

The BWR instrumentation tubes are thought 

to be more vulnerable to failure than the 

more massive and numerous control-rod 

drive penetrations. However, the drain line, 

which is located at the bottom center of the 

vessel and has no in-vessel structures, has 

the potential to be even more susceptible to 

ex-vessel tube failure. Melt expulsion from 

the reactor vessel into the BWR drywell will 

lead to ex-vessel core debris interactions, 

which will be somewhat different than 

ex-vessel interactions PWRs. The large 

reactor cavities of many PWRs will keep 

debris separated from the reactor 

containment boundary. However, the 

drywell floor of the Mark I BWR containment 

has a small sump. If the volume of debris 

expelled from the vessel exceeds the 

volume of the sump, molten debris can flow 

across the drywell floor and impinge directly

on the carbon steel containment boundary.

The presence of water on the drywell floor, 

as well as radiation and convective heat 

transfer from the debris to surrounding 

structures, could largely affect the integrity 

of drywell liner or the drywell head seals in 

the containment. 

Until now, modeling of the progression of 

severe BWR accidents has largely been 

speculation augmented with the results of a 

few small, stylized tests that did not attempt 

to represent all the features of the reactor 

core, the core support structures, the 

reactor vessel, or the containment. The 

inspection and dissection of the Fukushima 

Dai-ichi reactors, along with materials 

sampling at selected locations, will provide 

an opportunity to validate these modeling 

assumptions; and where the models are not 

valid, the data will provide the basis for 

improved modeling for accident progression 

and fission product transport. 

Post-accident forensic evaluations from the 

damaged Fukushima Dai-ichi plants have 

the potential to increase international 

reactor safety. Similar to what occurred in 

the TMI-2 post-accident examinations, 

Fukushima Dai-ichi evaluations will, by 

necessity, be guided by what is learned 

from initial observations into the reactor 

vessel and/or containment. To maximize 

the potential benefits of a Fukushima 

Dai-ichi reactor inspection effort, is the 

ACRS recommends that current 

international efforts be expanded and a 

consortium be formed to assist the 

Japanese and provide international input 

related to areas of interest for inspection. 

ACRS experience indicates that the types of 

data that need to be collected fall within 

these topical areas:

• initial stages of core degradation

• late stage core degradation

• behavior in the lower plenum

• degradation of upper core structures 

and radionuclide deposition

• observed materials and radionuclide 

deposition and behavior in the drywell

• observed materials and radionuclide 

behavior in the reactor building

The information needs in each of these 

categories are based on phenomenological 

modeling uncertainties. The ACRS expects

information, such as the operation of cooling 

systems (e.g., high-pressure coolant 

injection (HPCI), reactor core isolation 

cooling (RCIC), and isolation condenser (IC) 

and the operation of instrumentation, would 
13

also be addressed by other international 

efforts to understand severe accident 

management. This information will also be 

crucial as input for modeling the physical 

processes that need to be addressed.

Similarly, information on the timing and 

nature of seawater injection will be crucial 

inputs to the modeling. Even with 

international support, funding will be limited 

for obtaining the desired data. Hence, it is 

essential that this international effort 

develop a consensus with respect to 

prioritized data needs. 

As discussed within this report, the 

U.S. Nuclear Regulatory Commission 

(NRC) has a long history of successfully 

leveraging its resources by relying on 

collaborations in international research 

programs. The body of knowledge gained 

from NRC’s past experimental work and that 

obtained from international experimental 

programs are systematically incorporated 

into the MELCOR accident analysis code.

The ACRS recommends that the NRC 

proactively engage with the 

U.S. Department of Energy, Japanese 

research organizations, and others in the 

international community, to focus on the 

forensics of the Fukushima accident. Such 

efforts offer a unique opportunity to better 

understand BWR severe accident 

progression, and to develop better 

measures for mitigating future beyond

design-basis events.

15

4. ADVANCED REACTOR DESIGNS

Background

Since 2005, U.S. Nuclear Regulatory 

Commission (NRC) research on advanced 

reactor designs has focused on assessment 

of its research needs and its planned safety 

research to support the review of the Next 

Generation Nuclear Plant (NGNP), 

sponsored by the U.S. Department of 

Energy (DOE). DOE has interacted with 

reactor designers, potential process heat 

users, and industrial and international 

organizations to support the next-generation 

nuclear plant (NGNP) design 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 EPAct 2005 authorizes 

the NRC to collaborate with DOE in safety 

research related to licensing issues as the 

project proceeds through licensing to 

construction and operation. As DOE is 

reevaluating its research and development 

efforts for non-LWR (non-light-water 

reactor) advanced reactor concepts, the 

VHTR remains one of the major initiatives.

Since 2010, interest has grown in deploying 

small modular reactors (SMRs). Recent 

indications that applications might be 

received for one or more SMRs in the next 

couple of years with DOE matching support 

has generated a corresponding need to 

identify safety research topics to support 

 1 See Subtitle C: Next Generation Nuclear Plant 

Project.

review of licensing applications for these 

designs. SMR integral pressurized-water

reactors (iPWRs) vendors have been 

identified as most likely to submit 

applications in the near term. These include 

mPower and NuScale.

Artist’s rendition of iPWR plant designs 

for B&W mPower and NuScale PWRs

For both non-LWR advanced reactors and 

SMRs, the NRC staff has completed initial 
16

activities to provide the proper background 

information for the lead designs. Such 

activities included:

• Development of key information sources 

for the technologies.

• Conduct of a Phenomena-Identification 

and Ranking Table (PIRT) exercise for 

each design to identify the key 

safety-significant phenomena that 

require additional research and 

development. These key phenomena 

are associated with tools, standards, 

data, etc., required for the design and 

licensing review of each reactor 

technology.

Research conducted for advanced reactor 

designs is closely coordinated with severe 

accident and source term research and with 

research on human factors, digital 

instrumentation and control systems, and 

materials and metallurgy. 

Current Research Activities

VHTR ACTIVITIES 

The NRC staff has 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 DOE and 

its national laboratory contractors. Based on 

the 2011 HTGR [high-temperature 

gas-cooled reactor] NRC Research Plan, a 

number of research topics were considered 

by the staff to support licensing:

• plant safety analysis including 

thermal-fluids and accident analysis

• nuclear analysis

• fuel performance and fission product 

behavior 

• high-temperature materials performance

• graphite performance

• safety issues related to process heat 

applications

• structural analysis (with particular focus 

on high-temperature effects) 

The research plan was comprehensive. 

However, because the final VHTR design 

has yet to be determined, the plan focused 

on the research and development (R&D) 

aspects that are applicable to the range of 

possible VHTR reactor designs. One area 

that has received needed attention was the 

development of evaluation models and tools 

covering four areas: (1) thermal-fluid 

analysis, (2) nuclear analysis, (3) fuel 

performance, and (4) fission product release 

and transport. The staff also identified a 

number of data needs for the HTGR. 

Issues that specially deserve attention 

include steam-graphite oxidation, models for 

graphite dust production during operation 

and necessary limitations on dust 

accumulation, and approaches for dealing 

with uncertainty in the selection of 

mechanistic scenarios.

By 2012, DOE had submitted a series of 

white papers to the NRC on topics identified 

by research needs as key issues: defense 

in depth (DID); high temperature materials; 

fuel qualification; mechanistic source terms; 

structures, systems, and components 

(SSCs) safety classification; emergency 

planning; licensing basis event selection; 

and PRA. The staff reviewed the DOE

white papers and issued draft assessments 

on fuel qualification, mechanistic source 

terms, defense-in-depth, licensing basis 

event selection, and safety classification of 

SSCs. 

The staff also issued a final summary report 

for the following NGNP licensing issues: 

(1) licensing basis event selection 

(2) source terms 

(3) containment functional performance 
17

(4) emergency preparedness 

Since a final design had not been selected 

and no owner/operator had been identified 

for the NGNP project, the staff emphasized 

that statements in this assessment do not 

provide final regulatory decisions. 

SMR ACTIVITIES

Work on SMRs is in its early stages, but the 

staff has identified some generic safety 

issues that are likely to be important for 

licensing decisions. Technical issues 

related to thermal-hydraulic scaling, digital 

control and instrumentation, system 

integration, licensing basis events, and 

multi-module human-systems interface are 

especially important with respect to safety 

and licensing.

Assessment and Recommendations

The 2011 HTGR Research Plan is logically 

focused on generic issues that address 

licensing needs. However, the final VHTR 

design has yet to be identified. Given 

DOE’s decreased funding in non-LWR 

advanced reactor design and without an 

applicant or a specific design, existing 

research is properly being refocused on 

generic issues for advanced reactors. 

Safety research efforts on SMR designs is 

in early stages. The NRC has identified 

important issues for licensing of these 

designs. These efforts provide a significant 

input to the development of safety research 

necessary for future reactor licensing 

activities. Any future research efforts are 

being appropriately tied to the expected 

design certifications issues identified for 

mPower and NuScale.

19

5. DIGITAL INSTRUMENTATION AND CONTROL SYSTEMS

The current operating fleet of nuclear power 

plants continues to transition to digital 

technology for reactor protection and 

engineered safeguards systems. New 

reactor designs have proposed fully digital 

control rooms and small modular reactor 

(SMR) vendors will soon submit applications 

for design certifications using digital 

technology. Application of this technology 

must preserve the critical attributes of 

redundancy, independence, deterministic 

processing behavior, defense-in-depth, and 

diversity to ensure the reliable shutdown 

and actuation of protection and safeguards 

systems. The application of this technology 

also brings with it new concerns regarding 

control of access to the systems and 

cybersecurity.

The new systems using digital technology 

must be designed such that they can be 

reviewed and analyzed to verify these 

attributes—upon which safety depends—

against engineering deficiencies, common 

cause failures, unwanted intrusion, 

unwanted interactions, and hardware 

failures. 

Fundamentally, safety systems involve 

simple actuations and do not involve 

complex feedback and control functions.

Traditionally, safety systems were kept 

simple to preserve these properties. In 

contrast, digital systems, with 

interconnections across redundant 

divisions, across safety-grade and 

lower-grade systems, and across monitored 

and monitoring components, have 

increased considerably the potential to 

degrade a safety function and compromise 

safety division independence.

With these concerns and considerations in 

mind, the 2010-2014 Research Plan was 

organized into five major topic areas:

• safety aspects of digital systems

• security aspects of digital systems

•

• advanced nuclear power concepts

• knowledge management

• carry-over projects from the fiscal year

2005-2009 digital instrumentation and 

control (DI&C) research plan

Additionally, the Office of Nuclear 

Regulatory Research (RES) responds to 

emergent program offices support needs.

The current scope of research projects 

balances previously identified research 

needs with new specific user need support 

requests.

A recent RES reorganization combined the 

DI&C and electrical research into one 

branch: the Instrumentation, Control, and 

Electrical Engineering Branch. The 

significant new electrical research projects 

include electrical cable qualification and 

condition monitoring in extended license 

intervals, performance of nuclear power 

plant (NPP) vital 125 volts direct current 

(Vdc) station battery in normal and 

extended station blackout conditions, and 

susceptibility of NPPs to faults in offsite 

power grids. The ACRS plans to include 

the electrical research projects in its future 
20

review and evaluation of the NRC safety 

research program.

Safety assurance of digital systems 

continues to be a challenge for the 

U.S. Nuclear Regulatory Commission 

(NRC). NRC and the Electric Power 

Research Institute (EPRI) have 

independently evaluated and identified the 

limitations of Failure (Fault) Modes and 

Effects Analysis (FMEA) for regulatory 

assurance of complex logic in digital 

systems – particularly in the presence of 

interactions and feedback paths. The NRC 

and EPRI presented the results of these 

evaluations to the Advisory Committee on 

Reactor Safeguards (ACRS) on 

September 19, 2013. Both the NRC and 

EPRI have been researching a broader 

range of hazard analysis (HA) techniques 

for evaluating safety assurance. Office of 

New Reactors (NRO) is proposing a new 

Design Specific Review Standard (DSRS) 

for the mPower SMR and has included the 

use of hazard analysis to improve regulatory 

efficiency and effectiveness in the safety 

evaluation of a digital system. The ACRS 

encouraged the close NRC and EPRI 

collaboration on digital system research 

under the research memorandum of 

understanding (MOU).

One of the means for safety assurance of 

digital systems is through better practice in 

software verification and validation, 

configuration management, test 

documentation, software unit testing, 

software requirements specifications, and 

life cycle management. RES has completed 

new revisions to the Regulatory 

Guides - RG 1.168, 1.169, 1.170, 1.171, 

1.172, and 1.173. The revisions updated 

these guidance documents to the more 

current Institute of Electrical and Electronics 

Engineers (IEEE) standards. The ACRS 

reviewed and supported issuing these 

updated regulatory guides.

Another major challenge is quantifying the 

reliability of digital systems in probabilistic 

risk assessments (PRAs). Several projects 

have resulted in the issuing of NUREGs that 

benchmarked methods for reliability 

modeling and assessing traditional 

probabilistic risk assessment (PRA)

methods for DI&C systems. Another project 

is developing models of digital safety 

systems integrated with reactor safety 

analysis tools as well as for use in PRAs. 

Research Information Letter 1002, 

“Identification of Failure Modes in Digital 

Safety Systems – Expert Clinical Findings,”

documented recent work on utility of digital 

system failure modes in safety assurance. 

The ACRS noted the significant progress in 

analysis of digital system failure modes, but 

is concerned that the use of these results is 

not coordinated with digital system PRA 

research. 

The NRC continues to implement new 

regulatory oversight for cyber security. 

Current policy and regulatory guidance do 

not have enforcement authority during 

digital system safety reviews; the agency 

has recognized the issue and is taking steps 

to improve oversight in this area. Several 

initiatives have been developed by the 

program offices to ensure more effective 

internal collaboration between the Office of 

Nuclear Security and Incident Response 

(NSIR), Office of Nuclear Reactor 

Regulation (NRR), and Office of New 

Reactors (NRO). NSIR developed an 

interim framework and interoffice instruction 

to provide staff guidance and NRR is 

piloting earlier cyber review considerations 

during the Diablo Canyon digital protection 

systems licensing reviews. 

The use of digital technology for plant 

systems and networks for communication of 

data and control of actuation signals 

introduces significant vulnerabilities to 

control of access to critical plant controls 

and monitoring functions. These 

vulnerabilities are further exacerbated when 

bi-directional software communications 

outside of the plant and onto the Internet 

are allowed. These features also 

significantly increase the potential for 

compromise of other cybersecurity 
21

concerns. Analog systems generally 

required a physical presence to access 

systems and controls to alter equipment 

operating and output parameters and were 

more amenable to supervisory and 

procedural controls. The ACRS still is 

concerned that the review of safety system 

design via RG 1.152, “Criteria for Use of 

Computers in Safety Systems of Nuclear 

Power Plants,” explicitly prohibits the review 

of DI&C design features for their ability to 

provide satisfactory control of access and 

cybersecurity design during the licensing 

process. The control of access and 

cybersecurity features per RG 5.71, “Cyber 

Security Programs for Nuclear Facilities,”

may not be fully evaluated until after the 

design is complete (upwards of 3 or 4 years 

after licensing) and that review would have 

to be performed by NRC site inspectors that 

may not be experienced in design of digital 

systems and in cybersecurity threats and 

solutions. Thus, there is not an integration 

of design and cybersecurity review during 

the DI&C licensing process. The ACRS 

continues to see this nonintegrated 

approach as a vulnerability to achieving 

assuredly safe and secure DI&C safety 

systems. The ACRS agrees that review 

methods and framework development 

initiatives similar to those in progress as 

noted above are needed. However, 

regulatory guidance and rule development 

that focus on this problem should be 

developed to assure design solutions allow

only one-way hardware based (without any 

software involved) data flow to locations 

external to the plant in both the design of

new plants or backfits of DI&C technology 

into existing plants.

Recognizing the ongoing need for 

knowledge management in digital safety 

systems, RES has been consulting with

experts from other regulatory organizations, 

academia, other Federal agencies, and 

other industries. Learning from their 

knowledge, RES is improving its internal 

research capability.

In its 2012 report to the Commission, the 

ACRS noted that the events at Fukushima 

illuminated the need for improved, hardened 

instrumentation that can survive anticipated 

severe accident environments and provide 

critical plant condition information.

Examples include reactor pressure, 

temperature, and water level conditions 

inside the reactor vessel; spent fuel pool 

temperature and actual water level; real 

time hydrogen concentrations at key 

locations (not a sampling system); or the 

location of core materials. Such data are 

critical for the operators to correctly assess 

the plant status and implement appropriate 

severe accident mitigation strategies. 

Determining whether core cooling was 

being achieved was hampered by the lack 

of reliable instrumentation. In addition, this 

new instrumentation needs to be either 

non-electrical or have dedicated power 

sources that are available for weeks and 

possibly longer without recharging. The 

NRC accepted the ACRS recommendation 

in this area and RES is leading a Tier 3 

Fukushima Near-Term Task Force team for 

consideration of these needs. RES is 

collaborating with research efforts in DOE 

and EPRI on this topic.

The 2010- 2014 DI&C Research Plan is 

nearing its conclusion as a planning guide.

The ACRS would like to be informed when 

the 4-year plan for 2015-2018 will be 

developed so a briefing to the ACRS DI&C 

Subcommittee can be scheduled.

23

6. FIRE SAFETY

Background

The current focus of the fire safety research 

program is to support the U.S. Nuclear 

Regulatory Commission’s (NRC’s)

regulatory needs during licensee transitions 

to risk-informed, performance-based fire 

protection programs that meet the 

requirements of Title 10 of the Code of 

Federal Regulations (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 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

• fire risk assessment. 

Efforts are also underway on fire research 

knowledge management (KM). 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.

In 2007, the NRC, the Electrical Power 

Research Institute (EPRI), and the National

Institute of Standards and Technology 

(NIST) completed a collaborative study for 

verification and validation (V&V) of fire 

models used to analyze NPP fire scenarios. 

The results of this study were documented 

in the seven-volume NUREG-1824 report, 

“Verification and Validation of Selected Fire 

Models for Nuclear Power Plant 

Applications.” 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) study (NUREG/CR 

6978, “A Phenomena Identification and 

Ranking Table Exercise for Nuclear Power 

Plant Fire Modeling Applications”) was 

completed to assess the predictive 

capabilities of these fire models and to 

identify important fire modeling capabilities 

needed to be further improved.

International Testing Program for

High Energy Arc Faults (HEAF)

The NRC is leading an international 

Organization for Economic Cooperation and 

Development/Nuclear Energy Agency 

(OECD/NEA) testing program for HEAF. The 

primary objective of this project is to obtain fire 

data on the HEAF phenomena known to occur in 

NPPs through carefully designed experiments. 
24

In 2012, the Office of Nuclear Regulatory 

Research (RES) completed another 

collaborative project with the EPRI and 

NIST to develop an integrated “Nuclear 

Power Plant Fire Modeling Application 

Guide,” NUREG-1934, for the approved 

models. The Guide describes the 

capabilities of each fire modeling tool to 

evaluate specific phenomena during 

realistic fire scenarios, including limitations, 

precautions, and lessons learned from 

practical applications. The 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).

