I

I

6
AVAILABILITY OF REFERENCE MATERIALS

IN NRC PUBLICATIONS

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NUREG-1635, Vol. 6

Review and Evaluation of the

Nuclear Regulatory Commission

Safety Research Program

A Report to the

U. S. Nuclear Regulatory Commission

Manuscript Completed: March 2004

Date Published: March 2004

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

U. S. Nuclear Regulatory Commission

Washington, DC 20555-0001
NUREG-1635, Volume 6, has been

reproduced from the best available copy.
ABSTRACT

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

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

Safety Research Program being carried out by the Office of Nuclear Regulatory Research (RES).

This report focuses on that portion of the NRC research program dealing with the safety of existing

nuclear reactors and advanced light water reactor designs, AP1000 and ESBWR, submitted for

certification. In its review of the NRC research activities, the ACRS considered the programmatic

justification for the research as well as the technical approach and progress of the work. This

review attempts to identify research crucial to the NRC mission. It also attempts to identify

research activities that have made valuable contributions to the agency mission in the past, but now

have reached the point where additional research is not needed for efficient and effective safety

regulation. The review also attempts to identify areas where greater international cooperation in

research useful to the NRC could leverage resources of partners in the research and yield superior

technical products. The report does not address research on the vulnerability of existing nuclear

power plants to acts of sabotage and terrorism.

.M
TABLE OF CONTENTS

Page

Abstract ..... iii

Abbreviations ......... vii

1 Introduction ........... '.' 1

2 General Observations and Recommendations .......... ................ 3

3 Analysis and Evaluation of Operational Data ........... ................ 9

4 Containment Systems ........................................ ;.13

5 Digital Instrumentation and Control Systems .......... ................ 17

6 Fire Safety Research ........................................ 21

7 Reactor Fuel Research ....................................... 25

8 Neutronics and Criticality Safety .................................... 29

9 Human Factors and Human Reliability Research ........ ............... 31

10 Materials and Metallurgy ....................................... 35

11 Probabilistic Risk Assessment ..................................... 43

12 Radiation Protection . ....................................... 47

13 Seismic Research . ....................................... 49

14 Severe Accident Research ....................................... 51

15 Thermal-Hydraulics Research ..................................... 57

16 References ....................................... 65

V
TABLES

Page

1. Research Activities in Analysis and Evaluation of Operational Data ............. ................. 11

2. Research Activities in Containment Systems ..................................................... ..... 15

3. Research Activities in Digital Instrumentation and Control Systems .............. .................. 19

4. Fire Safety Research Activities .......................................................... 23

5. Research Activities in Reactor Fuel .......................................................... 28

6. Research Activities in Neutronics Analysis, Core Physics, and Criticality Safety ......... ...... 30

7. Research Activities in Human Factors and Human Reliability ..................... .................... 33

8. Research Activities in Materials and Metallurgy .......................................................... 38

9. Research Activities in Probabilistic Risk Assessment ..................................................... 44

10. Research Activities in Radiation Protection .48

11. Research Activities in Seismic Phenomena .50

12. Severe Accident Research Activities .54

13. Research Activities for Accident Consequence Models .55

14. Thermal-Hydraulics Research Activities .61

vi
ABBREVIATIONS

ACRS

ACNW

ACR-700

AEOD

ANS

ATHEANA

ATWS

BEIR

BMI

BWR

CAMP

CFD

CFR

CSARP

CSNI

DOE

EPA

EPIX

EPRI

ESBWR

FY

GALL

GSI

HERA

HRA

HSST

l&C

IAEA

IASCC

ICRP

IEEE

INPO

IPEEE

IRIS

ISI

ISO

LER

LERF

LBLOCA

LOCA

LWR

MACCS

MOV

MOX

NCRP

Advisory Committee on Reactor Safeguards

Advisory Committee on Nuclear Waste

Advanced CANDU Reactor-700

Office for Analysis and Evaluation of Operational Data

American Nuclear Society

A Technique for Human Event Analysis

Anticipated Transients Without Scram

Biological Effects of Ionizing Radiation

Bare Metal Inspection

Boiling Water Reactor

Code Applications and Maintenance Program

Computational Fluid Dynamics

Code of Federal Regulations

Cooperative Severe Accident Research Program

Committee on the Safety of Nuclear Installations

Department of Energy

Environmental Protection Agency

Equipment Performance and Information Exchange System

Electric Power Research Institute

Economic Simplified Boiling Water Reactor

Fiscal Year

Generic Aging Lessons Learned (Report)

Generic Safety Issue

Human Event Repository and Analyses

Human Reliability Analysis*

Heavy Section Steel Technology

Instrumentation and Control

International Atomic Energy Agency

Irradiation Assisted Stress Corrosion Cracking

International Commission on Radiological Protection

Institute of Electrical and Electronics Engineers

Institute of Nuclear Power Operations

Individual Plant Examination of External Events

International Reactor Innovative and Secure

In-Service Inspection

International Standard Organizations

Licensee Event Report

Large Early Release Frequency

Large-Break Loss-Of-Coolant Accident

Loss-of-Coolant Accident

Light Water Reactor

MELCOR Accident Consequence Code System

Motor-Operated Valve

Mixed Oxide

National Council on Radiation Protection

Vii
ABBREVIATIONS (Cont'd)

NDE Non-Destructive Examination

NEA Nuclear Energy Agency

NEI Nuclear Energy Institute

NFPA National Fire Protection Association

NMSS Office of Nuclear Material Safety and Safeguards

NRC Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

OECD Organization for Economic Cooperation and Development

PARCS Purdue Advanced Reactor Core Simulator

PFM Probabilistic Fracture Mechanics

PIRT Phenomena Identification and Ranking Table

PRA Probabilistic Risk Assessment

PSHA Probabilistic Seismic Hazard Analysis

PTS Pressurized Thermal Shock

PUMA Purdue University Multidimensional Integral Test Assembly

PWR Pressurized Water Reactor

RCS Reactor Coolant System

RES Office of Nuclear Regulatory Research

RG Regulatory Guide

ROP Reactor Oversight Process

RPV Reactor Pressure Vessel

SC! Secondary Containment Isolation

SDP Significance Determination Process

SNAP Symbolic Nuclear Analysis Package

SPAR Simplified Plant Analysis Risk Model

SSHAC Senior Seismic Hazard Analysis Committee

TRACE TRAC-RELAP Advanced Computational Engine

UCLA University of California, Los Angles

U.S. United States

USGS United States Geological Survey

VHP Vessel Head Penetration

viii
1 -INTRODUCTION

The Nuclear Regulatory Commission (NRC)

maintains a Safety Research Program to:

* Ensure its regulatory framework

has a sound technical basis

* Prepare for the anticipated

changes in the nuclear industry

that could have safety implications

* Develop improved methods for its

regulatory process

* Maintain an infrastructure of

expertise, facilities, analytical

tools, and data to support

regulatory decisionmaking

These missions for the research effort were

defined when the NRC was established and

there was limited experience with the

operation of nuclear power plants. The need

for research remains today, despite the

growth of experience with the plants, as:

* Nuclear power plants age and

encounter challenges of material

degradation not anticipated when

the plants were designed

* The NRC considers applications

for extending licenses and

increasing the operating power

levels of plants

* Reactor fuels are used to higher

levels of burnup, new claddings

for the fuels are introduced, and

mixed oxide (MOX) fuels are

considered for the disposal of

excess weapons-grade plutonium

* The NRC evolves its regulations

from a deterministic foundation to

a risk-informed basis that makes

greater use of 'best-estimate'

analyses

* New technologies such as

software-based digital

instrumentation and control (I&C)

systems are backfit into the

existing plants

* New light water reactor (LWR)

designs making more use of

passive systems are submitted for

certification

In this report, the Advisory Committee on

Reactor Safeguards (ACRS) presents its

observations and recommendations

concerning that portion of the NRC Safety

Research Program focused primarily on the

safety of existing nuclear reactors. This report

does consider research activities that support

the certification of the advanced LWRs,

AP1000 and ESBWR. It does not address

research to prepare the agency to certify the

ACR-700 design or other advanced reactor

designs such as the GEN IV design being

considered by the Department of Energy

(DOE) or the IRIS design. The ACRS

recently reported on the agency's plans for

research on these advanced reactor designs

[Ref. 1 ] and the progress in advanced reactor

research has not been sufficient to warrant

reexamination of these research activities at

this time. The present report does not

address research on nuclear waste and the

licensing of a permanent geological

repository for spent reactor fuel. The Advisory

Committee on Nuclear Waste (ACNW)

monitors this research and will report its

conclusions and recommendations

I NUREG-1 635
lI

separately. This report does not address

research on the vulnerability of existing

nuclear power plants to acts of sabotage and

terrorism. The ACRS will report separately

on the technical aspects of this research.

In its review of the NRC Safety Research

Program, the ACRS considered the

programmatic justification for the research as

well as the technical approach and progress

of the work. The ACRS supports research

that:

* Provides support to the

identification and resolution of

current safety issues

* Provides the technical basis for

resolution of foreseeable safety

issues

* Develops the capability of the

agency to independently review

risk-significant proposals and

submittals by the licensees

* Supports initiatives of the agency,

including the Reactor Oversight

Process (ROP) and the move to

risk-informed and performancebased regulation

* Improves the efficiency and

effectiveness of the regulatory

process

* Maintains technical expertise

within the agency and facilities in

disciplines crucial to the agency

mission and not readily available

from other sources

This review of the NRC Safety Research

Program, in particular, attempts to identify

research activities that have made valuable

contributions to the agency mission in the

past, but now have reached the point where

additional research is not needed for efficient

and effective safety regulation. This review

also attempts to identify areas where greater

international cooperation in research useful to

the NRC could leverage resources of

partners in the research and yield superior

technical products.

General observations and recommendations

concerning the NRC research activities are

presented in Chapter 2 of the report.

Other observations and recommendations

concerning the research activities in the

following technical disciplines are discussed

in Chapters 3 through 15 of this report:

* Analysis and Evaluation of

Operational Data

* Containment Systems

* Digital Instrumentation

Control Systems

and

* Fire Safety Research

* Reactor Fuel Research

* Neutronics and Criticality Safety

* Human Factors and Human

Reliability Research

* Materials and Metallurgy

* Probabilistic Risk Assessment

* Radiation Protection

* Seismic Research

* Severe Accident Research

* Thermal-Hydraulics Research

NUREG-1 635 2
2 GENERAL OBSERVATIONS AND RECOMMENDATIONS

Overall, the NRC has a well-focused, wellplanned Safety Research Program dealing

with existing reactors and advanced LWRs

submitted for certification. The research effort

may well be near the minimum needed to

support regulatory activities and agency

initiatives while maintaining technical

competencies crucial to the agency mission.

A very large fraction of the research is

focused on immediate user needs.

Resources available for exploratory research

to investigate other avenues of improved

safety and more efficient regulation are

minimal and may limit the agency's ability to

anticipate future needs.

Some of the research activities are especially

noteworthy. This includes the research being

done in probabilistic risk assessment (PRA)

to provide computational tools to support the

Reactor Oversight Process. Confirmatory

research resolving safety issues of reactivity

initiated accidents with high-burnup fuel is

also noteworthy. Human factors research at

NRC has been re-energized and holds the

promise of significant future contributions to

the regulatory process.

There are research challenges. The important

effort to provide the agency with a

consolidated, state-of-the-art, best-estimate

thermal-hydraulics computer code is such a

challenge. Though this effort has progressed

well, the time is approaching when the need

for this code will be critical. It may be

necessary to increase the effort and focus'

the thermal-hydraulics research on producing

this code and integrating it into the regulatory

process. Once this is accomplished, research

resources will be freed to improve the

predictive capabilities in thermal-hydraulics.

Further research in -the area of thermal-'

hydraulics especially in areas other than

large-break loss-of-coolant accidents (LBLOCAs) might be. facilitated 'by more

extensive international cooperation akin to

that being done in the area of severe accident

research.

Some research efforts have produced

valuable technologies, which can be

integrated into the regulatory process.

Further research in these areas should be

reduced substantially. An example of such

an effort is the research on the realistic

structural capacity of reactor containments.

This research has produced, benchmarked,

and validated computer codes that can be

used to assess containment designs and

estimate the' effects of both materials

degradation and construction errors.

