SU.S.NRC

United States Nuclear Regulatory Commission

Protecting People and the Environment

NUREG-1635

Vol. 8

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
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NUREG-1635

q• ,SNC Vol. 8

United States Nuclear Regulatory Commission

Protecting People and the Environment

Review and Evaluation of

the Nuclear Regulatory

Commission Safety

Research Program

A Report to the

U.S. Nuclear Regulatory

Commission

Manuscript Completed: March 2008

Date Published: March 2008

Advisory Committee on Reactor Safeguards

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 NuclearRegulatory Research (RES). 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 missions. The report also addresses the issue of long-term

sustained research at the NRC. This report does not address the research being done at NRC on

issues of reactor security or the threat of sabotage. The ACRS views on current work in the area of

security have been reported in separate documents. Two pertinent, interdisciplinary efforts, the

State-of-the-Art Reactor Consequence Analyses (SOARCA) Project and the study of sump screen

blockage are not addressed in this report. These projects are actively followed by the Committee.

The ACRS has been providing interim reports on the technical approach and progress of these

activities.

Paperwork Reduction Act Statement

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

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

Public Protection Notification

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

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

currently valid OMB control number.

iii

TABLE OF CONTENTS

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

ABBREVIATIONS.......................................................................i

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

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

3, LONG-TERM, SUSTAINED RESEARCH AT NRC ............................... 9

4. DIGITAL INSTRUMENTATION AND CONTROL SYSTEMS...................1

5. FIRE SAFETY.................................................................. 19

6. REACTOR FUEL............................................................... 23

7, NEUTRONICS ANALYSIS, CORE PHYSICS, AND CRITICALITY.............. 27

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

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

10. OPERATIONAL EXPERIENCE................................................. 39

11. PROBABILISTIC RISK ASSESSMENT......................................... 41

12. SEISMIC AND STRUCTURAL ENGINEERING................................ 45

13. SEVERE ACCIDENTS ..................................... I..................... 49

14. THERMAL HYDRAULICS...................................................... 53

15. ADVANCED NON-LWR DESIGNS............................................. 59

16. REFERENCES................................................................. 61

V

ABBREVIATIONS

ABR

ACRS

ACR-700

AEOD

AMPX

ANS

ASCE

ASP

ASTM

ATHEANA

ATWS

BWR

•CANDU

CAROLFIRE

CDF

CFD

CFR

COL

DOE

EPIX

EPR

EPRI

ESBWR

FAQ

FY

GDC

GNEP

GSI

GWd/t

HERA

HRA

HTGR

I&C

IAEA

IASCC

IEEE

IRIS

ISI

JAEA

JNES

LBLOCA

LER

LERF

LHS

LMR

LSTF

Advanced Burner Reactor

Advisory Committee on Reactor Safeguards

Advanced CANDU Reactor-700

Office for Analysis and Evaluation of Operational Data

A Modular Code System for Processing X-Sections

American Nuclear Society

American Society of Civil Engineers

Accident Sequence Precursor Program

American Society for Testing and Materials

A Technique for Human Event Analysis

Anticipated Transient Without Scram

Boiling Water Reactor

CANada Deuterium Uranium

Cable Response to Live Fire

Core Damage Frequency

Computational Fluid Dynamics

Code of Federal Regulations

Combined License

Department of Energy

Equipment Performance and Information Exchange System

Evolutionary Power Reactor

Electric Power Research Institute

Economic Simplified Boiling Water Reactor

Frequently Asked Questions

Fiscal Year

General Design Criteria

Global Nuclear Energy Partnership

Generic Safety Issue

Giga Watt day per metric ton

Human Event Repository and Analyses

Human Reliability Analysis

High Temperature Gas-cooled Reactor

Instrumentation and Control

International Atomic Energy Agency

Irradiation Assisted Stress Corrosion Cracking

Institute of Electrical and Electronics Engineers

International Reactor Innovative and Secure

In-Service Inspection

Japan Atomic Energy Agency

Japan Nuclear Energy Safety Organization

Large-Break Loss-of-Coolant Accident

Licensee Event Report

Large Early Release Frequency

Latin Hypercube Sampling

Liquid Metal-cooled Reactor

Large Scale Test Facility

vii
ABBREVIATIONS (Cont'd)

LOCA Loss-of-Coolant Accident

LWR Light Water Reactor

MACCS MELCOR Accident Consequence Code System

MCCI Molten Core Concrete Interaction

MeV Million Electron Volts

MOX Mixed Oxide

NDE Non-Destructive Examination

NEA Nuclear Energy Agency

NEI Nuclear Energy Institute

NEWT NEW Transport algorithm

NFPA National Fire Protection Association

NGNP Next Generation Nuclear Plant

NMSS Office of Nuclear Material Safety and Safeguards

NOAA National Oceanic and Atmospheric Administration

NRC Nuclear Regulatory Commission

NRO Office of New Reactors

NRR Office of Nuclear Reactor Regulation

NSIR Office of Nuclear Security and Incident Response,

NUPEC Nuclear Power Engineering Corporation

OECD Organization for Economic Cooperation and Development

PARCS Purdue Advanced Reactor Core Simulator

PA-UT Phased Array - Ultrasonic Technique

PCI Pellet Cladding Interaction

PFM Probabilistic Fracture Mechanics

PIRT Phenomena Identification and Ranking Table

PRA Probabilistic Risk Assessment

PSA Probabilistic Safety Assessment

PSHA Probabilistic. Seismic Hazard Analysis

PTS Pressurized Thermal Shock

PSI Paul Scherrer Institute

PUMA Purdue University Multidimensional Integral Test Assembly

PUREX Plutonium Uranium Extraction

PWR Pressurized Water Reactor

RBHT Rod Bundle Heat Transfer

RCS Reactor Coolant System

RES Office of Nuclear Regulatory Research

RG Regulatory Guide

ROP Reactor Oversight Process

RPV Reactor Pressure Vessel

SCALE Standardized Computer Analysis for Licensing Evaluation

SDP Significance Determination Process

SER Safety Evaluation Report

SNAP Symbolic Nuclear Analysis Package

SNL Sandia National Laboratories

SOARCA State-of-the-Art Reactor Consequence Analyses

SPAR Standardized Plant Analysis Risk Model

Vill
ABBREVIATIONS (Cont'd)

SRM Staff Requirements Memorandum

SSHAC Senior Seismic Hazard Analysis Committee

THIEF THermally Induced Electrical Failure

TRACE TRAC-RELAP Advanced Computational Engine

TRITON Transport Rigor Implemented with Time-Dependent Operation for Neutronic

Depletion

UREX Uranium Extraction

U.S. United States

USGS United States Geological Survey

V&V Verification &kValidation

I. INTRODUCTION

In this report, the Advisory Committee on

Reactor Safeguards (ACRS) presents the

results of its review and evaluation of the

Nuclear Regulatory Commission (NRC)

Safety Research Program. The NRC

maintains a Safety Research Program to:

* Ensure its regulations and regulatory

processes have sound technical bases

and these bases are refined as new

knowledge develops.

* Prepare for anticipated changes in the

nuclear industry that could have safety

implications.

* Develop improved methods to carry out

its regulatory responsibilities.

* Maintain an infrastructure of expertise,

facilities, analytical capabilities, and data

to support regulatory decisions.

The current research program, organized by

the Office of Nuclear Regulatory Research

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

and initiatives in the Offices of Nuclear

Reactor Regulation (NRR), New Reactors

(NRO), Nuclear Material Safety and

Safeguards (NMSS), and Nuclear Security

and Incident Response (NSIR). For the

purposes of this report, the ongoing research

has been examined in terms of the following

technical disciplines:

Digital Instrumentation and Control

Systems

Fire Safety

Reactor Fuel

* Neutronics Analysis, Core Physics, and

Criticality Safety

* Human Factors and Human Reliability

* Materials and Metallurgy

* Operational Experience

* Probabilistic Risk Assessment

* Seismic and Structural Engineering

" Severe Accidents

* Thermal Hydraulics

* Advanced Non-LWR Designs

This report does not address the research

being done at NRC on issues of reactor

security or the threat of sabotage. The ACRS

views on current work in the area of security

have been reported in separate documents.

Two pertinent, interdisciplinary efforts, the

State-Of-the-Art Reactor Consequence

Analyses (SOARCA) Project and the study of

sump screen blockage are not addressed in

this report. These projects are actively

followed by the Committee. -The ACRS has

been providing interim reports on the

technical approach and progress of these

activities.

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 15.

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:

I
* Identifies and resolves current safety and organizations (NRO,NRR, NMSS, NSIR) at

regulatory issues. NRC in 10 to 20 years.

* Provides technical bases for the

resolution of foreseeable safety issues.

" Develops the capabilities of the agency to

independently review risk-significant

proposals and submittals by licensees

and applicants.

* Supports agency initiatives, including the

move toward a much greater use of risk

information in the regulatory process and

to evolve NRC safety regulations to be

"technology neutral".

* Improves the efficiency and effectiveness

of the regulatory process.

* Maintains technical expertise within the

agency and associated facilities in

disciplines crucial to the agency mission

and that are not readily available from

other sources.

As requested by the Commission in its

June 22, 2007 Staff Requirements

Memorandum (SRM), the ACRS examined

the safety research program to identify any

gaps as well as ongoing research projects

that have progressed sufficiently to meet the

regulatory needs so that they can be curtailed

in favor of more important research activities.

The Committee finds the ongoing research

program well directed to address the shortterm needs of the agency. The Committee

has identified research that could benefit from

opportunities for collaborations with similar

activities under way outside the U.S.

The long-term, sustained research at the

NRC is discussed in Chapter 3. In that

Chapter, the ACRS suggests that in addition

to the work on specific technical issues that

are anticipated to arise, a fraction of the NRC

research program be devoted to developing a

regulatory infrastructure for much more

efficient and effective regulatory work by line

2
2. GENERAL OBSERVATIONS AND RECOMMENDATIONS

In this Chapter, the ACRS highlights major

components of the ongoing research activities

dealing with the safety of nuclear power

plants and presents its recommendations

concerning these activities. Some notable

generic aspects of the research activities are:

* NRC has succeeded over the last few

years in its effort to tie research activities

it undertakes to near-term issues being

confronted by the NRC line organizations

(NRO, NRR, NMSS, NSIR). About 67%

of research activities support specific

needs of these Offices. About 28% are

mandated programs such as the Accident

Sequence Precursor (ASP) program. The

remainder is administrative activities. An

especially useful innovation adopted in

some major research activities has been

the organization of Technical Advisory

Groups. Members of these advisory

groups consist of staff from the line

organizations. Often, the close

interactions of the Technical Advisory

Groups with the ongoing research

activities have positive impact on both the

scope and the pertinence of the research.

* Current NRC research activities are

positioned well to address most of the

research areas identified in a recent

Strategic Plan for Light Water Reactor

(LWR) Research and Development

prepared for the Department of Energy

(DOE) and the nuclear industry,

including:

- Greater use of risk information to

improve safety

- Support the development of a

regulatory process for deployment of

digital instrumentation and control

technology

Improve understanding of materials

degradation and plant aging to

support license extension to 80 years

NRC research will have to confront a

fourth thrust identified in the DOE/nuclear

industry Strategic Plan - extension of fuel

burnup to 85 GWd/t. Research the

agency has completed provides a suitable

basis for modifying existing regulations to

accommodate new fuels and alloy

claddings for extended bumup. Research

will be needed to search for possible

emergence of new fuel physics, cladding

degradation, and fuel-clad mechanical

and chemical interactions at such high

levels of burnup. As noted below, lack of

availability of suitable experimental

facilities will be a handicap. Building upon

the international collaborations developed

in the course of confirmatory research for

burnup extensions'from -30 GWd/t to the

current regulatory limit of 62 GWd/t may

be useful.

Collaborations with other countries are

being pursued in many of the research

projects. The nature and depth of

collaboration varies. In the case of severe

accident research, there is very

comprehensive collaboration on both

experiments and modeling. Similarly,

detailed collaboration is taking place in

fuels and materials research. In other

research areas, the collaborations are

simple information exchange.

Collaborations have the potential of

leveraging resources. More importantly,

collaborations expand the intellectual

basis for research and provide the NRC

staff with a kind of real time peer review of

its research plans.

Representatives of industry have

expressed an interest in closer

collaboration with NRC, including the joint

3
development of models and computer

codes. NRC has a Memorandum of

Understanding with the Electric Power

Research Institute (EPRI) supporting joint

experimentation. The ACRS has been

supportive of NRC collaboration with

industry on experiments provided that

NRC has an effective and early voice in

the design of the tests. The ACRS has

misgivings about collaboration between

industry and the NRC in the development

of models or computer codes and the

interpretation of test results. The

independent development of models has

been a significant route to discovery of

safety issues and safety margins. The

independence of analysis is a crucial

element in ensuring public confidence in

the regulatory process.

In the past, few of the research activities

undertaken by the agency had detailed

and updated research plans describing

the objectives of the research as well as

both tactics and strategy for meeting

these objectives. A notable exception has

been the planning of research in the area

of Digital Instrumentation and Control

Systems. It is noteworthy that several

other research activities are beginning to

document their research plans. Indeed,

there is work under way to document the

entire reactor safety research program.

We look forward to an opportunity to

review this overall research plan as well

as plans for individual research activities

as they are being developed.

The continued erosion of experimen'tal

capabilities in the U.S. is affecting the

NRC's research program. This loss of

experimental capabilities is especially

acute in the areas of fuel research and

neutronics research- Domestic hot cells

are aging and in short supply and test

reactors suitable for technical resolution

of LWR safety issues are scarce. The lack

of properly scaled experimental research

facilities also affects research into twophase thermal hydraulics.

There is a growing reliance on

computational methods to deal with

physical, chemical, and neutronics issues

affecting the safety of nuclear power

plants. Associated with this growth is a

demand to have a quantitative

understanding of the uncertainties

associated with computational results.

Many types of uncertainty may affect

results of computation, including

uncertainties in parameters, uncertainties

in the models, and uncertainties in the

completeness of the analysis. Techniques

have been developed for the quantitative

description of parametric uncertainties

such as Monte Carlo sampling and

Limited Latin Hypercube (LHS) sampling.

Such methods should be included in

computer codes being developed by the

agency.

The agency is vulnerable to the loss of

institutional knowledge in certain technical

areas where it is critical to maintain

competency and even excellence. The

ACRS calls attention to areas where this

vulnerability is being addressed well such

as neutronics research. New staff

members are brought into the field and

trained while more experienced staff

members are available for consultation.

This method is superior to simple exit

interviews and like for the preservation of

institutional knowledge and technical

capabilities.

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 15.

Digital Instrumentation and Control

The use of software-based digital safety

systems in nuclear plants is inevitable.

Demonstration of the reliability of these

systems is a challenge that merits research.

A well founded research plan is in place. But,

4
it must be recognized that the issue may not

be amenable to complete technical resolution.

The feasibility of technical resolution must be

demonstrated by generating research

products useful to the relevant line

organizations over the course of the current

five-year research plan.

Fire Safety

The agency has a much revitalized fire safety

research effort to prepare the line

organizations for evaluating licensees'

transition to the risk-informed, performancebased fire protection programs that meet the

requirements of 10 CFR 50.48(c), and the

referenced 2001 Edition of the National Fire

Protection Association (NFPA) Standard,

NFPA-805. The NFPA-805 Standard requires

that licensees use only fire models that are

acceptable to the NRC. This Standard further

requires that the fire models be verified and

validated and the fire models only be applied

within their limitations. Significant progress

has been made in the verification and

validation of fire models. Still, NRC does not

have the' phenomenological analysis

capabilities to support realistic assessment of

fire risk associated with operational events.