The NRC is continuing to update the fire 

modeling tools, expand the verification and 

validation (V&V) effort for the fire models, 

and to develop additional model input data. 

The NRC published Supplement 1 to 

NUREG 1805, “Fire Dynamics Tools (FDTs) 

Quantitative Fire Hazard Analysis Methods 

for the U.S. Nuclear Regulatory 

Commission Fire Protection Inspection 

Program,” that will include the THIEF

(Thermally Induced Electrical Failure) model 

and updated versions of the spreadsheets 

that are currently documented in 

NUREG-1805 in July 2013.

Fire Testing

This element of the research program 

includes three major projects.

CHRISTI-FIRE

The Cable Heat Release, Ignition, and 

Spread in Tray Installations during Fire 

(CHRISTI-FIRE) experimental program is 

an effort to develop a more realistic 

understanding of the burning behavior of 

grouped 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 are used to develop 

more realistic cable fire models for 

probabilistic risk assessments (PRAs) and 

to enhance the predictive capabilities of fire 

modeling codes.

During Phase 1 of CHRISTI-FIRE, a simple 

model of flame spread in horizontal tray 

configuration FLASHCAT (Flame Spread 

over Horizontal Cable Trays) was 

developed based on semi-empirical 

estimates of lateral and vertical flame 

spread and measured values of heat of 

combustion, heat release rate, and residue 

yield. The Phase 1 was completed in 2011 

and the results were documented in 

NUREG/CR-7010, “Cable Heat Release, 

Ignition, and Spread in Tray Installations 

during Fire (CHRISTIFIRE), Phase 1: 

Horizontal Trays,” Volume 1, issued in 

July 2012. 

During Phase 2 of the CHRISTIFIRE 

project, tests were conducted to examine 

flame spread on cables in vertical tray 

configurations and the impact of an 

enclosure on cable flame spread in multiple 

horizontal trays. The results of the Phase 2 

tests were documented in Volume 2 of 

NUREG/CR-7010, “Cable Heat Release, 

Ignition, and Spread in Tray Installations 

during Fire (CHRISTIFIRE), Phase 2: 

Vertical Shafts and Corridors,” issued in 

December 2013.

Future phases of CHRISTIFIRE are 

currently under development and include 

examining the effectiveness of various 

methods of protection for electrical cables.

SPENT FUEL SHIPPING CASKS SEAL 

PERFORMANCE UNDER 

BEYOND-DESIGN-BASIS FIRES

 

Small-scale fire tests were performed to 

evaluate the performance of spent nuclear 

fuel shipping cask seals during 

beyond-design-basis fires that exceed the 

manufacturers’ rated temperatures. The 

results of the first phase of testing were 

documented in NUREG/CR-7115,

“Performance of Metal and Polymeric O-ring 
25

Seals in Beyond-Design-Basis Temperature 

Excursions,” issued in April 2012.

The next phase will be small scale testing to 

further characterize different polymeric seal 

material and double O-ring seal 

configurations to investigate the effect of 

multiple seals in failure times and 

temperature exposures. This information 

will be used by NMSS to further develop risk 

insights related to the transportation of 

spent fuel.

HEAF TESTING

Experiments will be conducted to explore 

the basic configurations, failure modes and 

effects of HEAF (high energy arcing faults) 

events. The equipment to be tested 

consists primarily of switchgear and bussing 

components. The project will be conducted 

as part of a larger International OECD/NEA 

effort. RES will be leading the testing 

program for HEAF. Other international 

member countries participating in the 

project will provide equipment to be tested 

as well as technical expertise.

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 fire-induced circuit damage, 

particularly involving spurious signals 

caused by “hot shorts,” is a very significant 

effort in fire risk assessments.

The Direct Current Electrical Shorting in 

Response to Exposure-Fire 

(DESIREE-FIRE) project extended the 

Cable Response to Live Fire (CAROLFIRE) 

test program (NUREG/CR-6931) to examine 

fire damage to direct current (DC) circuits. 

The DESIREE-FIRE test program involved 

a series of both small- and 

intermediate-scale fire tests. Each test 

exposed one or more electrical control 

cables commonly used in the existing fleet 

of U.S. nuclear power plants (NPPs) to fire 

exposure conditions. Each test cable was 

connected to one of several circuit simulator 

units designed to mimic the behavior of 

typical NPP components. The simulated 

dc-powered control circuits included 

motor-operated valves, solenoid-operated 

valves of various sizes, and a medium 

voltage circuit breaker unit. Cable electrical 

performance is monitored throughout each 

test to determine both the timing and mode 

of circuit faulting behavior. The tests were 

performed in collaboration with EPRI under 

a Fire Research Memorandum of 

Understanding. The results were 

documented in NUREG/CR-7100, “Direct 

Current Electrical Shorting in Response to 

Exposure Fire (DESIREE-FIRE),” issued in

April 2012.

An electrical circuit PIRT that examined the 

results from industry-sponsored cable tests 

and NRC-sponsored CAROLFIRE and 

DESIREE-FIRE was completed in 

October 2012. The results were published 

in NUREG/CR-7150, Volume 1, “Joint 

Assessment of Cable Damage and 

Quantification of Effects from Fire 

(JACQUE-FIRE).” The PIRT panel has 

identified some preliminary areas for future 

research. Those include evaluating the 

fire-induced effects on instrumentation 

circuits, electrical panel/cabinet wiring, 

surrogate ground path failure mode, current 

transformers, and high conductor count 

trunk cables. The results of the PIRT were 

also used to start a PRA expert elicitation to 

advance the state-of-the-art in 

understanding the probabilities of hot 

shorting. This work has been completed 

and its results will be published as Volume 2 

of NUREG/CR-7150. 

Fire Risk Assessment
26

Integration of fire risk into a full-scope PRA 

framework is an important research activity. 

The requirements of NFPA-805 and the 

guidance in NUREG/CR-6850, 

“EPRI/NRC-RES Fire PRA Methodology for 

Nuclear Power Facilities,” provide 

cornerstones for the development of 

risk-informed, 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. Draft NUREG-1921, 

“EPRI/NRC-RES Fire Human Reliability 

Analysis Guidelines,” was published in 

July 2012. The Office of Nuclear Reactor 

Regulation/Office of Nuclear Regulatory 

Research (NRR/RES), in partnership with 

EPRI, finalized NUREG-1934 (EPRI 

1023259, “Nuclear Power Plant Fire 

Modeling Analysis Guidelines,” in 

November 2012. Workshops on Basic 

Concepts of Fire Analysis, Basic Concepts 

of Human Reliability Analysis, and Basic 

Concepts of Nuclear Power Plant PRA were 

conducted in 2012.

Research is underway related to testing and 

evaluation of Very Early Warning Fire 

Detection (VEWFD) Systems. Research 

will enable assignment of value for such 

systems so that the specific values may be 

used in a fire PRA. The research effort will 

address testing, literature reviews, and site 

visits to support evaluation of this 

technology.

Fire Research Knowledge Management

Fire research continues to be a 

rapidly-evolving 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. 

Relative to Knowledge Management, four 

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. This work has been 

expanded and published as 

NUREG/KM-0002 in May 2013.

• “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 (“Domestic Licensing of 

Production and Utilization Facilities”), 

guidelines for fire protection, fire 

inspection manuals and procedures, 

generic letters, bulletins, information 

notices, regulatory guides, and 

fire-related NUREG reports. This work 

has been updated and expanded and is 

in the process of being published as 

NUREG/KM-0003

• “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, 
27

and planned near-term research 

programs.

• “Methods for Applying Risk Analysis to 

Fire Scenarios (MARIAFIRES) 2008,” 

NUREG/CP-0194 July 2010. This 

report captures the 2008 NRC/EPRI 

joint training classes and provides a 

self-study program that can be viewed 

at the users’ convenience. In 

August 2013, NUREG/CP-0301 

MARIAFIRES 2010 was published in 

two volumes. These two volumes 

expanded the training to include a 

volume on fire HRA and a separate 

volume on Course Prerequisites. 

Additional work is currently ongoing on 

updating and expanding the self-study 

training.

These references are updated periodically.

ASSESSMENT AND 

RECOMMENDATIONS

The NRC’s fire safety research program 

continues to make progress at the forefront 

of developing understanding of fire events 

and the assessment of risk from fire 

damage in nuclear power plants. Structured 

collaboration with industry provides 

cost-effective 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 

user-identified needs. This process is 

responsive to immediate and near-term 

technical issues. It should remain an 

important part of integrated planning. 

However, where identification of need is 

clear and beneficial, research priorities and 

programs should more actively anticipate 

additional emerging applications and 

potential 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. The results from the 

CHRISTI-FIRE and DESIREE-FIRE projects 

have also provided valuable input for 

assessment of risk from fires. 

CHRISTI-FIRE has been successful in 

providing basis for development of the 

FLASHCAT model to predict flame spread 

in horizontal tray configurations. The NRC 

should continue to encourage and support 

additional testing and fire experiments, and 

fire prediction model development, through 

collaborative efforts with U.S. industry and 

international organizations.

Guidance for the assessment of fire risk 

during low power and shutdown operating 

modes was completed in September 2013 

and was published as NUREG/CR-7114, “A 

Framework for Low Power/Shutdown Fire 

PRA.”

As identified in NUREG/CR 6978, “A 

Phenomena Identification and Ranking 

Table (PIRT) Exercise for Nuclear Power 

Plant Fire Modeling Applications,” further 

research is recommended to evaluate 

performance of fire detection systems under 

complex geometries (e.g., highly congested 

spaces), performance of incipient detection 

systems, performance of fire sprinkler 

systems under highly obstructed conditions, 

and performance of fire sprinkler systems 

against a large oil pool fire.

New reactor designs rely heavily on 

integrated digital instrumentation, 

protection, and control systems. Extensions 

of operating licenses and obsolescence of 

analog equipment will 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. To that end, based on “Joint 

Assessment of Cable Damage and 
28

Quantification of Effects from Fire

(JACQUE-FIRE)” additional research 

projects should be initiated that address 

issues related to instrumentation and control 

circuits and panel wiring. 
29

Figure 1. Current NRC Research Activities in Fire Safety

Independent Evaluation of Licensees’ Submittals in the

Area of Fire Safety

Including Evaluating Licensees’ Transition to the Risk-informed, 

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 

• HEAF 

switchgear 

cabinet and bus 

fires 

• 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

• Evaluation of Very 

Early Warning Fire 

Detection System 

performance 

31

7. REACTOR FUEL

Background

Fuel integrity is an important contributor to 

nuclear plant safety. Fuel failures that may 

occur during normal operation or anticipated 

operational occurrences (AOOs) are not 

safety significant events because 

containment systems prevent release of 

radionuclides to the environment. Fuel 

failures, however, impact plant operation 

and economics. The various mechanisms 

capable of causing failures during normal 

operation have been identified, understood, 

and have been effectively controlled by fuel 

design, cladding materials, and operational 

improvements. The great majority of U.S. 

reactors operate without fuel failures.

Nuclear plant power uprates and higher fuel 

burnups, however, are steadily increasing 

the fraction of fuel operating near peak 

power levels, and the concentrations of 

fission products within the fuel rods. 

Concurrently, fuel cladding is being 

subjected to increased radiation damage, 

oxidation, and hydriding. 

Manufacturers are addressing these more 

demanding requirements with fuel rod and 

assembly design changes, and with the 

introduction of new fuel pellet, cladding, and 

assembly structural materials. Lead use 

assemblies in which chemical additives are 

being used to improve fission gas retention 

and mechanical properties of the uranium 

dioxide (UO2) fuel pellets have been 

licensed and are in operation. Similarly,

new fuel cladding materials are in the 

development and regulatory pipeline. New 

cladding materials have been developed 

with improved corrosion and hydriding 

resistance during normal operation, and 

resistance to embrittlement during loss of 

coolant accidents (LOCAs). To a significant 

extent, these improvements have been 

made by industry as a result of findings from 

the U.S. Nuclear Regulatory Commission’s

(NRC’s) research.

NRC-sponsored research continues to 

focus on events in which fuel elements can 

fail and impact public health and safety.

However, in view of fuel development 

trends, the NRC should maintain an 

adequate research program to assure its 

capability to evaluate and license future fuel 

designs. Continued participation in 

collaborative domestic and international 

research programs addressing the 

performance of new designs under normal 

conditions, AOOs, and severe accidents is 

essential in view of the limited U.S. nuclear 

test and examination facilities and declining 

research and development (R&D) budgets. 

ORNL High Burnup Fuel Test Device

System utilizes electro-magnetic motors, a 

U-frame design, and a horizontal loading, 

enabling pure reversible bending. The grip 

sections induce uniform bending moment across 

the gauge sections
32

Current Research 

Current NRC fuel research is addressing the 

properties of discharged fuel in response to 

Office of Nuclear material Safety and 

Safeguards (NMSS) user needs, and the 

performance of operating fuel in support of 

Office of Nuclear Reactor Regulation (NRR)

user needs. 

NMSS SUPPORT

High Burnup Spent Fuel Fatigue Testing 

for Transportation Applications

Research addressing spent fuel storage and 

transportation is being conducted at Oak 

Ridge National Laboratory hot cells.

Mechanical tests of high-burnup (up to 

65 gigawatt-days per metric ton (GWd/t))

fuel rods following extended storage under 

dry cask conditions are being performed to 

determine the extent to which slow 

degradation mechanisms (such as hydride 

reorientation) occurring during extended 

storage can affect fuel cladding static and 

cyclic properties. The stiffness, failure 

stress, and cyclic fatigue properties of 

irradiated fuel rod segments as well as 

unfueled cladding is being measured. The 

results of this research will provide an 

experimental basis for staff reviews of

applications and amendments for dry-cask 

storage and transportation of spent high

burnup fuel. A novel test apparatus was 

designed and qualified for these difficult 

measurements (see sidebar on page 31).

Calibration and benchmarking has been 

completed and testing of irradiated fuel is in 

progress.

Consequence Assessment of Fuel 

Failure on the Safety of Spent Nuclear 

Fuel Dry Storage and Transportation 

Packages

The consequence assessment has been 

completed and a NUREG documenting the 

impact of a range of postulated (damaged 

or failed) fuel configurations on criticality, 

dose rates, containment, and fuel 

temperatures in transportation and storage 

packages is under review. Although the 

study has evaluated a wide range of 

physically realizable scenarios, additional 

work may be needed to evaluate the 

likelihood of the scenarios and 

configurations. 

Extended Storage of High Burnup Fuel:

Cladding Stress Analysis

Mechanisms that may affect fuel cladding 

stress and mechanical properties during 

extended dry storage are being evaluated to 

determine whether they are sufficient to 

cause a cladding failure concern. Proposed 

sources of stress are plenum gas pressure 

buildup, phase changes and reorientation of 

the hydrides upon cooling from drying 

temperatures, and swelling of the fuel due 

to a buildup of helium decay products. This 

research has established that gas 

generation resulting from the decay of 

fission products is not sufficient to increase 

the rod internal pressure, even if 

100 percent of the decay gases are 

released to the rod voids volume. This 

research has also established that fuel 

lattice swelling saturates at about 1 

displacement per atom (dpa) at values in 

the range of ~0.3 and ~0.45 percent. 

Although spent fuel will accumulate 

between 0.1 and 5 dpa over a period of 100 

years of storage, saturation effects will limit 

increases on cladding stress. These values 

are now being incorporated into a modified 

version of the FRAPCON fuel performance 

code to determine cladding stress and 

creep rates during extended exposure. 

Experimental work has been completed at 

Argonne National Laboratory to develop an 

approach to determine the effects of 

zirconium hydride dissolution and 

reorientation during cask loading and drying 

on the ductility of spent fuel cladding. This 

work has yielded definitive results by 

demonstrating that a new metric (the 

radial-hydride continuity factor) of irradiated 

cladding correlates well with the measured 

ductile-to-brittle transition temperature 

(DBTT) measured by the ring compression 
33

test. Additional work using this approach is 

being pursued under U.S. Department of 

Energy (DOE) sponsorship.

NRR SUPPORT

The Studsvik Cladding Integrity Program 

(SCIP)

The NRC has participated in the 

internationally sponsored SCIP program for 

several years. The program has completed 

experimental and analytical research on the 

pellet cladding interaction (PCI) fuel failure 

phenomenon. The SCIP 1 and 2 programs 

have amassed an impressive database 

including 1,119 power ramp tests (800 

boiling-water reactors (BWR), 213 

pressurized-water reactors (PWR), 81 

pressurized heavy water reactors (PHWR), 

and 25 unspecified), as well as separate 

effects testing and modeling. Although 

limited work will continue on PCI, the SCIP 

research program will focus on the 

phenomenon of fuel fragmentation, axial 

relocation, and fuel dispersal from high 

burnup fuel following loss of coolant 

accidents (LOCA). The program will 

include: 

• Fuel rod testing at Halden and Studsvik

• Out-of-pile integral LOCA testing and 

separate effects testing to determine the

fine fuel fragmentation burnup threshold,

parameters influencing fine 

fragmentation, and the cladding strain 

threshold for fuel axial mobility

• Separate effects testing to investigate 

role of fuel microstructure on 

fragmentation, the effects of cladding 

overheating on post AOO cladding 

properties, and the effects of axial load 

on cladding failure

Fuel Fragmentation, Relocation, and 

Dispersal Research

NRC research was directed to determine 

the conditions that result in fuel 

fragmentation, relocation, and dispersal; the 

amount of fuel dispersed and its impact on 

post-LOCA fuel coolability; and then to

recommend appropriate regulatory actions.

Experimental and analytical work completed 

to date shows good progress in addressing 

the above questions. Available data has 

been incorporated into a new approach to 

core-wide LOCA modeling aimed at the 

prediction of fuel dispersal. Using steady 

state and transient fuel performance 

(FRAPCON & FRAPTRAN) and system 

thermal hydraulics (TRACE) codes, the 

likelihood and amount of fuel dispersal for a 

small set of BWRs and PWRs has been 

determined. 

Fuel Performance Codes FRAPCON and 

FRAPTRAN

A number of significant improvements have 

been made to the NRC’s fuel performance 

codes. For example, FRAPCON-3.5 to be 

released early in 2014 includes: updated 

cladding creep hydrogen pickup models, 

variable axial nodes and zoning for 

enrichment, Gadolinia, and annular pellets, 

expanded arrays for increased number of 

nodes and time steps, treatment of pellet 

chamfers, correlations for fuel-specific heat,

and high stress cladding creep. Similarly 

FRAPTRAN-1.5 to be released early in 

2014 includes: expanded arrays for 

increased number of nodes and time steps, 

treatment of pellet chamfers, advanced 

LOCA modeling (external plenum, double 

sided oxidation, thermal-hydraulic 

improvements), and clad ballooning models. 

In addition, research has provided statistical 

tools (ARM FRAPTRAN and ARM 

FRAPCON) which are widely used by NRR. 