Research on seismic engineering has

reached the point that the most significant

agency needs have been met and greater

reliance on engineering consulting firms, as

needed, may be adequate for the regulatory

process in the future. There is advantage in

maintaining modest research efforts to

accrue benefits from cooperative research on

seismology and seismic engineering.

Other observations and recommendations

concerning the research efforts 'are

summarized below and also discussed in

individual Chapters.

Analysis and Evaluation of Operational

Data

The ACRS supports the research activities

now under way in'the general area of analysis

and evaluation of operational data. Indeed,

these data collection and organizational

activities, including superior computerized

search capabilities, are essential to the

agency mission. However, the ACRS is

concerned about the vitality and planning of

continued efforts to use the database and

especially the opportunities being made to

3 NUREG-1 635
explore the database independently of

current research needs. Such independent

examinations of the database in the search

for unexpected interactions among

regulatory activities can always be deferred at

little cost. Continued deferral will deprive the

agency of the opportunity to utilize

operational data in many ways that may lead

to more effective and realistic regulatory

practices.

Containment Systems

Adequate computer models to assess the

structural capabilities of containments of

existing reactors are now available. They

have been benchmarked and validated by

comparison to well-scaled test data. The

effects of degradation as well as the effects

of construction errors can be evaluated

adequately by analysis without the need for

further experiments.

Debris accumulation in pressurized water

reactor (PWR) sumps is an important issue

that must be resolved. The planned work to

further explore the potential for chemical

interactions that generate suspensions,

which can lead to large head losses even with

low fiber loadings, is extremely important.

Research in this area needs to be expedited

and promptly reach the point licensees can

implement solutions to this issue.

Digital Instrumentation and Control

Systems

The current reliance on controlling the design

and development process of software, rather

than focusing on the software itself, has the

potential of straining agency resources in the

regulatory process. Although a controlled

development process is assumed to lead to a

highly reliable product, this reliability remains

unquantified. Complicating the regulatory

review of these l&C systems is the pace of

innovation in the digital electronics industry.

Innovations of revolutionary natures are being

made at rates far greater than those the

regulatory system can respond to.

Furthermore, threats to digital systems are

greatly expanded with the emergence of the

so-called "cyber security" threat posed by the

malevolent or just the prankster.

In general, the ACRS supports the research

activities now under way in the Digital l&C

Systems. These activities, if successful, will

provide tools to make the review process

more efficient and provide a basis for

including software reliability into PRAs, thus

reducing the need for relying on controlling

the design and development process, and

enhancing the ability of the agency to riskinform its regulations.

Fire Safety Research

The current NRC research efforts in Fire

Safety seem incongruent with the estimated

risk significance of fire. The limited fire

research effort is understandable since

resources for such research have been

diverted to respond to the events of

September 11, 2001. Now that these

responses are being completed, the agency

should revitalize its Fire Safety research

efforts and move the technical capabilities of

the agency to be more in line with risks being

ascribed to fire.

Reactor Fuel Research

The NRC is completing a confirmatory

research effort on reactivity insertion with

high-burnup fuel. A sustained NRC research

effort in reactor fuels behavior under accident

and off-normal conditions is needed

especially as competitive pressures force fuel

vendors to curtail their research efforts.

Continuing expertise and even additional

research may be needed if, as now expected,

licensees make requests to extend fuel

burnups up to and beyond 75 GWd/t. It is

quite likely that such industry proposals will

be substantiated by minimal experimental

NUREG-1 635 4
research. It is important for the NRC to have

a technically justified position on the

experimental research that is necessary to

support such proposals especially in light of

the "surprise" that accompanied extending

fuel burnups into the range of 50 to 60

GWd/t.

Neutronics and Criticality Safety

The Office of Nuclear Regulatory Research

(RES) is doing a commendable, costeffective job in maintaining the capabilities of

its neutronics and core physics codes and the

associated databases. Upgrades to these

codes to address both extended fuel burnup

and MOX fuels are needed and these needs

are addressed in the- current research

projects. The ACRS anticipates that these

extended capabilities will be tested

significantly in the certification of the

advanced reactor designs.

Human Factors and Human Reliability

Research

The attention NRC is paying to Human

Factors research is yielding dividends. The

Standard Review Plan Chapter 18 has

recently been updated to reflect

developments especially in the area of digital

l&C systems. A risk-informed method to

screen licensee submittals for human factors

review has the potential of better focusing

agency resources on risk- significant human

factors issues.

The issues surrounding human performance

in teams (such as the team of control room

operators) and human performance in

organizations, including the question of

organizational reliability, are important and

need to be considered in the ongoing Human

Performance Research planning.

The NRC is experimenting with an ambitious

methodology for estimating human reliability

that is to assess errors of commission as well

as errors of omission. This methodology

called ATHEANA focuses on the causes of

human performance ("performance shaping

factors"). The ACRS looks forward to

reviewing the results of the ongoing

ATHEANA application efforts and learning

whether the method has become easier to

apply, whether it yields insights significantly

better than those that could be derived from

simpler methods, and whether it includes a

quantification process for human reliability.

Of particular interest to the ACRS is the use

of expert opinion elicitation in these

applications to quantify the ATHEANA

results.

Materials and Metallurgy

Research in this area addresses primarily

radiation-induced embrittlement of ferritic

pressure vessel steels and austenitic core

structure materials, and the characterization

and monitoring of environmentally-assisted

cracking. The quantification of these

phenomena is central to maintaining the

integrity in LWR structures.

Projects dealing with Reactor Pressure

Vessel (RPV) Embrittlement should be

considered with respect to safety issues of

RPV integrity and with respect to maintaining

the core competency in irradiation

embrittlement. The staff needs to evaluate

what research should be retained in order to

meet these two needs.

The characterization of cracking kinetics

should be continued in the areas of

irradiation-assisted cracking of irradiated

stainless steels used in boiling water reactor

(BWR) core structures and of stresscorrosion cracking of nickel base alloys used

in PWR steam generators and in large

components in all LWR designs. This

information, in addition to the evaluation of

various inspection techniques, is required for

the NRC to maintain an independent

capability to examine integrity issues inherent

5 NUREG-1 635
to licensee plans concerning residual life,

inspection, and repair. Many of these

projects are being conducted within

international research programs, thereby

leveraging NRC resources.

The ACRS supports the "Proactive Materials

Degradation Assessment" Project recently

initiated by the NRC in order to move from

the economically inefficient reactive mode of

regulating materials degradation issues. The

plan to lead a phenomena identification and

ranking table (PIRT) effort with the industry

over the next two years, with expert elicitation

from international experts, should identify

potential materials degradations that may

occur in the future. This knowledge will

provide a better basis for future regulatory

decisions on timely inspection and repair

criteria.

Probabilistic Risk Assessment

The NRC has an impressive PRA research

program. The ACRS generally supports the

idea that the NRC needs to maintain its PRA

capabilities at or near the state-of-the-art

level. As the NRC moves toward a riskinformed regulatory system, it will need ever

greater amounts of risk information derived

from quantitative risk assessments. It will

want to maintain its own independent

capabilities for conducting PRAs at state-ofthe-art levels. A continued vital research

program in PRA is, then, crucial to the

agency mission.

Radiation Protection

The NRC research activities in the area of

Radiation Protection constitute a wellleveraged program that allows the NRC to

use quality data and information as the

foundation of its basic radiation protection

regulation, 10 CFR Part 20, "Standards for

Protection Against Radiation."

Seismic Research

The NRC has invested heavily in the

understanding of seismic issues and

development of its rules to ensure that

seismic risks are acceptably low. Licensee

submittals concerning risks associated with

external accident initiators such as seismic

events suggest these efforts have been

largely successful.

Much of the NRC research into seismic

issues involves cooperative efforts to address

essential seismic data. The ACRS supports

these efforts which are highly leveraged by

contributions from other partners.

The ACRS understands that there are needs

for additional work on the engineering of

components and structures to avoid seismic

damage. Unlike many other issues that are

dealt with by NRC research, there is in this

Country a very substantial engineering and

scientific community addressing this issue.

This community is independent of the NRC

licensees. Thus, the NRC-sponsored work is

not needed to maintain a core competency in

this area. If there are open issues associated

with licensee submittals such as the seismic

behavior of spent fuel casks or buried

structures, the NRC can rely on engineering

consulting firms on an as-needed basis to

deal with such issues.

Severe Accident Research

The ACRS is very supportive of the NRC

strategy to maintain and update its

capabilities for accident analyses. The ACRS

encourages the NRC to give serious

consideration to joining international

cooperative research agreements now being

proposed to conduct:

* Prototypic, in-pile tests of

radionuclide releases from, and

degradation of, reactor fuel

exposed to air for both reactor

NUREG-1 635 6
accident analyses and

spent fuel storage pool

accidents

* Prototypic, in-pile tests of

radionuclide releases from, and

degradation of, high-burnup fuels

* Prototypic, in-pile tests of

radionuclide releases from, and

degradation of, MOX fuel

The ACRS also supports programs to

maintain the MACCS code and looks forward

to reviewing results of the comparison of the

Gaussian plume model to state-of-the-art

dispersion models in the context of risk

assessments.

enthusiastically supported by the ACRS. The

ACRS understood this to be a challenging

undertaking. Indeed, the consolidation has

proved to be every bit as challenging as

anticipated. A version of the consolidated

code, TRACE, is now being tested. Efforts

should be focused on integrating TRACE into

the regulatory process even at the expense of

research to further improve the technical

capability of TRACE.

The ACRS encourages anticipatory research

to couple computational fluid dynamics (CFD)

methods with system codes such as TRACE.

This has the potential to enhance the

accuracy of thermal-hydraulic analyses.

Thermal-Hydraulics Research

The decision to consolidate NRC's thermalhydraulic modeling into a single code was

7 NUREG-1 635
3 ANALYSIS AND EVALUATION OF OPERATIONAL DATA

The NRC is moving toward a risk-informed

regulatory system that will require more

realistic estimates of vulnerabilities of nuclear

power plants. The ACRS has voiced its belief'

that the operational database for nuclear

power plants is a rich source of information

for improving the quality of quantitative risk

assessments and identifying omissions in

these assessments [Refs. 2 and 3]. For

several years following the 1979 accident at

Three Mile Island, then the Office for Analysis

and Evaluation of Operational Data (AEOD)

conducted independently motivated, detailed

evaluations of the operational database

[Refs. 4 and '5]. The products of these

evaluations were of direct use to the

quantitative risk analysis community.

In 1999, the function of the AEOD was

transferred to RES. It is of'considerable

interest to the ACRS to see that the intensely

inquisitive, independent evaluation of the

operational database is sustained in this

organizational structure. Furthermore, the

ACRS is anxious to see that results of such

independent investigations are used by other

NRC Offices.

Research Projects in the area of Analysis and

Evaluation of Operational Data' are listed in

Table 1. These Projects are primarily

devoted to the sustained acquisition and

organization of operational data from licensee

event reports (LERs), maintenance data and

the like. Currently, the motivation for the data

collection efforts appears to be primarily

anticipation of operational data needs of other

research initiatives by the agency to address

m -r .k Electrical Grd StabIlt '7 codjt

I '_ 7

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popagatilng6ioss"- of .'t'e' elect-rical.' g d 'in

ventral and Eastern United States that:'left

rI11lon's1 without . electrlcal . power.:;:'.Grid

vulneraeiity. be.ame -a topic-of Niumerous

iscussions ln' thee. .-news .- media.' -'The

ulnerabity had6een anticipatedbyji'hthe NRC

study, Vperating Experlenci Assessment

Effeats' 'of -Grid'Events :'onNuclear';Power 'Plant 'PerfonnceS, .' f l., -6: The stud'

earned nd events that- affected,`nuclear

iwer.'plantsf r1994 to 20d1..The events

ere chosen orexaiinmati6n With attentionito

the structural changes that have occurred In

the electrical at " . ^f. energy. , : .>!,m.ia, market ,,_'as'economc

d1eregiilath'n 'has. progressed. :Thb study

showed that the number Of r id events nay

d,6S d&e'asing Ahith time, .but theS dura'tons of

events were increasing -5a finding 'quited bjy

Aajonetwohrk. Though most events pose no

cxcejptihanIchallengesto nuclearpow`r plant

safetyissues 'are identified in the 'study tha't

o have the potential to increase .risk and

hallenge -the 'effectiveness ,ofJ the: cuurrent

gulations. :.me report docw;enting "the

study pro vides a baseline of grid 'erformance . .I ' . , .- I.:_ ' ' ' .,' ,.; ' ' -' ! ,- I2 ' .-I.:'! ' '_ ''! _ '-.. "

which can be used to 'gauge the.impact-of +,eregulatidn on, ',and changes in, grid

operations. K -' ¶

rhe report shows the public that the NRC Is

emerging'issues sthatdcan affect

owerpiant safety and how bpe'rationalidIa

can be< used toanticipat'ea'nd quantify

remerging issues.-.!:,':.-.-'... ;-,-:.''