The NRC needs to develop acceptable fireeffects models, including models of smoke

transport within plants and its impacts on

plant safety.

Reactor Fuel

Research has been completed on the effects

of burnup on fuel and cladding behaviour

under design basis accident conditions.

Results of this research can be used to

modify current regulations to better address

new fuels and claddings that will be

introduced by the industry to extend fuel

burnup. As more reactors begin operation at

extended power uprate, there is a need for

the NRC to develop and apply analytical

codes to evaluate the effects of pelletcladding mechanical and chemical interaction

on fuel integrity during abnormal operating

conditions. NRC needs to maintain expertise

in the area of reactor fuel and fuel cladding

since further developments in this field can be

expected. NRC may need to develop further

its expertise in safety of fuel reprocessing

facilities since such facilities may be the first

products of the DOE's Global Nuclear Energy

Partnership (GNEP) to be submitted for

licensing.

Neutronics Analysis, Core Physics, and

Criticality Safety

NRC has adequate computational capabilities

to meet many current needs for neutronics

analysis, core physics, and criticality safety

analysis. These capabilities have been

demonstrated in the identification of reactivity

transients associated with "checkerboard"

voiding in the Advanced CANDU Reactor

(ACR) -700. There is an adequate program to

maintain expertise in these areas that must

constitute a core competency of the agency.

However, available computational and

analysis methods are being challenged by the

complexity of advanced LWR cores such as

the core of the Economic Simplified Boiling

Water Reactor (ESBWR). Further upgrades to

the computer codes to address better these

complexities are needed if the capability to

perform independent assessments is to be

maintained. More extensive improvements to

the existing computational capabilities will be

needed to address advanced reactors that

use high temperature gas or liquid metal

coolants and involve neutron spectra with

larger epithermal and fast components.

Validation of the modeling is a crucial issue.

Much of the validation must now be done with

legacy data. Participation in international

benchmark activities certainly is some

compensation for the lack of active

experimental facilities in this area.

Collaborations on modern validation

experiments will be useful if major upgrades

to computational capabilities are found

necessary.

5
Human Factors and Human Reliability

Human factors and human reliability research

continues to generate results useful to the

regulatory organizations: An important

benchmark of the many human reliability

models is being planned. Results of this test

may help define future directions of human

reliability research at NRC.

Materials and Metallurgy

Based on the number and visibility of

programs, materials and metallurgy is the

most active area of research within the NRC.

This is appropriate in light of the continuing

emergence of crucial and unexpected

materials degradation in aging LWRs,

including the stress corrosion cracking of

control rod drive mechanism nozzles at

Davis-Besse and dissimilar metal weld issues

that have arisen at a number of plants.

Research programs are under way in five

areas:

" Environmentally Assisted Cracking in

LWRs

" Steam Generator Tube Integrity

" Non-destructive Examinations

* Proactive Materials Degradation

Assessment

" Reactor Pressure Vessel Integrity

All of the research activities seem well

founded. The studies of reactor pressure

vessel integrity are enhancing the capabilities

at the agency for probabilistic fracture

mechanics analysis to support more realistic

assessments of the behaviour of aging

pressure vessels. The ACRS is especially

pleased with the proactive materials

degradation assessment effort to identify

degradation mechanisms and vulnerable

components before phenomena become

manifest in plants. The ACRS is also pleased

with the work being done to compare various

methods for non-destructive examination

(NDE) of plant components and answer the

questions being raised about the reliability of

visual inspections of vulnerable components.

Operational Experience

The scope and the visibility of work to analyze

and evaluate operational data have dwindled

dramatically since the function of then Office

for Analysis and Evaluation of Operational

Data (AEOD) was transferred to RES. Aside

from mandated activities such as industry

trends analysis and the Accident Sequence

Precursor (ASP) program, the work is mostly

in support of other research efforts and

particular licensing efforts. No longer are

publicly available, in-depth, analyses of

particular systems or trends being produced.

The resources devoted to this activity are

approaching the minimum needed to sustain

a distinct work unit within the research at

NRC.

Probabilistic Risk Assessment

Probabilistic Risk Assessment (PRA)

research has been focused on the

applications of the available technology to

particular regulatory issues especially to the

Reactor Oversight Process. There has been a

very heavy demand for this type of support

from the line organizations. There is a

growing effort within the larger PRA

community to improve software. NRC needs

to devote some of its PRA research resources

to the development of next generation risk

analysis software to improve the accuracy of

risk assessment and to facilitate NRC review

of risk-informed applications from licensees.

Seismic and Structural Engineering

There is a resurgence of interest in the

threats posed by seismic events to new

power plants. The interest is driven by early

site permit applications, new design

certifications, and combined license (COL)

applications as well as re-evaluations of

seismic hazards especially in the central and

eastern U.S. , Further impetus for seismic

research has been driven by interest in the

performance-based seismic hazard standard

issued by the American Society of Civil

Engineers and used in the Clinton early site

6
permit application. The research program

has established a useful collaboration with

Japan. An excellent study of threats from

tsunamis, produced by undersea landslides,

on the eastern U.S. has been completed in

cooperation with the U.S. Geological Survey

(USGS). Other elements of the program are

well founded and will generate products

useful to the line organizations at NRC.

Severe Accidents

The ACRS supports the strategy developed

by the NRC staff to support regulatory

decisions associated with severe accidents.

This strategy involves development of

computer codes and analysis and evaluation

of experimental data. This approach can

successfully maintain and update the

modeling capabilities for severe accident

analyses. The NRC approach of leveraging

resources through international experimental

collaborations is particularly notable. The

planned program extensions and

continuations of these collaborations are well

worth the investment.

Thermal Hydraulics

The thermal-hydraulics research at NRC is

primarily focused on the development of the

TRACE computer code. TRACE development

is progressing well. The ACRS looks forward

to hearing results of the ongoing peer review

for TRACE.

There is a continuing and even growing need

to incorporate TRACE into the regulatory

process to deal with power uprates, and

greater use of best-estimate analyses in

licensing actions and new reactor design

certifications. It is important that TRACE be

brought into the regulatory process as

expeditiously as possible. The resources for

sustained code development of TRACE

appear minimal.

Experimental facilities to support and validate

TRACE are aging. They do not scale well for

validation of passive plant responses that are

central to many of the new LWR designs

being submitted for certification. NRC needs

to develop a strategy for providing highquality experimental data both to validate

TRACE predictions and to support

certification of advanced LWR designs.

The use of computational fluid dynamics

(CFD) methods in licensee applications will

continue to increase. NRC must be capable of

reviewing and independently validating

licensee proposals using these computational

methods. NRC may be better served by

joining international efforts to develop open

source CFD computer codes rather than

relying solely on commercial CFD codes.

Advanced Non-LWR Designs

The possibility of the need to certify or license

advanced reactors using gas-cooling or liquid

metal-cooling rather than conventional lightwater cooling pose a conundrum for NRC and

the planning of its research efforts. NRC has

some experience with these alternative

reactor technologies, but does not have in

place the technical infrastructure that would

be needed to support an effective certification

review of a reactor design based on such

technologies. The Office of New Reactors

(NRO) is dubious of the feasibility of

developing this technical infrastructure and

conducting research in parallel with a

certification review. The ACRS certainly

agrees with the NRO on this point. It is clear

that it will take some time to develop the

needed technical infrastructure at NRC. What

is not clear is when applicants might be able

to submit a certification application with an

adequate technical basis. This makes it

difficult to assign priority and allocate

resources for advanced non-LWR research.

The research staff has systematically

identified and prioritized the technical issues

associated with the certification of a gascooled reactor. This has been done based on

an expert opinion elicitation process that

identified important phenomena and ranked

these phenomena with respect to safety

7
importance and the , current level of

understanding. The results of the phenomena

identification and ranking will help to

communicate expectations for the technical

information needed for a defensible

certification application for a gas-cooled

reactor. Especially crucial will be use of the

results -to define areas where analyses must

be substantiated by pertinent, prototypical

experimental results.

NRC should initiate activities to assess the

state-of-knowledge in other advanced reactor

technologies before there is further loss of the

limited capability available (worldwide) in

some of these technologies. This effort can

provide a significant input to the development

of safety requirements for some of these

advanced designs. Multinational cooperation

at an early stage could lead to common safety

standards and a sound and more costeffective research program

8
3. LONG-TERM, SUSTAINED RESEARCH AT NRC

The research program that NRC sponsors is

now very closely coupled to the immediate

needs of the line organizations processing

licensing applications and other regulatory

matters. The assessments presented in this

report, overall, are indicative of a positive

appraisal of both the utility and the quality of

the applied research now being sponsored by

NRC. In its previous biennial reports on

review and evaluation of the NRC safety

research program, the ACRS noted the need

for long-term research not tied to the nearterm issues of the regulatory process. As

directed by the Commission, the staff has

undertaken some efforts to identify longerterm research. The development of a longterm research plan is a considerable

departure from the staff's focus in recent

years on immediate regulatory needs. The

ACRS has previously commented on several

of the specific long-term research activities

identified by the staff. By and large, the

topics in the long-term research plan

developed by the staff address regulatory

issues that will have to be confronted

eventually.

In this Chapter, the ACRS addresses the

scope of longer-term research that NRC

needs to consider. The Committee believes

that the primary role of long-term research

should be to address the technical capabilities

of the agency. and the way the agency

conducts its regulatory and safety mission.

The time horizon of interest is about 10 to 20

years. Two important themes fall within this

category of long-term research. One is

maintaining an infrastructure of technical

expertise; facilities, and analytical capabilities.

The second deals with the development of an

infrastructure for much more efficient and

effective regulatory work by line organizations

at NRC in 10 to 20 years.

The ACRS views on the scope of the longterm research the agency needs to consider

are based on a number of assumptions that

the Committee has made about the future:

* Use of nuclear power will grow in the U.S.

over the next twenty years.

" New reactors in the U.S. will be based on

LWR technology; Sodium-cooled and

gas-cooled reactors will become a part of

the NRC's regulatory workload.

* NRC staffing will not grow commensurate

with the growth in nuclear power.

* The average years of experience and

education of NRC staff will decrease over

the next ten years and then stabilize.

* Licensing and regulatory actions per plant

will be about the same in the future as it is

now.

* Licensee submittals will grow more

complicated and technically sophisticated

as licensees continue to utilize the margin

that exists between current operations

and regulatory limits.

* The support vendors and suppliers for the

nuclear industry will grow more

international in character than in the past.

* NRC will continue to risk inform its

regulations and regulatory processes to

achieve a greater focus on issues that are

important to safety. Greater use of

realistic rather than demonstrably

conservative safety analyses will be

encouraged.

* Fuel reprocessing and use of mixed oxide

(MOX) fuels in nuclear power plants will

become more economically and politically

attractive.

9
Maintaining an Infrastructure of Technical

Expertise, Facilities, and Analytical

Capabilities

The ACRS supports maintaining technical

expertise at the agency in areas where there

is limited availability of independent expertise

from the academic community and the private

sector. Technical expertise in these

disciplines (neutronics, fuel, PRA, fission

product chemistry, etc.) is needed by the

agency to make technically sound regulatory

decisions without imposing conservatisms

that are actually unnecessary when the stateof-the-art is examined. The ACRS also calls

attention to needs for access to experimental

facilities and state-of-the-art independent

analyses tools for the regulatory process. The

state-of-the-art in science and technology is

rapidly advancing. It is inevitable that some of

these technologies will also be utilized by the

existing fleet of reactors to improve

economics and safety. These new

technologies would include NDE to monitor

for impending failure, advanced safety

analysis tools (e.g., Multidimensional CFD

codes), next generation of PRA software,

innovative materials, development of highburnup fuels (85 GWd/t and enrichment

beyond 5%), etc. Although in many cases

introduction of these new technologies can

increase safety and efficiency, the use of new

technologies can also introduce new failure

modes.. NRC should assess its research

plans and develop a long-term program to

ensure that the agency is prepared to make

sound science-based safety decisions.

The topic of maintaining the infrastructure of

technical expertise, facilities, and analytical

capabilities to support regulatory decisions is

also addressed in Chapter 2 and in the

discussions of the various technical

disciplines of research in Chapters 4 through

15. The proposed long-term research topics

include:

Technical Infrastructure to Respond

to Innovative Fuels

There is widespread expectation that

the nuclear industry may push for

extending burnup limits to 85 GWd/t.

Certainly, a recent research strategy

document prepared for DOE and

nuclear energy industry proposes that

fuel burnups be extended to 85

GWd/t. In addition, innovative designs

utilizing advanced fuel pellet and

cladding materials are in the LWR

development pipeline. The nuclear

industry program for development of

high-burnup fuel is expected to take

about 10 years and involves

innovative fuels with uranium

enrichment above 5%.

NRC must develop and maintain

expertise in the area of reactor fuel to

respond to these proposed fuels and

claddings. The challenge in

maintaining what amounts to an

essential core competency of the

agency arises because of the limited

availability of expertise outside of the

agency that is independent of

licensees. Current manpower working

for the agency in this field either

directly or by contract is experienced

and there is a need to groom newer

generations in the field. A major

impediment in meeting this need is

the decline in this Country of in-pile

test facilities and hot cells for

examinations of irradiated fuels and

cladding. Collaborations with

international partners having these

facilities and capable of undertaking

pertinent studies will be essential for

NRC to maintain an adequate level of

expertise in reactor fuels.

Full-Height, Integrated, Scaled Test

Facilities for Passive Systems

Based on the expected number of

COL applications in the next few

10
years, it is very likely that in 10 to 20

years some of the operating reactors

would be of the advanced passive

LWR designs such as AP1000 and

ESBWR. It is perhaps not unlikely that

some unanticipated challenges may

arise with operating experience in

'such plants. To better understand the

safety impacts of events and to

maintain a high level of technical

expertise, NRC should consider the

value of having access to full-height,

integrated, scaled test facilities that

can simulate reactors with passive

systems. The understanding and

correlations developed in full-height

test facilities can be applied to reactor

systems more reliably, since they do

not depend on scaling analyses

required for reduced-height systems,

which are difficult to develop and

validate for two-phase flows. NRC

should explore potential international

interest in maintaining such facilities.

Technical Infrastructure to Support

the Independent Safety Evaluation

of non-LWRs, IRIS, and other

Unique Desiqns

Advanced non-LWRs, IRIS, and other

unique designs pose a challenge for

NRC and the planning of its research

efforts. The non-LWRs of interest to

NRC can be categorized as liquid

metal-cooled reactors (LMRs) and

high temperature gas-cooled reactors

(HTGRs). The Toshiba 4S (Super

Safe, Small, and Simple) is a LMR. In

addition, LMR technology is also

being considered for the Advanced

Burner Reactor (ABR) proposed

under the DOE GNEP initiative. The

Next Generation Nuclear Plant

(NGNP) prototype will also utilize

HTGR technology when the concept

is developed., NRC has some

experience with these alternative

reactor technologies, but does not

have in place the technical

infrastructure that will be needed to

support an effective certification

review of a reactor based on such

technologies.

As noted previously, the staff has

systematically identified and

prioritized the technical issues

associated with the certification of a

gas-cooled reactor based on an

expert opinion elicitation process that

identified important phenomena and

ranked these phenomena with respect

to safety importance and the current

level of understanding.