Future improvements in the capabilities of 

these codes are in progress with the 

planned introductions of FRAPCON 4.0 and 

FRAPTRAN 2.0 in 2016.
34

Accident Tolerant Fuel (ATF)

In response to events at Fukushima, an 

ATF development program is being 

implemented as a collaborative effort 

among national laboratories, industry, and 

universities within the United States. Fuels 

with enhanced accident tolerance are those 

that, in comparison with the standard 

uranium dioxide—zirconium (UO2 – Zr)

system, can tolerate loss of active cooling in 

the core for a considerably longer time 

period (depending on the light-water reactor 

(LWR) system and accident scenario) while 

maintaining or improving the fuel 

performance during normal operations. A 

wide variety of unconventional fuel and 

cladding materials and fuel designs are 

being evaluated for ATF development. 

Currently, the NRC is monitoring DOE and 

industry activities related to the 

development program. A down selection of 

the most promising ATF concept or 

concepts is scheduled for 2016, and at that 

time the NRC may need to engage in 

regulatory research. As the program makes 

progress, there will be a need for NRC 

engagement in the development of an ATF 

licensing strategy. This is a long-term 

program, and initial insertion of lead test 

assemblies into commercial reactors is not 

scheduled until 2022. 

Assessment and Recommendations

NMSS Support

Current research in support of spent fuel 

storage and transportation issues is well 

structured and making excellent progress. 

NRR Support

Current research addressing fuel 

fragmentation, axial relocation, and 

dispersal is very well structured, has made 

excellent progress, and is well on its way to 

generating experimental and analytical 

results to support regulatory decisions. In 

contrast, the very large BWR and PWR 

experimental PCI database assembled 

under the SCIP program should be 

incorporated into the NRC fuel performance 

codes or the Standard Review Plan. 

The FRAPCON and FRAPTRAN fuel 

performance codes meet most near-term 

agency needs. There is ample capability to 

assess risks of fuel failure due to fuel center 

melting, excessive mechanical strain, and 

departure from nucleate boiling events.

However, the NRC is still developing 

capability to quantitatively assess the more 

likely risk of PCI fuel failures during AOOs 

for current or future fuel designs. This topic 

is also important to consider if load following 

is to be considered in future LWR operation.

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.

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.
35

Figure 2. Current NRC Research Activities in Reactor Fuel

Reactor Fuel Research

Independent Evaluation of Licensees’ Submittals in 

the Area of Fuel Designs

(An important safety consideration of fuel integrity, and

a core competency essential to the agency mission)

Experimental Programs

Development and Maintenance 

of Analytical Capabilities

• OECD/NEA Studsvik Cladding 

Integrity Project (Sweden)

• LOCA Integral Testing at the 

Studsvik Nuclear AB laboratory 

(Sweden)

• In-pile LOCA Tests in OECD/NEA 

Halden Reactor Project (Norway) 

• OECD/NEA CABRI Water Loop 

Project at IRSN (France)

• High Burnup Cladding Program 

(ANL)

• Mechanical Testing of Spent High 

Burnup Fuel Rods (ORNL)

• Fracture Toughness Behavior of 

Hydrided Zircaloy Under ThruThick Crack Growth Conditions 

(PSU)

• FRAPCON Code 

Maintenance 

(PNNL)

• FRAPTRAN Code 

Maintenance 

(PNNL)

37

8. HUMAN FACTORS AND HUMAN RELIABILITY

All nuclear facilities and activities involve 

human operators and maintenance 

personnel. Understanding how people 

interact with machines is necessary to 

ensure safe operations. The staff’s 

research activities are focused on improving 

the NRC’s understanding of the issues 

affecting human performance and 

developing tools for analyzing performance, 

as well as developing systems to improve 

the operating environment to enhance the 

opportunities for improving performance.

Research funding continues to be primarily 

associated with user needs requests from 

the Office of Nuclear Reactor Regulation 

(NRR), Office of New Reactors (NRO), and 

Office of Nuclear Material Safety and 

Safeguards (NMSS). To be successful in 

meeting long-term research needs, the 

Office of Nuclear Regulatory Research 

(RES) staffs work with counterparts in those 

organizations to craft projects that meet 

both current and future needs. 

Current Research Activities

The Human Factors and Reliability Branch 

has adopted a goal-oriented framework to

organize its research program. The four 

program goals are as follows:

(1) Maintain the infrastructure of 

expertise, facilities, capabilities, and 

data.

(2) Ensure that Human Factors 

(HF)/Human Reliability Assessment 

(HRA) methods and programs have 

sound, up-to-date technical bases and 

guidance.

(3) Improve HF/HRA methods to reduce 

uncertainty and promote the 

state-of-the-art.

The NRC Human Performance Simulation 

Facility at the University of Central Florida

(4) Expand the HF/HRA infrastructure for 

new applications (anticipated changes 

in industry).

Figure 3 shows the program goals and lists 

major research topics in each area. 

Highlights of these programs are provided 

below.

Advanced Control Room Human Factors

New nuclear power plants (NPPs) will differ 

from the current plants in several ways that 

will change how operators interact with the 

plant: many employ passive designs, all will 

employ extensive use of digital 

instrumentation and control (DI&C) systems, 

and new human-system interfaces will be 

used. 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,” 

provided the results of this process, which 

led to a number of follow-on research 

projects to support advanced control room 

designs. 
38

These programs are examining whether 

new computerized procedures can provide 

adequate safety; developing measurement 

tools to evaluate workload, situational 

awareness, and teamwork; developing 

qualitative operator models to predict 

operator performance and errors; and 

developing a technical basis for regulatory 

decisionmaking regarding safety culture. All 

these projects have the potential to improve 

operator performance and provide clear

bases for regulation.

At the juncture of HF and HRA, the RES

has procured two desktop computer based

NPP control room simulators (one housed 

at NRC headquarters and the other at the 

University of Central Florida) and is 

developing a human performance test 

facility. These facilities will allow the staff to 

perform experiments rather than relying 

solely on collaboration in contracted 

facilities.

Support for Fukushima-related Initiatives

The reactor accidents resulting from the 

earthquake and tsunami at Fukushima in 

Japan and the observed and inferred 

activities by operators at the Dai-ichi plants 

have led to HF/HRA research in several 

areas. The feasibility and reliability of 

human actions in re-evaluated flooding 

events is being studied. Guidance is being 

developed for evaluating human 

performance implications of filtered vents. 

RES is also supporting the onsite

emergency response capabilities 

rulemaking.

Fitness-for-Duty Support

This ongoing effort provides scientific and 

technical support to four concurrent 

rulemakings, support to update Regulatory 

Guide 5.73 (“Fatigue Management for 

Nuclear Power Plant Personnel”), support to 

Title 10 of the Code of Federal Regulations

(10 CFR) Part 26 (“Fitness for Duty 

Programs”) implementation, and evaluation 

of the feasibility of performance indicators 

for fitness-for-duty programs.

Safety Culture Enforcement and Policy 

Statement Implementation

This ongoing effort provides support for 

safety culture inspections and support for 

assessment of the Safety Culture Policy 

Statement implementation.

Human Reliability Analysis

Since the earliest days of probabilistic risk 

assessment (PRA), researchers and 

practitioners have sought methods for the 

analysis of human performance during 

accident scenarios that offer face validity 

and clearly demonstrable accuracy. The 

result has been a proliferation of methods, 

none completely satisfying to all interested 

parties. 

The need for resolution was recognized by 

the Commission in a November 8, 2006,

staff requirements memorandum (SRM) 

directing the Advisory Committee on 

Reactor Safeguards (ACRS) to “work with 

the staff and external stakeholders to 

evaluate the different Human Reliability 

models in an effort to propose either a 

single model for the agency to use or 

guidance on which model(s) should to be 

used in specific circumstances.” In 

response, the staff initiated a series of 

research projects. The International HRA 

Empirical Study used the Halden 

Man-Machine Laboratory (HAMMLAB) to 

simulate accident sequences. It is a 

multinational, multi-team effort 

co-sponsored by the OECD Halden Reactor

Project, the Swiss Federal Nuclear Safety 

Inspectorate, and the Electric Power 

Research Institute (EPRI). The objectives

were to collect and analyze crew 

performance data, to separately apply HRA 

models to predict crew performance, and to 

evaluate these models on the basis of a 

comparison of the simulator data with the 

model predictions. Following a pilot study 

and a follow-on study, the staff organized a 
39

U.S. HRA Empirical Study, again using the 

expertise of HAMMLAB personnel, in a 

series of experiments run at a U.S. nuclear 

power plant.

The results of the three sets of experiments 

have been examined by an international 

group that developed a number of summary 

conclusions and a strong recommendation 

that all HRA analysts develop first a solid 

and thoroughly documented qualitative 

analysis to guide later quantification. The 

NRC staff has used these conclusions and 

recommendations as the starting point for 

developing a hybrid HRA method, borrowing 

the best aspects of other methods applied 

during the three empirical studies. The goal 

is to have a single method, or a limited set 

of alternative methods with a common 

qualitative underpinning, to apply to specific 

HRA problems.

Because of fixed schedule demands, the 

staff has performed parallel specific HRA 

methods development activities for NPP 

fires, fuel cask handling, low 

power/shutdown conditions, and medical 

treatments. Participants in these programs 

have been aware of the SRM-directed 

activity to develop a single methodology, so 

we hope that they can all be brought 

together, once the SRM-driven approach is 

refined and tested.

Assessment and Recommendations

The staff’s participation in domestic and 

international collaborations strengthens the 

human factors/human reliability research 

activities and should be encouraged. These 

collaborations have included Organization 

for Economic Cooperation and 

Development, EPRI, National Aeronautics 

and Space Administration, and professional 

organizations (e.g., American Nuclear 

Society, IEEE, and American Institute of 

Chemical Engineers). The staff has 

arranged for international experts to be 

involved in many of their research efforts.

The HF/HRA research program has evolved 

into a carefully coordinated series of 

projects that are extending knowledge and 

providing useful products to regulators. The 

Integrated Decision-tree Human Event 

Analysis System (IDHEAS) HRA 

methodology program has integrated 

experience from the International and U.S. 

HRA empirical studies to identify positive 

and negative aspects of existing HRA 

methods and developed a hybrid approach. 

The project has produced three reports: a 

very helpful volume on the cognitive 

foundation for HRA, a generic HRA 

methodology document, and an evolving 

IDHEAS method for internal events at 

power. The methodology needs to be 

tested on a large-scale application to ensure 

that it is practical. The guidance must be 

improved, especially to strengthen the 

qualitative analysis.

In discussions with staff from the Human 

Factors and Reliability Branch, we found 

that three areas could benefit from 

increased emphasis in the future: the work 

on medical applications should receive 

more emphasis, the development of HRA 

and HF approaches to support the Office of 

Nuclear Security and Incident Response 

(NSIR) applications needs to be explored, 

and the work on the effects of degraded I&C 

systems on human performance should be 

expanded. 

Finally, as the forensic efforts at Fukushima 

produce new understanding of severe 

accidents at boiling-water reactors (BWRs),

new research into ways to provide improved 

support to operators during such accidents 

will be needed. 
40

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

Human Factors and 

Human Reliability Research

Providing Safety Perspectives on the Impacts of 

Human Performance on Regulated Activities

Maintaining 

Current 

Infrastructure 

• Fitness for Duty 

Support

• Safety Culture 

Enforcement

• IDHEAS Users 

Manual

• Fitness for Duty –

Fatigue Management 

Guidance

• Joint EPRI/NRC-RES 

Fire HRA Guidelines

• NUREG on 

Understanding 

Human Error in 

Radiation Therapy 

and training materials

• Updating Control 

Room Design 

Guidance 

Improving 

HF/HRA Methods

Expanding HF/HRA 

Infrastructure for New 

Applications

• Fitness for Duty Revisions

• Assess Safety Culture

Policy Statement 

Implementation

• HRA Pilot for Medical 

Applications

• HRA Methodology for 

Level III PRA

• Advanced Control Room 

Human Factors

• Computerized 

Procedures

• Measuring 

Workload/Situational 

• Awareness/Teamwork

• Staffing Evaluation

• Degraded I&C

Developing Technical 

Bases and Guidance

Documents

• Fukushima 

Initiatives

• U.S. HRA 

Empirical Study

• IDHEAS HRA 

Method 

Development

• SACADA HRA 

Data Collection
41

9. MATERIALS AND METALLURGY

Background

There are 100 power reactors operating in 

the United States. While originally licensed 

for 40 years, all of the reactors that have not 

been shut down for economic or other 

reasons have applied for a license 

extension to 60 years. The continued 

increase in reliability, coupled with 

replacement of most large components 

(steam generators, pressurizers, etc.) and 

power uprates has demonstrated the value 

of existing plants. Improved knowledge of 

the aging mechanisms in play coupled with 

improved modeling and simulation methods 

has led to the realization that operating the 

existing plants beyond the license renewal 

period of 60 years is possible. However, 

operating beyond 60 years, termed 

“Subsequent License Renewal,” will require 

careful investigation of the potential for 

current degradation phenomena, or new 

phenomena that have not manifest 

themselves, to cause a decrease in 

reliability and safety. 

Materials and metallurgy continues to be an 

active area of research within the 

U.S. Nuclear Regulatory Commission 

(NRC). This is appropriate in light of the 

efforts required by the agency to address 

known and emerging materials degradation 

phenomena in aging light-water reactors 

(LWRs), and to monitor the effectiveness of 

licensees’ aging management programs. 

Due to the importance of long-term 

operation (license renewal and subsequent 

license renewal), it is critical that the NRC 

safety research programs in materials and 

metallurgy not only be following—and 

supplementing where necessary for 

safety—current issues related to 

environmental degradation, but also be 

forward looking to stay ahead of major 

issues that may be anticipated to be 

important for long-term operation. 

Extremely Low Probability of Rupture (xLPR) 

Project 

The NRC has initiated a multi-task cooperative 

research project (xLPR) with industry that will produce 

a flexible, modular, state-of-the-art probabilistic 

fracture mechanics code that can be used to assess 

the risk of pressure boundary rupture.

Current Research Activities

The materials and metallurgy research 

efforts are conducted by the Component 

Integrity Branch and the Corrosion and 

Metallurgy Branch within the Division of 

Engineering of the Office of Nuclear 

Regulatory Research (RES). Topical areas 

of research can be grouped into five broad 

categories:

(1) proactive management of materials 

degradation

(2) component integrity assessment 

(3) non-destructive examination (NDE)

(4) environmentally assisted cracking 

(EAC)

(5) steam generator related research

RES continues to make excellent use of 

domestic and international cooperative 
42

programs to accelerate progress, reduce 

cost, and resolve key issues related to the 

detection, understanding, and mitigation of 

materials degradation phenomena.

Proactive Management of

Materials Degradation

The nuclear industry and the NRC have 

often been surprised by unexpected 

materials degradation events. Unexpected 

materials degradation in nuclear power 

plants has led to a heightened interest by 

the NRC and nuclear power industry in 

developing a proactive approach to 

materials degradation management. The 

NRC initiated the project “Proactive 

Materials Degradation Assessment” 

(PMDA) to identify materials and systems in 

LWRs where degradation can reasonably 

be expected to occur in the future. With 

such knowledge, the inspection and 

monitoring programs at plants could be 

reviewed and modified as needed to provide 

early identification of incipient degradation.

A comprehensive assessment of the 

likelihood and safety significance of possible 

environmental degradation mechanisms for 

approximately 1,900 boiling-water reactor 

(BWR) and pressurized-water reactor 

(PWR) components was completed in 2008. 

The results of this effort have been 

documented in NUREG/CR-6923, “Expert 

Panel Report on Proactive Materials 

Degradation Assessment.” 

The Advisory Committee on Reactor 

Safeguards (ACRS) supported the initial 

vision of the proactive materials degradation 

assessment (PMDA) project and the later 

expansion of its original scope to include 

concrete structures and electrical cable 

insulation. The NRC, in collaboration with 

the U.S. Department of Energy (DOE), has 

undertaken an expansion of the scope 

originally investigated in NUREG/CR-6923 

to identify any additional degradation 

phenomena that must be considered for 

extension of operation beyond 60 years. 

The results of this effort are being 

documented in a five-volume 

NUREG/CR-7153, “Expanded Materials 

Degradation Assessment” (EMDA). The 

analyses have focused on the following 

areas: (1) core internals and piping 

systems, (2) reactor pressure vessel steels, 

(3) concrete civil structures, and (4) 

electrical power and instrumentation and 

control cabling and insulation.

The EMDA process has resulted in:

• Determining the status of the current 

knowledge base and updating the 

PMDA-developed information.

• Identifying gaps in knowledge for 

systems, structures, and 

components (SSC) or materials that 

warrant future research.

• Identifying potential new forms of 

degradation.

• Identifying and prioritizing the

research needs.

Assessment: The program has identified 

key areas of concern for subsequent license 

renewal operation (time frames of 60-80

years). The current research programs will 

likely be adequate to address most areas of 

concern. There is one area that is not 

currently being addressed that should be 

further evaluated. There is recent research 

that indicates that the exposure of stainless 

steels and nickel-based alloys to 

high-temperature water in the absence of 

radiation damage can result in a significant 

decrease in tearing resistance and fracture 

toughness. If confirmed, this effect, in 

combination with radiation-induced 

toughness reduction, may become a safety 

issue in the future.

COMPONENT INTEGRITY ASSESSMENT

The topical areas of research within the 

component integrity assessment addresses 

the development of a probabilistic fracture 
43

mechanics tool (xLPR), the integrity of 

dissimilar metal welds, and reactor pressure 

vessel (RPV) integrity. In addition to 

research on metallic components, work is 

also being done to confirm the acceptability 

of high-density polyethylene (HDPE) piping 

for long-term safety-related applications.

Development of the xLPR code: General 

Design Criterion 4 (GDC-4), “Environmental 

and Dynamic Effects Design Bases,” of the 

Appendix A to Title 10 of the Code of 

Federal Regulations (10 CFR) Part 50, 

“General Design Criteria for Nuclear Power 

Plants,” allows local dynamic effects of pipe 

ruptures to be excluded from the design 

basis if it is demonstrated that the 

probability of fluid system piping rupture is 

extremely low under conditions consistent 

with the design basis. Leak before break 

(LBB) must be assured. However, if EAC is 

present, there is the possibility that for some 

conditions these requirements may not be 

met. The purpose of the xLPR program, a 

cooperative effort between the NRC and 

industry, is to develop the methodology 

which, when combined with inspection, will 

assure that these criteria will be met. The 

xLPR program has developed a probabilistic 

framework for the assessment of LBB 

issues in primary system SSCs. Although 

initially focused on resolving the primary

water stress-corrosion cracking (PWSCC) 

challenge for PWRs, the intent of the 

extremely low probability of rupture (xLPR)

project is to develop a fully probabilistic 

approach applicable to a range of active 

degradation mechanisms associated with 

both BWRs and PWRs.

Assessment: The xLPR program is 

developing a quantified, fully quality 

assurance (QA) certified, solution to the 

LBB issue. It represents the state-of-the-art 

in probabilistic fracture mechanics. Use of 

xLPR will allow for assessment of 

component life for current, extended, and 

subsequent license renewal periods. The 

program is meeting and exceeding its goals.

Reactor Pressure Vessel Integrity: 

Reactor pressure vessel embrittlement has 

been identified as a potential life-limiting 

issue for operating in the 60-80 year time 

frame. As such, it is important that the NRC 

safety research program supports the 

evaluation of pressure vessel embrittlement 

as the plant ages. 