1 Note that occupational exposure

data and reports are discussed in Chapter

12, "Radiation Protection," of this report.

9 NUREG-1 635
regulatory issues. Campaigns for the

immediate future and the longer term to

utilize the archives of operational data are no

longer readily apparent from the research

program. An undertaking that is apparent and

could be of particular regulatory significance

is the search for safety-significant interactions

among the ongoing regulatory actions to:

* Extend the license period of nuclear

power plants

* Increase the operating power levels of

the plants

* Use fuel in the plants to higher levels of

burnup

* Evolve plant operations to do more

online maintenance

The ACRS is concerned that this search for

interactions among the various operational

changes has stagnated for lack of

management support.

In summary, the ACRS does not critique the

research Projects now under way in the

general area of Analysis and Evaluation of

Operational Data. Indeed, these data

collection and organizational activities

including superior computerized search

capabilities are essential to the agency

mission. It is evident that useful products are

still being generated in this effort (See

sidebar success story on Electrical Grid

Stability). The ACRS is concerned about the

vitality and planning of continued efforts to

use the database and especially the

opportunities being made to explore the

database independently of current research

needs. Such independent examinations of the

database in the search for unexpected

interactions among regulatory activities can

always be deferred at little cost. Continued

deferral will deprive the agency of the

opportunity to utilize operational data in many

ways that may lead to more effective and

realistic regulatory practices.

NUREG-1 635 10
Table 1. Research Activities in Analysis and Evaluation of Operational Data2

Project Title Comment

A9134 Sequence Coding Search This is an important Project for analyzing industry

System trends and establishing a data source for

Database from LERs used for NRC measuring industry performance.

studies

G6810 Management of Precursors This is a short-term grant for a two-day workshop

Study for detecting, analyzing, and benefitting from

knowledge of accident precursors.

J8258 International Common-Cause This is an useful collaborative effort to maintain

Exchange Project cognizance of developments in the realistic

Maintaining awareness of quantification of common-cause failures in PRA.

developments in common-cause

failures within the international

reactor safety community

Y6215 Operational Events for This important Project seeks to identify and rank

Accident Sequence Precursor those operating events that were most significant in

Program terms of the potential for inadequate core cooling and the potential for core damage.

Y6546 Industry Trends Program These important Projects produce plant-specific and industry-wide estimates, summary tables,

graphs, and charts from operational data to support

Y6636 Support for Industry Trends the Industry Trends Program and various data

analysis activities of the NRC.

Y6626 Access to INPO's EPIX Database on failures of key components; database

System for SPAR models used in the ROP.

Y6214 Integrated Data Coding and Improved database management; only a small

Analysis Methods effort remaining.

Y6468 Reactor Operating Experience Collects, codes, and maintains operational data for

Data for Risk Application reactor systems and components, initiating events,

common-cause failures, and fire events for use in

regulatory activities. Utilizes methods and

procedures developed under Y6214.

Y6406 Assess and Improve Potential for identification of unanticipated, safetyRegulatory Effectiveness significant interactions among regulatory activities.

(See also Y6522 in Probabilistic Risk

Assessment, Table 9) .-

2See also Y6123, Retrospective Risk Analysis of Human Reliability in Operating Events,

Table 7.

I1I NUREG-1635
4 CONTAINMENT SYSTEMS

it_-Liz-t-M__It__oCt-_ .- Ace

The defense-in-depth safety philosophy has

led .to encasement of power reactors in

robust containment buildings. Containment is

the final physical barrier to the uncontrolled

release of radioactive materials to the

environment in the event of an accident.

Comparisons of the environmental releases

of radioactivity during the accidents at

Windscale, Three Mile Island, and Chernobyl

demonstrate the value of robust

containments.

The structural capabilities of containments for

nuclear power plants in this Country are far

greater than might be inferred from their

design bases. To realistically assess these

capabilities and demonstrate their capacity to

provide protection in the event of severe

accidents, the NRC, along with other

countries and agencies, has established a

technically sophisticated program. This

program involved testing of quarter-scale

containment models to provide databases for.

validating analytical structural models.

uRound robin" comparisons of the predictions

of these models to the data obtained in the

tests were conducted.

This effort to develop validated technology for

the quantitative characterization of

containment capabilities has been both

essential and quite successful. The

technology for predicting containment

responses to loads can now be transferred to

the capable hands of users of the

information such as line regulatory

organizations, codes and standards

organizations, and risk analysts.

Research studies to date have been on Uasdesigned" structures. Many containments

have suffered some degradation to either the

concrete shell or the steel liner and it is

important to understand how degradation

771

l','' tI Is...

oRf cntainment model dur' in

V5 ~tr ; ~ 'Ae 

tetdone in cooperation: with Japan is

on faseries attie'sts 'that have 'be'en dn

-b becmrk anhdyvalidatestructuralcapacity

uter 'codes. 7 yiall, e, 'scle

~ontaineht, mdels~. re foud to Jila

~ha~ twce he design 'pressure.,. Cim- u

coes jvaidted,..S by .w:the' tsig4:: arused;.'to

s o-rak eictins`o : 'l

-Rbbptures of co con ianimoelts undri

'cid'ent condtions' ,'

bne'f asend oftest tht hae ben 'don

13 NUREG-1 635
affects the structural margins attributed to

containments in risk analyses. Because the

computer codes for structural analyses of

containments have been benchmarked and

validated, the effects of degradation, as well

as the effects of construction errors, can be

evaluated adequately by analysis without the

need for further experiments. Projects Y61 64

and W6684 (See Table 2) address such

questions and these efforts should be

continued. However, the value of general

study performed under Project W6684 is

questionable.

The aging management of containment

structures is being addressed in the current

activities associated with license extension. If

further work is needed in this area, it should

be done by the licensees in support of their

applications.

Project W6593, "Effect of Aging and

Emerging Issues on MOV performance,"

should be concluded. Active components

such as motor operated valves (MOVs) are

included in ongoing maintenance and testing

programs which should be adequate to detect

any potential aging issues. Further studies

could improve reliability, but this should be

the responsibility of the licensees or the

vendors.

The ACRS has pointed out that the

technology base is still not complete [Ref. 7].

The planned work to further explore the

potential for chemical interactions that

generate suspensions, which can lead to

large head losses even with low fiber

loadings, is extremely important. The Project

Y6041 may deserve more management

attention. Research in this area needs to be

expedited and focused to support the

regulatory process and provide a basis for

licensees to implement solutions to this

issue.

Assessment of containment structures for

entombment, Project Y6331, can be

subsumed within decommissioning projects.

This does not appear to be a particularly

pressing issue.

Terminating or transferring some of the

Containment Systems Projects (See Table 2)

should free resources available to the agency

to address containment/confinement issues

and security vulnerabilities of reactor

containments for advanced reactors. The

ACRS believes it would be prudent for the

NRC to examine the safety issues associated

with underground siting of future reactors and

other new nuclear facilities.

Debris accumulation in PWR sumps is an

important issue that must be resolved.

NUREG-1 635 14
Table 2 Research Activities in Containment Systems

Project Title Comment

J6043 Inspection of Aged/Degraded Containment This work should be concluded. If

additional information is needed, it

should be provided by licensees

who have the responsibility to

demonstrate adequacy.

W6851 Review Guidance for Lightning This effort should be concluded.

Y6757 Containment Capacity Studies Development of technology to

estimate capacities of containments

to sustain pressure loads has been

successful and should now be

brought to closure.

W6593 Effects of Aging and Emerging Issues on This work should be concluded.

MOV Performance These are active components and

ongoing programs should be

adequate to detect any potential

aging issues.

W6684 Assessment of Aged and Degraded Margins associated with real,

Structures and Components degraded structures must be

understood for realistic analyses.

Such issues are best addressed in

terms of more specific conditions

such as in Y6164. The value of a

general study is questionable.

Y6041 Assess Debris Accumulation on PWR This is an important issue but one

Sump Performance that has been slow to resolve; research to support regulatory

initiatives should be expedited and

focused to support the regulatory

process and provide a basis for

licensees to implement solutions to

this issue.

Y6164 Structural Risk-informed Assessment This work should be continued. It

Containment Degradation provides quantification of loss of

margin as a result of degradation.

Y6331 Long Term Containment Stabilization Studies for entombment of nuclear

Concrete Structures facilities as permitted by regulations. This does not appear

to be a particularly pressing issue.

15 NUREG-1 635
5 DIGITAL INSTRUMENTATION AND CONTROL SYSTEMS

The move to software-based, digital l&C

systems is taking place for advanced nuclear

power plants 'and more slowly for existing'

nuclear power plants. Analog, electromechanical systems in existing plants'are

difficult to maintain as replacement parts of

suitable quality become less readily available.

Furthermore, there is a widespread belief that

software-based digital electronic systems

offer superior levels of control, information

density, and reliability.

Certainly, digital systems have inherent

reliability that is potentially much greater than

electro-mechanical systems. There have

been, however, notable instances of

catastrophic failures of digital systems.

Usually, these failures have been traced to

omissions in the definitions of the

requirements and specifications for the

software that controls the system. The

challenge in the regulation of digital l&C

systems is the difficulty in comprehensive

testing of these systems due to their diverse

capabilities. At this time, in order to ensure

adequate quality and reliability of softwarebased digital systems, the NRC relies

primarily on the manpower-intensive

monitoring of the design and development

processes rather than focusing on the

product. It is assumed that a controlled

production process will lead to a highly

reliable product. This reliability, however,

remains unquantified.

The reliance on process monitoring has the

potential of straining agency resources in the

regulatory process if highly capable tools are

not available to the agency. Complicating the

design and development of such regulatory

tools is the pace of innovation in the digital

electronics industry. Innovations of

revolutionary natures are taking place at rates

far greater than those to which the regulatory

framework can respond. Furthermore, threats

to digital systems are greatly expanded with

the emergence of the so-called 'cyber

security" threat posed by the malevolent or

just the prankster.

There is a vast literature on software

reliability. This literature surrounds two

schools of thought. One school holds that

software reliability can be quantified and

methods can be developed for doing so. The

other school of thought views software

reliability as meaningless by itself.

Embedded software failures are only one

contribution to the unreliability of the overall

system. The staff has judged that these

models have not reached a level of maturity

that allows them to be used in the regulatory

process. The ACRS agrees.

The NRC research activities in Digital l&C

Systems are listed in Table 3. Project K6907

is intended to provide a much needed

assessment of the merits and demerits of the

modeling methods of the two schools of

thought. Within this Project, the staff plans to

develop and test methods and models for

integrated assessment of digital system

reliability. Project Y6591 is developing and

testing methods for evaluating software

reliability. Project Y6332 is an investigation of

development of PRA models for digital

systems and Project Y6472 is to investigate

the integration of digital systems into the

PRAs for current generation nuclear power

plants. These Projects, if successful, will

provide the basis for including software

reliability in PRAs, thus reducing the need for

relying on controlling the design and

development process and enhancing the

ability of the agency to risk-inform its

regulations.

Studies of digital system performance and

reliability appear to involve a larger issue than

any one agency can undertake in the face of

17 NUREG-1 635
an innovative digital electronics industry. The

staff is working in several areas to try to tap

into this wider research, including serving on

the steering committee of the DOE l&C and

Human Machine Interface advisory

committee. The staff is also in the process of

developing an international collaborative

program and a potential research effort in

this area is digital system reliability modeling.

The RES staff also chairs the OECD/NEA

COMPSYS group, which is charged with

developing nuclear power plant specific

digital systems failure data. The ACRS

encourages these activities. However, it is not

evident that the Project at Halden (Y6349)

directly addresses the NRC needs.