NRC should initiate activities to

assess the state of knowledge in

other technologies before there is

further loss of limited capability

available (worldwide) in some of these

technologies. The state-of-knowledge

report should identify significant gaps

in knowledge that might be necessary

for future licensing of some of these

unique designs. Specific focus should

be given to understanding severe

accident behavior and crucial areas of

safety (e.g., reactor fuel, hightemperature materials). This effort can

provide a significant input to the

development of safety requirements

for some of these designs.

Multinational cooperation at an early

stage could lead to common safety

standards and a sound and more

cost-effective research program

Technical Infrastructure to Support

Safety Evaluation of Fuel

Reprocessing Facilities

Reprocessing of nuclear fuels may be

an emerging technology that merits

research attention by NRC. Aqueous

reprocessing of irradiated fuels has

been done for many years within the

nuclear weapons community using

the PUREX process. Though familiar,

this process has not been trouble free

I1
and there are many known hazards.

NRC is gaining some exposure to the

associated safety issues of "red oil,"

hydroxylamine nitrate, ammonium

nitrate and the like through its review

of the construction authorization

application for the DOE's MOX Fuel

Fabrication Facility.

Aqueous reprocessing of spent

reactor fuels is widely thought to be

the first technology that will emerge

for licensing from the DOE's GNEP

initiative.

Modifications of the PU REX process -

the so-called UREX processes - are

being considered by DOE for its

GNEP initiative. Undoubtedly, safety

issues similar to those of the PUREX

process and new safety issues will

arise in these new processes.

Even more challenging will be the

pyrometallurgical reprocessing of

spent nuclear fuel also being

considered by DOE as part the

longer-term plans of its GNEP

initiative. The large-scale

implementation of pyrometallurgical

reprocessing of commercial fuels has

never been undertaken in the past.

NRC will need to develop expertise in

the safety of fuel reprocessing

technologies well before licensing

applications involving use of such

technologies are submitted for review.

Technical Infrastructure to Support

Materials Degradation Assessment

Long-term research on materials

degradation will be essential to

anticipate and address known and

emerging issues as the Nation's

LWRs extend their operation for 60

years or longer.

Research will be needed to confirm

that the various materials, fabrication,

and water chemistry changes that

have been implemented in LWRs to

mitigate known environmentally

assisted degradation mechanisms will

be effective for the extended life of

the plants.

* Technical Infrastructure for

Independent Assessment of

Multidimensional CFD Analyses

The ACRS anticipates that licensees

will increasingly employ multidimensional CFD capability to resolve

problems which the current generation of thermal-hydraulic codes, such

as TRACE, are unable to do. The

NRC has limited confirmatory

capability to check such submittals.

Such capabilities will also undoubtedly be needed for advanced

reactors, such as HTGRs, and for

evaluation of new fuel designs.

The NRC currently has very little effort

in the area of CFD other than

application of commercial CFD codes.

Although these are used in the

process industry for qualitative

indications of phenomena, they are

validated to a much less rigorous

standard than codes for nuclear use,

and the source codes are not

available. They appear to include a

number of ad hoc fixes to improve

stability and robustness which may

affect their predictive capability for

situations that cannot be studied

experimentally.

The international nuclear community,

on the other hand, has instituted the

development of multidimensional CFD

capabilities. CFD codes are being

developed and validated to standards

of reliability and accuracy required for

use in evaluations of nuclear systems.

NRC should join in this effort and

provide adequate resources to allow

productive participation.

12
. Next Generation of PRA Software

The agency's main PRA tool is the

SAPHIRE code that has been

developed in the last twenty years

and is the basis for the Standardized

Plant Analysis Risk (SPAR) models

that are used extensively in the

Reactor Oversight Process. The

ACRS Subcommittee on Reliability

and Probabilistic Risk Assessment

was briefed recently by representatives from EPRI and an

international group of researchers that

have formed the Open PSA Initiative.

Two issues with the present state of

PRA software were identified. First, it

is hard to check the completeness

and correctness of the logic of the

PRA models produced using current

PRA codes because these models

are code dependent. Second, there is

disturbing evidence that, under certain

circumstances, the probabilistic

calculations may be flawed.

There are efforts under way to

develop a standard PRA representation format that would promote

independence between the logic

models and the individual codes.

Such a standard would facilitate

significantly the review of the PRA

logic, a benefit of great value to the

agency. The staff should participate

actively in the international activities

related to the development of the next

generation of PRA software.

In its 2006 report to the Commission

on the NRC Safety Research

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

ACRS stated that "the staff needs to

review the literature concerning

Binary Decision Diagrams and

evaluate the need to adopt this

technology." The ACRS stated further

that "the growing importance of the

SAPH IRE code and the SPAR models

in the regulatory process warrants

such an investigation." The ACRS

continues to believe that the staff

should undertake such an

investigation and participate actively

in the international activities related to

the development of the next

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

of the potential impact of the new

software on regulatory

decisionmaking. The results of such

an evaluation would guide the

establishment of a longer-term

research project in this area.

Development of a Regulatory Infrastructure for Much More Efficient and

Effective Regulatory Work

The second theme within the general topic of

long-term research at NRC deals with the

productivity of the line organizations handling

regulatory issues and the growth of nuclear

power in the U.S.

The growth of nuclear power seems inevitable

at this juncture based on applications for early

site permits and the recent submittal of

several COL applications. This presumes,

however, that no major reactor accident

occurs in this Country or pertinent accidents

occur in other countries with well developed

nuclear energy capabilities. A focus on LWR

technologies also seems likely, since

alternative technologies have not yet been

demonstrated adequately and several modem

LWR designs have been certified. More

certifications of advanced LWRs are

anticipated. However, the NRC will likely

need to develop safety requirements and an

infrastructure to deal with safety matters

relating to certain non-LWR designs (e.g.,

LMR, HTGR) and unique LWR designs (e.g.,

IRIS). Constraints on the growth of the size of

the NRC staff seem likely based on the

political challenges to growth. Loss of

experience within the staff is simply a reality

of demographics. The seriousness of this

loss will depend, in a significant measure, on

the competition for manpower by the nuclear

13
industry itself. Stabilization of the staff

demographics may be optimistic. Even if the

loss of experience is not realized, the staff will

have to confront more numerous and more

complex licensing actions. Growing technical

sophistication and complexity of licensee

submittals seems likely since the "easier"

changes to utilize margin between operations

and regulatory limits are being made now by

the licensees. Other assumptions such as

the concern over the growing use of

international vendors and suppliers as well as

the development of reprocessing are

extrapolations of current trends.

Together, the assumptions lead to the

unavoidable conclusion that the NRC staff will

have to be both more productive and more

technically sophisticated in the future. Riskinforming the regulatory processes can

relieve some of the burdens on the staff, but

increasing use of realistic analyses in the

place of demonstrably bounding analyses for

risk-important regulatory actions imposes

additional, different burdens.

To achieve greater productivity in line

organizations at NRC, the ACRS sees no

alternative to the provision of superior

technology to the staff in these organizations.

Useful, superior technology will have to come

from NRC research. But, this technology will

have to be useful to the staff within line

organizations that may or may not have indepth familiarity with the bases of the

technology. That is, the research will have to

yield computational and simulation technology

that can be routinely used without the close

assistance of the developers of the technical

capabilities. The ACRS foresees a time when

individual members of NRC staff can access

routinely such things as:

* Comprehensive PRA for individual plants

that addresses all modes of operations

and all initialing events.

* Comprehensive neutronics and thermal

hydraulic analyses of particular nuclear

steam supply systems.

* Simulations of control rooms of particular

nuclear power plants.

" Expert systems that replace the current

Standard Review Plan.

* Corrosion analyses, including stress

corrosion cracking and probabilistic

fracture mechanics of reactor pressure

boundary materials and components

* Virtual reality simulations of particular

plants.

* Finite element structural models.

* Fire-effects models of particular

structures, systems, and components

* Human reliability analyses

* Intrusion simulations

In addition, the ACRS foresees the availability

to individual staff within line organizations of

"user friendly", searchable databases dealing

with such things as:

* Operational events

" Component reliability data

* Failure modes and effects analyses of

systems, structures, and components

* Seismic hazards

* Weather at licensee sites

The specific computational and simulation

capabilities that will be needed to facilitate

greater staff productivity in the processing of

regulatory and the licensee applications are

not the focus of ACRS.. What the ACRS

recommends is that some fraction of the NRC

research effort be devoted to the

development of tools, databases, and

decision-support capabilities that will be

needed in the future. A first step that might

be taken in such research is a systematic

examination of line organization activities to

identify areas where new information

technologies can be used to facilitate

14
regulatory, review. A second step might be

the review of existing computational,

simulation and database resources to

determine how these resources might be

made more "user friendly," robust, and less

dependent on technical support. A third step

might be the development of strategies not to

just preserve knowledge but to make the

knowledge broadly available throughout the

agency.

The long-term research that the agency

needs to undertake to prepare itself for a

future that makes vastly more use of

information technology will have to be

centered within RES. The research cannot

be done, however, independent of the line

organizations. RES has proved itself many

times capable of coordinating well

multidisciplinary research projects through its

Action Plan structure. The ACRS feels such

an organization of the research with

involvement of NRR, NRO, NMSS, NSIR, and

regional personnel could be effective. In any

case, the communications' between the

research effort and the line organizations

must be at the staff level if effective products

welcome by the users are to be produced.

15

4. DIGITAL INSTRUMENTATION AND CONTROL SYSTEMS

Three issues bedevil regulations and the

inevitable transition from analog to softwarebased, digital instrumentation and control

technology for reactor safety systems. First,

the development of methods for the

comprehensive testing of the reliability of

digital systems has not reached to a level of

maturity that allows them to be used in the

regulatory process. It may not even be

possible to comprehensively test the reliability

of such systems. Consequently, the industry

(IEEE) standards that are endorsed by the

NRC focus on the process of design and

development of software-based digital

systems. The assumption is, of course, that a

good controlled production process will lead

to a highly reliable product. The quality of the

product is then very dependent on the quality

of the requirements established for the

system. The focus on process leads to the

second challenge faced by NRC in dealing

with digital systems. Process monitoring and

reviews are very manpower intensive. They

can be the limiting step in a system review.

Furthermore, the focus on the development

process frustrates the licensees' desire to use

commercial off-the-shelf digital systems for

safety functions in nuclear power plants. The

third challenge faced by the NRC is how to

portray the reliability of digital systems in

PRAs.

Digital systems of interest to the NRC are

safety systems such as the reactor protection

system. By and large, the safety systems

involve simple actuations and do not involve

complex feedback and control functions. But,

software-based digital systems offer the

designer much greater flexibility and

functionality than analog systems. In taking

advantage of the potential of digital systems,

the designer creates complexity in the

circuitry, logic, and software. Interconnections

among safety and control systems, some of

A Highly Integrated Control Room

which are subtle and even unintentional,

create vulnerabilities to common-cause

failures that are not random in nature. To a

very real extent, the difficulties associated

with the safety review of digital systems are

burdens the nuclear industry is placing upon

itself.

NRC has been well aware of all of these

challenges for some time. A detailed research

plan1 has been developed to confront these

issues. The staff has formed a Steering

Committee consisting of senior managers to

provide oversight and guidance on six key

technical and regulatory issues and to

interface with the industry:

* Cyber security

* Diversity and Defense In Depth

* Risk-informing Digital Instrumentation

and Control

* Highly Integrated Control Room -

Communications

* Highly Integrated Control Room -

Human Factors

• Licensing Process Issues

See NRC Digital System Research Plan

FY2005-2009.

17
Each key issue is assigned to a Task Working

Group that reports to the Steering Committee.

The staff has identified specific deliverables

and due dates for these deliverables.

The ACRS has been meeting frequently with

the staff and providing reports to the

Commission on progress in these issues on a

regular basis. In general, the ACRS agrees

with the staffs approach to the development

of a process that will facilitate the deployment

of digital instrumentation and control

technology for new and operating reactors.

18
5. FIRE SAFETY

The fire safety research program is

focused on the NRC's regulatory needs

as it prepares the line organizations for

evaluating licensees' transition to the

risk-informed, performance-based

fire protection programs that meet the

requirements of 10 CFR 50.48(c), and the

referenced 2001 Edition of the National

Fire Protection Association (NFPA)

Standard, NFPA-805. The NFPA-805

Standard requires that licensees use only

fire models that are acceptable to the

NRC. This Standard further requires that

the fire models be verified and validated

and the fire models only be applied within

their limitations. There is a close coupling

between the ongoing research programs

and the regulatory needs.

The "small-scale" (Penlight) radiant

heating test facility at SNL, shown testing

two multi-conductor cables for the US

NRC's CAROLFIRE program

Research projects within the pr

be grouped into three technica

* Fire Risk Assessment

* Fire Modeling

* Fire Testing

Each of these technical

discussed below.

FIRE RISK ASSESSMENT

In 2004, the NRC amende

protection requirements in

50.48(c) to permit existin

licensees to voluntarily a

maintain a fire protection pro

meets the requirements of

Edition of the NFPA-805 Stan

alternative to the existing deterr

protection requirements. As c

2008, 27 sites involving 42 nucl

plants have announced their

transition to NFPA-805 requiren

scope fire risk assessment.

performed for each of the tra

nuclear power plant.

ogram can The NRC research efforts in the fire risk

I areas: assessment area are aimed at developing

standards and associated guidance to

assess the quality of such fire risk

assessments and providing inspectors

with the tools and knowledge necessary

to assess the validity of the fire protection

areas is licensing bases for the plants.

During the past few years, RES in

cooperation with EPRI has taken some

important steps to consolidate the fire

d its fire PRA research and development activities.

10 CFR This effort has led to the publication of

g reactor NUREG/CR-6850, "EPRI/NRC-RES Fire Jopt and PRA Methodology for Nuclear Power

gram that Plants," which provides a structured

the 2001 framework for the overall fire risk

Jard as an assessment, along with recommended

ninistic fire practices to address specific aspects of

3f January the analysis. Current RES activities in the Janaryor fire risk assessment area include support

intent to for projects aimed at:

•ents. Full s will be 0 Development of ANS standards for

insitioning fire PRAs

* Supporting NRR in the

implementation of 10 CFR 50.48(c)

19
through participation in NFPA-805

pilot plant visits.

Development and delivery, in

collaboration with EPRI, of training on

NUREG/CR-6850 and conducting

User's Group meetings to provide

joint interpretations of fire PRA issues

such as frequently asked questions

(FAQs) as well as issues arising from

non-pilot-plant application of

NUREG/CR-6850.

FIRE MODELING

Fire models are the phenomenological

basis for fire risk analysis. 10 CFR

50.48(c), and the referenced-2001 NFPA805 Standard requires that "only fire

models that are acceptable to the

authority having jurisdiction shall be used

in fire modeling calculations." NFPA-805

Standard also requires that the fire

models be verified ahd validated, and be

applied only within their domains of

validity. These requirements of NFPA805 create, then, the agency need to

review, assess, and validate fire models.

RES and EPRI sponsored a collaborative

project for verification and validation

(V&V) of selected fire models used by the

nuclear industry. The results of this

project have been documented in

NUREG-1824 (EPRI 01o1999)"

Verification and Validation of Selected

Fire Models for Nuclear Power, Plant

Applications." This research fulfills NRR's

current needs in this area, 'inasmuch as it

provides the necessary V&V for

implementing NFPA-805 Standard and

supports the credibility of fire models used

in the SDP.