The reactor pressure vessel integrity 

research has led to revision of several NRC 

regulations, such as the pressurized thermal 

shock rule in 10 CFR 50.61, “Fracture 

Toughness Requirements for Protection 

against Pressurized Thermal Shock events,” 

and 10 CFR 50.61a, “Alternate Fracture 

Toughness Requirements for Protection 

against Pressurized Thermal Shock 

Events,” as well as related American 

Society of Mechanical Engineers (ASME) 

and American Society for Testing and 

Materials (ASTM) codes and standards. 

Work is now in progress to revise other 

rules that address pressure vessel integrity 

such as Appendix G, “Fracture Toughness 

Requirements,” and Appendix H, “Reactor 

Vessel Material Surveillance Program 

Requirements,” to 10 CFR Part 50. 

The NRC is also sponsoring the Reactor 

Embrittlement Archive Project (REAP) at the 

Oak Ridge National Laboratory (ORNL) to 

create a Web-based RPV steel 

embrittlement archive database for the 

entire U.S. fleet of 104 commercial NPPs. 

The REAP project is intended to provide an 

important tool for reactor pressure vessel 

(RPV) integrity assessment within the NRC 

regulatory environment.

Assessment: The RPV integrity programs, 

as currently structured, are meeting goals. 

However, these programs will need to be

reassessed in light of the recommendation 

that will come from the EMDA analysis.

HPDE: HDPE piping has been used for 

more than 10 years in nonsafety-related 

service water piping in nuclear power 

plants. HDPE piping has very attractive 
44

characteristics in the areas of corrosion and 

ease of installation/fabrication. Special 

exemptions have been applied for and 

granted for use in selected safety-related 

applications, but this has been rare due to 

issues related to joining, the inspection of 

“welded” joints, and aging-related 

phenomena that are not fully understood. 

ASME Code Case N-755-1, “Use of 

Polyethylene (PE) Class 3 Plastic Pipe 

Section III, Division I and Section XI” 

describes the requirements for use of HDPE 

in Class 3 buried pipe systems. The current 

NRC research programs are being 

conducted in collaboration with EPRI and 

with the Nuclear Energy Standards 

Coordination Collaborative (NESCC). The 

goals of these programs are to confirm the 

proposed requirements for use in ASME 

Class 3 safety-related applications and to 

support the Office of Nuclear Reactor 

Regulation (NRR) and the Office of New 

Reactors (NRO) in ASME code actions and 

relief requests. Confirmatory research is 

also being conducted related to fusion joint 

integrity and nondestructive evaluation

(NDE) methods. 

Assessment: The level of effort in this area 

is appropriate for the need. HDPE will see 

much more extensive use in the future as 

service water and other related systems are 

repaired or replaced. The maintaining of 

joint integrity and verification by NDE 

methods are the key issues for their use in 

safety-related systems.

Non-destructive Examination

Improvement of NDE resolution and 

sensitivity will be of increasing importance 

for extended operation where margins may 

be reduced. There are currently three 

ongoing research programs: (1) 

Effectiveness and Reliability of NDE for 

Vessels and Piping; (2) Evaluation of 

Ultrasonic Testing (UT) in Lieu of 

Radiological Testing (RT) for Repairs and 

Modifications; and (3) Program to Assess 

Reliability of Emerging Nondestructive 

Techniques (PARENT).

The project “Effectiveness and Reliability of 

NDE for Vessels and Piping” is a renewal of 

a previous program of the same name. As 

plants age and the possibility that heretofore 

unknown degradation mechanisms may 

manifest themselves, it will be critical that 

more accurate and detailed characterization 

of degradation be achieved. The industry, 

through EPRI, has made significant strides 

in the development of new techniques and 

application of these techniques to areas 

where previous techniques have been 

inaccurate. This is especially true for welds 

where stress corrosion cracking (SCC) is an 

active degradation mechanism. It is 

important that the NRC maintain 

competence in this area and maintain the 

expertise to evaluate new techniques. 

The project “Assessment of Reliability of 

Emerging NDE Techniques” is a 

compliment to the effectiveness and 

reliability project above. The project 

focuses on the reliability of inservice 

inspection (ISI) techniques for degraded 

welds. A cooperative agreement between 

the NRC and the French IRSN is being 

implemented to develop techniques for 

inspection of coarse grained materials in 

dissimilar metal welds.

An additional research project is underway 

to evaluate the use of phased array UT to 

detect PWSCC leakage between the reactor 

upper head penetration and the reactor 

pressure vessel head such as has occurred 

at the North Anna nuclear power station. 

The leakage path was detected by UT and 

confirmed by destructive examination of the 

nozzle. In addition, the pattern of boric acid 

deposits observed with phased array UT

was in agreement with observation during 

destructive examination at North Anna.

Assessment: Accurate NDE methods are 

critical to assurance of system integrity and 

complement the probabilistic fracture 

mechanics methods development in the 
45

xLPR program. New NDE techniques 

promise to reduce the uncertainty in 

characterization of defects, which will 

improve the results of component life

evaluation. This will be especially important 

for operation beyond 40 years. The current 

programs will keep the staff current in the 

latest developments. These programs are 

meeting their goals.

Environmentally Assisted Cracking

Environmentally assisted cracking 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. The topical areas of research within 

the environmentally assisted cracking (EAC)

umbrella include irradiation assisted 

degradation, primary water stress corrosion 

cracking (PWSCC), environmental fatigue, 

spent fuel storage system degradation, 

containment liner corrosion, and used fuel 

neutron absorber issues.

Irradiation Assisted Degradation

/PWSCC: As has been discussed above, 

SCC is the dominant degradation mode for 

pressure boundary and structural materials 

in the primary system. The addition of the 

radiation environment aggravates the 

situation by causing radiation hardening, a 

reduction in ductility and toughness, and 

radiation induced segregation (RIS) that 

increases susceptibility to PWSCC. In 

addition, for cast materials there is the 

potential for a combination of thermal aging 

and neutron exposure to reduce toughness 

in these materials. Radiation effects, 

including hardening and swelling, are also 

an issue for extended operation beyond 40 

years. However, it is believed that this 

phenomenon is adequately understood and, 

as such, will not likely be an issue for 

extended operation. Research programs in 

this area include the study of: (a) cracking 

of stainless steels in low corrosion potential 

(e.g., PWR) environments, (b) the effect of 

irradiation on void swelling-material 

exposed in the BOR-60 reactor, and (c) 

crack growth and fracture toughness of 

irradiated case stainless steel. Most of 

these programs have recently terminated. 

The key results of these programs include

the following:

(1) In low potential environments (PWR) 

both cold worked and non-cold worked 

stainless steels exhibit increasing SCC 

crack growth rates with increasing 

fluence from 5 to 25 displacements per 

atom (dpa).

(2) Fracture toughness of the same 

materials decreases with increasing 

fluence up to 8 dpa with toughness 

J values at crack initiation below

100 kJ/m2

.

(3) For cast stainless steel that has been 

thermally aged, low dose (0.08 dpa) 

decreased the fracture toughness in 

addition to the decrease from thermal 

aging alone.

In addition to the program on stainless 

steels, a research project is underway on 

PWSCC of nickel-based alloys. The 

purpose of this project is to help in the 

development of a safety evaluation criteria 

for this material when used as pressure 

boundary. The focus has been on Alloy 690

and its weld metals (Alloy 52 and 152)

compared to Alloy 600 and its weld metals 

(Alloy 82 and 182). The results have shown 

that Alloy 690 is very resistant to PWSCC 

unless subjected to significant cold work. 

The results have also shown that dissimilar 

weld dilution zones may be susceptible to 

PWSCC.

Assessment: PWSCC is the dominant 

degradation mechanism for the primary 

pressure boundary. While replacement 

components will be fabricated with more 

resistant materials using Alloy 690 and its 
46

weld metals, this form of degradation will 

continue to be an important issue especially 

for extended operation. It is thus very 

important that the staff maintains expertise 

in this area and that the NRC participates in 

research in this area. Research in PWSCC, 

via collaboration or partnering with industry, 

should continue. The international 

community has established a collaborative 

body that is now called the International 

Cooperative Group on Environmentally 

Assisted Cracking (ICG-EAC). It has been 

in operation for several decades and was 

co-founded by the NRC. The group meets 

once a year and brings together the world’s 

experts in this area. The NRC is currently a 

member and should remain so.

 

Environmental Fatigue: Environmental 

fatigue has been identified as a potential 

issue for extended operation (EMDA). 

Research in this area includes updates to 

the existing environmental fatigue 

evaluation methodology as outlined in 

Regulatory Guide 1.207, “Guidelines for 

Evaluating Fatigue Analyses Incorporating 

the Life Reduction of Metal Components 

Due to the Effects of the Light-Water 

Reactor Environment for New Reactors,” 

and NUREG/CR-6909, “Effect of LWR 

Coolant Environment on the Fatigue Life of 

Reactor Materials.” 

Assessment: The level of effort in this area 

is appropriate for the need.

Spent Fuel Storage Systems 

Degradation: The dry storage canister 

systems in use today were designed with 

the expectation that within approximately 20 

years the spent fuel would be transferred to 

a permanent storage facility. Materials 

choices were made with this assumption. 

Conventional austenitic stainless steel 

(almost exclusively Type 304) was selected 

as the pressure boundary material. 

However, with the new requirements that 

dry storage systems be available for time 

periods greater than 80 years, the materials 

choice has now been demonstrated to have 

been a poor one. Painful experience has 

shown that the use of this material in a 

chloride-containing environment can be 

problematic if great care is not taken to 

eliminate tensile stresses. Moreover, for 

exposure in salt environments, salt 

deliquescence can take place at 

temperatures in the 50-60°C range and 

humidity values that are normally present. 

Chloride concentrations in the thousands of 

parts per million (ppm) are easily 

achievable. This chemistry, in combination 

with non-stress relieved welds, will 

eventually result in SCC of this material.

The realization that the reliability of dry cask 

storage systems will be a concern has 

prompted the development of a number of 

research programs that are being 

conducted in collaboration with DOE.

These include: (a) SCC of Dry Storage 

Canisters, (b) Vacuum Drying of Spent Fuel 

Canisters, and (c) Functional Monitoring of 

Dry Cask Storage Systems.

The results of this research have confirmed 

that for environments that can easily be 

achieved with canisters at sites where salt 

can deposit on the material, SCC will occur. 

The overall conclusions of the 

chloride-exposure research are as follows:
47

(1) For simulated sea salt, SCC on Type 

304 stainless steel is possible and has 

been observed between 35 and 80°C 

when the relative humidity is higher than 

20-30 percent.

(2) SCC initiation has been observed at salt 

quantities as low as 0.1 g/m2 or a metal 

strain as low as 0.4 percent. The extent 

of cracking increases with increasing 

salt quantity or strain.

(3) Sensitized material is more susceptible 

to SCC than material in the as-received 

(non-sensitized) condition.

(4) No SCC was observed for material 

exposed to non-chloride bearing 

species.

In addition to SCC-related research, there 

are programs that address vacuum drying 

and monitoring of canisters.

Assessment: Much of the research in this 

area is being conducted by DOE as a part 

of Nuclear Energy University Programs 

(NEUP). These include modeling of 

canister life, drying and development of 

monitoring systems. The NRC staff is also 

participating in the industry sponsored 

Extended Storage Collaboration Program 

(ESCP) that is the umbrella for research in 

this area. The direct NRC funded research 

in this area is making a significant 

contribution to the knowledge base and 

should be continued. Close monitoring of 

the non-NRC funded programs should be 

maintained and additional research

conducted if significant gaps are observed.

Containment Liner Corrosion:

Containment liner corrosion has occurred in 

cases where unique conditions have been 

present. Corrosion is almost exclusively 

from the outside when a foreign object

(piece of wood, etc.) has been left in contact 

with the steel liner during construction. 

Perforation of the containment can result in 

leakage during a design-basis loss of 

coolant accident. The purpose of this 

research is to determine if additional 

inspections should be required in 

accordance with ASME Section XI, 

Subsection IWE. The conclusions of this 

research include the following:

(1) The corrosion cell that can be 

developed will support a high corrosion 

rate and a significant corroding area.

(2) Through wall corrosion is initiated from 

foreign objects left in the concrete 

during construction.

(3) The leak rate is controlled by the size of 

the hole in the containment when the 

hole is small (~ 10 mm2

).

(4) Radioisotope releases would be largely 

restricted by the narrow gap between 

the liner and the containment wall.

Assessment: The level of effort in this area 

is appropriate for the need.

Neutron Absorber Degradation: Neutron 

absorber is used in the spent fuel pool to 

allow for higher density racking and prevent 

criticality. The boron is imbedded in either a 

metal or phenolic resin based material. 

Degradation of the materials results in a 

loss of boron from these materials. 

Additionally, modeling of this degradation 

has been difficult with a lack of correlation 

between experiment and model results in 

many cases. The research programs in this 

area are designed to determine if future 

regulatory actions are needed.

Assessment: The level of effort in this area 

is appropriate for the need.

Steam Generator Related Research

Stress corrosion cracking has been 

problematic for steam generators with mill 

annealed Alloy 600 tubing. While thermally 

treated Alloy 600 tubed steam generators 

have experienced minimal SCC, the 
48

consensus is that these steam generators 

will eventually experience SCC as well. 

Operation beyond the 40 year initial license 

period adds additional concern that tube 

degradation in Alloy 600 tubed steam 

generators will continue to be an issue.

Essentially all of the original mill annealed 

Alloy 600 tubed steam generators have 

been replaced with either thermally treated 

Alloy 600 or Alloy 690 tubed units. 

Alloy 690 has been shown to be extremely 

resistant to SCC and no instance of SCC in 

these units has been reported to this date. 

However, as the degradation due to 

excessive wear in the San Onofre steam 

generators has demonstrated, although due 

to excessive thermo-elastic vibration, the 

performance of steam generator tubing 

must continue to be monitored.

The focus of the NRC-sponsored research

in this area has been to provide additional 

data to support the technical basis for 

nondestructive evaluations, tube integrity,

and consequential tube rupture (CSGTR). 

The research builds on extensive previous 

research and will support technical 

evaluations of licensee submissions and 

inspections of steam generator tubes. The 

results of this program will be documented 

in several NUREGs or technical letter 

reports. These reports will include the 

following:

1. “Consequential SGTR Analysis for 

Westinghouse and Combustion 

Engineering Plants with 

Thermally-Treated Alloy 600 and 690 

Steam Generator Tubes” 

2. “Algorithms to Automatically Analyze

Eddy Current Data”

3. “Creep and Leak Rate Models for Alloy 

690 Steam Generator Tubes”

4. “Stability of Circumferential Flaws in 

Once-through Steam Generator Tubes 

Under Thermal Loading during LOCA, 

MSLB, and FWLB”

5. “Leak Rates and Burst Pressures for 

Flaws in the U-bend Region of Steam 

Generator Tubes”

6. “Assessment of Eddy Current Methods 

to Detect and Size Flaws in the U-bend 

Region of Steam Generator Tubes”

7. “Development and Validation of Models 

for Predicting Leakage from Degraded 

Tube-to-Tubesheet Joints during Severe 

Accidents”

8. “Evaluation of Examination Guidelines 

for Pulled Steam Generator Tubes”

Assessment: The current industry 

inspection and integrity models have been 

confirmed by the NRC-sponsored research. 

Moreover, there is now a very large 

database to support analysis of licensee 

submittals. Alloy 690 tubed steam 

generators are exhibiting excellent behavior 

and the inspection programs have been in 

place for over 20 years. Thus, while the 

current programs are expected to produce 

valuable insight, there is little justification for 

further NRC-sponsored research in this 

area with the exception of special cases that 

may arise in the future.
Figure 4. Current NRC Research Activities in Materials and Metallurgy

Materials and Metallurgy 

Research

Component Integrity 

Assessment

• Development of 

probabilistic fracture 

mechanics code, xLPR

• RPV integrity

• Acceptability of HDPE 

piping for long term 

safety related applications

License Renewal & 

Aging Management

Evaluating the Adequacy and Effectiveness 

of NDE and ISI Methods and Programs

Developing Risk-info

Regulation of Reac

Int

Nondestructive 

Evaluation

Environmentally 

Assisted Cracking

S

• 

• 

• 

• 

• Irradiation Assisted 

Degradation/PWSCC 

• Environmental Fatigue

• Spent Fuel Storage 

Systems Degradation

• Containment Liner 

Corrosion

• Neutron Absorber 

Degradation

Proactive Management 

of Materials 

Degradation

• Expanded Materials 

Degradation 

Assessment 

(EMDA) 

• Evaluation of the 

effectiveness & 

reliability of NDE of 

vessels & piping 

• Assessment of 

emerging NDE for 

dissimilar metal 

welds 

49

51

10. NEUTRONICS AND CRITICALITY SAFETY

Background

The technical capability to conduct 

neutronics and criticality evaluations in 

concert with state-of-the-art computational 

analysis tools is an essential core 

competency of the U.S. Nuclear Regulatory 

Commission (NRC). This capability 

connects with numerous research and 

regulatory activities within the agency and 

thereby continues to be of particularly high 

importance. Specific areas that define the 

focus of these research activities in 

neutronics and criticality safety include:

• Fission reactor neutronics analyses 

which include moderator, coolant, and 

fuel types for current and advanced 

designs and configurations and address 

steady-state and transient conditions,

• Nuclide generation and depletion 

computations to evaluate reactor and 

spent nuclear fuel decay heat powers, 

radioactive source behaviors, and 

radionuclide inventories,

• Radiation transport and attenuation 

applied to material damage, dosimetry, 

activation, and shielding evaluations for 

radiation protection, and

• Nuclear criticality safety, preventing and 

mitigating critical configurations outside 

reactors, for a variety of applications.

Important categories of recent and future 

direct application activities are:

• Confirmatory analyses for licensing 

reviews of operating light-water reactor 

extended power uprate applications, 

including licensing approval of advanced 

modes of operation for boiling-water 

reactors (BWRs),

• Burnup credit methodology development 

for spent nuclear fuel criticality analysis 

for pressurized-water reactor fuel,

Power oscillation visualization during 

simulated ATWS with instability

TRACE/PARCS methodology is being used to 

simulate complex transients such as anticipated 

transients without SCRAM with instability 

(ATWS-I). Visualization tools have been 

developed to analyze and display the evolution 

of the power oscillations during ATWS-I and to 

study the oscillation contour.

• Certification of new light-water reactor 

(LWR) core designs and confirmation of 

applicant accident analyses, especially 

for integral advanced reactors and small 

modular reactors, and

• Spent fuel criticality analysis 

methodologies for advanced and long 

term fuel storage analysis for spent fuel 

pools and canisters.

Related Office of Nuclear Regulatory 

Research (RES) programs supported and 

monitored by this research area include the 

fuel cycle oversight process, fuel rod 

mechanical and thermal analysis, and high 

burnup fuel performance and safety 

evaluations. An important component of 

this RES work is continuing training 
52

programs for agency staff on the use of 

these computer codes and application 

methods for performing licensing analysis 

and related review work.