Two Projects (Y6371 and Y641 0) address

the issue of aging of cable systems in existing

nuclear power plants. Demonstration that

cables in a power plant have not aged to the

point that plant safety is threatened is the

responsibility of the licensee. The NRC

needs to understand the issue well enough to

ensure that adequately realistic conditions for

aging rather than unnecessarily conservative

conditions are taken into consideration and

that proposed technical responses to the

issue are adequate. Consequently, these

studies can be brought quickly to a resolution

sufficient for regulatory needs.

Similarly, the study of electromagnetic

conditions at nuclear power plants (Y6272)

has been under way for sometime. An

adequate understanding should be soon

available to review licensees' arguments that

electromagnetic conditions do not threaten

plant safety. Therefore, this Project should

be concluded in FY 2004.

Several Projects address the issue of cyber

security of digital systems (Y6729, Y6712,

Y6733, and Y6734). Again, the issue of cyber

security seems bigger than what any one

government agency can confront, least of all

an agency as small as the NRC. RES should

explore the possibilities of leveraging the

resources it has for cyber security by joining

with other government agencies in the study

of this issue.

NUREG-1 635 18
Table 3. Research Activities in Digital Instrumentation

and Control Systems

Project Title Comment

K6907 Digital System Performance Program is intended to provide an unbiased

and Reliability. assessment of the merits and demerits of

A survey of the various modeling modeling methods in the two main schools of

methods for digital systems thought concerning software reliability. The staff

plans to develop within this activity a tool to

assess the reliability of an integrated digital

system based on fault-tolerant system methods.

Y6475 Wireless Wireless communication may resolve some fire

Anticipatory research on the potential issues but is very susceptible to cyber attack.

challenges and regulatory issues of There is interest within the nuclear industry. in

using wireless communication using this technology and NRC research is being

technologies in nuclear power plants done in anticipation of an industry proposal.

Y6590 Review IEEE Standards for An essential activity since these standards will be

Endorsement the bases for regulatory review of the

IEEE Standards are the bases for performance and reliability of digital systems.

most Regulatory Guides dealing with

issues of digital instrumentation and

control

Y6332 Digital Systems Risk Essential to developing the technology to include

Investigation of digital l&C system digital systems reliability in PRAs.

analysis methods

Y6371 Risk Associated With Cable A simple effort that ought to be concluded in FY

Aging 2004.

Y641 0 Collaborative Research on Progress should be monitored to ensure

Wire System Aging research does not stray beyond needs of the

Collaborative research on current regulatory process and does not enforce

conditions, residual lifetime and fire overly conservative constraints on licensees.

risk of cables

Y6272 Characterize Electromagnetic Second year of a two-year effort. Should be

Conditions at Nuclear Power concluded in FY 2004.

Plants

Y6349 Halden Environmentally Despite the title, this is I&C; it is not evident that

Assisted Cracking the current Halden program in this area directly addresses NRC needs.

Y6472 Risk-importance of Digital To investigate integration of digital systems in

Systems PRAs.

Y6591 Software Reliability Code Development of measures to quantify the

Measurements reliability of safety system software.

. .

19 NUREG-1635
Table 3. Research Activities in Digital Instrumentation

and Control Systems3 (Continued)

Project Title Comment

Y6729 Classification of Digital Vulnerability to cyber attack

System Vulnerabilities

Y6712 Safety System Isolation Study Generically qualified systems vulnerability to cyber attack

Y6733 Protocol Robustness Analyses Anticipatory research on the vulnerability of the

industry to cyber attack

Y6734 Security Tool Vulnerability Adequacy of commercial off-the-shelf tools for

Case Study preventing cyber attack

3 See also Y6651, Effects of switchgear aging on energetic faults, Table 4

NUREG-1 635 20
6 FIRE SAFETY RESEARCH

As the NRC moves from a deterministic

regulatory process to one based on an ever

greater use of information from quantitative

risk assessment, it is inevitable that some

safety issues of the past will wane in

importance and other issues will be found to

be of greater importance than previously

anticipated. Several industry-sponsored fullscope risk assessments performed about 20

years ago, as well as the insights gained

recentlyfrom the Individual Plant Examination

of External Events (IPEEE) Program [Ref. 8],

have shown that fire is a more important

accident initiator than might have been

concluded in the past. The risk from fireinitiated accidents is significant at some

nuclear power plants in both an absolute

sense and in comparison to accident initiators

internal to the plant equipment and

operations.

The importance of fire-initiated accidents was

made profoundly evident by the 1975 fire at

the Browns Ferry plant. Requirements for fire

protection beyond common industrial fire

protection practices were imposed on

licensees as backfits and compliance with

these requirements has been achieved at

considerable expense. The agency invested

heavily in experimental and analytical tools

for the analysis of fire consequences at

nuclear power plants. Since these early

efforts, progress has not been commensurate

with the major strides made in the

quantification of the probabilities of accidents

caused by internal initiators or the prediction

of the progression of such accidents. Fire

risks today are typically assessed using what

may be considered at best semiquantitative

bounding methods. Fire progression models

used in safety assessments are decades old

and do not compare favorably with the stateof-the-art methods available in other technical

4

_7_7n r-=5'7 we

f ' * ~b~ftad*, iv s'p ,b f *a lure1. ':ih tii

. _, .. , Tires.

;. .' . .. - . : C s

-, - . -.- in Fires -i n.

At4 , -, '' ia t 'l. ' ' ~ a- ' f''i;}'a_'.''!--I

nv~jestlgate the pro~i6ilities of .tailures, shtks

'to6 -the';- -groihd "' ,,aiid '.ihot"f 'i'sh'orts v >ift

h~stmumentati6on"cotro,andpowrcircuits of

ru,:,rpower plants during fires. IR7T'6NRC

Pals6'completed a' 'reiew. of cableo fire testing

in~d operational expence t' i wvith ,senous

nucle'arplant fires incduding The Bro'wns F~err

f'ire ;ii .1975 ;an'd;several 'fires --'i foreigri

'1Luclemar p ower , plants'i < NUREG/CR ,6834

M\>f 9)) Si-nce; then,'.',..a ,mbe~thod -for

.incorporating .-circuit, analyses..-into n.rsk

assessments is being used by NRCand EPRI

66 ire quantify" nsk at.two-representatwe US fr clear, power plants.;~t*'., :i ft-, {s I.. .t' ;-4.eo;.....

'' ; -.

21 NUREG-1 635
areas. Fire-induced faults in l&C systems

(hallmarks of the Browns Ferry fire) are still

assessed manually and incompletely despite

the long-held belief that such analyses could

be effectively done by computer. Currently,

there is no capability to include fire as a

consequence of accidents initiated by other

means in PRAs.

The current NRC research efforts in fire

safety (See Table 4) seem incongruent with

the estimated risk significance of fire. Fire

modeling benchmark and validation studies

(Y6688) are intended to show where agency

models stand relative to the state-of-the-art

models rather than advancing the modeling

capabilities. The fire risk 'requantification'

studies being done cooperatively with the

Electric Power Research Institute (EPRI) will

hone approximate risk assessment methods

rather than making these methods

commensurate with risk assessment methods

available for accidents initiated by internal

events. Lack of adequate risk information for

fire comparable to that provided by the

NUREG-1 150 study of internal initiators [Ref.

10] will continue to hamper the significance

determination process for deficiencies in the

licensees' fire protection programs especially

if these programs are guided by the National

Fire Protection Association (NFPA) standard

805 [Ref. 11] rather than Appendix R to 10

CFR Part 50, "Fire Protection Program for

Nuclear Power Facilities Operating Prior to

January 1, 1979."

The limited fire research effort in comparison

to the estimates of risk posed by fire is

understandable since resources for such

research have been diverted to respond to

the events of September l1, 2001. Now that

these responses are being completed, the

agency should revitalize its fire safety

research efforts and move the technical

capabilities of the agency to be more in line

with the risks being ascribed to fire. The

revitalized efforts should include:

* Automated methods

associated circuits analysis

* Improved models of

progression and damage

* Better integration of fire

including induced fire, in

PRAs

for

fire

risk,

plant

* Development of methods to

estimate the reliability of human

performance under fire conditions

NUREG-1 635 22
Table 4. Fire Safety Research Activities4

Project

Y6037 Fire Risk Assessment

* Tools for circuit failure mode and likelihood

analysis

* Tools for fire detection and suppression

analysis

* IEEE-383 rated cable self-ignited fire

frequency analysis: feasibility study

* Fire modeling toolbox: input data and

assessment

* Experience from major fires

* Cable failure mode likelihood studies

* Fire risk 'requantification' studies

* Fire significance determination process

* Fire risk assessment tools precursor analysis

* Support for ANS fire PRA standard

A Program with a large number of

important elements for fire safety.

Activities support the ROP and

improved risk analysis of nuclear

power plants.

Y6688 Fire Model Benchmarking and Validation Assesses the current status; should

be used to define what

improvements agency needs in its

modeling capabilities.

Y6651 Effects of Switchgear Aging on Energetic Deals with a significant fire

Faults protection issue.

4 See also Collaborative Research on Wire System Aging (Y6410) in Table 3 and LowPower and Shutdown Risk Study - Level 1 (W6904) in Table 9.

23 NUREG-1 635
7 REACTOR FUEL RESEARCH

Reactor Fuel Research at the NRC is

completing a quintessential example of riskinformed, confirmatory research. By now, the

basis of the work is well known though it still

provides an object lesson in the need to

sustain technical expertise in selected areas

of research crucial to the agency's mission.

In the past, the NRC research provided data

and models of fuel behavior under off-normal

and accident conditions while fuel vendors

and licensees focused their attention on fuel

behavior during normal operations. When

typical fuel burnups at the end of life were

less than about 30 GWd/t, a point was

reached in the NRC work that understanding

of fuel behavior was thought to be sufficiently

sophisticated and there was a confidence that

licensee proposals to use fuel at higher

burnups could be evaluated by extrapolation

with existing fuel models. As a result, the

NRC-sponsored research in fuel performance

was greatly curtailed in favor of other

priorities. End-of-life fuel burnups crept

upward even though it was known in the

reactor fuel community that microstructural*

changes in the fuel occurred .at average

burnups exceeding about 52 GWd/t.

Testing of high-burnup fuels in France and

later in Japan and Russia [Ref. 12] showed

that high-burnup fuels were susceptible to

damage and there was the potential for lossof-fuel coolability during reactivity-initiated

accidents. In response to these experimental

findings, the NRC made the regulatory

decision that adequate safety would be

preserved by limiting fuel burnups to 62

GWd/t unless requests for higher burnups

were supported by persuasive evidence of

adequate fuel capability. A confirmatory

research program was initiated to

substantiate this regulatory decision.

Eaillooned and ruptured high-bumup'...

-fuel 'ro'dfollowing a LOOA test and

'- ..'.ibbseque n 'tteg.

MVC is using these tests of high-bumup fuel,

along'' with extensive: 6metallurgica ;and

rna'teriais .;t£stin'g, toe con firm' '.~rgula'toiry

decisions bn'the safe use of 'reactor fuels at

bumups of 4'to'62 OWdAf.tThese'testsare

art of alargr internatlo'hal effdrt'toexamine

the effects-of highfuel burnup' and newhfuel

lladings on reactor behaviol: under design

basis. and -.s'evere, i;:accident chditios

Festinig. ould.- )idcusde I Integral ,4ests';. of

egradatfon andfissioh Priduct release.. -I

Though the technical issues presented by

high-burnup fuel dealt with design basis

accidents, the scope of the confirmatory

research was, designed by, using risk

information. This risk information pointed to

PWR reactivity accidents initiated by control

rod withdrawal, :LOCAs and anticipated

transients without scram (ATWS) in BWRs to

be of risk concern with respect to high-burnup

fuel. On the other hand, BWR rod drop

reactivity accidents were found not sufficiently

probable to merit explicit experimental

investigation.

25 NUREG-1 635
- :

This risk-significant research has also been

well organized. Technical issues and

considerations were identified using a

prestigious panel of international experts in

comprehensive phenomena identification and

ranking efforts. Both the vulnerabilities of the

fuel and the probabilities of threats have been

considered in these efforts. The research

itself was organized in several cooperative

agreements with other Nations, which greatly

leveraged the resources the NRC was able to

devote to this effort.

Research on high-burnup fuel behavior

during reactivity-initiated accidents is now

reaching a conclusion with the definition of

usable criteria for fuel performance. The fuel

performance computer codes, FRAPCON

and FRAPTRAN, have been upgraded to

treat fuel taken to elevated levels of burnup.