NRC does not have adequate models of

fire effects, including the effects of smoke

that may be transported and deposited

substantial distances from the fire

locations. A first step taken by the staff in

the development of fire effects modeling

has been the conduct of a Phenomena

Identification and Ranking Table (PIRT)

process. This process is used to identify

the important phenomena that arise

during fires and judge whether current fire

models adequately simulate or

characterize these phenomena. Future

work will also develop probability

distributions for the uncertainty associated

with each model; these distributions will

be useful for the fire risk assessments.

FIRE TESTING

Confirmatory testing is a critical element

of the fire safety research program. The

ability to determine risk due to fire

damage of instrument, control, and power

cables in nuclear power plants has been a

concern for many years. The issue of

cable hot-shorts (including spurious

actuation) has been a source of

uncertainty for the NRC and licensees. In

order to better understand this issue,

Nuclear Energy Institute (NEI) and EPRI

jointly conducted a series of cable fire

damage tests that were witnessed by the

NRC staff in 2001. Data from these tests,

as well data from previous tests available

in literature, led the NRC to issue

Regulatory Issue Summary (RIS) 2004-

003, Revision 1, "Risk-Informed Approach

for Post-Fire Safe Shutdown Circuit

Inspections." This RIS provides guidance

for NRC inspectors in deciding which

causes of fire-induced hot-shorts are

important to safety and should be

considered during inspections. The RIS

specifically describes four categories of

concerns that should be considered

during inspections (called 'Bin-1 items').

Other issues whose importance still

needed to be determined were also

described in the RIS and are referred to

as 'Bin-2 items'. These included intercable shorting for thermoset cables, intercable shorting between thermo-plastic

and thermo-set cables, configurations

involving three or more cables, multiple

spurious actuations in control circuits, and

prolonged fire-induced hot-shorts that

20
could impair the ability of a plant to experience gained from pilot application

achieve hot shutdown. NRC identified the of NUREG/CR-6850.

need for empirical testing to provide

additional data in certain cable

configurations to support further

development of guidance and modeling

capabilities.

To improve understanding of the Bin-2

items, tests were conducted in 2006 to

examine cable response to live fire. Both

small-scale and intermediate-scale openburn fire tests were conducted. The test

results are documented in two volumes of

a draft report (NUREG/CR-6931)

prepared by Sandia National Laboratories

(SNL). The first volume, "Cable Response

to Live Fire (CAROLFIRE) Volume 1:

General Test Descriptions and the

Analysis of Circuit Response Data,"

contains the small-scale and

intermediate-scale test results for electric

circuit failures (i.e., "hot shorts" data). The

second volume, "Cable Response to Live

Fire (CAROLFIRE) Volume 2: Cable Fire

Response Data for Fire Model

Improvement," contains thermal test data

from these same experiments aimed at

improving cable fire response models. As

a part of CAROLFIRE project, a simple

model was developed to predict thermally

induced electrical failure when a given.

interior region of the cable reaches an

empirically determined threshold

temperature. The description and

validation of the Thermally Induced

Electrical Failure (THIEF) model are

reported in volume 3 of NUREG/CR6931, "Cable Response to Live Fire

(CAROLFIRE) Volume 3: Modeling."

Efforts in the fire safety research area

include the support of fire incident records

exchange. Data from fire events at foreign

nuclear power plants are collected and

analyzed to provide insights into their

causes, means of prevention and

consequence mitigation, and their

applicability to U.S. power plants. The

effort also provides guidance on the

quantification of frequencies based on

21

6. REACTOR FUEL

For nuclear reactors, the fuel integrity is

one of the most important safety

considerations. The fuel and its metallic

cladding are the first barriers in the

defense in depth against inadvertent

release of radionuclides.

Over the last 25 years, nuclear reactor

fuel has evolved from a specialty product

of individual, high-technology vendors to a

commodity product. Indeed, it is not

uncommon now for commercial reactors

to use fuel from different suppliers. The

commodity nature of fuel has affected

research on fuel behavior especially in

off-normal circumstances. Vendor focus

is now on incremental operational

improvements that have the potential to

capture larger fractions of market share.

This focus is especially evident in the

development of new cladding alloys such

as the M5 alloy in which niobium replaces

tin as the main alloying agent. Small

alloying changes can dramatically affect

cladding performance under accident

conditions as is evident in comparing the

performance of the M5 alloy and the

nominally same composition alloy, El 10.

The suitability of new fuel-clad systems

for use in power reactors is assessed

typically by analyses and results obtained

with lead-test assemblies located within

normally operating reactor cores. Seldom

do either vendors or reactor operators

validate predictions of fuel performance

during design basis accident conditions

with experimental data. Experimental

data on the performance of fuel and

cladding systems under severe accident

conditions are never provided by

licensees. Severe accidents, of course,

pose the bulk of the risk to the public

health and safety. Experimental data

Oxidation of a foreign niobiumbearing cladding alloy under LOCA

conditions

obtained in research sponsored by the

NRC have shown that cladding

interactions with the fuel plays a central

role in the rate and extent of core

degradation and fission product release

under severe accident conditions.

The NRC staff and its contractors are in

the process of completing experimental

studies of high-burnup fuel and cladding

behavior under design basis accident

conditions. The staff has carried out inpile tests of fuel behavior during reactivity

insertion events and out-of-pile tests of

fuel behavior under design basis loss of

coolant conditions. These experimental

investigations have been conducted using

an impressive combination of national and

international collaborations. Results of

the research have led to well considered

proposals for changes in the regulations.

23
The proposed changes would make the

regulations more realistic and could

decrease burden on both the staff and

licensees, especially as new fuel and

cladding systems are proposed.

Implementation of these proposed

changes to the regulations has been slow.

Acceptance of the changes to the

regulations now awaits results of

additional testing that could be regarded

as confirmatory in nature.

The staff has also completed revisions of

its fuel performance computer codes.

FRAPCON and FRAPTRAN. These

computer codes are used to

independently confirm analyses done by

the vendors and other licensees. The

modifications allow the computer codes to

be used to evaluate fuels taken to.

burnups of up to 62 GWd/t and to

evaluate the performance of MOX

(plutonia-urania) fuels. MOX fuels will be

used in the Catawba reactor as part of a

DOE program to isotopically dilute excess

weapons-grade plutonium. (The NRC

staff is also examining the severe

accident behavior of high-burnup and

MOX fuels in its severe accident research

program. See Chapter 13 on Severe

Accidents.)

The upgraded fuel performance models

do appear to meet most, near-term

agency needs. However, the staff does

not have the analytical capability to

assess the risk of Pellet Cladding

Interaction (PCI) fuel failures during

anticipated operational occurrences. This

fuel failure mechanism can compromise

fuel integrity as plants continue to uprate

power, increase fuel burnup, and

introduce new fuel designs. The staff

should develop its analytic capabilities in

the near term to assess the fuel

vulnerability to PCI during operational

transients. There is a wealth of

experimental data on the PCI

phenomenon that can be used to develop

and validate a PCI failure model.

Advances in simulation .under way in the

national laboratories are making it

possible to examine fuel performance in

vastly more detail than is done with either

FRAPCON or FRAPTRAN. Whether

such detail is needed will depend critically

on what efforts are made by licensees to

extend fuel burnups beyond the current

regulatory limit of 62 GWd/t and the

amount of experimental data provided to

support these proposed changes to

regulatory limits. There is widespread

expectation that the nuclear industry may

push for extending burnup limits to 85

GWd/t. Certainly, a recent research

strategy document prepared for DOE and

nuclear energy industry proposes that fuel

burnups be extended to 85 GWd/t. There

appears to be some confidence within the

nuclear industry that such extensions of

fuel burnup can be done by extrapolating

the currently available bases of fuel

performance data and models.

Emergence of new physics complicated

such extrapolations of fuel performance

for burnups beyond 40 GWd/t. This

necessitated the experimental research

on fuel behavior under accident

conditions that the NRC is currently

completing. It is not evident that no new

phenomena will arise in connection with

the extrapolation of fuel burnup to 85

GWd/t. Consequently, there will be a

continuing need for the agency to

independently evaluate the safety of

proposed changes in the nature and

burnup limits of reactor fuels.

Lead-test assemblies of MOX fuels have

recently emerged from their first cycle of

irradiation in the Catawba reactor. These

MOX fuels are being tested as part of a

DOE program to dispose of excess

weapons-grade plutonium by using it as

reactor, fuel. MOX fuels have never

before been used in U.S. commercial

nuclear power plants though they are

being used in foreign reactors. It appears

that the staff does not have plans for any

research examinations of these novel

fuels either after the first cycle of

24
irradiation or after subsequent cycles of

irradiation. In light of the limited NRC

experience with MOX fuels, the ACRS

recommends that there be a research

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

assemblies planned by DOE.

Because of license extensions, power

uprates, and the prospect of additional

new reactors, it is anticipated that the

vendors will introduce new fuel-cladding

systems. NRC must develop and

maintain expertise in the area of reactor

fuel to respond to these proposed fuels

and claddings. The challenge in

maintaining what amounts to an essential

core competency of the agency arises

because of the limited availability of

expertise outside of the agency that is

independent of licensees. Current

manpower working for the agency in this

field either directly or by contract is

experienced and there is a need to groom

newer generations in the field. A major

difficulty in doing so is the decline in this

Country of in-pile test facilities and hot

cells for examinations of irradiated fuels

and cladding. Collaborations with

international partners having these

facilities and capable of undertaking

pertinent studies may well be essential for

NRC to maintain an adequate level of

expertise in reactor fuels.

Reprocessing of nuclear fuels may be an

emerging technology that merits research

attention by NRC. Aqueous reprocessing

of irradiated fuels has been done for

many years within the nuclear weapons

community using the PUREX process.

Though familiar, this process has not

been trouble free and there are many

known hazards. NRC has gained some

exposure to the associated safety issues

of "red oil," hydroxylamine nitrate,

ammonium nitrate and the like through its

review of the construction authorization

application for the DOE's MOX Fuel

Fabrication Facility. Modifications of the

PUREX process - the so-called UREX

processes - are being considered by

DOE for its GNEP initiative. Undoubtedly,

safety issues similar to those of the

PUREX process and new safety issues

will arise in these new processes. Even

more challenging will be the

pyrometallurgical reprocessing of spent

nuclear fuel also being considered by

DOE as part the longer-term plans of its

GNEP initiative. The large-scale

implementation of pyrometallurgical

reprocessing of commercial fuels has

never been undertaken in the world. NRC

will need to develop expertise in the

safety of fuel reprocessing technologies

well before licensing applications

involving use of such technologies are

submitted for review. Aqueous

reprocessing of spent reactor fuels is

widely thought to be the first technology

that will emerge for licensing from the

DOE's Global Nuclear Energy Partnership

initiative.

25

7. NEUTRONICS ANALYSIS, CORE PHYSICS, AND CRITICALITY

Both the public and other government

agencies expect the NRC to have expertise in

the fields of neutronics analysis and criticality

safety that is at or very near the state-of-theart. Neutronics analysis is, of course, basic to

the safe design and operation of nuclear

power plants. Shielding analysis and

criticality safety are essential for the safe use

of special nuclear materials. Shielding

analysis and criticality safety will become

more important to the regulatory decisions

especially if the Nation elects to move to the

processing of spent nuclear fuel.

The fundamental phenomenologies of

neutronics, shielding analysis, and criticality

safety are well established, particularly for

issues that arise in connection with the

current fleet of operating nuclear reactors and

nuclear facilities. The current NRC research

in these fields focuses on treatment of

improved data and the development of

computational capabilities to address diverse

circumstances in greater detail. Examples of

issues where improved capabilities have been

needed recently include:

" Burnup credit for spent nuclear fuel

" Neutronics of reactor cores with MOX

(plutonia and urania) fuels for the isotopic

dilution of weapons-grade plutonium

" Nuclear effects of "checkerboard" voiding

of the ACR-700 core

The NRC staff has formed effective

partnerships with investigators at Oak Ridge

National Laboratory and Purdue University to

establish high-quality computational

capabilities to support agency needs in the

areas of neutronics, shielding, and criticality

/1D0zaJ

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SCALES- -OW

NRC Nuclear Analysis Codes for Reactor

Physics

safety. Computational capabilities available

to the staff include:

" AMPX (A Modular Code System for

Processing X-sections)

The AMPX code system has been

upgraded recently to permit fuller use of

the recent ENDF/B-VI database as well

as European and Japanese databases.

* SCALE 5 (Standardized Computer

Analysis for Licensing Evaluation)

This modular set of codes includes:

- ORIGEN-S for radionuclide

depletion analysis

- TRITON (Transport Rigor

Implemented with Time-Dependent

Operation for Neutronic Depletion)

- NEWT (NEW Transport Algorithm)

* PARCS (Purdue Advanced Reactor

Core Simulator)

Coupled with the TRACE thermalhydraulics code, PARCS provides three27
dimensional, multiple group reactor core

neutronics analysis.

These computational capabilities appear to be

adequate for most of the current agency

needs, including core load analyses and the

certification of modern LWR designs utilizing

fuel to burnup levels less than 62 GWd/t.

Though the neutronics analysis codes have

capabilities, they are being challenged by the

growing complexity and asymmetry of new

generations of LWR core and fuel designs.

Also, the codes do not have capabilities to

routinely assess uncertainties associated with

calculated results. The demand for rigorous

uncertainty analyses. is growing within all

fields where computer calculations are used

to support safety analyses and regulatory

decisions. As licensees seek to use ever

more of the margins between current

operations and regulatory limits through the

use of ever more realistic safety analyses, the

need for uncertainty analysis will grow.

Validation of code predictions with

experimental data is quite limited. Validations

are done now using legacy data obtained,

typically, 20 to 30 years ago. Code-to-code

comparisons must increasingly be used to

provide validations of analyses. Neutronics

data suitable for more definitive validation will

be crucial should licensees seek to extend

fuel burnup beyond the current regulatory limit

of 62 GWd/t. In light of the limited and

declining nuclear research capabilities in the

U.S., collaborations with other countries may

be important for the development of new

validation data.

Rather substantial improvements in

neutronics computational capabilities will be

required for analysis of gas-cooled reactors

where epithermal (0.1 to 1.0 Mev) portions of

the neutron spectrum are more important and

graphite geometry effects are complicated.

The NRC staff has outlined well the

anticipated needs in this area and has

identified several ongoing benchmark

activities within the international community.

Because of the very lengthy lead times

associated with computer code upgrades, it

will be useful for NRC to participate in these

international benchmark activities.

Computational capabilities available for

criticality safety assessments appear

adequate for safety review of the MOX Fuel

Fabrication Facility and current licensee

activities. None of these are finding it

necessary to press much beyond current

expectations with respect to criticality safety.

Criticality analysis may be more demanding

for safety analysis of commercial fuel

reprocessing systems such as the aqueous

UREX process being considered in the GNEP

initiative. This modification of the familiar

PUREX process will entail new designs and

materials that have not been evaluated for

criticality issues. An alternative fuel

reprocessing method being considered in the

GNEP initiative involves pyrometallurgy.

Large scale implementation of

pyrometallurical processes for nuclear fuels

have not been done in the past and the

criticality analysis of such processes may

entail substantial advances in the current

state-of -the-art.

Adequate computational capabilities are, of

course, essential to the agency's work.

Manpower skilled in the use of the computer

codes and knowledgeable concerning the

important aspects of neutronics, radiation

shielding, and criticality safety issues are

even more important. It is essential that the

NRC have available to it knowledgeable and

experienced manpower in these areas and

that this manpower maintains awareness of

findings in. the fields beyond just the

improvement of computational capabilities.