The primary computational resources 

available to the staff for neutronics and 

criticality analysis include the SCALE suite 

of computer codes maintained in 

cooperation with the U.S. Department of 

Energy (DOE) through Oak Ridge National 

Laboratory (ORNL). The Purdue Advanced 

Reactor Core Simulator (PARCS) code 

provides three-dimensional core depletion 

and reactor kinetics that can be used in 

conjunction with the TRACE (TRAC/RELAP 

Advanced Computational Engine) computer 

code for thermal-hydraulics analyses. 

Current Research Activities

The research programs are defined and 

supported by user needs within the agency 

and/or by NRC obligations under the Energy 

Policy Act. The program plans are 

well-developed and documented 

appropriately. Specialized areas of 

particular importance at this time in the 

areas of reactor performance analysis, 

methods development, training programs, 

and analysis methodology support are listed 

below:

• Evaluate recent analyses applications 

using TRACE/PARCS to support staff 

evaluations of the Maximum Extended 

Load Line Limit Analysis Plus ( 

MELLLA+) protocol for extended power 

uprates in BWRs for near term 

improvement opportunities in analysis 

protocol, code capability enhancement, 

and benchmarking requirements.

• Maintain the PARCS code package and 

expand training of NRC staff in use of 

the code and its limitations. Provide 

user support for the PARCS code 

package for use in analysis of LWR 

designs.

• Augment the recently completed 

pressurized-water reactor (PWR) spent 

nuclear fuel burnup credit methodology 

and applications to develop the 

regulatory basis for burnup credit for 

BWR fuel in spent fuel pools, dry 

storage canisters, and transport 

systems. This work is a long term 

research program element and includes:

o Determining reactivity influences due 

to variations in input parameters for 

BWR burnup credit criticality safety 

evaluations.

o Supporting isotopic analyses 

validation (SCALE/TRITON 

depletion calculations) with available 

measured isotopic composition data.

o Evaluating available critical 

experimental data to support BWR 

fuel critical evaluations.

• As was done within the PWR burnup 

credit analysis and application methods 

research program, develop a logical and 

consistent approach to identification and 

treatment of uncertainties in the analysis 

for BWRs.

• Provide continued support for the 

development and use of benchmark 

data for validation of burnup credit 

criticality analyses, including the 

MINERVE database from France that is 

part of the Organization for Economic 

Cooperation and Development/Nuclear 

Energy Agency (OECD/NEA) Burnup 

Credit Benchmark Exercise. Identify 

any additional improvement 

opportunities from this work for the PWR 

analysis program and assure program 

application to the BWR methods 

development and analyses 

requirements. Focus should include 

technical basis development and 

support for expanded guidance on 

burnup credit in spent nuclear fuel 

storage and transport applications. 
53

• Develop a user guide and deliver staff 

training for the application of 

SCALE/TRITON in the generation of 

cross section libraries for the PARCS 

code.

• Upgrade the SCALE code package 

capabilities in the areas of in-core power 

peaking, reactor stability and control, 

core monitoring, fuel burnup, 

radionuclide inventories for accident 

source terms, safe shutdown, decay 

heat power, radiation sources and 

attenuation, and nuclear criticality 

safety. Modify SCALE architecture to 

enable parallel processing in analysis of 

accident sequences.

Lower Priority Research Activities 

(related to NGNP, MOX, and advanced 

fuel cycles)

As with other research areas, some of the 

program elements have been slowed or 

deferred as a result of resource redirection 

due to Fukushima response activities. The 

following activities have been affected 

based upon programmatic and priority 

changes since the 2012 ACRS (Advisory 

Committee on Reactor Safeguards) review 

and evaluation of the program:

• Develop nuclear data libraries for Next 

Generation Nuclear Plant (NGNP)

designs for use in the SCALE code. 

Investigate reactor neutronics for the 

prismatic and pebble-bed designs for 

NGNP. Enhance computational 

capabilities to enable prediction of 

nuclide composition, decay heat, 

criticality and associated uncertainties 

for gas reactor fuel at burnup of 80 and 

160 gigawatt days per metric ton 

(GWd/t). Enhance radiation transport 

methods and databases to assess 

shielding and evaluate irradiation 

degradation of the structural integrity of 

graphite and metal components in 

NGNP.

• Perform criticality analyses to support 

licensing of the Savannah River mixed

oxide (MOX) fuel fabrication facility.

• Investigate criticality issues of fuel with 

greater than 5 percent enrichment in 

uranium-235 (235U).

Assessment and Recommendations

The prioritization to achieve progress on key 

research areas in this program is 

appropriate and should be effective to 

achieve the most important research 

progress and results. Accordingly, we agree 

with those elements deferred as described 

above.

Expanded application capability, improved 

accuracy, and computational efficiency 

demands on neutronics and criticality 

analyses should be achieved in both near 

term and long term research programs. 

Licensees of existing and planned reactors 

continue to optimize core designs using 

improved analytical techniques, some of 

which require full core, coupled approaches 

to safety analyses. Current analyses and 

related regulatory issues continue to 

demonstrate the need for additional 

uncertainty and sensitivity analyses to 

improve understanding and regulatory 

interpretations of the analysis results. In 

addition similar capabilities must be applied 

to assure appropriate safety analyses for 

integral small modular and large advanced 

reactors. 

As a prime example the TRACE/PARCS 

analyses used to evaluate BWR 

performance in complex ATWS transients 

initiating from the MELLLA+ operating 

regime demonstrate the value of advanced 

methodology and high quality modeling 

application. The analysis methodology and 

visualization techniques (displayed in the 

insert figure) elevate the capability of the 

RES staff to better compare and evaluate 

results, identify and explain sensitivities, 

and communicate findings with applicants 

and licensees.
54

The burnup credit methodology developed 

for PWRs to improve isotopic composition 

analyses and criticality predictions 

represents a key success. The importance 

of developing a logical and consistent 

approach to evaluate appropriate 

uncertainties was recognized and an 

effective program and methodology was

devised. 

A similar quality approach designed to 

develop analysis methods and their 

technical bases for BWR applications is 

being implemented. The staff has identified 

the objectives that are critical to the success 

of this program. This work should also 

continue to incorporate and support 

international benchmark validation derived 

within the OECD/NEA MINERVE program. 

Further, the staff has ongoing plans to 

modify and extend the existing NRC codes 

and methods to accommodate new spent 

fuel pool, spent fuel canister, and core

configurations, including those for current 

and advanced light water reactors as well 

as small modular reactors.
55

11. OPERATIONAL EXPERIENCE

Background

Operational experience (OpE) data can 

provide an important source of information 

which can be used to refine and improve the 

effectiveness of the regulatory process. 

Since passage of the National Energy 

Policy Act in 1977, the U.S. Nuclear 

Regulatory Commission (NRC) has explicit 

responsibility to collect and analyze OpE 

data. The NRC decided at that time to use 

data from a voluntary industry program; 

initially the Nuclear Plant Reliability Data 

System, and currently known as the Institute 

of Nuclear Power Operations (INPO) 

Consolidated Events System (ICES). 

Data are also obtained from Licensee Event 

Reports (LERs) submitted directly to the 

NRC.

With the increased use of risk insights to 

inform the regulatory process, OpE data, 

when properly documented and analyzed, 

are important to assure the risk insights 

have an adequate basis for use in 

decisionmaking.

Finally, OpE data provide a measure of 

regulatory effectiveness and input to the 

required annual report to Congress on 

significant operating events.

Sources and uses of operating data and analyses in NRC regulatory programs
56

Current Research Activities

The Office of Nuclear Regulatory Research 

(RES) Data Collection, Analysis and 

Trending Program (Program) interfaces are 

shown on page 58 (Figure 5). The program 

is ongoing, and it evolves as required to 

meet changing needs. An important user of 

program output is the Accident Sequence 

Precursor Program that results in an annual 

SECY which addresses: (1) the evaluation 

of precursor data trends and (2) the ongoing 

development of associated risk models. 

Other users include the Reactor Oversight 

Program and Industry Trends Program. 

Data analysis and summarization are 

currently performed by Idaho National 

Laboratory. The Program process and 

results are well displayed on numerous 

NRC Web site pages, so that much of the 

information is made widely available both 

within and outside the NRC. A few of the 

topics on the current Reactor OpE Results 

and Databases page include the following 

as examples:

• Fire Protection

• Fitness-for-Duty Programs

• Multiple/Repetitive Degraded 

Cornerstone Program

• Buried Piping Activities

• Loss of Offsite Power

• Industry Performance of Relief Valves

• Common-Cause Failure Insights

In its 2012 report to the Commission on the 

review and evaluation of the NRC Safety 

Research Program, the Advisory Committee 

on Reactor Safeguards (ACRS)

emphasized the importance of OpE data to 

the enhancement of planning for the 

operational response to an emergency 

event. This work is ongoing in support of 

Near-Term Task Force (NTTF)

Recommendation 8. The Human Factors 

Information System (HFIS) provides a 

limited OpE data reference concerning 

human performance. 

Assessment and Recommendations

The primary sources of operating 

experience data are the INPO’s EPIX

(Equipment Performance and Information 

Exchange System), LERs, and inspection 

reports. These sources provide information 

operational failures. When these data are 

used to support risk analysis, they must be 

augmented with estimates of the successes 

(all demands, both successes and failures, 

must be considered, i.e., data are needed 

on the number of demands and times of 

operation). The demands are highly 

plant-specific, involving operating practices, 

the details of testing (e.g., exactly which 

portion of the equipment is tested), and 

details of the surveillance procedures (e.g., 

a pump test often requires operation of a 

number of associated valves generating 

additional demands on those valves).

Many plant-specific probabilistic risk 

assessments (PRAs) have documented 

thorough success and failure data and the 

NRC’s Handbook of Parameter Estimation 

for Probabilistic Risk Assessment 

(NUREG/CR-6823) provides guidance on 

this process. Lacking accurate 

plant-specific success data, it must be 

recognized that there is substantially greater 

uncertainty in the count of successes 

compared to the count of failures. Likewise, 
57

any generic industrywide failure rate data 

should include substantial uncertainty due 

to plant-to-plant variability in the number of 

successes.

Some years ago, the former Office for 

Analysis and Evaluation of Operational Data 

(AEOD) published a number of reports

(NUREG/CR-5500, Volumes 1-11) that 

investigated this issue and provided 

industrywide failure rate estimates for a 

variety of equipment. RES should examine 

the impact of the uncertainty in success 

data in their applications and analyses. 

Expanding the previous work of AEOD with 

research to fully incorporate these 

uncertainties may be appropriate.
58

Figure 5. 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
59

12. PROBABILISTIC RISK ASSESSMENT

Background

Probabilistic risk assessment (PRA) is an 

essential technology for the U.S. Nuclear 

Regulatory Commission (NRC). The use of 

risk information in the regulatory process for 

operating reactors is well-established. 

PRAs are being used to inform design 

decisions for new reactors and will be used 

in future regulatory oversight of those 

designs. Risk significance evaluations for 

emergent issues provide an important 

perspective to inform NRC policies, 

priorities, and regulatory decisions. The 

proposed transition to a more fully 

integrated risk management regulatory 

framework will increase reliance on risk 

information throughout all elements of NRC 

decisionmaking. The NRC must have 

state-of-the-art PRA capabilities to support 

these regulatory functions.

In its 2012 report to the Commission on the 

review and evaluation of the NRC Safety 

Research Program (NUREG-1635, Vol. 10), 

the Advisory Committee on Reactor 

Safeguards (ACRS) noted that much of the 

recent research work in this area has 

focused on applications of existing PRA 

models and data to support the reactor 

oversight process for the current reactor 

fleet. Extensions of PRA scope and the 

development of new methods have not 

been priorities. That emphasis continues to 

guide much of the current PRA research 

and planned activities. Technical issues 

and challenges identified during the 

development of a full-scope Level 3 PRA for 

a multi-unit operating reactor site will 

provide an important stimulus and focus for 

needed extensions of risk assessment 

methods and modeling practices.

Current Research Activities

The Division of Risk Analysis of the Office of 

Nuclear Regulatory Research (RES) has 

adopted a goal-oriented framework to 

organize its research programs. The four 

fundamental program goals are:

(1) Support the Reactor Oversight and 

Operating Experience Programs

(2) Remove obstacles to implementation of 

risk-informed regulation

(3) Expand PRA infrastructure to 

encompass new and advanced reactor 

concepts and designs

(4) Support continuous advancement in 

PRA state-of-the-art and 

state-of-practice
60

The ACRS has noted that this type of 

framework “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.” 

The ACRS continues to encourage that 

broad perspective and an emphasis on 

future research needs.

The Division of Risk Analysis of RES has 

developed close working relationships with 

its counterparts in the Office of Nuclear 

Reactor Regulation (NRR) and the Office of 

New Reactors (NRO) to define specific user 

needs and priorities for short-, 

intermediate-, and long-term research 

projects. Memoranda of understanding with 

organizations such as the National 

Aeronautics and Space Administration 

(NASA) and Electric Power Research 

Institute (EPRI) have been used effectively 

to share knowledge and resources. 

Cooperative research agreements and 

grants with universities provide an important 

channel for the development of advanced 

analysis methods. The ACRS supports 

these efforts and encourages further 

collaboration for mutual benefits in areas of 

focused technical needs.

Figure 6 shows the program goals and lists 

the current research topics in each area. 

The following items briefly summarize 

selected research projects that extend the 

scope of current NRC-supported risk 

models and analysis methods.

SPAR Models Extension and Support

Standardized Plant Analysis Risk All 

Hazards Models (SPAR-AHZ): Limited 

models for external initiating events have 

been developed for 20 plants. This total 

includes four more plants since the 2012 

ACRS review of this program. The models 

for 17 plants are based primarily on 

information derived from the Individual Plant 

Examination of External Events (IPEEE) 

analyses. The models for three plants (D.C. 

Cook, Shearon Harris, and Vogtle) are more 

comprehensive. They have been validated 

by comparisons with more recent 

plant-specific information and design 

features, including internal fire analyses that 

have been performed for transition to 

National Fire Protection Association 

(NFPA)-805 at D.C. Cook and Shearon 

Harris. A project is underway to validate the 

SPAR-AHZ models and create new models 

for other plants.

Low Power and Shutdown (LPSD) Models: 

The current scope of the SPAR models 

includes shutdown operating states for only 

eight plants. No extension of the SPAR 

modeling effort for shutdown modes has 

occurred since the 2012 ACRS review of 

this program.

Level 2 Models: Limited-scope Level 2 PRA 

models have been developed for Peach 

Bottom Unit 2, Sequoyah, and Surry Unit 2. 

These models include an evaluation of 

severe accident progression and 

containment functions to support the 

quantification of large early release 

frequencies. The Peach Bottom Level 1 

and Level 2 models have been linked 

through the SAPHIRE software to develop 

an integrated capability model for that plant.

Digital I&C PRA

The digital Loop Operating Control System 

(LOCS) for the Advanced Test Reactor 

(ATR) at the Idaho National Laboratory is 

being used as an example system to focus 

two quantitative software reliability research 

activities.

A Bayesian Belief Network (BBN) 

methodology has been proposed to facilitate 

quantitative evaluation of software reliability. 

The methodology develops causal 

relationships between quality attributes of 

the software development process and 

reliability of the product software. The 

results of the BBN provide a prior estimate 

for the software failure rate, including an 

evaluation of the uncertainty. If testing or 

operational data are available, a Bayesian 

update of the prior can be performed to 
61

produce a posterior distribution for the 

software reliability.

Statistical testing of the Loop Operating 

Control System (LOCS) software will 

provide quantitative measures of the 

software reliability to perform its intended 

functions in the context of the success 

criteria and boundary conditions that apply 

in the pilot model. The test data may be 

used directly to quantify software failure 

modes, or the data may be used to update 

the BBN estimates.

Support for Fukushima Near-Term Task 

Force Initiatives

Two research projects support resolution of 

issues that were identified in the Fukushima 

Near-Term Task Force (NTTF) report 

(SECY 11-0093) and subsequent 

Commission guidance.

RES is supporting the development of 

policies and guidance for an integrated 

regulatory framework that implements the 

risk-informed, performance-based, 

defense-in-depth principles that are 

proposed by the NTTF (Recommendation 1) 

and the Risk Management Task Force 

(NUREG-2150).

In accordance with NTTF 

Recommendation 3, a project is underway 

to perform a feasibility study for evaluating 

the risk from seismically-induced fires and 

floods.

SPAR Models for New Reactor Designs 

This project supports the development of 

SPAR models for each certified new reactor 

design. Models for internal events that 

occur during full power operation have been 

completed for the AP1000, US-APWR

(advanced pressurized-water reactor), 

US EPR (evolutionary power reactor), and 

both versions of the advanced boiling-water 

reactor (ABWR) (GE and Toshiba). A 

SPAR model has not yet been developed 

for the economic simplified boiling-water 

reactor (ESBWR) design. The AP1000 

models have been extended to include 

seismic events, fires, and internal flooding. 

Efforts are currently focused on 

development of a shutdown model for the 

AP1000.

Support for Advanced Light Water Reactor 

Reviews 

Current research activities for advanced 

reactor designs are limited to the 

development of guidance for the technical 

adequacy of PRAs that will be performed to 

support design-specific licensing reviews for 

small modular reactors. A pilot study to 

apply the risk-informed licensing framework 

proposed in NUREG-1860, “Feasibility 

Study for a Risk-Informed and 

Performance-Based Regulatory Structure 

for Future Plant Licensing, Volumes 1 and 

2,” for a specific advanced reactor design 

has been cancelled.

Level 3 PRA Project

This project involves the development of a 

full-scope comprehensive Level 3 PRA for 

an operating plant site. The PRA scope 

covers all plant operating modes, spent fuel 

storage pools, and onsite dry cask storage 

facilities. The analyses will include internal 

and external hazards. Site-level risk from 

accidents that involve multiple radiological 

sources will also be evaluated (i.e., the 

reactor units, spent fuel pools, and dry 

storage casks).

The project is benefitting substantially from 

voluntary support by Southern Nuclear 

Operating Company and an existing 

memorandum of understanding with EPRI. 

The PRA is being performed for Vogtle 

Electric Generating Plant Units 1 and 2. 

Available plant-specific models and data are 

being used as a basis for development of 

full-scope NRC models. EPRI is providing 

additional input and reviews through its 

participation in the project Technical 

Advisory Group.
62

This project is a major research effort. It 

accounts for a considerable reorganization 

of PRA research priorities since the 2012 

ACRS review of this program. The project 

will advance the state of knowledge about 

plant-level and site-level risk beyond the 

landmark NUREG-1150 study, “Severe 

Accident Risks: An Assessment for Five 

U.S. Nuclear Power Plants,” performed 

more than 20 years ago. It will also 

integrate substantial advancements in 

understanding and modeling of internal 

hazards, external hazards, severe accident 

phenomena, containment performance, and 

offsite consequences. The project intends 

to use current state-of-the-practice methods 

and modeling techniques, with limited 

extension to new research. However, this 

project will identify needs for future research 

efforts to refine the treatment (i.e., reduce 

the uncertainty) of issues that are currently 

evaluated with rudimentary models or 

excessively conservative assumptions.