High-Burnup Fuel Research efforts are now

focusing on fuel behavior under LOCA and

ATWS conditions. In connection with these

research efforts (See Table 5), the ACRS

makes the following recommendations:

* The NRC should withdraw from the

CABRI water loop Project (W6832) if

tests planned in this area cannot be

changed to involve more appropriate

pulse shapes and energies for

identifying cladding failure thresholds.

* Single rod tests of high-burnup fuel

behavior under LOCA conditions

should continue and should include the

"quenching" phase of the hypothesized

scenario. Researchers should consider

whether bounding peak clad

temperature conditions should be

imposed on the single rods or whether

more realistic but limiting single rod (or

assembly) temperature scenarios are

more meaningful.

* If single rod tests continue to show

behavior that is not extraordinarily

different than anticipated based on

low-burnup fuel behavior,

considerations should evolve from

issues of fuel embrittlement to issues

of fuel coolability. The NRC should

consider joining the international

efforts being planned for in-pile tests of

multi-rod fuel bundles under LOCA

conditions.

* Research on mechanical properties

and oxidation kinetics of high-burnup

cladding, including Zirlo and M5

cladding (Y6367) should continue. This

research will allow amending 10 CFR

50.46, "Acceptance Criteria for

Emergency Core Cooling Systems for

Light-Water Nuclear Power Reactors,"

so that alternative fuel claddings that

are important both to the nuclear

industry and to the NRC can be used.

Research on El 10 cladding (Y6723)

should not be an emphasis for the

NRC since there are no indications

that the licensees will use this

cladding. Limited efforts to understand

why this alloy performs poorly in

comparison to the compositionally

similar alloy M5 might be of use, but

should not require continued use of

El 10 in tests to characterize alloy

response to accident conditions.

* Knowledge and understanding that has

been gained in the resolution of the

reactivity-initiated accident issues as

well as in the resolution of the LOCA

issues of high-burnup fuel may make it

possible to resolve by analysis the

ATWS issues of high-burnup fuel

without explicit testing of ATWS

scenarios. This possibility should be

explored by RES.

NUREG-1 635 26
* Plans to upgrade fuel performance

codes (Y6580) to address MOX fuel

should be pursued. The MATPRO

database documentation should be

updated to reflect modern data actually

being used in the fuel performance

codes.

Understanding of reactor fuel is crucial to the

NRC mission. It is widely anticipated that the

nuclear industry will be proposing further

changes in fuel cladding and fuel burnup.

Consequently, the agency must maintain

expertise in the reactor fuel area. Expertise

outside the agency and its contractors are

probably not sufficiently independent of fuel

vendors and licensees to be used on an as

needed basis to review fuel proposals by

licensees. A sustained NRC research effort in

reactor fuel behavior under accident and offnormal conditions is especially needed as

competitive pressures force fuel vendors to

curtail their research efforts. Continuing

expertise and even additional research may

be needed if, as now expected, licensees

make requests for extending fuel burnups up

-to and beyond 75 GWd/t. It is quite likely that

such industry proposals will be substantiated

by minimal experimental' research. It is

important for the NRC to have a technically

justified position on the experimental

research that is necessary to support such

proposals especially in light of the "surprise"

that accompanied extending f uel burnups into

the range of 50 to 60 GWd/t.

27 NUREG-1 635
Table 5 Research Activities in Reactor Fuel

Project Title Comment

W6832 CABRI Water Loop This Project's focus has changed

at the behest of partners and no

International program for in-pile testing of high-burnup longer meets regulatory needs in

fuel behavior during reactivity-initiated accidents reactivity accidents. NRC should

withdraw from this activity if the

tests planned for this facility

cannot be changed to involve

more appropriate pulse shapes

and energies for identifying

cladding failure thresholds.

Y6195 Dry Cask Storage License for High-Burnup Behavior of high-bumup fuel in

Fuel dry cask storage.

Y6367 High Burnup Cladding Performance An essential Project to verify

efficacy of the existing fuel

Testing of high-burnup fuel clad properties damage criteria for high-burup fuel operation. Results provide

access to data from foreign

programs.

Y6580 Fuel Code Applications for High Burnup Fuel FRAPCON and FRAPTRAN are

essential regulatory tools and

Upgrade the fuel behavior codes FRAPCON and need to be kept up to date as

FRAPTRAN for high burnup fuel licensees propose ever more

aggressive use of fuel.

Y6586 Fuel Code Assessment for MOX Fuel FRAPCON and FRAPTRAN will

be used extensively to assess

Modification of FRAPCON and FRAPTRAN for MOX proposals for using MOX fuel in

fuel to dispose of excess weapons grade plutonium commercial PWRs.

Y6723 International Agreement on Fuel Behavior Research on boron dilution

and Materials Science Research accidents and on El 10 cladding. Since there are no indications

that the licensees will use the

El1 0 cladding, research on this

cladding should not be an

emphasis for the NRC. Limited

efforts to understand the poor

performance of this alloy

compared to the similar alloy M5

might be of use.

NUREG-1 635 28
NEUTRONICS AND CRITICALITY SAFETY

Neutron transport modeling and criticality

safety are closely related to fuel safety but

are distinguished here.'These technical areas

are also crucial to the agency mission. It is

essential for public confidence, if nothing

else, that NRC has the capability to

independently ensure neutronic safety of

reactor cores and fissile materials. The tools

that the NRC uses for these independent

evaluations must be maintained at the stateof-the-art level.

The NRC research Projects in Neutronic

Analysis, Core Physics, and Criticality Safety

are listed in Table 6. The PARCS code, an

analytical tool for neutronics analysis, is also

included in Chapter 15, Thermal-Hydraulics

Research, of this report.

maintaining the capabilities of its neutronics

and core physics codes and the associated'

database. Upgrades to these codes to

address both extended fuel bumup and MOX

fuels are needed and these needs are

addressed in the research Projects. 'The

ACRS anticipates that these extended

capabilities will be tested significantly in the

certification of the advanced reactor designs.

The effort to provide guidance for highenrichment fuels (Y651 0) reflects a belief that

licensees will indeed propose eventually to

extend fuel burnups beyond 75 GWdIt as

discussed in Chapter 7, Reactor Fuel

Research, of this report. Indeed, DOE is

sponsoring research to examine fuels

enriched to 7 percent.

The ACRS assessment is that RES is doing

a commendable, cost-effective job in

29 NUREG-1 635
Table 6 Research Activities in Neutronics Analysis,

Core Physics, and Criticality Safety

Project Title Comment

Y6320 NEWT Lattice Code Lattice physics cross-sections for

MOX and high-bumup fuels.

Y6403 Reactor Core Analysis To compare the PARCS code

predictions for MOX to French

data.

Y6587 Reactor Analysis for High-B urnup Fuel Analyses with PARCS code of

high-burnup fuel for transients

and accidents involving rapid

reactivity changes.

Y6685 Upgrade Neutronic Code - Nuclear Fuel To upgrade ORIGEN source

Comp/Safety Assessment model for MOX and conventional

fuels.

Y6771 MOX Neutronics To upgrade PARCS code for

MOX fuel

Y6510 Extend Fuel Enrichment Anticipatory research for code

and guidance for fuel

enrichments of 5-10%.

Y6517 High-Burnup Source Term for Storage To develop chemical and

radioassay data for high-burnup

fuel to revise guidance on decay

heat and shielding for transport

and storage casks.

NUREG-1 635 30
9 HUMAN FACTORS AND HUMAN RELIABILITY RESEARCH

The ACRS continues to believe that human

performance issues will be important for the

continued safe operation of nuclear power

plants. The quantification of human reliability

under accident conditions will continue to be

among the most challenging -aspects of

quantitative risk assessment. Consequently,

it is important for the NRC to maintain an

active research program both in the areas of

Human Factors and Human Reliability

Analysis (HRA). The ACRS believes that

issues surrounding human performance in

teams (such as the team of control room

operators) and human performance in

organizations, including the question of

organizational reliability, are important and

need to be considered in ongoing Human

Performance Research planning. These are

both important to usafety management', a

concept that Chairman Diaz discussed

recently [Ref. 13].

The NRC research activities in the areas of

Human Factors and Human Reliability

Analysis are listed in Table 7. Human Factors

is a broad discipline that extends well beyond

the nuclear arena. The NRC research

strategy with respect to human factors is to

remain aware of pertinent developments in

the field and based on this awareness,

provide tools and guidance to line regulatory

organization to facilitate the regulatory

process. The NRC remains aware of

developments in the international nuclear

community through its participation in the

international cooperative Halden Project

(B7488) and the ACRS continues to support

participation in this Project. The NRC's

capabilities in the area of human factors are

likely to be tested in the near future as plant

staffing requirements are challenged certainly

by advanced reactor designs proposed for

certification and, perhaps, even by existing

reactors searching for competitive advantage

in a deregulated energy market. The NRC

must have defensible bases for its plant

staffing requirements for nuclear power plant

operations and security. Research is

beginning to address these'staffing issues

(Y6630). The attention'NRC is paying to

Human Factors Research is yielding

dividends. The Standard Review Plan

Chapter 18, "Human Factors Engineering,"

has recently been updated to reflect

developments especially in the area of digital

systems. A risk-informed method to screen

licensee submittals for human factors review

has the potential of better focusing agency

resources on risk-significant human factors

issues.

The Human Reliability Research at the NRC

is undertaking a potentially significant effort to

establish the basis for crediting operator

actions at nuclear power plants (Y6022). Risk

assessments now credit only proceduralized

actions for which operators are trained. It is,

however, well known that operator actions

outside of normal procedures can provide

significant safety benefit. This research holds,

then, the promise of adding greater realism to

quantitative risk assessments.

Quantification of human reliability remains a

troublesome feature of risk assessments.

Numerous plausible approaches have been

devised that often yield disparate results.

The database for the validation of these

approaches remains distressingly limited.

Uncertainties from both parameters and the

models are seldom evaluated. The results of

the benchmark exercise conducted by the

Ispra Laboratory of the 'European Union

[Ref. 14], although admittedly fairly old by

now, are troubling. These results show that

the choice of HRA model has a'significant

impact on the results. This is one of the few

areas in Level 1 PRA in which model

uncertainty is significant. The ACRS has not

seen a critical review by the NRC of the

31 NUREG-1 635
merits and disadvantages of the existing

models. Often in the regulatory process

simple, transparent methods such as THERP

that deal only with human errors of omission

or industry-developed models that have not

received extensive review are used. The

regulatory process seems to accept these

simple methods in the absence of guidance

on the necessary levels of rigor and accuracy

needed for quantification of human reliability.

Acceptance of such simple models without

substantiation by a robust database

undermines any proposal to develop credible

human reliability models.

The ACRS notes that RES is preparing a

document on its assessment of "good

practices" that will support all aspects of

HRA, including the quantification process

(W6994). These descriptions of good

practices are needed to supplement

information provided in the ASME PRA

standard. The ACRS understands that the

second phase of this project (planned for

2005) will be a review and evaluation of

existing HRA approaches for their capability

to meet the "good practices" when employed

to address different regulatory applications.

The ACRS anticipates that this review will

address the important issue of model

uncertainty.

because of both its cumbersome, manpowerintensive character and its lack of

quantification. The NRC is now applying this

methodology to two important issues:

* Human performance during fires

* Human performance issues associated

with pressurized thermal shock

The ACRS looks-forward to reviewing the

results of these application efforts and

learning whether the method has become

easier to apply and whether it yields insights

significantly better than those that could be

derived from simpler methods. Of particular

interest to the ACRS is the use of expert

opinion elicitation in these applications to

quantify the ATHEANA results.

The staff needs to show how the DavisBesse incident has affected the set of

performance shaping factors and what

guidance is being provided to the ROP.

The NRC is developing an events database

called Human Event Repository and Analyses

( HERA ) (Y6123). This database may be a

key step toward much improved analyses of

human reliability in nuclear reactors. It should

provide an useful basis for evaluating the

model uncertainties mentioned above.

The NRC is experimenting with an ambitious

methodology for estimating human reliability

to assess errors of commission as well as

errors of omission. This methodology called

ATHEANA (W6994) focuses on the context

that drives human performance

("performance shaping factors"). The ACRS

has been critical of this approach in the past

NUREG-1 635 32
Table 7. Research Activities in Human Factors

and Human Reliability

Project Title Comment

Y6123 Retrospective Risk Analysis of Valuable Project to develop HERA database

Human Reliability in Operating on human reliability from past operational

Events experience.