Steps taken by NRC management to bolster

the manpower within research assigned to the

areas of neutronics, shielding, and criticality

appear adequate. The challenge the agency

has is to keep manpower current with the

states-of-the-art in these fields.

28
8. HUMAN FACTORS AND HUMAN RELIABILITY

Human performance plays a critical role in the

safe operation of nuclear plants. Operating

experience shows that human performance

issues have been contributors to many

accidents and unsafe conditions experienced

by the current generation of operating plants.

Human performance issues are likely to

continue to have an important impact on plant

safety at existing reactors and for new

designs. The staff needs to be able to

evaluate the treatment of operator action in

risk-informed licensing applications that

include the quantification of human reliability

under accident conditions. The staff will also

need revised guidance and methods to review

new reactor designs that are likely to depend

on a higher degree of automation than current

designs. Therefore, it is essential that the

NRC maintain an expertise in the areas of

human factors and human reliability analysis

(HRA).

Human Factors Research

Human factors research at NRC recently has

been focused on support of the regulatory

process, regulatory applications, and

advanced reactors.

A research project now completed has

resulted in the inclusion of safety culture

reviews in the inspection process. Attributes

of safety culture have been developed and

identified as components of the Reactor

Oversight Process "crosscutting issues."

Detailed guidelines for independent

assessments of safety culture have been

developed, and inspectors are being trained

in their implementation. Inspector feedback is

generally positive. This is likely to lead to

significant improvements in the NRC

inspection program and in the Reactor

Oversight Process.

The project entitled "Impact of Operator

Workload on Human Performance" is focused

on improving the regulatory process. This

effort is intended to assess the impact of

operator overload on performance. The plan

includes the development of licensing

requirements as well as inspector guidance

and techniques for reviewing the impact of

workload on operator performance and plant

safety. This is an important project that could

strengthen the regulatory process and

deserves support both for current reactors

and advanced reactors. The success and

continuation of this project depends on the

cooperation of the industry and this

cooperation may not be forthcoming.

The MOX Fuel Fabrication Facility being

constructed at Savannah River is highly

dependent on operator action during both

normal operation and upset conditions. A

novel, computer-based permissive system

has been proposed for enhancing the quality

of operator actions at the facility. A research

project is in place to provide human factors

review and inputs to the safety evaluation of

the fabrication facility. Because of the

dependency of the MOX Fuel Fabrication

Facility on operator action, this human factors

review is likely to be a critical input to the SER

and deserves to be supported.

Advanced reactor designs are introducing

much greater automation than current

designs. Digital instrumentation and control

systems and new human-system interfaces

are likely to be incorporated into the new

designs. These new features are likely to

affect human performance in different ways in

new designs than in current designs. There is

some concern that errors of commission by

operators may become an issue for the

advanced designs. A project entitled "Human

Factors of Advanced Reactors" is intended to

develop regulatory guidance and analytical

techniques for the review of human factors

issues associated with new and advanced

reactor designs. The ACRS views this project

29
as essential to prepare the staff for its review

of new and advanced reactor designs.

Current research in the human factors area

includes a continuing international

collaborative research program at the Halden

project. As described later in this Chapter,

the Halden Project will host a HRA

benchmark exercise that the ACRS views as

very important. The ACRS continues to be

supportive of this collaborative program and

recommends continued NRC participation.

Human Reliability Analysis

Human reliability analysis remains one of the

important frontiers of PRA. The quantification

of human reliability continues to be a

challenge in risk assessments. Many

techniques have been developed, but

analysts can obtain widely different results for

human failure probabilities. In mid 1980s, the

Ispra Joint Research Center of the European

Commission organized a benchmark exercise

where many teams used a number of HRA

models available at the time to estimate the

probability of the operating crew not

responding correctly to a transient. The

results produced by the teams using the

same HRA model differed by orders of

magnitude. The results produced by a single

team using a number of HRA models also

differed by orders of magnitude.

Although these results are fairly old now, not

much progress has been made to improve

this situation since that exercise was

performed. Many techniques have been

developed and are being used by the industry

to support risk-informed licensing

applications. Even within the NRC, the staff

uses multiple approaches to HRA for actions

following an initiating event. A systematic

comparison of such diverse methods and

their results has not been performed.

Recently, the Commission requested ACRS

to "work with the staff and external

stakeholders to evaluate the different HRA

models in an effort to propose either a single

model for the agency to use or guidance on

which model(s) should be used in specific

circumstances." The staff and EPRI are in the

process of developing a plan that is intended

to lead to an integrated approach to evaluate

various HRA models. The objective should be

to develop a common understanding of the

relative importance of factors affecting human

performance and ways in which they could be

integrated into analyses. Under a project

entitled, "HRA Method Benchmarking Using

Simulator Data," the staff is organizing an

HRA Empirical Study to perform model-tomodel comparison to assess the strengths

and weaknesses of HRA models. Various

operator crews will run scenarios similar to

those appearing in PRAs at the simulator in

Halden, Norway. Teams of analysts will then

analyze the human actions appearing in these

scenarios using different HRA models. The

results will provide insights on the validity of

the assumptions that the teams made and on

how the models were applied.

The HRA Empirical Study by itself will

probably not be sufficient to develop

meaningful quantitative estimates of error

probabilities. Additional evidence needs to be

collected from operating experience. The

research project entitled "Human Event

Repository and Analysis (HERA)" continues

to support the effort to collect and analyze

human performance data. These data can be

used to enhance the insights gained from the

Empirical Study. Continued support of HERA

is important to improve the ability to develop

meaningful quantitative estimates of the

probability of errors and to reduce the large

uncertainties associated with the modeling of

human performance.

Human reliability modeling introduces large

uncertainties in PRAs. The staff needs

guidance in its review of the HRA models

used by the industry in licensing applications.

A Technique for Human Event Analysis

(ATHEANA) is the main NRC tool to estimate

human reliability. It is claimed that ATHEANA

is even capable of estimating reliability with

respect to errors of commission. These are

30
particularly of interest for modern designs that

rely heavily on passive safety systems and do

not require operator intervention in the event

of abnormal events. ATHEANA has been

tested in recent applications to fire safety and

to pressurized thermal shock. The research

project entitled "HRA Application and

ATHEANA Maintenance" will support

evaluation of current HRA models, the

development of an ATHEANA user's guide,

and the development of NUREG-1852,

"Demonstrating the Feasibility and Reliability

of Operator manual Actions in Response to

Fire." This effort is clearly needed to provide

guidance to the staff in reviewing licensing

applications and to demonstrate the strengths

and capabilities of ATHEANA. ATHEANA is a

state-of-the-art model but is complicated to

use. Applications of ATHEANA will ultimately

determine if the benefits of its use outweigh

the disadvantages introduced by its

complexity. The research project on "HRA

Methodology for Fire Analysis" will provide

further application and demonstration of the

effectiveness of ATHEANA.

31

9. MATERIALS AND METALLURGY

Based on the number and visibility of

programs, materials and metallurgy is the

most active area of research within the NRC.

This is appropriate in light of the efforts

required by the agency to address continuing

emergence of unexpected materials

degradation in aging LWRs and to monitor the

effectiveness of aging management

programs. As plants age, known degradation

mechanisms will continue to affect

components important to safety, and new

degradation mechanisms may develop. The

agency must develop the technical

capabilities to minimize the risk of such

surprises and to assess the effectiveness of

the industry initiatives to deal with materials

degradation.

Current materials and metallurgy research

activities are grouped in five areas:

• Environmentally Assisted Cracking in

LWRs

" Steam Generator Tube Integrity

" Non-destructive Examinations

• Proactive Materials Degradation

Assessment

• Reactor Pressure Vessel Integrity

The research activities in the five areas are

appropriate and address key materials issues.

The results of these research activities will

improve the agency's ability to independently

evaluate licensees' efforts to prevent or

mitigate environmentally assisted stress

corrosion cracking and other environmental

degradation mechanisms. RES is making

excellent use of domestic and international

cooperative programs to accelerate progress,

reduce costs, and resolve key issues related

to the detection, understanding, and

mitigation of materials degradation

phenomena. These include the Program for

the Inspection of Nickel Alloy Components,

the Stress Corrosion Cracking and Cable

IL Cra _ 2Sigr

J71 ~

..... 2

4•

ck

tal

k

. . . .. . . II I II I II II I I • I II I .

Schematic view of flaw detection at far

side of weld using phased array

ultrasonic (PA-UT) technique

Aging Program, the OECD Pipe Failure Data

Exchange, the Halden Reactor Project, and

the Cooperative irradiation-assisted stress

corrosion cracking (IASCC) Program (CIR II).

While the research being performed is

laudable, materials research planning

documents should be improved. It is difficult

to determine the priorities, relationships, and

schedule for completion of the various tasks

within the five principal research areas from

documents made available to the ACRS.

33
Environmentally Assisted Cracking in

LWRs

Environmentally assisted cracking is a

complex phenomenon influenced by applied

and residual stresses, water chemistry,

radiation exposure, material composition and

microstructure, and fabrication history.

Unexpected cracking continues to arise in

nuclear power plants as metal components

age and radiation exposure increases. In

recent years, IASCC has occurred in

components internal to the vessels of boiling

water reactors (BWRs) and pressurized water

reactors (PWRs). Stress corrosion cracking

of reactor vessel penetrations and dissimilar

metal welds has occurred in both reactor

types. Although the industry has implemented

improved materials, fabrication processes,

and water chemistries to prevent and mitigate

thesedegradation mechanisms, they continue

to occur in unmitigated components.

Environmentally assisted cracking can lead to

serious secondary damage as observed on

the Davis-Besse control rod drive mechanism

nozzles. It is clear that the NRC staff must

maintain capabilities to evaluate licensees'

analyses of the active degradation

phenomena in their plants, and the

effectiveness of implemented or proposed

mitigation methods. The research projects

now under way are designed to ensure that

the NRC has the necessary technical

understanding of the root causes of the

various environmental degradation

phenomena, their underlying mechanisms,

and the long-term reliability of mitigation

methods. The effectiveness of stress

corrosion cracking mitigation methods used in

BWRs and PWRs should be confirmed by

long-term and aggressive confirmatory

testing. Some methods may lose

effectiveness over time, while others

(individually or in combination) may

demonstrate the ability to protect critical

components from stress corrosion cracking

for the life of the plant.

Environmentally assisted cracking of reactor

materials is an international concern. The

project entitled "CIR-II Cooperative

Agreement" involves NRC collaboration with

the international community to develop a

mechanistic understanding and a predictive

model of IASCC. This understanding is

required to ensure that current mitigation

methods will remain effective as plants age

and, possibly, to identify more effective

countermeasures.

The project on "Environmentally Assisted

Cracking of LWRs" includes tasks addressing

IASCC and primary water stress corrosion

cracking (PWSCC) of nickel based alloys.

Researchers at the Halden Reactor are

measuring IASCC initiation and growth of

relevant materials in both BWR and PWR

environments. This effort includes tests of

neutron-irradiated specimens to improve the

understanding of IASCC initiation and the

influence of stress relaxation on crack growth

and arrest. It also provides data on the

performance of electrochemical potential

probes and monitoring techniques in radiation

environments. This work is essential and

should be continued.

The project on "Investigation of Stress

Corrosion Cracking in Selected Materials" is

intended to develop a better understanding of

the PWSCC mechanism affecting PWRs.

Understanding the root cause and underlying

mechanisms of this phenomenon is essential

for effective long-term mitigation.

Research addressing environmental effects

on fatigue of 'steels used in LWRs has been

completed. NUREG-6909 Rev. 1 and

Regulatory Guide 1.207 have been issued

providing designers and regulators with

quantitative adjustment factors to account for

the effects of environment on fatigue life of

reactor materials.

34
Steam Generator Tube Integrity validated experimentally.

Rupture of steam generator tubes in PWRs

can lead to accidents that allow radioactive

materials released from the core to bypass

the reactor containment and enter directly into

the environment. Severe accidents involving

containment bypass can be risk dominant in

some PWRs. Degradation has been observed

on both primary and secondary sides of

steam generator tubes, and many different

phenomena have been observed including:

general corrosion and wastage, denting,

crevice corrosion, pitting, intergranular attack,

and stress corrosion. Careful water chemistry

control by licensees as well as the

introduction of new tubing materials and

improved designs in replacement steam

generators had mitigated most of these

degradation mechanisms. Stress corrosion

cracking is now the dominant threat to the

integrity of steam generator tubes.

NRC research is concentrated on the

assessment and improvement of nondestructive crack detection methods, the

development of analytical models to predict

initiation and growth of stress corrosion

cracks, and the development of analytical

models to predict leak rates and rupture of

degraded steam generator tubes.

The reliability of current manual eddy current

testing methods used in in-service inspection

is being assessed by Comparing the results of

past human inspector round robin tests on

mockups with known flaws with automated

eddy current testing inspections.

A variety of advanced NDE and signal

analysis techniques are being evaluated for

inspecting original or repaired steam

generator tubes.

An analytical model based on improved

understanding of the mechanisms of initiation

and propagation of stress corrosion cracks in

nickel based alloys, and the influence of

crevice conditions will be developed and

An analytical model to predict leak rates or

rupture of degraded steam generator tubes

under normal or postulated accident

conditions will be developed and validated

experimentally.

Non-Destructive Examinations

Various NDE methods are relied upon to

monitor the integrity of reactor Coolant

systems. These include ultrasonic testing,

eddy current testing, penetrant testing,

radiographic testing, and visual testing. The

reliability and effectiveness of these methods

can vary considerably depending on

component geometry, materials, and types of

defects- There are two primary projects

addressing NDE technology.

The project on "Evaluation of Reliability of

NDE Techniques" is focused on quantifying

the reliability of NDE techniques used in

power plant in-service inspection programs.

This task was initiated in 2007 to evaluate the

accuracy and reliability of the NDE methods

and to provide recommendations to the staff

to improve the effectiveness and adequacy of

in-service inspection programs. This

research, which is expected to be completed

in 2012, covers the entire spectrum of NDE

techniques used in reactor construction, inservice inspection, and repairs. In particular,

the effectiveness of in-service inspection

techniques on materials with coarse-grained

microstructures such as cast austenitic

stainless steels, dissimilar metal welds,

overlays, and claddings will be evaluated.

Because conventional ultrasonic testing is

ineffective for cast materials, promising

advanced ultrasonic testing methods,

including phased array and synthetic aperture

focusing techniques are being evaluated.

The project on "Cooperative Activities Reactor

Coolant System Pressure Boundary

Components" is a shorter-term effort planned

for completion 2008. It is focused on

comparative non-destructive and destructive

35
examinations of control rod drive mechanism

nozzles and J-groove welds recovered from

the North Anna 2 reactor vessel head.

Cracking of vessel head penetrations has

called into question the adequacy of visual

testing methods routinely used during inservice inspection of these components. This

research is testing real components

containing real defects under ideal laboratory

conditions to assess the effectiveness of

various NDE techniques, and to provide

detailed information on flaws in J-groove

welds. Research completed on one

penetration has revealed that eddy current

testing was the most effective technique for

detecting through-wall PWSCC. However,

complexities in component geometry and

PWSCC cracks made interpretation difficult

and there was little overlap in flaw detection

using different NDE techniques. Results of

the in-service inspection of plants, although

not in full agreement with the results of

laboratory tests, did provide sufficient

information to identify the penetration as

needing repair. Future research will gather

additional NDE reliability data by destructively

examining 11 additional crack-like indications

found using eddy current testing techniques.