Assessment and Recommendations

The program structure in Figure 6 facilitates 

examination of the fundamental goals of 

each project and the balance among 

research priorities. For example, the PRA 

research program managers indicated that 

approximately 40 percent of all resources 

are currently allocated to projects in support 

of goal (1), approximately 30 percent are 

allocated to goal (2), approximately 5

percent are allocated to goal (3), and 

approximately 25 percent are allocated to 

goal (4)2

. These allocations are more 

balanced among the four fundamental 

research goals than was the case when the 

2012 ACRS review was conducted.

The current programs under goal (1) are 

influenced primarily by a strong RES 

emphasis on responsiveness to immediate 

user needs and short-term requirements to 

 2 These estimates are very approximate. They are 

noted here only to illustrate broad relative allocations 

for the purposes of this report.

support existing PRA model infrastructure 

for the Reactor Oversight Process. 

Maintenance and updating of SPAR 

models, extension of those models to 

include a consistent level of detail for each 

operating reactor, comparisons with industry 

PRA models, and compilation of operating 

experience data are important tasks to 

support risk-informed decisionmaking and 

balanced reactor oversight. Those support 

activities for existing SPAR models do not 

typically require extension of fundamental 

analytical knowledge, methods, or modeling 

techniques beyond the current capabilities 

of trained PRA practitioners. 

In its 2012 review of this program area, the 

ACRS recommended that RES should 

examine the feasibility for increased sharing 

of SPAR model maintenance and support 

activities among the regional offices, 

headquarters staff, and contractors, with 

increased technical responsibilities 

allocated to risk engineers in the regional 

offices. The ACRS continues to encourage 

that diversification of PRA knowledge and 

modeling capabilities, allowing RES to focus 

more effectively on advancement of 

state-of-the-art risk assessment methods 

and practices.

The current programs under goal (2) 

address a variety of intermediate- to 

long-term research needs. Resources are 

allocated about equally between the 

development of methods and modeling 

capabilities for focused issues (e.g., digital 

instrumentation and control (I&C) systems 

and software, seismically-induced fires and 

floods) and support for broader regulatory 

initiatives (e.g., a Risk Management 

Regulatory Framework, PRA standards and 

technical guidance). The framework in 

Figure 6 provides a structure for examining 

the overall goals and direction of the PRA 

research program. Therefore, for a clear 

understanding of RES priorities, it is 

important to assign specific projects to the 

most appropriate category. There is a 

conceptual overlap between some of the 

projects in this area and the objectives for 
63

goal (4). For example, the projects for 

evaluating digital I&C systems and 

seismically-induced fires and floods may be 

more appropriately associated with 

extension of current state-of-the-practice 

methods and modeling techniques. The 

ACRS supports those projects and agrees 

that they are important for improved 

understanding of these difficult issues. 

However, the ACRS recommends the 

original intent of each element in the 

goal-oriented framework should be 

examined carefully to ensure that the 

distribution of projects and resources 

accurately reflects the overarching RES 

priorities.

The current programs under goal (3) have 

been influenced very significantly by the 

compound effects from limited RES 

resources and a slowdown in the 

development and certification of new reactor 

designs. These pragmatic considerations 

have affected several near-term research 

priorities. Cancellation of the planned pilot 

study of a risk-informed licensing framework 

for advanced reactors may be an ill-advised 

decision. Performance of early pilot studies 

for proposed methods and regulatory 

practices provides valuable experience and 

feedback to focus research for necessary 

refinements. Small modular reactor designs 

that include several units at a single site 

with common external infrastructure 

introduce additional regulatory challenges 

that must be addressed to understand and 

manage the risk from potential single-unit, 

multiple-unit, and site-level accidents. 

Analytical capabilities, modeling techniques, 

and regulatory tools to support a 

risk-informed licensing process for those 

reactors must be developed and tested well 

in advance of the first applications. The 

ACRS recommends that RES should 

pursue a pilot study to apply the 

risk-informed licensing framework proposed 

in NUREG-1860 for a specific new small 

modular reactor design. The study should 

address site-level issues that affect accident 

progression, mitigation options, emergency 

planning, and offsite consequences. It 

should be coordinated with the availability of 

information about the selected reactor 

design and with evolving activities to 

support development of an integrated Risk 

Management Regulatory Framework.

Almost all of the programs under goal (4)

support development of the Level 3 PRA. 

The ACRS continues its strong support for 

this project. It will provide an integrated risk 

model context to define the scope and 

requirements for additional 

intermediate- and long-term research. 

Some of the difficult technical issues are 

known and are being addressed by current 

projects, but at a relatively low level of 

effort. Other specific challenges will emerge 

as the Level 3 PRA models are assembled 

and analysts discover inevitable unexpected 

problems. Consistent assessment of the 

site-specific risks from internal and external 

hazards; accidents that affect stored spent 

fuel; events that occur during plant 

shutdown; and integration of the Level 1, 

Level 2, and Level 3 PRA models will also 

address issues that have been identified in 

the wake of the accidents at the Fukushima 

Dai-ichi site. The ACRS recommends that 

RES should use evolving knowledge and 

insights from the Level 3 PRA project to 

risk-inform initiatives for specific research 

that may be proposed as a direct 

consequence of NRC and industry 

responses to the Fukushima Dai-ichi 

accidents.

RES has completed an important update to 

NUREG-1855, “Guidance on the Treatment 

of Uncertainties Associated with PRAs in 

Risk-Informed Decision Making,” which 

contains practical guidance on methods to 

systematically identify, document, and 

quantify sources of uncertainty. However, a 

disciplined assessment of uncertainty is not 

yet applied consistently throughout the 

agency. The ACRS recommends that RES 

should initiate efforts to ensure that an 

appropriate characterization of uncertainty 

is performed in all agency analyses. Explicit 

acknowledgement, quantification, and 

communication of uncertainties will be a key 
64

element for rational understanding of public 

health risk and its contributors as PRA 

results are expanded to integrate the risks 

from hazards such as fires, floods, seismic 

events, and other extreme natural 

phenomena. A clear understanding of the 

uncertainties and their sources is also 

essential for risk-informed regulatory 

decisionmaking and rational selection 

among possible risk management options.

Substantial efforts have been made to 

improve security against physical and cyber 

attacks on our operating nuclear power 

plants. New plants will further integrate 

security protections from the initial stages of 

their designs. Despite these efforts, 

concerns remain that the applied security 

controls may not be allocated optimally to 

cope with the full spectrum of potential 

threats. The ACRS recommends that the 

fundamental risk assessment framework 

and analysis techniques should be applied 

to address these concerns and to assess 

proposed protection strategies for future 

evolving threats. An integrated risk 

assessment process can also provide a 

perspective that more fully addresses 

issues that affect both reactor safety and 

physical security. RES should initiate a 

research project and pilot applications to 

examine the use of risk assessment 

methods to inform security programs and 

practices.

The ACRS encourages active collaboration 

among RES, other Government agencies, 

industry research organizations such as

EPRI, and universities to effectively share 

knowledge and experience that contribute to 

intermediate- and long-term research 

objectives. These collaborative programs 

are often the first to suffer in times of 

declining resources. Although modern risk 

assessment is generally characterized as a 

“mature” science, further research is 

needed to develop practical methods, 

models, and data for realistic evaluation of 

several difficult issues. Collaborative 

programs and tailored research projects 

should remain an integral part of the RES 

initiatives to address these issues.

During 2013, decisions were made to 

permanently shut down four U.S. reactors. 

Others may follow over the next few years. 

These actions raise questions about 

whether the NRC has fully addressed the 

risks from decommissioning activities. The 

Nation continues to debate options for 

disposal of used nuclear fuel and high level 

wastes. Recently proposed revisions to 

Title 10 of the Code of Federal Regulations

(10 CFR) Part 61, “Licensing Requirements 

for Land Disposal of Radioactive Waste,”

have focused attention on the scope of 

performance assessments for sites that 

provide long-term storage of low level 

wastes. The ACRS recommends that RES 

should initiate a project to examine how 

quantitative risk information can be used to 

inform regulatory decisions in these 

interrelated disciplines. These activities do 

not align directly with research programs 

that are intended to support operating 

reactors, new reactors, or advancement of 

risk assessment methods. Therefore, RES 

support for reactor decommissioning and 

waste disposal activities may represent a 

fifth goal in the general framework of 

Figure 6.
65

Figure 6. Current NRC Research Activities in PRA

Probabilistic Risk Assessment 

Research Goals

Independent Capability to Provide Quantitative Assessment of Risk for 

Regulatory Activities

(An essential technology for NRC as it integrates the use of risk information into 

the regulatory process)

(1) Support the Reactor 

Oversight and Operating 

Experience Programs

• Standardized Plant 

Analysis Risk (SPAR) 

Models Extension and 

Support

• SAPHIRE Support and 

Development

• Accident Sequence 

Precursor (ASP) 

Program and 

Regulatory Issues 

Support

• Digital I&C PRA

• PRA Standards 

Committees and 

Regulatory Guidance

• Risk Management 

Regulatory Framework

• Seismically-Induced 

Fires and Floods

(3) Expand PRA 

Infrastructure to 

Encompass Advanced 

Reactors Designs

(4) Support Continuous 

Advancement in PRA 

State-of-the-art and Stateof-practice

• Level 3 PRA Project

• Automated Reliability 

Prediction System

• Advanced Knowledge 

Engineering Tools

• Integrated SPAR Model 

Development

(2) Remove Obstacles to 

Implementation of Riskinformed Regulation

• New Reactor SPAR 

Models

• Support Advanced 

Light Water Reactor 

Reviews

67

13. RADIATION PROTECTION

Background

The U.S. Nuclear Regulatory Commission 

(NRC) continues to maintain a program of 

research related to radiation protection in 

the areas of:

• risks from radiation

• the sciences of internal and external 

dosimetry

• 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.

Additionally, the NRC continues to 

participate in international standards setting 

efforts to exchange technical information to 

the benefit of all participating organizations. 

Current Research Activities

The current NRC research activities in the 

area of Radiation Protection are depicted in 

Figure 7. The research is focused on 

development and maintenance of health 

effect/dose calculation tools, emerging 

health effects and dosimetry research, and 

participation in a number of national and 

international collaborative radiation 

protection activities. There are also some 

efforts on preparation of exposure and 

abnormal occurrence reports. These are all 

essential activities and need to be 

sustained.

Biokinetic Model

RES is working closely with other Federal agencies to 

support and monitor the work of Oak Ridge National 

Laboratory (ORNL) on development of biokinetic and 

dosimetric models and dose coefficients for 

occupational and public exposure to radionuclides 

that are based on International Commission on 

Radiological Protection (ICRP) Publication 103 

recommendations. 

DEVELOPMENT AND MAINTENANCE OF 

HEALTH EFFECTS/DOSE CALCULATION 

TOOLS

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 by adding the 
68

ability to model three dimensional sources 

(cylinders, spheres, and slabs) with 

materials placed between the source and 

skin (including air gaps that attenuate the 

beta particles). In addition, the code 

incorporated a user interface that greatly 

simplified data entry and increased 

efficiency in calculating skin dose. 

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 

overestimated the dose with increasing 

photon energy. RES recently completed 

further enhancement of the code, including 

replacement for the code’s photon dose 

algorithm. 

The current version of the VARSKIN code 

does not accurately predict beta dose from 

some radionuclides for some exposure 

conditions. The accuracy of the code 

decreases with skin depth. In order to 

address these issues, RES is currently 

sponsoring a research project to replace the 

current beta dose model in order to enable 

the code to more accurately model beta 

dosimetry resulting from contamination on 

the skin or on protective clothing covering 

the skin. The secondary objective of this 

project is to further enhance the code’s 

functionality. RES also plans to develop a 

training module for using the code.

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. 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 7 environment.

RADTRAD 4.0: The RADionuclide 

Transport and Removal And Dose 

Estimation (RADTRAD) code uses a 

combination of tables and numerical 

models, based on simplified source term 

parameters, to determine the 

time-dependent dose at user-specified 

locations for a given accident scenario. It 

also provides the inventory, decay chain, 

and dose conversion factor tables needed 

for the dose calculation. The RADTRAD 

code can be used to assess occupational 

radiation exposures (typically in the control 

room); to estimate site boundary doses; and 

to estimate dose attenuation due to 

modification of a facility or accident 

sequence. RADTRAD was rewritten in 

2009 as a plug-in to the Symbolic Nuclear 

Analysis Package (SNAP). SNAP removes 

the need for analysts to use the text-based 

entry methods by providing a powerful, 

flexible, and easy-to-use graphical user 

interface (GUI). 

RES is currently sponsoring a research 

project to add more features to the 

RADTRAD plug-in to SNAP before the final 

release of RADTRAD 4.0. This project will 

also provide support for development of the 

combined RADTRAD 4.0 user manual, 

theory manual, and verification & validation 

(V&V) documentation as a NUREG/CR 

report. 

REIRS: The REIRS database is a 

comprehensive system used to compile and 

analyze occupational radiation exposure 

reports, which licensees submit on an 

annual basis as required by 

10 CFR 20.2206, “Reports of Individual 

Monitoring.” This information is used to 

produce the annual publication 

NUREG-0713, “Occupational Radiation 
69

Exposure at Commercial Nuclear Power 

Reactors and Other Facilities.”

RES is currently sponsoring a research 

project on expanding the current REIRS

database to include additional NRC and 

Agreement State licensee information and 

conducting analyses of international 

occupational radiation exposure databases.

PARTICIPATION IN NATIONAL AND 

INTERNATIONAL RADIATION 

PROTECTION ACTIVITIES

RES is actively engaged in monitoring and 

participating in a number of national and 

international committees (e.g., the National 

Council on Radiation Protection and 

Measurements, the National Academy of 

Sciences, the International Commission on 

Radiological Protection, and the NEA 

Committee on Radiation and Public Health). 

NRC research is also well leveraged by 

working with a number of collaborating 

agencies, including the U.S. Environmental 

Protection Agency. Such activities promote 

consistency and coherence in regulatory 

applications of radiation protection and 

health effects research among NRC 

programs, as well as those of other Federal 

and State regulatory agencies.

EMERGING HEALTH EFFECTS AND 

DOSIMETRY RESEARCH

Radiation Worker Health Studies: The NRC 

has entered into an interagency agreement 

with the U.S. Department of Energy’s 

(DOE’s) Office of Science in a Low Dose 

Radiation Research Program to study the 

health effects of more than 1 million 

radiation and atomic veterans. The One 

Million U.S. Radiation Worker study is 

designed to provide information on risk 

following low dose rate exposures. The 

significance of this research is considerable 

because it applies directly to existing 

concerns about and standards for chronic 

radiation exposure. The collaborating or 

cooperating agencies include DOE, U.S. 

Department of Defense, U.S. Department of 

Veteran Affairs, National Cancer Institute, 

and National Aeronautics and Space 

Administration.

Research began on the nuclear power 

worker cohorts in fall 2012. The NRC 

expects to see the results in late 2014. The 

results of this study will provide valuable 

information for future radiation protection 

standards-setting bodies and any resultant 

occupational radiation dose standards.

 

Analysis of Cancer Risks in Populations

Living Near NRC-Licensed Nuclear 

Facilities: Efforts are underway to have the 

U.S. National Academy of Sciences conduct 

a study analyzing the cancer risk of 

populations living near NRC-licensed 

facilities, including power reactors and fuel 

cycle facilities (e.g., fuel enrichment and 

fabrication plants). This study will update 

and expand on the 1990 U.S. National 

Cancer Institute report, “Cancer in 

Populations Living near Nuclear Facilities.” 

The staff plans to use this report as a 

scientifically defensible resource to aid in 

addressing continued stakeholder concerns 

about perceived elevated cancer rates in 

populations near reactors, including cancer 

incidence (i.e., being diagnosed with cancer 

but not necessarily dying from the disease).

The NRC and NAS agreed on a two-phase 

approach: (1) preparation of a scoping 

study to determine the best methodology, 

the best approach, and the potential 

limitations for performing the cancer 

incidence and mortality epidemiology study 

and (2) conduct of the actual study. 

The NAS Phase 1 committee completed 

their report in May 2012 and recommended 

two alternative approaches for assessing 

cancer risks. The committee also 

recommended a pilot study of seven nuclear 

facilities to assess whether the approaches 

could work on a larger scale. The NRC has 

engaged with the NAS to perform the 

Phase 1 recommendations and expects the 

pilot studies to be completed in 2015.
70

Assessment and Recommendations

The staff has developed an appropriate and 

robust research program in the areas of 

radiation protection noted above. This 

program includes radiation protection of 

workers and the public and radiological 

assessments related to radiation exposure 

and health risks around NRC-licensed 

nuclear facilities.

The ACRS is supportive of RES’s 

participation in national and international 

efforts relating to radiation protection. 

These collaborations provide excellent 

opportunities for the NRC staff to benefit 

from the work of other organizations in the 

United States and around the world.
71

Figure 7. Current NRC Research Activities in Radiation Protection

Radiation Protection

Research

• Analysis of Cancer 

Risks in Populations

Living Near NRCLicensed Nuclear 

Facilities

• Radiation Worker 

Health Studies

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

Participation in 

National and 

International Radiation 

Protection Activities 

Emerging Health Effects 

and Dosimetry Research 

• VARSKIN Skin 

Computer Code 

• Radiological 

Toolbox 

• Radionuclide

Transport, 

Removal, and Dose 

(RADTRAD)

• Radiation Exposure 

Information and 

Reporting System 

(REIRS)

• International Commission on 

Radiological Protection, ICRP

• National Council on Radiation 

Protection and Measurements, 

NCRP

• NEA Committee on Radiation 

Protection and Public Health, 

CRPPH

• Institute De Radioprotection Et 

De Surete Nucleaire, IRSN

• Joint Coordinating Committee for 

Radiation Effects Research, 

JCCRER (a bilateral government 

committee representing agencies 

from the United States and the 

Russian Federation)

73

14. NUCLEAR MATERIALS AND WASTE

Background

The NRC continues to maintain a program 

of research related to nuclear materials and 

radioactive waste topics associated with 

licensing, facility siting, facility 

environmental performance, and the 

decontamination and decommissioning of 

licensed facilities.

Current Research Activities

Research activities in the general areas of 

nuclear materials and waste reflect the 

current hiatus in the resolution of policy 

issues. By in large, the current activities 

address issues encountered or anticipated 

in the inspection and monitoring activities.

Among such current activities are:

Stress Corrosion Cracking Analysis of Dry 

Cask Storage Materials: Evidence of dust 

deposition on surfaces of dry cask storage 

containers has been found. In locations 

near coasts the deposited material includes 

chlorides that could sustain stress corrosion 

cracking of the cask material should they 

hydrate in humid environments. The 

research attempts to discover the stress 

corrosion cracking potential of these 

deposits.

Spent Fuel Cask Thermal Testing Phase 2:

The current efforts focus on the vulnerability 

of elastomeric materials used for sealing 

storage casks to sustained periods of 

modestly elevated temperatures. Results of 

this work could also be pertinent to issues of 

head seals for the Mark I and Mark II 

boiling-water reactors.