Y6832 Human Reliability Analysis for Identifies human reliability issues for NMSS -

Byproduct Material tools of use for quantification of risk.

B7488 HALDEN Reactor Project This is a long-term Project in Human

Performance that is contributing to NRC work

in human factors.

Y6022 Credit for Operator Action This is a small effort to provide guidance for

NRC reviewers. It has the potential of adding

greater realism to PRA.

Y6630 Development of Regulatory An important proactive measure in anticipation

Guide and Analytical Techniques of trends in current reactors and themes in the

for Assessing Nuclear Power development of more advanced reactors.

Plant Staffing

W6994 Review, Application and This is a very ambitious effort to characterize

Refinement of ATHEANA human performance. At issue is whether

ATHEANA can be applied practically to the

issue of reactorsafety. This effort will also yield

guidance for performing and reviewing human

reliability analyses and supplementing

information provided in the ASME PRA

standard.

33 NUREG-1635
10 MATERIALS AND METALLURGY

Research Projects in the area of Materials

and Metallurgy are listed in Table 8. These

Projects can be grouped into the following

broad categories:

* Reactor Pressure

Embrittlement (W6953,

Y6378, W6986, Y6396,

Y6737)

VesselI

Y6533,

Y6638,

* Neutron Fluence Determination

(Y6391,Y6742)

* Environmentally Assisted Cracking

(Y6536,Y6588, K6266, Y6270,Y6388,

Y6722)

* Monitoring and Inspection (Y6534,

Y6604, Y6649, Y6869, Y6882,

Y6909)

* LOCA Frequencies and International

Standards (Y6296, Y6538, Y6744)

* Proactive Materials Degradation

Assessment (Y6919)

These Projects represent a well-balanced

program that address current regulatory

needs, including the need for the Office of

Nuclear Reactor Regulation (NRR) to

maintain an independent analytical capability.

in various areas of materials and metallurgy.

Capabilities are transitioned to NRR as they

are developed, and new projects are

introduced as their need becomes apparent.

The ACRS recommendations on various

Projects in the Materials and Metallurgy area

are given below and summarized in Table

8. Many of these Projects are part of

collaborative programs, involving about 65

International Organizations, thereby giving

good leverage of the NRC resources. -

e g o ~ t:'; .4 "

k '; S j > * v ' .; z ~~~~~~~i ,v rLL 

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-n*ik1r i..ts*$ ,\,m; io-e- i 'v,< -{eRi C t ,-; %s ._

Exampl ot,. a.. P , .... ^t'!.*red'm.natlAt ,I4. ;' *- Ite-r-ranulaX Stress C roio Crac

-/~~ a. I; -'4 A- ;--4:

Fig~ J .; ' ' .................................... , . X -0

The Projects dealing with reactor pressure

vessel embrittlement should be considered

both with respect to safety issues of reactor

pressure vessel integrity and with respect to

maintaining whatever core competency the

agency needs in irradiation embrittlement.

The staff needs to evaluate what research

should be retained in order to meet these two

needs. The remaining technical issue

involves high-neutron fluence on the lowcopper, high-nickel pressure vessel material

found in just a few plants (W6953, Y6533).

Work in Y6533 will support risk-informing the

pressure- temperature limits in 1OCFR Part

50, Appendix G, "Fracture Toughness

Requirements." Project Y6638 provides

support for revising Regulatory Guide (RG)

1.99, "Radiation Embrittlement of Reactor

Vessel Materials."

Pr~oject Y6378

participation in

Energy Agency

ensures continued NRC

the International Atomic

(IAEA) deliberations on

35 NUREG-1635
pressure vessel integrity and will contribute

to the preservation of core competency.

Peer review of the technical basis for the

reevaluation of the PTS screening criterion

(Y6737) is most important and should be

continued.

Work in Project W6986 should be

terminated since its objectives have been

met and fracture mechanics technology

development will be consolidated in Y6533.

The development of analytical capabilities

for neutron fluence (Y6391,Y6742),

especially in the core regions near the

bottom and top of the core, are important

since cracking of major welded

subassemblies in BWRs has occurred in

these regions. The ability of the NRC to

independently check the licensees'

arguments for disposition and repair of these

components with complex geometries is

important. Hence, the uncompleted tasks

of these Projects should be continued in

FY2004.

Corrosion, especially environmentally

assisted cracking, is a complicated technical

issue involving the interplay of chemistry,

corrosion science, metallurgy, and

mechanics. The current concerns about

irradiation effects on cracking of core

components further complicate this situation.

The management of environmentally

assisted cracking incidents in LWRs is the

prime responsibility of the licensees.

However, the NRC must have independent

capabilities to assess the licensees'

positions regarding the structural integrity of

reactors and their proposed mitigation

actions. Indeed, recent events at the Indian

Point and Davis-Besse nuclear plants have

demonstrated that licensees' corrosion

control programs merit continued scrutiny

and monitoring. Consequently, the NRC

should have active programs in the areas of

materials degradation data collection (in

order to assess the kinetics of degradation),

monitoring and inspection (e.g., NDE), and

codification.

A particularly vulnerable element of the

reactor coolant system in PWRs is the steam

generator. Rupture of steam generator tubes

can lead to accidents that allow radioactive

materials released from the core to bypass

the reactor containment and enter into the

environment. Risk analyses show that severe

accidents involving containment bypass can

be risk dominant at some PWRs. Modes of

corrosion (e.g. cracking, denting) of the

steam generator tubes have changed over

the years as changes have been made in

both the materials used for the tubes and in

the PWR coolant chemistry. The complexity

of these phenomena and the risk significance

of tube failures make it important for the

NRC to monitor these issues. This need does

not abate as further changes from alloy 600

tubes to alloy 690 tubes are made by the

industry. Although alloy 690 is more resistant

to primary water stress corrosion cracking,

the complexity of the environments possible

on the secondary side of steam generators

makes it impossible to preclude the potential

for stress corrosion cracking. Therefore, the

Projects (Y6588, Y6536) associated with the

NRC's Steam Generator Tube Integrity

Program should be continued.

In recent years, environmentally assisted

cracking in LWRs has raised other concerns

such as irradiation-assisted stress corrosion

cracking (IASCC) of stainless steel core

internals of BWRs and stress corrosion

cracking /general corrosion of vessel head

penetration (VHP) subassemblies in PWRs.

These latter incidents have been the subject

of several Bulletins and an Order.

Consequently, it is appropriate for the NRC

to sponsor research to independently

evaluate licensees' plans to mitigate these

phenomena. Projects Y6270 and Y6388

should be continued in FY2004 since they

NUREG-1 635 36
provide the essential cracking kinetics data

required for in-service inspection (ISI)

regulations. Project K6266 should also be

continued since it relates to the

quantification and understanding of IASCC

issues; The value-,of this research is

enhanced by collaboration with several

International Organizations. Project Y6722

was instigated by the Davis-Besse incident

and several of the tasks have been

completed. This Project should be

continued to complete the remaining tasks

and to ensure adequate independent

analysis of the licensees' extensive program

spawned by this incident.

Nondestructive examination methods are

crucial for monitoring the integrity of the

reactor coolant system. These methods for

detecting corrosion and cracking have

improved greatly in recent years largely due

to NRC research activities. However, the

reliability of these methods in the field

continues to be an issue. Sustained NRC

research (Y6534, Y6604, Y6649, Y6869,

Y6882, Y6909) on the evaluation of these

nondestructive examination methods is

crucial to the agency mission. Projects

Y6649, and Y6869 merit increased attention;

these relate to the inspection methodologies

(technique and frequency) that should be

used for VHP assemblies. In the latter

project, a technical basis for reevaluation of

requirements related to reactor coolant

leakage will be developed.

failure database. This is a further input

needed for risk-informing 1 OCFR50.46.

Project Y6296 involves NRC participation in

the development of international standards

and should be continued.

Over the years, the NRC has been surprised

by a number of materials degradation issues.

The ACRS supports the "Proactive Materials

Degradation Assessment" Pr6ject (Y6919)

recently initiated by the NRC in order to move

from the economically inefficient reactive

mode of regulating materials degradation

issues. The plan to lead a PIRT effort with

the industry over the next two years, with

expert elicitation from international experts,

should identify potential material

degradations that may occur in the future.

This knowledge will provide a better basis for

future regulatory decisions on timely

inspection and repair criteria.

A major NRC regulatory initiative is riskinforming 1 OCFR50.46 and redefining largebreak LOCA in terms of LOCA frequency

distributions. Part of this work involves the

development a new probabilistic fracture

mechanics (PFM) code for piping. The

current PFM code, PRAISE, has been a

valuable tool, but an improved code is

needed to better address materials

degradation modes. Thus, Project Y6538

should be continued. Project Y6744

maintains NRC access to the CSNI pipe

37 NUREG-1635
Table 8. Research Activities in Materials and Metallurgy

Project Title I Comment

Reactor Pressure Vessel Embrittlement

The remaining technical issue involves

W6953 Heavy-Section Steel Irradiation Program high-neutron fluence on the low-copper, high-nickel pressure vessel material

found in just a few plants. Work in Y6533

will support risk-informing the pressureY6533 SST-temperature limits in 10CFR Part 50, Y6533 H55T-3 Appendix G.

Y6378 International Pressure Vessel Technology Representation of NRC at IAEA Division

Cooperative Program of Nuclear Power on RPV structural

integrity issues.

W6986 Fracture Mechanics Technology for LWR Should be terminated since its objectives

Materials have been met and fracture mechanics

technology development will be

consolidated in Y6533.

Y6396 Radiation Embrittlement Damage Analysis Longer term confirmatory research with

& Prediction current emphasis on embrittlement at

end-of-life fluences.

Y6638 Statistical Analysis of RPV Steels Longer term support for rev 3 of RG 1.99

Y6737 Peer Review of PTS Technical Basis A critical Project at the end of a very

successful program on reevaluation of the

PTS screening criterion.

Neutron Fluence Determination

Y6391 Boiling Water Reactor Fluence Support for development of independent

NRC analysis of Industry code for neutron

fluence. Uncompleted tasks of this

Project should be continued in FY2004.

Y6742 BWR Reactor Vessel Samples Measurement of helium and boron

concentrations in BWR core samples and

correlation to fluence. Critical to weld

repair decisions. Uncompleted tasks

should be continued in FY2004.

NUREG-1 635 38
Table 8. Research Activities In Materials and Metallurgy

(Continued)

Project Title | Comment

Environmentally Assisted Cracking

Y6536 PWR Primary System Components' Addresses the potential for steam

Behavior Under Severe Accident Loads generatorbypass in severe accidents and

should be continued.

Y6588 Steam Generator Integrity Program -3 Addresses potential degradation modes

for new, degraded or repaired tubes and

an assessment of their consequences (e.g. leak rates) and control (e.g., ISI)

techniques. Should be continued.

K6266 CIR-Il Cooperative Agreement Cooperative program, organized by EPRI

to predict the irradiation assisted cracking

characteristics of new and existing LWR

core materials. Good leverage of NRC

funds. Should be continued.

Y6270 Environmentally Assisted Cracking Evaluation of IASCC and fracture

behavior of stainless steels used BWR

core structures. Relevant to license

renewal. Should be continued.

Y6388 Environmentally Assisted Cracking in Quantification of a variety of cracking

LWRs phenomena (corrosion fatigue, IASCC) to

develop independent life prediction

capability for low alloy steel, stainless

steel and nickel base alloys in LWRs.

Should be continued.

Y6722 Degradation of RPV in Boric Acid Project instigated by Davis-Besse

incident. Several tasks have been

completed. Should be continued to

complete the remaining tasks.

39 NUREG-1 635
Table 8. Research Activities in Materials and Metallurgy

(Continued)

Project |Title I Commentl

Monitoring and Inspection

Y6534 Piping - Non-Destructive Examination Cooperative international program to

(NDE) Reliability develop NDE methods. Should be

continued.

Y6604 Evaluation of the Reliability of NDE Evaluation of effectiveness, reliability and

Techniques adequacy of advanced NDE methods

necessary for e.g. cast stainless steels.

Should be continued.