Based on work completed to date, guidance

has been provided to licensees to perform

more rigorous inspections, and work will be

completed in 2008 to support NRR

rulemaking.

These projects are responsive to the NRC's

needs and should be continued.

Proactive Materials Degradation

Assessment

The nuclear industry and the NRC have often

been surprised by unexpected material

degradation events. The project "Proactive

Material Degradation Assessment" is an NRC

initiative to identify materials and systems in

LWRs where degradation can reasonably be

expected to occur in the future. With such

knowledge, current inspection and monitoring

programs at plants could be reviewed and

modified as needed to provide earlier

identification of incipient degradation before it

affects plant safety.

The staff has completed phase 1 of the

project. A comprehensive assessment of the

likelihood and safety significance of possible

environmental degradation mechanisms has

been completed for approximately 1900 BWR

and PWR components, and NUREG/CR-6923

documenting this work has been issued.

Phase 2 of the project will establish

agreements with industry and international

organizations to define research tasks

addressing the identified issues of greatest

concern. The Zorita Internals Research

Program is such an activity which will

examine high fluence core internal

components which have operated for over 30

years. This research will provide valuable

data on the potential for IASCC in PWR

environments.

This program should be augmented by a

long-term experimental effort to confirm or

correct the most safety-significant predictions

produced in the proactive assessment. In

particular, research should confirm that the

various materials, fabrication and water

chemistry changes that have been

implemented in LWRs to mitigate known

environmentally assisted , degradation

mechanisms will be effective for the life of the

plants. In addition, potential materials

degradation phenomena identified by the

Proactive Materials Degradation Assessment

Program should be confirmed experimentally

in order to justify proactive regulatory actions

and to avoid unexpected failures in operating

plants.

The ACRS admires the vision of this

Proactive Materials Degradation Assessment

Program and supports its continuation.

Reactor Pressure Vessel Integrity

The integrity of the reactor pressure vessels

has been studied for decades. Licensee

obligations to ensure the structural integrity of

36
the reactor pressure vessel during both

routine operations and postulated upset

conditions, are codified in three general

design criteria (GDC 14, GDC 30, and GDC

31) as well as in 10 CFR 50.61 and the

appendices G and H to 10 CFR Part 50. The

technical bases for these requirements were

largely established in the1980s.

Significant progress has been made in

completing and closing six research projects

on reactor pressure vessel integrity. This

completed research has led to updates in

several NRC regulatory documents as well as

ASME and ASTM codes and standards.

Revisions in progress to Pressurized Thermal

Shock (PTS) screening criterion in the PTS

rule and the associated regulatory guides and

Appendices G and H to 10 CFR Part 50 are

likely to provide great benefit by removing

excessive conservatisms in current

requirements and allowing longer life of

reactor pressure vessels.

The staff is now focusing pressure vessel

research to incorporate probabilistic fracture

mechanics in analytical models and to

improve understanding of the embrittlement

mechanisms controlling the properties of

current and future pressure vessels.

The Integrated Component Integrity task will

concentrate on the development and

validation of a modular tool to perform

probabilistic assessments of any reactor

pressure boundary component, Success in

this endeavor will provide the NRC with a

flexible computational tool enabling efficient

analyses of new materials degradation

phenomena such as those observed at Davis

Besse and Wolf Creek. It will also provide a

systematic basis for risk-informed

assessment 'of any pressure boundary

component in new and advanced reactors.

The Integrated Embrittlement task will focus

on the development and validation of a

methodology to systematically identify and

quantify embrittlement mechanisms before

they occur in service. Specific issues to be

addressed include embrittlement predictions

at high fluence, and the influence of nickel

content, flux effects, and attenuation effects

on mechanical properties of current and new

alloys of reactor pressure vessels.

37

10. OPERATIONAL EXPERIENCE

Operating experience provides an

invaluable source of information that is

used by the NRC to identify safety

significant weaknesses in plant design,

operation, or equipment. An important use

of this information is to assess the overall

state of reactor safety and to determine

the effectiveness of the regulatory

process.

Two of the most important ongoing

research activities associated with

operational experience are the Accident

Sequence Precursor (ASP) Program, and

the Industry Trends Program Support.

The ASP is used to monitor the agency's

performance against the NRC Strategic

Plan goals to ensure the industry is

maintaining appropriate levels of safety.

The Industry Trends Program Support

provides trends and other data used to

enhance inspection activities, evaluate

risk-informed applications, evaluate the

need for changes to the regulatory

requirements, and provide the public with

operating trends developed independent

of licensee-sponsored assessments.

These two programs provide important

information necessary to maintain a

strong and effective regulatory process.

Two projects intended to improve the

efficiency and accuracy of the NRC's

significance assessments of findings and

events at nuclear power plants are

"Procedure Development for External

Events" and "SDP/ASP Standardization."

The procedure for external events will

provide generic guidance for the

calculation of risk from external events

(fire and seismic) and evaluation of

findings in ASP and SDP assessments.

The SDP/ASP Standardization effort will

attempt to make the procedures used in

the SDP and ASP programs consistent.

71qrWyi

Uses of Operational Data and

Analyses in NRC Regulatory

Programs

This project will develop analysis

guidelines for operating events during

low-power/shutdown conditions and

calculation of large, early release

frequency (LERF) for all event types for

use in SDP and ASP. Providing

consistent guidance for the two programs

and including more guidance for fire,

seismic, and low-power/shutdown

conditions should enhance the efficiency

and accuracy of the staffs evaluations.

One of the key uses of operational

experience is in the assessment and

improvement of regulatory effectiveness.

In the research project to

"Assess/Improve Regulatory Effectiveness," the operating experience will be

used as one of the tools to determine

whether regulatory changes and other

NRC actions have accomplished their

intended objectives. The effectiveness

reviews and subsequent actions will

support implementation of high level

performance based guidelines, develop

decisionmaking methodologies for NRC

activities, and develop a technical basis to

identify and reduce unnecessary

regulatory burdens. Effectiveness

39
reviews are important in maintaining an

efficient and effective regulatory process.

The scope and the visibility of work to

analyze and evaluate operational data

have dwindled dramatically since the

function of then AEOD was transferred to

RES. Aside from mandated activities

such as ASP Program and Industry

Trends Program Support, the work is

mostly in support of other research efforts

and particular licensing efforts. No longer

are publicly available, in depth analyses

of particular systems or trends being

produced. The resources devoted to this

activity are approaching the minimum

needed to sustain a distinct work unit

within the overall NRC research program.

40
11. PROBABILISTIC RISK ASSESSMENT

In this era of risk-informed regulation, NRC

must have state-of-the-art PRA capabilities.

Certainly, NRC has been responsible for the

development of methods in wide-spread use

today. In recent years, as NRC revises both

its regulations and its regulatory processes to

be risk informed, much of the research work

has been in the area of applications.

Development of methods has not been a

priority. The extensive use of risk information

by both the industry and the staff in regulatory

decisionmaking, the reviews of new reactor

designs, and articles published in the

literature have identified areas where

methodological advances are needed.

There are many computer codes used by the

PRA community to perform the necessary

calculations. The industry uses codes such

as CAFTA, RISKMAN, and RiskSpectrum.

The agency's main tool is the SAPHIRE code

that has been developed in the last 20 years

and is the basis for the SPAR models that are

used extensively in the Reactor Oversight

Process. The ACRS Subcommittee on

Reliability and Probabilistic Risk Assessment

was briefed recently by representatives from

EPRI and an international group of

researchers that have formed the Open PSA

Initiative. Two issues with the present state of

PRA software were identified. First, it is hard

to check the completeness and correctness of

the logic of the PRA models produced using

these codes because these models are code

dependent. Second, there is disturbing

evidence that, under certain circumstances,

the probabilistic calculations may be flawed.

The ACRS was informed that there are efforts

under way to develop a standard PRA

representation format that would promote

independence between the logic models and

the individual codes. Such a standard would

facilitate significantly the review of the PRA

logic, a benefit of great value to the agency.

Pending the implementation of a fully

integrated PRA representation standard, the

agency should explore other methods to

improve the transparency and efficiency of its

PRA review process. Regulatory activities

involve increasingly extensive reviews of

probabilistic analyses for risk-informed

applications at existing plants, and PRAs that

support new reactor licensing efforts.

Substantial improvements in staff review

efficiency, understanding of the plant-specific

PRA models, and reductions in the number of

requests for additional information may be

achieved through implementation of "PRA

viewer" technology. This technology is

currently available in some commercial PRA

software platforms such as RiskSpectrum and

RISKMAN. This technology allows a reviewer

to examine all parts of a PRA including the

event tree model, the fault tree models, the

logic rules, data, results, as well as

uncertainty and sensitivity analyses. The

reviewer cannot change the actual models or

data, and cannot independently re-quantify

the PRA results. However, all facets of the

actual model used to perform a submitted

analysis are directly available, at the

reviewer's desktop. Information of this detail

should allow a capable reviewer to confirm

quickly key technical elements of the models

and data without the need for reference to

cumbersome summary reports or timeconsuming requests for additional supporting

details. The agency should explore the

feasibility of developing this technology,

including necessary protocols to control the

exchange of application-specific PRA models

and data.

Most current-generation codes produce

results such as CDF, LERF, and importance

measures using a set of accident sequences

that is left over after a truncation process in

which sequences with frequencies below a

user-defined cutoff value have been deleted

from the model. Recent studies have

indicated that the choice of the cutoff value

41
may affect the results significantly, especially

the risk achievement worth which is used

extensively in risk-informed decisionmaking

by the agency. New calculation methods,

such as those based on Binary Decision

Diagrams and EPRI's Direct Probability

Calculation (DPMTM), are. exploring the

possibility of eliminating the approximations in

current methods.

In its 2006 report to the Commission on the

NRC Safety Research Program (NUREG1635, Vol. 7), the ACRS stated that "the staff

needs to review the literature concerning

Binary Decision Diagrams and evaluate the

need to adopt this technology." The ACRS

stated further that "the growing importance of

the SAPHIRE code and the SPAR models in

the regulatory process warrants such an

investigation." The ACRS continues to

believe that the staff should undertake such

an investigation and participate actively in the

international activities related to the

development of the next generation of PRA

software. A short-term product could be an

evaluation of the potential impact of the new

software on regulatory decisionmaking. The

results of such an evaluation would guide the

establishment of a longer-term research

project in this area.

The quantification of uncertainties is an

essential element of risk-informed

decisionmaking. The uncertainties in the

values of input parameters to the PRA are

usually handled well and there are many

software packages that propagate these

uncertainties to the output quantities.

Uncertainties in the models themselves due

to questionable or plausible alternative

assumptions are still not included routinely in

PRAs. Regulatory Guide 1.174, Revision 1,

contains a discussion of these uncertainties

but does not provide acceptable methods for

handling them. There is a need for specific

guidance on what we mean by model

uncertainties, how to identify them, and how

to quantify or manage them in

decisionmaking.

Regulatory Guide 1.174,. Rev. 1, states that

the impact of alternative model assumptions

can be investigated through sensitivity

studies. Of course, if all such studies lead to

a conclusion, that the decision is insensitive to

them, the issue of model uncertainty goes

away. If, however, some studies have a

significant impact on the decision, the

analysts would have to evaluate the

uncertainties associated with the underlying

assumptions and factor them in the

decisionmaking process. Such studies are

rarely, if ever, done. The NRC staff should

provide guidance on what a sensitivity study

should entail and how its results ought to be

used.

A systematic approach to the evaluation of

the predictive capability of mechanistic

models is being developed for compartment

fire models. Preliminary results are

documented in NUREG-1 824. The staff is to

be commended for this effort which should

provide insights useful to the general handling

of model uncertainties.

The uncertainty issues discussed above

should be investigated in the context of

regulatory decisionmaking. The ultimate

objective is not the development of the most

rigorous quantification method but, rather, the

management of uncertainty in the integrated

decisionmaking process. The ACRS views

this to be the proper context for these

investigations.

The development of efficient computational

tools for the propagation of uncertainties in

mechanistic models (usually computer codes)

is desirable. This will be particularly important

when the need arises to evaluate the

unreliability of passive systems in advanced

reactor designs. If, for example, one wishes

to use Monte Carlo simulation to propagate

these uncertainties, the long running times

(e.g., hours) of the codes inhibit the rigorous

estimation of output uncertainties. These

issues have been encountered in other

contexts such as the NUREG-1 150 study and

the performance assessments for nuclear

42
waste repositories. The staff should take

advantage of these past efforts.

The development of risk information for

regulatory decisionmaking requires often the

elicitation of expert opinions. A recent

important example is the evaluation of the

frequencies of LOCAs of various sizes in the

context of the efforts to risk-inform 10 CFR

50.46 (NUREG-1829). Another important

study (jointly done with DOE and EPRI and

reviewed by a National Research Council

Committee) focused on expert opinion

elicitation in the assessment of seismic risk

(NUREG/CR-6372). Of course, the first

studies to formalize the use of expert opinions

in nuclear plant risk assessments are

contained in NUREG-1 150 and its supporting

documents. Numerous other examples

involving expert opinion utilization can be

found within the agency including examples in

performance assessments for waste

repositories, development of phenomena

identification and ranking tables, and the

ATHEANA model for human reliability

assessment. Few of these studies build on

previous NRC-sponsored approaches. The

staff should combine the best attributes of

these studies and develop a systematic

methodology for the elicitation and processing

of expert opinions that should be used

agency-wide. If there is a need for parts of

the general methodology to be different for

different applications, these differences

should be identified.

PRA issues related to HRA and digital

instrumentation and control are discussed in

Chapters 4 and 8 of this report.

43

12. SEISMIC AND STRUCTURAL ENGINEERING

Seismic safety in the design, construction,

and operation of nuclear facilities has been an

,ongoing and evolving issue since shortly after

the inception of civilian nuclear facilities.

Probabilistic risk analyses show that seismic

events can be important contributors to the

overall risk of both current and advanced

nuclear reactors.

Initially, seismic hazards were addressed by

"deterministic" regulation and guidance. In

the early and mid-1990s, however, the NRC

began to evolve its siting and design review

processes toward the use of a "probabilistic"

seismic hazard approach that could better

address both aleatory and epistemic

uncertainties. Probabilistic seismic hazard

analyses (PSHAs) are performed by the NRC

in order to independently validate seismic

hazard levels provided in nuclear power plant

applications. These assessments have revealed

differences between the results of NRC

analyses and those performed by industry.

These differences are attributed to differences

in seismic source characterization, ground

motion prediction, and approaches to the

PSHA process. Research is planned to

provide the technical basis for resolving these

differences and developing appropriate

guidance.

Seismic source characterization is a key area

for the seismic research program because it

is a major contributor to the uncertainty in the

calculation of seismic hazards, particularly in

geographical areas that tend to have rare

(although possibly large) events and have

limited seismic instrumentation such as the

central and eastern U.S.

Current research related to seismic source

characterization includes a study of the

maximum magnitude (Mm.) appropriate for

seismic sources in the central and eastern

U.S. and further characterization of the East

Tennessee Seismic Zone (ETSZ). The ETSZ

issue stems from the review of early site

permit application of Vogtle. Work is also

planned to enhance the understanding of the

seismic processes and characteristics of the

New Madrid and Charleston earthquake source

zone. Much of this work involves cooperative

efforts between the NRC and the United

States Geological Survey (USGS). These

research efforts are well leveraged and

important for providing the data needed by

the staff for an independent review of the

seismic source parameters used by

applicants.