Analysis of Vacuum Drying of Spent 

Nuclear Fuel Storage Canisters: It is 

difficult to remove liquid water from 

interstitial locations throughout fuel 

assemblies and the like. Residual water in

storage casks will react with cladding on the 

fuel and other metals to form hydrogen that 

can further embrittle the fuel cladding. This 

research is directed at determining how 

vacuum drying can be better used to 

remove residual water from storage 

containers.

Dry Storage System Design

Hydride Reorientation Effects on Cladding 

Fracture: Zirconium alloy fuel cladding 

absorbs hydrogen during normal fuel 

operation. The absorbed hydrogen 

precipitates as zirconium hydrides when the 

fuel is cooled below normal fuel operating 

temperatures. The precipitated hydrides 

can affect the fracture properties of the 

cladding during storage.

Assessment and Recommendations

The research activities listed above have 

been prompted by current regulatory and 

monitoring efforts and appear to be useful 

expenditures of NRC resources. 

Electrochemical separation has been 

suggested as a means to facilitate the safe 

disposal of waste. The NRC staff has a 

research program to better understand the 
74

possible separations processes and 

hazards that could arise in the possible use 

of electrochemical preprocessing of waste 

for disposal. This is of necessity a very 

exploratory effort. It would be useful for the 

staff to have access to the ASPEN code for 

the exploration of flow sheet alternatives 

and associated hazards. Very detailed work 

in this area should await more definitive 

proposals from potential licensees.
75

15. SEISMIC AND STRUCTURAL ENGINEERING

Background

Nuclear power plants are designed to cope 

with large seismic events – up to at least the 

safe shutdown earthquake (SSE).

Earthquakes of greater magnitude have the 

potential to damage the power plant and 

physical barriers to radionuclide release.

Earthquake damage can be a common 

cause mechanism for the failure of systems 

to prevent accidents and systems to 

mitigate accidents – extending even to the 

final line of defense-in-depth of public 

evacuation and emergency planning.

The massive damage to the Fukushima 

Dai-ichi nuclear complex in Japan following 

a severe earthquake and tsunami in 

March 2011 has highlighted the importance 

of seismic and structural engineering to the 

safety of U.S. nuclear power plants. Less 

well known, but also significant, was a 

severe earthquake in 2007 at the 

Kashiwazaki-Kariwa plant in Japan. The 

latter event was even more severe in terms 

of ground acceleration at the plant site, 

although less damage to the plant occurred. 

In both instances, the earthquakes 

exceeded the maximum design basis 

earthquakes to which the plants were 

designed, thus calling into question the 

probabilistic seismic hazard assessments 

used to develop site specific seismic 

hazards at those plants. Research was 

already underway by the U.S. NRC, in 

conjunction with U.S. Geological Survey 

(USGS) and National Institute of Standards 

and Technology (NIST), to develop 

improved seismic hazard assessments at 

nuclear plants in the central and eastern 

United States (CEUS), and this work is 

enhanced and supplemented by the current 

NRC seismic and structural research plan. 

The seismic research is supplemented by 

research in the areas of improved tsunami 

modeling, integrated tsunami warning for 

U.S. coastlines, and evaluation of tsunami

impact on new and existing power plants. 

Structural Engineering research is also 

being performed in the areas of concrete 

and steel containment vessel design and 

degradation, irradiation effects on concrete, 

Alkali-Silica Reaction degradation of 

concrete structures, and modeling of 

transportation of spent nuclear fuel. The 

Office of Nuclear Regulatory Research 

(RES) is participating in several 

international cooperative efforts directed at 

review and assessment of the Fukushima 

and Kashiwazaki-Kariwa events and 

improved understanding of earthquake and 

tsunami hazards. These programs will 

support user needs in the Office of Nuclear 

Reactor Regulation (NRR), Office of New 

Reactors (NRO), and Office of Nuclear 

Material Safety and Safeguards (NMSS). 

Current Research Activities

Tsunami Hazard

Several research projects are underway 

directed at better understanding and 

definition of the Tsunami Hazard at U.S.

nuclear power plants. Under one such 

project, “Tsunami Hazard Modeling and 

Integrated Tsunami Warning for US 

Coastlines,” NRC staff will be trained in the 

use of existing National Oceanic and 

Atmospheric Administration (NOAA)

tsunami warning and modeling tools. A 

second project, “Tsunami Landslide Source 

Probability and Potential Impact on New 

and Existing Power Plants,” will develop 

probabilistic methods to evaluate 

landslide-based tsunami sources. These 

projects together will support a new 

regulatory guide on tsunami hazard 

assessment. Finally, the staff will 

participate in an International Seismic 

Safety Centre Tsunami/Seismic Project that 

will provide a mechanism for sharing 

international experience on the impact of 
76

flooding on nuclear power plants, including 

Fukushima.

Seismic Hazard Assessment

There are currently several projects 

underway to improve assessment of the 

Seismic Hazard at new and existing nuclear 

power plants. The development of a new 

seismic source characterization (SSC) 

model for the central and eastern United 

States (CEUS) has been completed through 

the joint sponsorship of the NRC, 

Department of Energy (DOE), and the 

Electric Power Research Institute (EPRI). 

The existing, validated PSHA (Probabilistic 

Seismic Hazard Analysis) software will be 

modified to add capability to perform 

seismic hazard calculations using the 

CEUS-SSC model and the EPRI ground 

motion relationships, and a related project 

will further refine and inform the 

uncertainties in these models. The NRC 

continues to sponsor the Next Attenuation 

Relationship for the Central and Eastern 

North America (NGA-East) project. The 

new ground motion characterization (GMC) 

model resulting from this project will replace 

the EPRI GMC model currently used for 

new nuclear power plants in the CEUS. 

The NGA-East Project is being sponsored 

cooperatively by the NRC, DOE, EPRI, and 

USGS. This project is expected to end in 

2015.

Other seismic research projects will study 

paleoliquefaction as an aid in predicting 

large, but rare, earthquakes in the CEUS 

region; evaluate current state of seismic 

monitoring in the CEUS, improve the staff’s 

site response capability; develop a 

two-dimensional non-linear soil response 

analysis capability as well as a 

multidimensional soil constitutive model; 

and support reviews of Fukushima 

Near-Term Task Force (NTTF) 2.1 

submittals.

Seismic Hazard Assessment for the Central and Eastern United States

The NRC continues to sponsor projects in support of an updated assessment of seismic hazard 

in the central and eastern Unites States (CEUS). These projects will result in an advanced 

regional model for probabilistic seismic hazard analysis (PSHA) for critical facilities in CEUS
77

Geotechnical Seismic Engineering

Three research projects are underway (or 

have recently been completed) in the area 

of geotechnical seismic engineering. 

Recently completed projects include 

engineering evaluation of post-liquefaction 

residual strength, and an evaluation of the 

applicability of current structural, 

geotechnical, and seismic regulations to 

advanced design Small Modular Reactors 

on individual foundations. Projects currently 

underway include an evaluation of 

numerical tools for assessing soil-structure 

interaction (SSI) of deeply embedded 

nuclear power plant structures and a 

computational platform for addressing SSI 

under non-traditional seismic input loads.

Structural Seismic Engineering

Structural seismic engineering refers to 

analyzing the response of structures once 

the seismic ground motions are defined. 

NRC RES projects in this field include 

completion of a lessons learned report on 

the Kashiwazaki-Kariwa earthquake with 

specific emphasis on issues that may 

impact U.S. plants; analysis of two key 

issues identified in base-isolated structures; 

research in support of walkdowns of 

operating U.S. facilities for seismic and 

flood hazards; collaboration with the Japan 

Nuclear Engineering Society on seismic 

issues; and improvement of the staff’s ability 

to estimate failure correlations for 

structures, systems, and components in 

seismic probabilistic risk analyses (PRAs).

Structural Engineering and Materials

Several instances of containment 

degradation have been observed in U.S.

nuclear power plants. RES is pursuing 

several projects to provide better tools for 

addressing such degradation, which 

includes corrosion of steel liner or shell, loss 

of prestress, degradation due to repairs 

associated with steam generator 

replacement, and Alkali Silica Reaction 

(ASR) degradation of concrete structures. 

Several new and ongoing research projects 

are directed at better understanding these 

mechanisms and modeling their effects on 

service life of the structures. In addition, 

research is underway to provide tools to 

better address new design concepts 

proposed for safety-related steel-concrete 

structures and the vulnerability of such 

structures to missile impacts and other 

threats.

Structural Analysis

Finally, research is underway to help 

maintain the NRC’s state-of-art capability in 

nonlinear structural analysis to assess such 

topics as beyond-design-basis accidents 

and security threats. RES is participating in 

international cooperative research on 

impact testing, as well as analysis of drop 

tests of spent nuclear fuel transportation 

casks. 

Assessment and Recommendations

The NRC seismic and structural research 

program in support of regulatory activities is 

being conducted under a well-developed 

research plan that has been broadly 

reviewed for both technical quality and 

programmatic elements. The program 

funds state-of-the-art work via contracts to 

several renowned organizations in the field, 

including national labs and universities, as 

well as international cooperative programs 

and collaborative research with other 

governmental agencies such as DOE. The 

ACRS believes that this program will 

adequately support the NRC staff’s 

capabilities to evaluate potential risks to 

U.S. nuclear plants due to events such as 

the severe earthquakes and tsunami that 

impacted Japanese reactors and to assure 

the continued safety of new and operating 

U.S. nuclear power plants.

79

16. SEVERE ACCIDENTS AND SOURCE TERM

Background

The U.S. Nuclear Regulatory Commission 

(NRC) uses its severe accident expertise 

and analysis capabilities to support 

regulatory decisions for operating nuclear 

power plants and for certifying new and 

advanced reactor designs. Severe accident 

analysis tools also help the staff in its 

transition to a more risk-informed regulatory 

framework. In addition, severe accident 

analysis tools are used for better 

understanding of the progression and 

potential radiological releases due to such 

accidents.

The Office of Nuclear Regulatory Research 

(RES) long-term severe accident and 

containment response evaluation 

development plan focuses on two areas: 

maintenance and development of severe 

accident analysis tools and continued 

collaboration in international experimental 

research programs. The NRC made 

considerable investment to achieve the 

current levels of severe accident 

understanding. However, recent NRC 

investments have been limited to a level 

that only allows for continuation of required 

analysis and risk-informed activities. 

Current Research Activities

Figure 8 illustrates current NRC severe 

accident and source term research 

activities. Current activities are focused on 

development and usage of the MELCOR 

code and other severe accident analysis 

tools and on participation in collaborative 

severe accident research programs. There 

are also some efforts underway on severe 

accident knowledge management.

The NRC leverages its resources by relying 

on international collaborative research 

programs in the Pacific Rim (e.g., Japan 

and Korea) and Europe (e.g., France,

Switzerland, and Germany). Knowledge 

gained from the NRC’s past experimental 

work and from recent international 

experimental programs are systematically 

incorporated into MELCOR. 

MELCOR Investigations of Effectiveness of 

Filtered Vents

In developing recommendations related to 

filtered containment venting, NRC completed 

MELCOR evaluations for a BWR Mark I plant to 

assess the effectiveness of filtered venting on 

cesium release for cases with various mitigating 

options, such as operation of Reactor Core 

Isolation Cooling System (RCIC), wetwell 

venting (vent), core sprays (CS), and Drywell 

sprays (DW spray). 
80

Organized by the NRC, the international 

Cooperative Severe Accident Research 

Program (CSARP) is an annual forum for 

exchanging severe accident research 

findings. Twenty seven foreign countries 

currently participate in CSARP. One 

significant outcome of this effort is the 

adoption of the MELCOR code by most of 

the other countries and institutions as an 

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 accident 

progression in current nuclear power plants, 

new and advanced reactor designs, and 

some nonreactor systems such as spent 

fuel pools. Since the accident at Three Mile 

Island in 1979, significant advances have 

been made related to severe accident 

phenomena as a result of extensive 

domestic and international research.

Updates to its models have allowed the 

MELCOR code to become the repository of 

this improved understanding. Efforts to 

enhance the Symbolic Nuclear Analysis 

Package (SNAP) plug-in for MELCOR have 

also continued, with improvements to 

streamline coupling between SNAP and 

MELCOR.

The MELCOR code, along with the 

MELCOR Accident Consequence Analysis 

Code (MACCS2) for modeling atmospheric 

dispersion and radiological consequence 

calculations, were extensively used in the 

State of the Art Reactor Consequence 

Assessment (SOARCA) project. Several 

recent modifications enhanced MELCOR 

capabilities for the SOARCA effort. These 

include (1) new default parameters, either 

as input records or sensitivity coefficients, 

that best capture experimental observations 

and improve code numerical robustness, (2) 

addition of a model to simulate the 

thermo-mechanical collapse of fuel rods 

with highly oxidized cladding at high 

temperatures, and (3) enhancement of the 

SPARC pool scrubbing model to treat 

fission product vapors. As part of the 

SOARCA project, a draft of 

NUREG/CR-7008, “MELCOR Best 

Modeling Practices,” has been developed to 

provide users guidance in selecting code 

options.

MELCOR and MACCS2 were used 

extensively to perform a large number of 

calculations in support of the response to 

the Fukushima Near-Term Task Force

(NTTF) recommendation on containment 

vent filtration. It is anticipated that 

MELCOR calculations will also be used in 

support of implementing other NTTF 

recommendations.

The spent fuel pool (SFP) models in 

MELCOR have been developed over the 

past 10 years supported by experimental 

data. MELCOR was used in the zirconium 

fire experiments for boiling-water reactor 

(BWR) assemblies, and the predictions 

showed good agreement with data for the 

initiation and propagation of zirconium fire 

(NUREG/CR-7143, “Characterization of 

Thermal-Hydraulic and Ignition Phenomena 

in Prototypic, Full-Length Boiling Water 

Reactor Spent Fuel Pool Assemblies After a 

Postulated Complete Loss-of-Coolant 

Accident”). The code was also used for the 

recent Spent Fuel Pool Study (SFPS).

To support probabilistic risk assessment 

(PRA) and severe accident analyses for 

NRC design certifications, MELCOR 

analyses were performed for several new 

reactor designs (e.g., EPR (evolutionary 

power reactor), ABWR (advanced 

boiling-water reactor), ESBWR (economic 

simplified boiling-water reactor), US-APWR

(advanced pressurized-water reactor), and 

AP1000). In addition to these new LWR 

(light-water reactor) designs, NRC will use 

MELCOR to provide confirmatory analysis 

of small modular reactors, including the 

iPWRs (integral pressurized-water reactors)

that will be submitted to the NRC for 

licensing review. MELCOR models were 

developed for two iPWRs (e.g., the NuScale 

and mPower plants), and some preliminary 

severe accident simulations were completed 
81

to support NRC reviews of iPWR design 

certifications. 

MELCOR analysis also supports projects 

with the Division of Engineering and 

Division of Risk Analysis to estimate the risk 

from consequential steam generator tube 

ruptures in Combustion Engineering (CE) 

plants. The MELCOR model derives part of 

its input from FLUENT analyses. The 

output is combined with that from FLUENT 

to provide a time-dependent spatial 

distribution of temperature in the steam 

generator tubes for use in calculating 

failure. The MELCOR analyses also 

provide fission product release estimates.

Efforts are underway to document results 

from this effort.

Various workshops and user training 

programs are conducted each year to 

support MELCOR users. For example, the 

MELCOR Code Assessment Program 

(MCAP) is an annual technical review 

meeting that focuses on MELCOR code 

development and assessment, providing a 

forum for users to present and discuss their 

experiences. In addition, several 

organizations conduct annual European 

MELCOR User Group (EMUG) meetings.

Collaborative Severe Accident Experimental 

and Analyses Research Programs 

The NRC participates in several 

collaborative experimental severe accident 

research programs, which continue to 

provide key data for MELCOR model 

development and assessment. Table 1 

summarizes the objectives and scope of 

current and recently completed international 

collaborative severe accident research 

programs. NRC participation in an 

international program is evaluated using 

established criteria to ensure that the 

expected value of such collaboration to the 

NRC is well worth the investment. 

Data from these experimental collaborations 

have the potential to significantly impact 

severe accident assessments, including: 

the release of iodine, cesium, and 

molybdenum based on PHÉBUS-FP and 

VERDON tests; containment iodine 

behavior in the PHÉBUS-FP, BIP2, and 

STEM-EPICUR tests; ruthenium release 

from fuel based on data from the VERDON 

and START tests; debris coolability based 

on data from the QUENCH-DEBRIS tests; 

melt spreading and core-concrete 

interaction phenomena based on results 

from the MACE and OECD-MCCI 

(Organization for Economic Cooperation 

and Development—molten core concrete 

interaction) tests; containment spray 

effectiveness based on results from several 

SARNET programs; and fuel-coolant 

interaction phenomena based on tests 

conducted in the FARO, KROTOS, and 

TROI facilities. MELCOR and other severe 

accident analysis code models have been 

revised to reflect these new results, and 

analyses completed with updated models 

provide essential input for decisions related 

to plant siting, equipment qualification, 

control room habitability, and emergency 

planning. For example, MELCOR models 

with updates to reflect recent findings from 

PHÉBUS-FP are currently being used to 

revise the Alternative Source Term.

In addition to collaborative experimental 

programs, NRC has and continues to 

participate in a number of international 

collaborative analytical activities. These 

include: (1) reactor applications of the 

fuel-coolant interaction tools under the 

OECD/NEA [Nuclear Energy 

Agency]-SERENA program; (2) benchmark 

exercise of core-concrete interaction codes 

against OECD/NEA-MCCI experiments at 

ANL; (3) OECD/NEA Benchmark Study of 

the Accident at the Fukushima Daiichi 

Nuclear Power Station (BSAF); (4) 

European Research on Containment 

thermal-hydraulics and Severe Accident 

Management (ERCOSAM); and (5) 

OECD/NEA Filtered Containment Venting 

Systems (FCVS) project. In addition, the 

NRC continues work-in-kind participation in 

the European Severe Accident Research 

NETwork of excellence (SARNET). Such 

efforts ensure that the NRC is cognizant of 
82

analysis approaches taken by other 

international agencies and contribute to 

multinational convergence of evaluation 

methodologies and the success of the 

multi-national design evaluation program 

(MDEP). In addition, analysis activities help 

identify areas where further code 

enhancements may be warranted.

Severe Accident Research Knowledge 

Management

As a part of its commitment to the agency 

effort in knowledge management (KM), RES 

has initiated efforts to collect and catalogue 

experimental data and models used to 

describe severe accident phenomena for 

submittal to the NRC Agencywide 

Documents Access and Management 

System (ADAMS).

Over the years, the NRC invested heavily in 

severe accident research through major 

experimental and model development 

studies to achieve the current 

understanding of the progression and the 

radiological consequences of severe 

accidents. RES KM efforts are essential for 

preserving the information and insights 

gained from this research. 

The NRC has also been participating in the 

SARNET2 project, which includes meetings 

and computer forums where different 

organizations can share their work and 

databases for preserving project 

information. 