Y6649 Alloy 600 Cracking - Phase II Merits increased attention. Relates to

inspection methodologies to be used for

VHP assemblies.

Y6869 Barrier Integrity Research Program Merits increased attention. Provides

technical basis for reevaluating the

requirements of reactor coolant leakage.

Y6882 Technical Assessment of Bare Metal Evaluation of the inspection techniques to

Inspection Techniques be used for PWR bottom head

penetrations. Should be continued.

Y6909 Examination of North Anna Unit 2 Head Establish, via inspection of nozzles from

North Anna Unit 2 and Davis-Besse, the

development of the leak path as

cracking/degradation occurred. Relates

to adequacy of BM_

NUREG-1 635 40
Table 8. Research Activities in Materials and Metallurgy

(Continued)

Project Title I Comment

LOCA Frequencies and International Standards

Y6296 ISO Participation Ensure NRC participation in the

development of international standards.

Should be continued.

Y6538 Technical Development of LOCA Support for risk-informing 1 OCFR 50.46

Frequency Distributions and redefining LBLOCA. Should be

continued.

Y6744 CSNI Piping Data Base Cooperative program to maintain NRC

access to CSNI pipe failure database.

Should be continued.

Proactive Materials Degradation Assessment

Y6919 Materials Degradation PIRT Project recently initiated by the NRC. The

plan to lead a PIRT effort with the industry

over the next two years, with expert

elicitation from international experts,

should identify potential material

degradations that may occur in the future.

This knowledge will provide a better basis

for future regulatory decisions on timely

inspection and repair criteria.

U - -

41 NUREG-1 635
11 PROBABILISTIC RISK ASSESSMENT

As the NRC moves toward a risk-informed

regulatory system, it will need ever greater

amounts of risk information derived from

quantitative risk assessments. It will want to

maintain its own independent capabilities for

conducting PRAs at state-of-the-art levels. A

continued vital research program in PRA is,

then, crucial to the agency mission.

The NRC research activities in the area of

PRA are listed in Table 9. A major thrust is

the support for the ROP which is a major

NRC initiative in the use of risk information for

monitoring the operations of nuclear power

plants. Support for the development of the

SAPHIRE code suite and the SPAR models

is important and continuing (Y6394, W6355,

W6467, Y6153, Y6553). Also under way in

support of the ROP is the development of

objective, risk-based, performance indicators

(J8263, Y6370). A candidate metric called

the Mitigating System Performance Index is

being tested. There is also continued support

for the significance determination process

(Y6553) which is an important step in the

assessment of findings from the ROP that

continues to challenge the agency.

Another major element of the current PRA

Research is the expansion of the method into

the analysis of dry cask storage of spent

fuel(Y6423, Y6502, Y6612). Such storage of

spent fuel will become ever more necessary

as the development of a permanent geologic

repository for spent fuel continues to be

delayed.

Research activities of importance are also

under way to support risk-informed revisions

to 10 CFR 50.46 (W6224, Y6538).

In addition to the many demands on

resources for the application of PRA

methods, resources are still made available

for developing agency capabilities in PRA

(K6007). The ACRS generally supports the

idea that the NRC needs to maintain its PRA

capabilities at or near the state-of-the-art

level. A detailed examination of the tasks

funded within this developmenteffort shows

that many overlap with distinct efforts in other

fields. Some items with substantial overlap

include the study of turbulent mixing in

complex geometries and software reliability.

Greater coordination among research within

the NRC may eliminate the need for such

overlapping research efforts.

The ACRS would prefer to see work to

expand the scope of the NRC PRA

capabilities. Work to develop capabilities to

analyze risk during low-power and shutdown

conditions at nuclear power plants (W6904

and Y6103) is very important. The ACRS

strongly supports these efforts that will extend

PRA capabilities toward the eventual goal of

handling all modes of plant operation. Work

to better integrate fire-risk assessment,

including fires induced by other accident

initiators, is a need not now being met

despite evidence of risk significance coming from the IPEEE Program.

Altogether the NRC has an impressive PRA

Research Program. This Program is all the

more impressive since many of the agency

resources in risk analysis are being applied to

security issues that are not addressed in this

report.

43 NUREG-1635
Table 9. Research Activities in Probabilistic Risk Assessment'

Project Title Comment

Y6423 Dry Cask PRA Expanded application of PRA

into an issue that will be of

Y6502 PRA for Dry Cask Storage - Peer Review growing importance over the next few years.

Y661 2 Reliability of a Secondary Containment Radionuclide release paths for

Isolation (SCI) accident involving drop of a dry

spent fuel storage cask.

J8263 Reactor Oversight Process Support Essential support for the Reactor

Oversight Process.

Y6394 Maintain and support SAPHIRE Code and

Library for PRA

W6355 SPAR Model Development: Low

Power/Shutdown

W6467 SPAR Model Development: Level 1 Rev 3

Y6153 SPAR Model Development: Level 2ILERF

Y6553 Significance Determination Process

Front-end Interface for SPAR Models

Y6370 Development of Risk-based Performance Exploratory activity that has

Indicators yielded a candidate metric that is

now being tested.

Y6522 Synergistic Effects of Power Uprates Important effort to identify issues

not revealed in the examinations

of individual changes to the

licensing bases of nuclear power

plants.

K6007 Probabilistic Assessment and Applications Program of several activities to

improve PRA capabilities.

* Model and parameter uncertainty

* Software reliability Greater coordination among

* Thermal-hydraulic uncertainty research within NRC may

* Turbulent mixing within complex geometries eliminate the need for

* AP1000 LBLOCA uncertainty analysis overlapping research efforts in

* Importance analysis this Project.

* Causal modeling

5 See also Y6332, Digital Systems Risk, Table 3

NUREG-1 635 44
Table 9. Research Activities in Probabilistic Risk Assessment

(Continued)

Project Title Comment

W6144 Technical Support in Risk Assessment Staff augmentation

W6224 Risk-informing 10 CFR Part 50 Support for the effort to riskinform 10 CFR Part 50

Y6538 Technical Development of LOCA Frequency

Distributions

W6904 Low-Power and Shutdown Risk Study - Essential study of one of the

Level 1 important elements of risk posed by nuclear power plants during

low-power and shutdown

Y6103 Low-Power and Shutdown Risk Study- operation that has been

Level 2 inadequately studied in the past

W6970 Support to Develop Consensus PRA Necessary NRC participation in

Standards standards setting activities

Y6335 Risk-informed Initiatives for Nuclear Safety goals and guidance for

Materials nuclear materials and waste

Y6430 PTS Risk Assessment Activities in support of the

agency's reevaluation of the PTS

screening criteria

Y6486 Severe Accident Initiated Steam Generator Essential element of the agency

Tube Rupture Sequences strategy to resolve issues of

steam generators

L

45 NUREG-1 635
12 RADIATION PROTECTION

The NRC research activities in the area of

Radiation Protection are listed in Table 10.

These activities constitute a well-leveraged

program that allows the NRC to use quality

data and information as the foundation of its

basic radiation protection regulation, 10 CFR

Part 20. A major thrust of these activities is

the collection of occupational exposure data

(Y6133 and Y6698). Another important thrust

is the support for. commissions (ICRP and

NCRP) that establish radiation protection

standards (G6071, G6251, G6589). There

are also activities that support dose

assessment by the NRC (Y6433, Y6460,

Y6470). These are all essential activities for

the agency mission and should be sustained.

A potentially important activity (Y6330) would

harmonize requirements from multiple

agencies for complex site assessments. This

is' a laudable joint effort with the

Environmental Protection Agency (EPA) to

develop common databases for use in

environmental dose assessment calculations.

Another cooperative effort (Y6470) is yielding

a modeling platform that gives access to a

wide range of system and process models as

well as databases used by other government

agencies. This platform will eventually yield

a powerful environmental modeling capability

for use in the evaluation of complex,

contaminated sites.

47 NUREG-1 635
Table 10. Research Activities in Radiation Protection

Project Title Comments

G6071 BEIR - VIl

The Phase II study deals with low doses and low dose NRC contribution to the

rates. The BEIR-VIl effort is to conduct a comprehensive development of consensus

reassessment of health risks resulting from exposure to standards in radiation

ionizing radiation since the 1990 BEIR-V [Ref. 15] report protection.

G6251 Radiation Protection Standard Development ICRP support; ICRP is

ICRP task groups will provide information on radiation- considering the revision of

associated risks of cancer and severe heredity effects, ICRP-60 [Ref. 16] by 2005 that

application of recommendations in terms of secondary simplifies the current system of

limits, and the principles of optimization and radiological protection.

management of radiological protection

G6589 NCRP Study on Radiation Protection Issues NCRP support.

Y6133 Collection and Analysis of Occupational Radiation Analysis of required licensee

Exposure Data submittals on occupational

exposures.

Y6298 Technical Bases Information Clearance Information to support Commission decisions on

rulemaking.

Y6330 Support for Interagency Cooperative Research Interagency harmonization of

complex site evaluations.

Y6407 NUREG-1640 [Ref. 17) Finalization - Clearance Development of conversion

from dose to concentration for

materials

Y6433 Extremity Code Update the VARSKIN code to

This Project is to correct errors and limitations in the confirm licensee skin dose

VARSKIN Mode 2 computer code identified by licensees assessments.

Y6460 Update Codes for Assessing Radiation Doses Update dose coefficients

Y6470 FRAMES Software Development Well-leveraged effort to

provide agency a dose

assessment tool for complex

situations. Good example of

interagency cooperation.

Y6698 Support NEA Information System on Occupational Essential data for agency

Exposures mission.

Data are used to compare U.S. and foreign nuclear

power plant initiatives and identify best practices during

maintenance and operational procedures

NUREG-1 635 48
13 SEISMIC RESEARCH

Seismic events produce a baseline risk for

nuclear power plants. The NRC has invested

heavily in the understanding of seismic issues

and development of its rules to ensure that

seismic risks are acceptably low. Licensee

submittals concerning risks associated with

external accident initiators such as seismic

events suggest these efforts have been

largely successful. Issues concerning the

estimation of seismic hazards of power plant

sites in the eastern United States based on

methodologies developed independently by

EPRI and NRC have been resolved. The

NRC research activities in the seismic area

are listed in Table 11.

Much of the NRC research into seismic

issues involves cooperative efforts to address

essential seismic data. This work is highly

leveraged by contributions from other

partners. There is advantage in maintaining

modest research efforts to accrue benefits

from cooperative research on seismology and

seismic engineering.

Projects W6829 and Y6718 are addressing

specific issues that should be done by

licensees. It is not clear that the NRC needs

an independent assessment of these issues.

Unlike many other issues that are dealt with

by the NRC research, there is in this Country

a very substantial engineering and scientific

community addressing issues of seismic

hazards and the engineering of structures to

avoid seismic damage.'By'and large, this

community is independent of the NRC

licensees. Thus, the NRC-sponsored

research is not needed to maintain a core

competency in this area. If there are open

issues associated with licensee submittals

such as the seismic behavior of spent fuel

casks or buried structures, the NRC could

rely on use of engineering consulting firms on

an as-needed basis to deal with such issues.

, ~s.. ...-A Succ~ssStorY , -7 TZT ¢--

Probabilistic Seismi'c Hazard Anilysis:..