The prediction of ground motions for a given

magnitude and distance has always

constituted a significant source of uncertainty

in seismic hazard results. Uncertainty in

these relationships can lead to discrepancies

in the hazard levels. A very successful

collaborative effort, known as the "NextGeneration Attenuation Relationship" project

or the "NGA-West" project produced a set of

consensus relationships that are now viewed

as the state-of-the-art. Follow-on work is

planned on an "NGA-East" project. One result

of the revaluation of seismic hazards for rock

sites has been predicted motions in the highfrequency range that exceed the enhanced

spectrum in Regulatory Guide 1.60, "Design

Response Spectra for Seismic Design of

Nuclear Power Plants," used for some

certified nuclear power plant designs. The

effects of these motions depend on the

ground motion, i.e., the degree to which two

points in the soil are moving in phase at each

frequency. Methods for incorporating

incoherency effects into seismic design of

nuclear facilities to address the effects of

high-frequency motion have been proposed

by the industry. Although significantwork has

been undertaken to review the proposed

methodology (as part of the development of

Regulatory Guide 1.208, "A Performance45
Based Approach to Define the Site-Specific

Earthquake Ground Motion," and updates to

the related Standard Review Plan sections), a

more comprehensive validation program is

warranted.

In an effort to standardize approaches to

PSHA, the NRC sponsored the research

documented in NUREG/CR-6372, "Senior

Seismic Hazard Analysis Committee

(SSHAC) Recommendations for Probabilistic

Seismic Hazard Analysis: Guidance on

Uncertainty and Use of Experts" (referred to

as the "SSHAC guidelines"). This document

provides guidance for performing PSHA using

four different levels of complexity depending

on project needs.

Although the SSHAC guidelines provide a

framework for PSHAs, the document does not

describe in detail how to conduct PSHAs.

Subsequent to the publication of NUREG/CR6372, practical experience in conducting

PSHAs in accordance with the SSHAC

guidelines has been gained at Yucca

Mountain, in the Swiss PEGASOS Project,

and in the EPRI CEUS Ground Motion

Project. This experience has not been

captured in a form that could benefit an

organization conducting or reviewing a major

PSHA effort. Development of additional

guidance based on this experience is planned.

Recently, the NRC and the nuclear industry

have accepted a risk-informed, performancebased approach to the definition of sitespecific earthquake motion. The American

Society of Civil Engineers (ASCE) has

provided some of the technical basis for the

performance-based approach through its

Standard ASCE 43-05, "Seismic Design Criteria

for Structures, Systems, and Components

in Nuclear Facilities." Although the staff has

developed Regulatory Guide 1.208, for the

performance-based approach to be

acceptable, additional research is planned to

complete the development of the technical bases

for reviewing a performance-based approach.

The Sumatran earthquake in December 2004

(magnitude =9) and the associated

devastating Indian Ocean tsunami focused

considerable attention on structures and

facilities that are sited on or close to the

coastline. The intensity of an extreme

tsunami event could exceed known historical

events considered in the design bases of

nuclear power plant structures or other nuclear

facilities located close to the coastline. In the

past, design of coastal facilities did not

explicitly address sources known as

"submarine landslides," which can trigger

extreme tsunami events. The NRC is

currently working with the National Oceanic

and Atmospheric Administration (NOAA) and

the USGS to review the existing state-ofknowledge for the tsunami hazard

assessment, mitigation, and landslide

mechanics. This collaboration has produced

an excellent assessment of the state-ofknowledge concerning submarine landslides

that could affect future nuclear power plant

sites on the east coast of the U.S.

The NRC routinely participates in cooperative

research activities with the International

Atomic Energy Agency (IAEA), the

Organization for Economic Cooperation and

Development (OECD) and its Nuclear Energy

Agency (NEA), and other organizations to foster

the exchange of data and analysis results. In

addition to ongoing activities, the recent

earthquake at the Japanese KashiwazakiKariwa (K-K) Nuclear Power Plant provides a

unique case history and an opportunity to

assess the accuracy of analytical tools that

are typically used in seismic design of nuclear

power plants and other nuclear facilities.

In August 1999, the NRC signed a 5-year

collaborative agreement with the Japanese

Agency for Natural Resources and Energy,

Nuclear Power Engineering Corporation

(NUPEC). The collaboration was extended for

an additional 5 years (expiring in August 2009)

with the'successor to NUPEC, the Japan

Nuclear Energy Safety Organization (JNES).

46
The research conducted and in-kind

information exchanged under this collaborative

agreement leverages NRC resources to

obtain earthquake test data for scale-model

structures, multi-axial shear walls, piping, and

equipment fragility tests. Collaboration with

JNES has proven to be an economical way for

the NRC to obtain the results of large-scale test

programs that are not available anywhere else

in the world. In addition, this research program

provides opportunities to interact and

exchange information with other Japanese

organizations, thereby ensuring that the NRC

staff remains cognizant of all ongoing seismic

research in Japan.

47

13. SEVERE ACCIDENTS

In the last three decades, the NRC has

conducted substantial research in severe

accident phenomena through experiments,

data analysis and evaluations, as well as the

development of an integrated, systems-level

computer code (MELCOR) for use in

analyzing severe accident progression. The

NRC requires such expertise and analysis

capabilities to help support regulatory

decisions for current and future nuclear power

plants as well as to help the staff in its

transition to a more risk-informed regulatory

framework.

Support of regulatory needs implies sufficient

understanding of severe reactor accidents to

estimate whether risks reach a level to

threaten adequate protection of the health

and safety of the public. The NRC invested

heavily in severe accident research to

achieve the needed understanding. Once this

need was met, the NRC curtailed its

investments in severe accident research to

needed levels for on-going analysis and riskinformed activities.

Severe accident research is continuing

internationally, with substantial experimental

programs in the Pacific Rim (Japan and

Korea) as well as Europe (France and

Germany). The NRC has developed an

effective strategy to maintain its leadership in

severe accident analytical capabilities as well

as to maintain its state-of-the-art knowledge

and understanding of key severe accident

phenomena by collaborations in international

research programs. The knowledge gained

from NRC's past experimental work and

current ongoing international efforts are

systematically incorporated into MELCOR

accident analysis code. In addition, the NRC

staff has entered into a number of

international cooperative experimental

research programs to obtain key data for

validating the MELCOR code as well as

improving its accuracy and realism.

PHEBUS-FP

MELCOR Code Development and Usage

The MELCOR code was originally developed

as a PRA tool, and thus, models the reactor

coolant system and safety systems as well as

containment systems in a less detailed

manner than more mechanistic thermalhydraulic and fuel rod models. A major

current activity is to consolidate the key

physical models and capabilities of more

detailed severe accident codes into

MELCOR. This effort will provide an efficient

state-of-the-art code for severe accident

analyses. In addition, the MELCOR code has

been adopted by a number of international

nuclear safety organizations (e.g., IBRAERussia, PSI - Switzerland) and universities

(e.g., Purdue, Texas A&M) for use in severe

accident analyses. Finally, the MELCOR code

and its fission product source term transport

and consequences model (MACCS) is being

used as an integral part of the SOARCA

project. MACCS is a widely accepted tool

used for consequence analysis and its

continued development and support is

important to the agency for both safety and

security regulatory applications. The ACRS is

supportive of this effort.

49
Collaborative Severe Accident

Experimental Programs

Collaborative severe accident research

programs that the NRC has joined are making

good progress and some of the key

accomplishments are noted below.

ARTIST: This program involves an

experimental study being completed at

the Paul Scherrer Institute (PSI) in

Switzerland. The objective is to measure

the aerosol removal on the secondary

side of steam generators during accidents

in a PWR, given that containment bypass

occurs due to a steam generator tube

failure. Such bypass accidents are often

risk dominant for PWRs. Based on initial

results, the key areas where additional

data are needed are: gas jet behavior

across the tube bank; particle

resuspension as well as particle inertial

impaction, and turbulent deposition under

certain flow conditions.

MASCA: The MASCA experimental

program as well as RASPLAV, its

predecessor, was begun in Russia at the

Kurchatov Institute to understand the

behavior of reactor core debris in the

lower plenum of the reactor pressure

vessel. The technical objective is to

determine if the core debris can be

retained within the reactor pressure

vessel with water flooding on the outside

of the vessel. The major experiments

have been completed. Results of the

experiments are now being incorporated

into the MELCOR computer code. There

is an interest in maintaining capabilities

for follow-on experiments that may be

needed for advanced reactor designs. No

commitment to such a follow-on program

has been made.

OECD-MCCI: This program is an

international collaborative experimental

study being conducted at the Argonne

National Laboratory and is focused on

investigating the heat transfer to an

overlying water layer to cool molten core

materials interacting with concrete and its

viability for long-term coolability. The

MCCI project has completed a second

phase of experimentation and consists of

two experimental efforts: Small-scale

Water Ingression and Crust Strength

Tests and Core Concrete Interaction

Tests.

The Small-Scale Water Ingression and

Crust Strength tests focus on quantifying

the heat loss to an overlying water layer

by mechanisms other than conductionlimited heat transfer, and to measure the

strength of the crust formed during

flooding of the melt. The Core Concrete

Interaction tests are focused on resolving

uncertainties in axial versus lateral power

splits and respective concrete ablation

rates. The tests attempt to replicate as

closely as possible conditions at plant

scale and provide data to verify and

validate predictive codes. These tests

allow water flooding after partial ablation

to obtain debris coolability data at later

stages in the MCCI process. Results of

the tests need to be incorporated into the

MELCOR computer code.

PHEBUS-FP: This program consists of

five large-scale, in-pile integrated tests of:

fuel degradation; fission product release;

radionuclide transport through a model of

a reactor coolant system; and aerosol

behavior in a model containment. These

tests have been designed to allow

validation of core degradation and fission

product release and transport models.

Although these tests have been

completed in 2006, data analysis and

associated computer simulations'are still

on-going. These experiments have

proved to be important in providing

physical insights on fission product

behavior, validating MELCOR modeling of

radionuclide release from degrading

reactor fuel, and validating the alternative

,source term used in 10 CFR Part 100

plant analyses. NRC has also joined a

50
second phase of the program that will

conduct separate-effects tests to better

understand key source term phenomena.

For example, this program will address

the containment chemistry of radioactive

iodine, as well as investigate fission

product release from mixed-oxide fuels

and high-burnup fuels under accident

conditions.

The ACRS encourages the strategy that the

NRC staff has developed to support

regulatory decisions for severe accidents via

computer code development and

experimental data analysis and evaluation.

This approach can successfully maintain and

update its modeling capabilities for severe

accident analyses. The planned program

extensions and continuations of these

collaborations are well worth the investment.

This type of collaboration in experimental

research could be emulated in other NRC

research areas such as thermal-hydraulics

and digital instrumentation and control.

51

14. THERMAL HYDRAULICS

Thermal-hydraulics research aims at

improved prediction of pressures,

temperatures, and flow rates in nuclear

systems and containments during normal

operating and accident conditions. Such

predictions require understanding and

modeling of flow and heat transfer in complex

geometries. The flows may involve more than

one phase, for example, vapor-liquid or liquidsolid flows. Of particular importance in the

regulatory process is prediction of thermalhydraulic conditions such as fuel clad

temperatures and containment pressures in

anticipated transients and postulated

accidents. Independent evaluation of

licensee's thermal-hydraulic analyses has

long been a key element in NRC design

certification process.

Historically, early thermal-hydraulic analyses

of nuclear power plants employed very

conservative bounding assumptions, assuring

large safety margins with regard to allowable

temperatures and pressures over a wide

range of accident and operating conditions.

With time, the database of experimental

findings and the confidence in analytical

predictions have grown, allowing the NRC to

consider submittals from licensees employing

best-estimate thermal-hydraulic analyses

together with estimates of uncertainties. As

these analyses have grown ever more

sophisticated, it has been necessary for the

NRC to continue development of state-of-theart thermal-hydraulic computational tools and

more sophisticated understanding of

important thermal-hydraulic phenomena. To

this end, the NRC maintains competence in

the thermal-hydraulics field and capability to

conduct confirmatory analysis through its

research program.

OECDINEA ROSA Program

In 2005, under the auspices of the OECD-NEA,

the NRC joined 19 other international

organizations in an agreement with the Japan

Atomic Energy Agency (JAEA) to conduct thermalhydraulic experiments using the ROSA-V

configuration of the Large Scale Test Facility

(LSTF), a full-height, 1/48 volume-scaled model of

a four-loop pressurized water reactor.

The major elements of the current NRC

thermal-hydraulics research program fall into

three general areas:

* The TRACE computer

development and validation

code

* Experimental studies of thermalhydraulic phenomena

53
* Thermal hydraulics of sump screen

blockage

There are also smaller efforts directed

towards maintenance of legacy codes such

as RELAP5, participation in international

benchmarking activities foW thermal-hydraulic

sub-channel codes, and addressing BWR

ECCS suction strainer concerns arising from

possible gas entrainment (GSI-193).

The main features of the first two elements of

the thermal-hydraulics research program are

discussed below. The issue of sump screen

blockage is being actively followed by the

Committee. The ACRS has been providing

reports on the technical approach and

progress of research in this area.

TRACE Computer Code Development and

Validation

In the mid-1990s, a prudent decision was

made that the several primary reactor system

thermal-hydraulic codes that were in use at

that time be consolidated into a single code.

The several codes included RELAP5 (for

LOCA), TRACP (for PWR LOCA), TRACB

(for BWR LOCA), and RAMONA (for BWR

stability).

The models, correlations, and solution

methodologies in these codes did not reflect

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

review and modification. It was also

recognized that they had been designed at a

time when computer capabilities were limited

and included many structural aspects, such

as memory management, that were no longer

needed and increased the cost of continued

code maintenance and development. The

availability of graphical user interfaces and

their wide acceptance also suggested the

desirability of incorporating similar capability

into the NRC codes. All these considerations

led to extensive code consolidation, model

improvements and implementation efforts,

culminating in the development and validation

of the TRACE computer code.

TRACE is intended to serve as the main tool

for the confirmatory analyses of a broad

range of thermal-hydraulic problems for

current and future reactor designs. It has the

potential to offer significantly enhanced

capabilities for state-of-the-art analyses of

thermal-hydraulic issues. Several important

technical issues, such as core stability and

ATWS behavior, involve coupling between

neutronics and thermal hydraulics requiring

that TRACE be properly coupled to neutronics

codes like PARCS, an activity that is still

ongoing. TRACE also has the capability to

interface with the CONTAIN code for

containment response analysis as well as

with other computational tools, including

MATHCAD. Applications include analyses in

support of certification of new reactor designs

and the regulatory review of power uprates for

currently operating reactors.

Much of the work in the period under review

here has gone towards documentation,

validation, and the peer review of TRACE. All

these are current work items, and high priority

is being given to completing these tasks

expeditiously. Efforts are also being made to

use TRACE in support of the reviewing the

ESBWR and EPR design certification

applications. These designs have some novel

thermal-hydraulic features that may challenge

TRACE capabilities.

The staff has responded positively to the

previous ACRS recommendations related to

TRACE, and the required activities are now

under way, including initiation of the peer

review. It remains to be seen whether the

work will be completed on a schedule that

enables incorporation of the TRACE code into

the regulatory process, particularly with

regard to reviewing several power uprates

and design certification applications currently

under consideration.