Assessment and Recommendations

The Advisory Committee on Reactor 

Safeguards (ACRS) continues to endorse 

the strategy that the NRC staff has 

developed to support regulatory decisions 

for severe accidents via computer code 

development validated by experimental data 

analysis and evaluation. This approach has 

successfully allowed the NRC to maintain 

and update its modeling capabilities for 

severe accident analyses. 

The ACRS also continues to support the 

NRC approach to leverage resources by 

obtaining data through participation in 

international experimental collaborations. 

Planned program extensions and 

continuations of these collaborations are 

well worth the investment and essential for 

NRC to keep abreast of recent data 

affecting source term evaluations and 

analyses of the effects of proposed accident 

mitigation strategies. The ACRS plans to 

review the process used by the staff to 

determine that new data are sufficient to 

warrant improvement to severe accident 

analysis models.

The ACRS notes that models in the NRC’s 

system-level code, MELCOR, were 

developed from a database that is primarily 

focused on in-vessel testing with 

significantly more PWR-specific 

experiments. Hence, it is anticipated that 

model deficiencies associated with 

BWR-specific phenomena, the impact of 

salt water addition, and ex-vessel 

phenomena will be identified as more 

information from the events in the 

Fukushima Dai-ichi nuclear complex in 

Japan become available. As such 

deficiencies are identified, ACRS 

recommends that the NRC expand its

current severe accident research program 

to actively engage in efforts to obtain the 

required data to enhance and validate 

models that are found to be deficient. 

The ACRS also endorses RES efforts to 

support agency KM efforts. The ACRS 

encourages RES to seek collaborations with 

other organizations, such as Electric Power 

Research Institute (EPRI), U.S. Department 

of Energy (DOE), and SARNET2, to 

leverage research funding to preserve and 

consolidate experimental data. 
83

Figure 8. Current NRC Research Activities Related to Severe Accident and Source Term 

Severe Accident and 

Source Term Research

Independent Evaluation of Licensees’ Submittals and 

Support for Risk-Informed Rulemaking involving Severe 

Accident and Source Term Evaluations

MELCOR Development and 

Severe Accident Code 

Evaluations

Collaborative Severe Accident 

Experimental and Analysis Programs

• MELCOR Development, 

Maintenance and User Support 

• MELCOR, MACCS2 and Other 

Code Analyses for Design 

Certification, Pre-application, and 

other evaluations (e.g., filtered 

venting, spent fuel pool study, 

CSGTR, SOARCA)

• Fukushima Analyses (MELCOR, 

MELTSPREAD, TEXAS, MACCS2, 

etc.)

• OECD/NEA-STEM

• OECD/NEA-SERENA

• OECD/NEA Benchmark Study of the 

Accident at the Fukushima Daiichi 

Nuclear Power Station (BSAF)

• OECD/NEA Filtered Containment 

Venting System (FCVS)

• ERCOSAM

• OECD/NEA-MCCI-2

• PHEBUS-ISTP (IRSN)-(VERDON)

• QUENCH Program (KIT)

• SARNET2

Severe Accident

Knowledge 

Management
84

Table1. International Collaborative Severe Accident Experimental Research Programs with 

NRC Participation

Program /Performing 

Organization (Country)

Objectives and Scope 

OECD/NEA BIP2

(Behavior of Iodine 

Project),

AECL (Canada)

BIP (completed) was initiated to provide separate effect tests and modeling studies of iodine 

behavior in a containment following a severe accident. The BIP-2 program is performing more 

detailed evaluations of iodine absorption on painted containment surfaces and organic iodine 

formation. 

OECD/NEA-STEM 

(Source Term Evaluation 

and Mitigation)

IRSN

(France)

STEM was initiated to address remaining issues in source terms. The project focuses on iodine 

behavior under radiation in the EPICUR facility (aerosols, long term), on ruthenium chemistry in the 

primary circuit of a water cooled nuclear reactor and on the state-of-the-art on iodine interaction with 

paint. It involves two experimental projects: The Experimental Program for Iodine Chemistry Under 

Irradiation (EPICUR) and Study and Transport of Ruthenium in the primary circuit (START). 

EPICUR, a facility used under PHEBUS-ISTP that consists of a closed gas loop in which iodine 

samples can be irradiated, is being used to further study iodine release behavior from iodine loaded 

paint and from iodine aerosols. The START experiment consists of a controlled-temperature-profile 

heated tube in which different ruthenium species are introduced. The chemical form of deposits and 

transported species are observed. 

OECD/NEA - SERENA 

(Steam Explosion

Resolution for Nuclear 

Applications)

KROTOS Experiments, 

CEA (France) 

TROI Experiments, 

KAERI (Korea) 

The SERENA program was established to assess the capabilities of fuel-coolant interaction (FCI) 

computer codes to predict steam explosion-induced loads in reactor situations. Phase II of the 

SERENA project is using the complementary features of the TROI (KAERI) and KROTOS (CEA) 

corium facilities. Experimental research is supplemented with analytical activities so as to improve 

the FCI codes and strengthen confidence in their applicability to severe accident scenarios. 

The SERENA program has been completed. OECD proposed a Technical Opinion Paper on the 

subject. One follow-on national program on corium-water interaction (ICE) is underway at IRSN. 

KAERI continues to perform additional experiments in the TROI facility. The NRC continues to 

follow these programs through bilateral agreements with the respective organizations.

OECD/NEA- MCCI (Melt 

Coolability and Concrete 

Interaction) Project, ANL 

(United States )

The focus of the original project, MCCI-I (completed), was to investigate the coolability of molten 

core materials, interacting with the containment structural concrete, by an overlaying water layer.

The second phase of project (MCCI-II), carried out from 2006 to 2010, helped bridge data gaps not 

fully covered in previous tests. MCCI-II also addressed the effectiveness of design features in new 

LWR designs for augmenting coolability, e.g., CCI tests for examining the melt behavior with 

underlying cooled refractory basemat, similar to the EPR core retention concept. 

The NRC is participating in one joint EdF-IRSN/CEA-NRC test to confirm early water ingress 

phenomena. Data will be used in a benchmark effort by participants.

OECD/NEA 

PHEBUS -FP and ISTP

IRSN (France)

PHEBUS-FP Program (completed in 2006) consisted of a series of in-pile integral experiments of: 

fuel degradation; fission product release; radionuclide transport through a model of reactor coolant 

system; and aerosol behavior in model containment. The NRC also participated in the 

PHEBUS-ISTP follow-on program, which is made up of several separate effects projects to pursue 

specific aspects of PHEBUS-FP findings. With the exception of VERDON, all PHEBUS-ISTP 

experimental projects have been completed.

OECD/NEA Sandia Fuel 

Project (SFP)

(United States)

This project provided experimental data relevant for hydraulic and ignition phenomena of prototypic 

LWR fuel assemblies. The proposed experiments focused on thermal-hydraulic and ignition 

phenomena in pressurized-water reactor (PWR) 17x17 assemblies and supplemented earlier results 

obtained for boiling-water reactor (BWR) assemblies. Code validations based on PWR and BWR 

experimental results considerably enhance severe accident code capabilities. This project to 

evaluate BWR experimental work has been completed, and results are documented in 

NUREG/CR-7143. MELCOR evaluations of PWR experiment results are underway.
85

17. THERMAL HYDRAULICS

BACKGROUND

Evaluation of thermal and hydraulic 

phenomena and their coupled effects on 

nuclear safety have always been central to 

the conduct of the NRC’s regulatory 

mission. Of particular importance has been, 

and continues to be, the capability to 

independently confirm thermal-hydraulic 

analyses in submittals from licensees or 

applicants. 

The thermal-hydraulic analyses are 

inherently complex because the phenomena 

are difficult to model, occur at large 

geometric scales, and usually result in 

strongly nonlinear behavior, which is difficult 

to model and solve. Early thermal-hydraulic 

approaches to analyses of nuclear power 

plants, therefore, adopted simple models to 

describe the phenomena with ‘conservative’ 

assumptions, to assure that large safety 

margins could be built into estimates of 

temperatures and pressure limits. With 

time, the experimental database and the 

confidence in the capability of more 

sophisticated models that better predict 

thermal-hydraulic behavior have grown. 

The NRC now evaluates license submittals 

from licensees employing such 

“best-estimate” thermal-hydraulic analyses 

together with estimates of uncertainties in 

predicting the limits. To reduce these 

uncertainties, analyses continue to grow 

ever more sophisticated partly enabled by 

computational capability, which continues to 

grow. This in turn allows for more detailed 

models that better portray the physical 

systems. It is necessary for the NRC to 

continue thermal-hydraulic development 

both to retain state-of-the-art capability and 

to keep pace with the sophistication of 

future submittals. Thus, the NRC, through 

its research program, maintains 

thermal-hydraulics competence and 

capability to conduct appropriate 

confirmatory analyses for safety 

evaluations.

Rod Bundle Heat Transfer (RBHT) facility

Through the joint efforts of the United States 

Nuclear Regulatory Commission and 

Pennsylvania State University, an experimental 

Rod Bundle Heat Transfer (RBHT) facility was 

designed and built. The purpose of the facility is 

to collect fundamental rod bundle 

thermal-hydraulic data that can be used to 

develop and validate models for 

thermal-hydraulic codes used for safety 

analyses of nuclear power plants.

CURRENT RESEARCH ACTIVITIES

The organization of the main NRC research 

activities in thermal-hydraulics is depicted in 

Figure 9. The foci of the research activities 

are primarily on the development of the 

TRACE computer code for confirmatory 

analyses of a 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), currently using 

commercial tools.
86

TRACE Computer Code Development 

and Validation

In the mid-1990s, a 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 

loss-of-coolant accident (LOCA)), TRAC-P 

(for PWR LOCA), TRAC-B (for BWR 

LOCA), and RAMONA (for BWR stability). 

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. 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 meant 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 

Anticipated Transient without Scram 

(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 has been completed. The integration, 

validation, and assessment of the 

TRACE/PARCS coupled code has been 

recently completed 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 MATLAB. 

Thermal-hydraulic system codes, including

TRACE, solve an intertwined structure of 

approximate conservation equations and 

empirical correlations, which have become 

increasingly sophisticated. It is now 

common practice in many vendor analyses 

for the nuclear industry to employ at least 

three fields—droplets, continuous liquid,

and continuous vapor—to capture core 

thermal-hydraulics behavior. The oil-gas 

industry already uses more sophisticated 

models such as the four-field description 

recommended by the peer-review group. 

All this has the potential to put the empirical 

correlations used on a firmer physical basis 

and reduce uncertainties. 

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. Because of the high 

uncertainties that can exist in a calculation, 

modifications are being made to TRACE so 

that the code uncertainty can be 

quantified. The modifications are intended 

to be general, and enable the staff to 

evaluate uncertainty methods proposed by 

applicants as well as allow the staff to 

statistically determine the uncertainty in 

calculations made with TRACE and 

TRACE/PARCS.
87

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. The assessment 

of the applicability of TRACE for 

confirmatory analyses of safety-significant 

thermal-hydraulic phenomena in the 

AP1000, APWR, and EPR designs has also 

been completed to validate the use of 

TRACE in the design certification process. 

Work is also ongoing to develop TRACE for 

applicability to iPWR designs such as 

NuScale and mPower. The pre-applicability 

assessments indicate that the TRACE and 

PARCS codes can be effectively used to 

perform thermal-hydraulic confirmatory 

analyses on iPWR designs. 

In 1985, RES developed the International 

Code Assessment and Application Program 

(ICAP) to assess and improve its 

thermal-hydraulic computer codes. In the 

early 1990’s, ICAP developed into the Code 

Application and Maintenance Program 

(CAMP). The CAMP agreement involves 

monetary and in-kind technical 

contributions. The technical contributions 

include sharing code experience and 

identifying areas for code and model 

development. The CAMP provides 

members with TRACE code. As TRACE 

matures, CAMP continues to be an 

important contributor to its assessment. 

Experimental Studies of 

Thermal-hydraulic Phenomena

Thermal-hydraulic phenomena involved in 

normal and accident conditions for 

light-water reactors (LWRs) are complex, 

and often involve the difficult-to-model 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 

two-phase flows in complex geometries, 

where large-scale tests are the primary 

means of confirming the validity of these 

predictions. In view of this, NRC-RES has 

maintained two relatively large-scale 

experimental facilities:

• the PUMA facility at Purdue University 

for BWR-related issues 

• the RBHT facility at Pennsylvania State 

University for PWR emergency core 

cooling issues 

The Purdue University Multidimensional 

Integral Test Assembly (PUMA) is a 

medium-sized, reduced height scaled facility 

and has been used in the past to perform 

integral LOCA tests of interest for the 

ESBWR design. Tests have also been 

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 are used 

to support the technical assessment of 

Generic Safety Issue (GSI) 193, “BWR 

ECCS Suction Concerns.”

The Rod Bundle Heat transfer (RBHT) 

facility at Pennsylvania State University 

(PSU) was developed to address issues 

related to emergency core cooling. The 

RBHT experimental program has been 

performing separate experiments using a 

full length rod bundle. The tests focus on 

steam cooling and reflood 

thermal-hydraulics, including the importance 

of spacer grids and droplet coalescence and 

breakup behavior in determining peak clad 

temperature (a key regulatory figure of 

merit). These results emphasize the need 

to move toward more detailed modeling of 

dispersed fields as recommended by the 

TRACE peer-review group. 

An alternate approach to modeling 

dispersed and continuous fields explicitly is 

to develop so-called “closure relationships” 

for the evolution of interfacial area in 

two-phase flows. An exploratory program to 

this end has been going on at Purdue 

University to complement their experimental 
88

program for some time. Tests have 

performed with the emphasis of bubble 

interfacial area transport in pipes, annuli, 

and rod bundles. 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 

developed for incorporation into the TRACE 

code, potentially improving its accuracy and 

reliability. An experimental research 

program at PSU is also in place to enhance 

the database for interfacial area models. 

Development of interfacial area transport 

models may have some value as 

demanding research projects to educate the 

next generation of thermal-hydraulic 

experts. However, whether implementing 

such models in the TRACE code should be 

prioritized over development of the four-field 

model recommended by the TRACE 

peer-review group should be carefully 

assessed and reviewed. Since results from 

the interfacial area transport research have 

been slow in coming, the strategy for 

utilizing interfacial transport models in 

TRACE requires examination going forward.

In parallel to these efforts, the NRC is 

collaborating with international groups in 

undertaking experiments in facilities abroad,

as noted below:

OECD/NEA ROSA-2: The NRC is 

participating in the OECD/NEA ROSA-2 

Program 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 Program was completed in 

2012 and has provided 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 are 

being used to assess predictive capability of 

thermal-hydraulic analysis codes including 

TRACE. 

OECD/NEA-PKL-3: The OECD/NEA-PKL3 

Program began in 2012 and is expected to 

conclude in 2015. The PKL facility is a 

full-height, 1:145 power and volume scaled 

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

PKL-3 program has been developed to 

address new series of topics, to include 

beyond-design-basis accidents with 

significant core heatup, accident during 

shutdown conditions, and a followup to the 

boron-precipitation test conducted under 

PKL2. This program provides an extensive 

database for use in the further development 

and validation of thermal-hydraulic codes.

Development of Multidimensional CFD 

Capabilities

The NRC currently has a modest but 

productive effort in the area of 

computational fluid dynamics (CFD) using 

commercial CFD codes from ANSYS Inc. 

(FLUENT) and CD-Adapco (STAR-CCM+).

These codes are user-friendly and are 

applied to a wide variety of industrial 

problems, with reasonably accurate results 

for steady-state single-phase flows or 

multiphase flows which are homogeneous 

and in equilibrium. In such applications,

these commercial codes can provide 

three-dimensional single-phase or 

homogeneous equilibrium two-phase flow 

information not available from system code 

thermal hydraulic simulations. On the 

negative side, some of the code details 

cannot be readily subjected to peer-review 

because of commercial constraints. 

Nonetheless, their CFD predictions have 

been useful for multidimensional analyses 

of certain phenomena where such effects 

are significant, and have played a role in 

resolution of a number of broad technical 

issues, such as induced steam generator 

tube failures, distribution of injected boron in 

the ESBWR, spent fuel pool analyses, and 

predicting the scale-up behavior of 

advanced accumulators. 
89

ASSESSMENT AND 

RECOMMENDATIONS

Excellent progress has been made in 

developing and incorporating the NRC’s 

systems thermal-hydraulics code, TRACE, 

into the regulatory process. Further 

development should focus on 

implementation of the four-field 

thermal-hydraulics model, as recommended 

by the TRACE peer-review group. The 

research program on interfacial area 

transport should be phased out as the 

results are of limited value for TRACE 

applications.

The ACRS is supportive of the agency’s 

active participation in international 

collaborative efforts, as they take advantage 

of facilities that are of a scale and capability 

that do not currently exist in the United 

States. They also draw on the expertise of 

international partners, who have continued 

to maintain a high level of capability in 

thermal-hydraulics. Complementary 

development of unique U.S. facilities to 

support confirmatory accident analyses for 

new fuel and reactor designs should be 

seriously considered. Integral tests would 

enhance confidence in regulatory analysis 

of new reactor designs. Such integral tests

cannot be carried out in any existing 

facilities.

The NRC currently has modest, but 

productive, efforts in the area of 

computational fluid dynamics (CFD) through 

the use of commercial CFD codes, which 

have played a role in improving the 

technical bases for certain licensing 

decisions. In this direction, licensees are 

and will continue to capitalize on the 

extraordinary advances in computing power 

and computational science to address many 

critical safety issues, e.g., prediction of the 

behavior of full-scale components, such as 

advanced accumulators, based on 

small-scale experiments. The agency 

should maintain independent confirmatory 

capabilities that keep pace with such 

developments in industry. To this end, the 

NRC should evaluate various options, which 

could include the following: First, 

participation and cost-sharing in programs 

with international partners to develop 

next-generation, high-fidelity, well-validated 

CFD simulation tools suitable for 

independent confirmatory analyses; 

Second, perhaps in conjunction with the 

first, formation of an NRC-US university 

consortium to develop such capability,

leveraging the world-leading CFD expertise 

in several U.S. universities; Third, perhaps 

in conjunction the other options, building on 

one of the excellent open-source CFD 

platforms. While such platforms are still 

less user-friendly than commercial codes, 

they are being increasingly developed in 

this direction as they offer significant 

advantages in transparency, independent 

verification, and incorporation of advanced 

capabilities for transient multidimensional 

thermal-hydraulics computations.
90

Figure 9. Current NRC Research Activities in Thermal-Hydraulics Research

Thermal-Hydraulics 

Research

• Development of 

computational 

capability with 

STAR-CCM+ and 

FLUENT CFD codes

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 an 

uncertainty 

methodology for 

TRACE

• Assessment of 

TRACE

• Support of the 

CAMP program

• T-H calculations for 

EPUs and new 

t

• RBHT test 

program at Penn 

State

• T-H Experiments 

in Purdue 

University 

• Participation in 

OECD/NEA 

PKL-3 Project 

• Participation in 

OECD/ROSA 

Program
91

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92

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April 2012.





UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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OFFICIAL BUSINESS
NUREG

-1635

Volume 11

Review and Evaluation of the Nuclear Regulatory 

Commission Safety Research Program