Experimental Studies of Thermal-hydraulic

Phenomena

Thermal-hydraulic phenomena involved in

normal and accident conditions for LWRs are

54
complex, and often involve the difficult-tomodel flow of two-phase mixtures (steam and

water). Predictions from computer codes of

such phenomena need extensive

experimental validation, and there are many

effects, such as those involving

multidimensional flows in complex geometries

where large-scale experiments are the

primary means of confirming the validity of

these predictions. In view of this, NRC-RES

has maintained several large, -complex

experimental facilities:

* APEX facility at Oregon State

University for PWR-related problems

* PUMA facility at Purdue University for

BWR-related problems

* RBHT facility at Penn State

University for PWR emergency core

cooling problems

In parallel, the NRC has collaborated with

international groups in undertaking

experiments in facilities abroad, notably the

Japanese ROSA facility.

APEX is a medium-sized, reduced-height test

facility that has been used in the certification

of the AP600 and the AP1000 reactor

designs. In the period under review, it has

been used to assess the predictive capability

of thermal-hydraulic codes for heat transfer in

steam generators under partially voided

primary system conditions, particularly reflux

condensation in the steam generator. The

facility is also being used to study

condensation on the primary side of steam

generators. This is of interest for the EPR

design, which incorporates controlled

secondary-side cool-down for pressure

reduction in small-break LOCAs.

Since APEX is a reduced-height facility,

properly scaled natural convection tests are

difficult to define. Consequently, the NRC is

participating in the OECD ROSA program to

study steam generator behavior during

depressurization and in natural circulation

conditions. The full-height ROSA facility is

able to provide both integral and separateeffects experimental thermal-hydraulic data.

NRC is collaborating in OECD/NEA programs

that also cover a range of other important

thermal-hydraulic phenomena, such as

thermal stratification in horizontal legs during

emergency core coolant injection, the effects

of water hammer phenomena, and smallbreak LOCA without scram. Such

collaborative efforts are to be commended, as

they take advantage of facilities that are of a

scale and capability that do not currently exist

in the U.S. Furthermore, they draw on the

expertise of international partners, who have

continued to maintain a very high level of

capability in thermal-hydraulics field.

The PUMA facility is also of reduced height,

and is being used to perform integral LOCA

tests of interest for the ESBWR design. In

particular, the beyond-design-basis passive

safety system response of the ESBWR

design is to be evaluated and provide data for

TRACE assessment.

In addition, an exploratory research program

to develop so-called "closure relationships" for

the evolution of interfacial area in two-phase

flows is being undertaken at Purdue

University. It is expected that when the data

encompass the range of flow regimes

expected in two-phase flows, then a model of

interfacial area evolution will be incorporated

into the TRACE code, potentially improving its

accuracy and reliability in the next-generation

version. Results from this program have been

slow in coming and the strategy for utilizing

them in TRACE still remains to be elucidated.

The RBHT facility at Penn State University

was developed to address issues related to

emergency core cooling, including the

development of a better understanding of

reflood and rewetting in realistic, bundled

geometries. Currently, this facility is being

used to provide data for the development of

spacer grid models for the TRACE computer

code. The effects on heat transfer of both

entrained drops and non-condensable gases

are also being determined.

55
It is clearly desirable that the NRC maintain

capability to assess thermal-hydraulic

phenomena in both existing and future reactor

designs. In view of this, maintenance of

research programs in U.S.. experimental

facilities is justified. Participation in

international programs utilizing large-scale

facilities abroad is also to be encouraged.

The advantage of using the full-height

international facilities is that the

understanding and correlations developed

can be applied to reactor systems more

reliably and do not depend on scaling

analyses required for reduced height

systems, which are difficult to develop and

validate for two-phase flows.

The accelerated efforts to document, validate,

and peer review TRACE and enable its

incorporation for confirmatory analysis in the

regulatory process is laudable. The review of

several BWR extended power uprate

applications require coupled

neutronic/thermal-hydraulic confirmatory

analyses, both for operational and ATWS

instability evaluations. These will prove to be

a challenge in the near future. Furthermore,

future design certifications will require the

capability to finely nodalize riser structures in

steam generators and chimneys to capture

oscillatory phenomena and refluxing which

will necessitate excellent numerical stability

and computational efficiency.

In this regard, it should be noted that largescale parallel computer clusters are becoming

available at low cost. Undoubtedly, licensees

and vendors will take advantage of such

capabilities using their own codes. It is

therefore desirable that the NRC also institute

long-term programs to take advantage of

massively parallel distributed memory

systems to speed run times and allow finer

resolution in the numerical models used for

confirmatory analysis. The development of

codes to be scalable on large, parallel

machines is therefore highly encouraged.

The ACRS anticipates that licensees will

increasingly employ multidimensional

computational fluid dynamic (CFD) capability

to resolve problems which the current

generation of thermal-hydraulic codes, such

as TRACE, are unable to do.

Multidimensional effects are important in

downcomers, plena, and in large pipes in the

vicinity of discharge locations, e.g. automatic

depressurization systems. The NRC currently

has very little effort in the area of CFD other

than some testing of commercial CFD codes.

Although these are used in the process

industry for qualitative indications of

phenomena, they are validated to a much

less rigorous standard than codes for nuclear

use, and -the source codes are not available.

They appear to include a number of ad hoc

fixes to improve stability and robustness

which may affect their predictive capability for

situations that cannot be studied

experimentally. NRC has proposed in the

NPHASE effort to develop a CFD model

suitable for regulatory applications. The

ACRS questions the feasibility of developing

such a code within the resource constraints of

the agency. The international nuclear

community, on the other hand, has instituted

the development of multidimensional CFD

capabilities. CFD codes are being developed

and validated to standards of reliability and

accuracy required for use in nuclear systems

evaluations. NRC should join in this effort and

provide adequate resources to allow

productive participation.

CFD analyses are already being submitted for

a number of issues such as particle drop out

-during containment debris transport following

loss-of-coolant accidents. The NRC has no

confirmatory capability to check such

submittals. Such capabilities will also

undoubtedly be needed for advanced

reactors, such as high-temperature gascooled reactors, and for evaluation of new

fuel designs.

The experimental facilities being maintained

by the NRC are aging and require many

assumptions regarding scaling to interpret the

data obtained due to their reduced-height

design. To some extent, this problem can be

56
alleviated by participation in international

programs in which full-height facilities are

utilized. NRC participation in such programs

is therefore to be encouraged and should be

increased in the future.

57

15.- ADVANCED NON-LWR DESIGNS

Advanced reactors using gas-cooling or liquid

metal-cooling rather than conventional light

water cooling pose a conundrum for NRC and

the planning of its research efforts. The nonLWRs of interest to NRC can be categorized

as liquid metal-cooled reactors (LMRs) and

high temperature gas-cooled reactors

(HTGRs). The Toshiba 4S (Super Safe,

Small, and Simple) is a LMR. In addition,

LMR technology is being considered for the

Advanced Burner Reactor (ABR) proposed

under the DOE GNEP initiative. The Next

Generation Nuclear Plant (NGNP) prototype

will also utilize HTGR technology when the

concept is developed. NRC has some

experience with these alternative reactor

technologies, but does not have in place the

technical infrastructure that will be needed to

support an effective certification review of a

reactor based on such technologies. NRC is

currently considering how to address the

evolving demands related to non-LWRs, IRIS,

and other unique designs. The Office of New

Reactors (NRO) is dubious of the feasibility of

developing this technical infrastructure and

conducting research in parallel with a

certification review. The ACRS certainly

agrees with the NRO on this point. It is clear

that it will take some time to develop the

needed technical infrastructure at NRC. What

is not clear is when applicants might be able

to submit a certification application with an

adequate technical basis. This makes it

difficult to assign priority and allocate

resources for advanced non-LWR research.

The research staff has identified well the

technical issues associated with the

certification of a gas-cooled reactor. They

have done this based on an expert opinion

elicitation process that identified important

phenomena and ranked these phenomena

with respect to safety importance and the

current level of understanding. It will be

important for NRC to use results of the

phenomena identification and ranking to

communicate expectations for a defensible

certification application for a gas-cooled

reactor. Especially crucial will be defining

areas where analyses must be substantiated

by pertinent, prototypical experimental results.

NRC should initiate activities to assess the

state of knowledge in other technologies

before there is further loss of limited available

capability (worldwide) in some of these

technologies. The state-of-knowledge report

should identify significant gaps in knowledge

that might be necessary for future licensing of

some of these unique designs. Specific focus

should be given to understanding severe

accident behavior and crucial areas of safety

(e.g., reactor fuel, high temperature

materials). This effort can provide a significant

input to the development of safety

requirements for some of these designs.

Multinational cooperation at an early stage

could lead to common safety standards and a

sound and more cost-effective research

program

59

16. REFERENCES

1. U.S. Nuclear Regulatory Commission,

"Review and Evaluation of the Nuclear

regulatory Commission Safety Research

Program," Advisory Committee on

Reactor Safeguards (ACRS), NUREG1635, Vol. 6, March 2004.

2. U.S. Nuclear Regulatory Commission,

"Review and Evaluation of the Nuclear

regulatory Commission Safety Research

Program," Advisory Committee on

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

3. Idaho National Laboratory/Nuclear Power

Industry, "Strategic Plan for Light Water

Reactor Research and Development: An

Industry-Government Partnership to

Address Climate Change and Energy

Security," INL/EXT-07-13543, November

2007.

4. Staff Requirements Memorandum, Dated

November 8, 2006, Subject: Meeting With

Advisory Committee on Reactor

Safeguards, October 20, 2006.

5. Staff Requirements Memorandum, Dated

June 22, 2007, Subject: Meeting With

Advisory Committee On Reactor

Safeguards, June 7, 2007.

6. Memorandum to The Commissioners

from Luis A. Reyes, Executive Director for

Operations, Subject: SECY-07-0068,

"Candidate Agency Long-Term Research

Activities for Fiscal Year 2009," April 6,

2007. (Official Use Only Document -

Sensitive Internal Information - Limited to

NRC Unless the Commission Determines

Otherwise)

7. Report Dated May 16, 2007, from William

J. Shack, Chairman, ACRS, to Dale E.

Klein, Chairman, NRC, Subject:

Development of An Integrated Long-Term

Regulatory Research Plan.

8. Staff Requirements Memorandum, Dated

November 8, 2006, Subject: Meeting With

Advisory Committee on Reactor

Safeguards, October 20, 2006.

9. U.S. Nuclear Regulatory Commission,

"Cable Response to Live Fire

(CAROLFIRE) Volume 1: Test

Descriptions and Analysis of Circuit

Response Data,"NUREG/CR-6931 ,Vol. 1,

SAND2007-600N1, Final Prepublication

Draft, December 2007.

10. U.S. Nuclear Regulatory Commission,

"Cable Response to Live Fire

(CAROLFIRE) Volume 2: Cable Fire

Response Data for Fire Model

Improvement," NUREG/CR-6931, Vol. 2

SAND2007-600N2, Final Prepublication

Draft, December 2007.

11. U.S. Nuclear Regulatory Commission,

"Cable Response to Live Fire

(CAROLFIRE) Volume 3: Modeling,"

NUREG/CR-6931, Vol. 3, NISTIR 7472,

December 2007.

12. U.S. Nuclear Regulatory Commission,

"Demonstrating the Feasibility and

Reliability of Operator Manual Actions in

Response to Fire," NUREG-1852,

October 2007.

13. U.S. Nuclear Regulatory Commission,

"Effect of LWR Coolant Environments on

the Fatigue Life of Reactor Materials,"

NUREG/CR-6909, ANL-06/08, February

2007.

14. U.S. Nuclear Regulatory Commission,

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," March 2007.

61
15. U.S. Nuclear Regulatory Commission,

"Expert Panel Report on Proactive

Materials Degradation Assessment,"

NUREG/CR-6923, BNL-NUREG-77111-

2006, February 2007.

16. Code of Federal Regulations, Title 10,

Part 50, Appendix A, General Design

Criteria for Nuclear Power Plants.

17. Code of Federal Regulations, Title 10,

Part 50, Section 61, Fracture Toughness

Requirements for Protection against

Pressurized Thermal Shock Events.

18. Code of Federal Regulations, Title 10,

Part 50, Appendix G, Fracture Toughness

Requirements.

19. Code of Federal Regulations, Title 10,

Part 50, Appendix H, Reactor Vessel

Material Surveillance Program

Requirements.

20. U.S. Nuclear Regulatory Commission,

Regulatory Guide 1.174, "An Approach for

Using Probabilistic Risk Assessment in

Risk-Informed Decisions on Plant-Specific

Changes to the Licensing Basis,"

Revision 1, November 2002.

21. U.S. Nuclear Regulatory Commission,

"Verification and Validation of Selected

Fire Models for Nuclear Power Plant

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

EPRI 1011999, May 2007.

22. U.S. Nuclear Regulatory Commission,

"Severe Accident Risks: An Assessment

for Five U.S. Nuclear Power Plants,"

NUREG-1 150, December 1990.

23. Code of Federal Regulations, Title 10,

Part 50, Section 46, Acceptance Criteria

for Emergency Core Cooling Systems for

Light-Water Nuclear Power Reactors.

24. U.S. Nuclear Regulatory Commission,

"Estimating Loss-of-Coolant Accident

(LOCA) Frequencies Through the

Elicitation Process," NUREG-1829, Draft

Report for Comment, June 2005.

25. U.S. Nuclear Regulatory Commission,

"Recommendations for Probabilistic

Seismic Hazard Analysis: Guidance on

Uncertainty and use of Experts,"

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

1997.

26. U.S. Nuclear Regulatory Commission,

Regulatory Guide 1.160, "Monitoring the

Effectiveness of Maintenance at Nuclear

Power Plants," Revision 2, March 1997.

27. U.S. Nuclear Regulatory Commission,

Regulatory Guide 1.208, "A PerformanceBased Approach to Define the SiteSpecific Earthquake Ground Motion,"

March 2007.

28. Code of Federal Regulations,

Part 100, Reactor Site Criteria.

Title 10,

29. Generic Safety Issue (GSI)-193, "BWR

ECCS Suction Concerns."

30. American Society of Civil Engineers and

Structural Engineering Institute,

ASCE/SEI Standard 43-05. "Seismic

Design Criteria for Structures, Systems,

and Components in Nuclear Facilities,"

2005.

62
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(9-2004) (Assigned by NRC, Add Vol., Supp., Rev.,

NRCMD 3.7 and Addendum Numbers, If any.)

BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse) NUREG-1635, Vol. 8

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED

Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program - A MONTH YEAR

Report to the U.S. Nuclear Regulatory Commission March 2008

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

Hossein Nourbakhsh Technical Report

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor,

provide name and mailing address.)

Advisory Committee on Reactor Safeguards

U. S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type 'Same as above'; if contractor, prOvide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission,

and mailing address.)

Same as above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

.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 (RES). 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 missions. The report also addresses the issue of long-term sustained research at the NRC. This report

does not address the research being done at NRC on issues of reactor security or the threat of sabotage. The ACRS views on

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

State-of-the-ACRS Reactor Consequence Analyses (SOARCA) Project and the study of sump screen blockage are not

addressed in this report. These projects are actively followed by the Committee. The ACRS has been providing interim reports

on the technical approach and progress of these activities.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT

advanced non-LWR designs observations and recommendations unlimited

digital instrumentation and control systems operational experience 14. SECURITY CLASSIFICATION

fire safety probabilistic risk assessment (This Page)

human factors and human reliability research seismic and structural engineering unclassified

long-term, sustained research severe accidents (This Report)

materials and metallurgy thermal hydraulics unclassified

neutronics analysis, core physics, and criticality

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (9-2004) PRINTED ON RECYCLED PAPER