','Guidance.-on' Unce'ifaInty and Use'of-:'

~ariy^ lndustry-sponsored ,~, PRAs, -such-as

.thbse'foi:Zion ~and Indian Point, s'howedthe

Wigniicance of-seismic contrbtutiohs"

ror' inuclear ' psowersplants.; 'Later studies

confirmed-these finding1s,.`'A'idifficulty 'iin

fruantifying 'seismticrsk,'especiallyf orplants

east 'of the; ocky Mountains,w th

sparsity'of data. Thisled oextensive use of

expert~'-judgiment .;-with :,-attendan't iarge

lunceJtainties,: .in.'.the. ,inIyses. srnportant robabilis tic ,-,teismic :z hazard ',! analyses

sponsored-,by the -NRC-andbji.EP.RI in the

9--980s hig' -different.expe.rts andiiffererii .tS L '., ' o'!s6. *I' ,r . ii

ihe, .controversyK associatedi with .fthes6

ivbrgent results was not conducive to public

confidence -lh the regulat or. proess.o

rear'ess thlssltuation, NRC, the Departmenr

"Eergy,- nf aid. ,EPRI, "flm eJd the "Senior

;eismic", Hazard ..Analysls.i.iCommittee

SSHAC) top'r6ide.a procedure for btaining'

heprdd~icible tesults. In a two-volume report" .SSHAC'ldentified -keyissues 'egarding the

h&fantification :612-; unc'ertainties ',anhd.,' the, btIlization tof expertopbinion." ^;: ,.>ic;k.*X5>

A National Researclh co 'ciuc Panel revie!wed

*his work ard concluded tha't. ".-';the SSHAC'

epor's disussionsandrecommendations on'

.vncertainty and theuse-'ofexperts are quite

rndependent'of PSHA~and-can be applied to

other -types- of lrisk..anslysis.' 'The '4Panel

beiieves that theSSHAC report makes a solid

cdntribution'. to, the 'methodologyl of :hazard

analysis, -especially -in the, use of expert

opinion [Ref v'18],,A,^ '-t.-i''",':^, '

49 NUREG-1 635
Table 11. Research Activities in Seismic Phenomena

Project Title Comments

W6081 Japanese Collaboration on Seismic Issues Five-year cooperative agreement with Japan. Highly leveraged and

should be sustained

W6829 Seismic Behavior of Spent Fuel Storage Why should NRC do this work

Cask Systems rather than licensees?

Y6166 Cooperative Geoscience Research Continuing work on seismic

hazard of central and eastern

U.S. Adequate understanding

now available for regulatory

processes. The need for this

work is questionable.

Y6233 Garner Valley Downhole Seismic Array Ongoing data collection of limited

Operation and Analysis of Data potential impact on regulatory processes.

Y6481 Senior Seismic Hazards Analysis Committee Expert opinion elicitation for 10-

(SSHAC) Method year update of a semiquantitative analysis for 10 CFR Parts 72

[Ref. 1 91 and 100 [Ref. 20].

Y671 8 Soil-Structure Interactions for Buried It is not apparent that NRC needs

Structures to conduct this research even in

anticipation of advanced reactor

technology. It is only necessary

that NRC have available to it the

resources necessary to

adequately and independently

review licensee submittals

concerning this issue.

Y6796 IAEA Coordinated Research Project on This is primarily an effort to

Seismic Ground Motion acquaint other Nations through IAEA with NRC data and

information on ground motion

from nearby earthquakes of

moderate intensity.

Y6797 Evaluation of USGS 2002 Seismic Hazards Support for SSHAC method.

Assessment Provides seismic hazard

information for central, eastern

and other regions of U.S.

NUREG-1 635 50
14 SEVERE ACCIDENT RESEARCH

,_. In the past, the NRC invested very heavily in

the experimental and analytical

characterization of severe reactor accidents

that dominate the risk to the public posed by

the use of nuclear reactors to produce

electrical energy. A very substantial

technology has been established to

understand the progression of reactor

accidents and the radiological consequences

of such accidents. Once its immediate needs

were met to understand severe reactor

accidents sufficiently well to estimate risks to

the level of confidence needed and to provide

assurance of adequate protection, the NRC

substantially curtailed its investments in

severe reactor accident research. The NRC

Severe Accident Research activities other

than those directly associated with advanced

reactors are listed in Tables 12 and 13.

Research on Severe Accidents has been

continuing in other Countries because they

have more restrictive acceptance criteria than

adequate protection. Substantial programs

are under way in both Europe and Japan.

The NRC has developed quite an effective

strategyto maintain the technologyforsevere

accident analyses and to update this

technology with research results from

international programs. The body of

knowledge coming from the NRC's past work

and ongoing work are systematized in

useable form in the MELCOR accident

analysis code. At the same time, the NRC is

entering into international cooperative

research programs to obtain data for

validating the MELCOR code and improving

its accuracy and realism. Current agreements

include the following:

%R.:eblistic Reactor Accident Analyses H

MELCOR Prediction of PHEBUS Data

Hydrogen Production FPT-1

0.18

{t|- MELCORl

* * PHEBUS Data U0.12

0.06

0.00 2t

-0.061

8000 12000 16000

Time (seconds)

MELCOR Prediction of PHEBUS Data

Iodine Release Fraction FPTA1 . 1.0 ] z | z . . . I

o 0.8-

1t

U

U. 0.6-

410

0 0.4

. 2

aMELCOR

| * PHEBUS Data I

nnP-. . . . I . . . I

8000 12000 16000 20000

Time (seconds)

RES.':is>sic6eedingg'~iri~achieving,^-greater

uracy and;realisi' ii~. its iiseverereactor

a'c cident .£ea naly y~s,. . 7 7w-'. , xfor-.~ . 1plant' ;risk

sments;by_.validatfig MELCO?,F model

predictions , with' data from -several

jntemationab . cooperative experimental

programs. Figures above show campansons

bf MELCOF code predictions with 'datafrom

thb HEBUS- FPT-'1in-pile testof irradiated

b- a 'c tor f u el d e gra 'd atio n . A; ' ' *,- *.-l,+~S' -2, d

51 NUREG-1 635
0 MASCA * OECD-MCCI

This is an experimental study underway

in Russia on the behavior of reactor core

debris in the lower plenum of a reactor

pressure vessel. Results of these studies

will be needed to evaluate safety claims

related to the certification of both the

AP1000 and ESBWR designs.

* ARTIST

This is an experimental study under way

in Switzerland to measure the aerosol

removal on the secondary sides of

steam generators during accidents at

PWRs that bypass reactor containments.

Such bypass accidents are often risk

dominant for PWRs. The high risks

associated with such accidents may

stem from conservatism in the aerosol

decontamination assumed in accident

analysis models for steam generators.

Test results are expected to provide the

basis for more realistic analyses of these

accidents.

* PHEBUS-FP

This Program consists of five largescale, in-pile integrated tests of fuel

degradation, fission product release,

radionuclide transport through the

reactor coolant system, and aerosol

behavior in the containment. These tests

have been designed to validate reactor

accident models. Exceptionally

prototypical results for code validation

are being produced by this international

program (see sidebar). Additional

information is being provided by

supporting separate-effects experimental

programs, such as the French Program

VERCORS, to investigate fission

product release from MOX and highburnup fuels under accident conditions.

This is an experimental study of the

viability of using an overlying layer of

water on reactor core debris that has

escaped the reactor coolant system and

is interacting with structural concrete of

reactor containments. This research will

provide data for improved or new models

of core debris coolability for systemslevel accident analysis codes. Improved

models will be used in the risk analysis

of existing plants and the safety

evaluation of advanced reactor designs.

This Program is to be completed in

FY2005.

These highly leveraged international

programs are providing the bases for

validating the NRC's accident analysis code,

MELCOR. The MELCOR code has proven its

utility to the regulatory process in recent

years in connection with resolving issues

such as the need for hydrogen igniters in ice

condenser and Mark IlIl containments and

risk-informing 10 CFR 50.44, "Standards for

Combustible Gas Control System in LightWater-Cooled Power Reactors." MELCOR

analyses are also important for the

certification of the AP1000 and ESBWR

advanced LWR designs. MELCOR analyses

will likely be instrumental in the certification of

other advanced reactor designs.

The ACRS is very supportive of the strategy

the NRC has developed to maintain and

update its capabilities for accident analyses.

The ACRS encourages the NRC to give

serious consideration to joining international

cooperative research agreements now being

proposed to examine:

* Prototypic, in-pile tests of radionuclide

releases from, and degradation of,

reactor fuel exposed to air for both

reactor accident analyses and spent fuel

storage pool accidents

NUREG-1 635 52
* Prototypic, in-pile tests of radionuclide

releases from, and degradation of, highburnup fuels

* Prototypic, in-pile tests of radionuclide

releases from, and degradation of, MOX

fuel

The ACRS also supports programs to

maintain the MACCS code' and looks forward

to reviewing results of the comparison of the

Gaussian plume model to state-of-the-art

dispersion models in the context of risk

assessments.

o Sound technical basis for evaluating

energetic fuel-coolant interactions

* Aerosol behavior in steam generator

secondaries and other pathways that

bypass reactor containments

* Effects of high burnup on core-melt

progression and fission product release

The ACRS foresees the following

challenges in the area of severe accident

analysis:

* Assessment of in-vessel retention with

respect to materials interactions

* Spent fuel pool accident progression

assessment especially the effects of

hydrides, effective ignition temperature,

extent of pool involvement, and releases

of fission products and actinides, as well

as transport energetics

* Evaluation of severe accident

uncertainties for PRA applications

53 NUREG-1 635
Table 12. Severe Accident Research Activities

Project Title Comment

Y6312 MASCA Program Supports certification of the AP1000

International cooperative research on core and ESBWR designs. The MASCA

debris interactions with the pressure vessel experimental efforts will be completed

head soon.

Y6313 OECD- Melt Coolability and Concrete International cooperative research on

Interactions Program core debris coolability during interactions with concrete.

Y6668 Analytical Support for Y6313

Y6321 Benchmark MOX Fuel Release Source Research to define fission product

Term Experiments release from MOX fuel.

Y6377 MELCOR Code Development and Repository for research results

Assessment generated in cooperative international

programs. Code is being used

Y6802 MELCOR Severe Accident Code extensively for revising regulations and

Development and Assessment certifying advanced LWRs.

Y6504 Steam Generator Fission Product Effort to add realism to predictions of

Retention releases during bypass accidents.

Y6607 Support for ARTIST Tests of Fission

Product Retention in Steam Generators

Y6571 CSARP Meeting Useful annual meeting to review

cooperative research results.

Y6512 Tests of Severe Accident Phenomena Phenomenological tests and analyses

in Oxidizing Medium for spent fuel pool accidents.

Y6664 Containment Analysis and Experiments Support for AP1 000 certification.

for AP1000

Y6668 Containment Analysis Support Containment analysis of advanced

reactors.

Y6848 High Burnup Fission Release Data Receipt and analysis of data from

French VERCORS tests.

Y6696 AP 1000 Severe Accident Phenomena Severe accident phenomena

Evaluation assessment of AP1 000.

Y6721 Agreement with IBRAE-RAS on Nuclear Very cost-effective effort to modernize

Safety Analysis Codes the MELCOR code.

NUREG-1 635 54
Table 13. Research Activities for Accident Consequence Models

Project Title Comment

Y6785 Plume Model Adequacy Evaluation Test the assumption that simple

plume treatments in MACCS

code are adequate by comparing

with the state-of-the-art

dispersion model.

Y6628 MACCS Uncertainty Assessment for Support for emergency planning.

Consequence Models

Y6469 Evaluation of Radionuclide Pathways and Upgrade information on uptake

Uptakes pathways.

55 NUREG-1 635
15 THERMAL-HYDRAULICS RESEARCH

The NRC regulatory process makes

extensive use of thermal-hydraulic analyses

for certifying advanced reactor designs,

processing changes in the licensing bases of

existing reactors, and investigating thermalhydraulic safety issues. To meet the many

needs for.such thermal-hydraulic analyses,

the NRC developed and maintained several

thermal-hydraulic transient analysis codes:.

Development and assessment of these codes,

require separate-effects experiments to

provide data for developing specialized

models as new operating regimes and

phenomena are encountered. Further, as

new reactor concepts are proposed, it is

necessary to obtain data for assessing the

applicability and accuracy of agency codes

for the different nuclear steam supply

systems. The NRC has satisfied these needs

through the agency staff and contractors, and

by developing and maintaining specialized

thermal-hydraulic test facilities at universities

and national laboratories. In addition, the

NRC has joined international efforts on the

limited basis of exchanging data when

opportunities have arisen.

The burden to maintain, modernize, and'

support several thermal-hydraulic transient

analysis codes was significant. A prudent

decision was made to consolidate the agency

analysis capabilities into a single code now

called TRACE. Consolidation was envisaged

as an ambitious effort to provide capabilities

in one code that could reproduce results'of

the other NRC codes, some of which had

unique features. There was also the desire to

modernize the consolidated code to some

extent and certainly to rationalize the many

correlations and approximations devised over

The PUMA Thermal-Hydraulics Test - Facility at Purdue Universiy

This' Experimental -Facility provides -data for

the development and Validation of thermalhydrau'lic models of passive safety systems in

advanced light 'water reactor. designs.

i. - ;. ; ; -;

57 NUREG-1 635
the years to obtain computational results

consistent with data, especially in the

complicated area of two-phase flow.

Ultimately, the consolidated code can be

used for independent evaluation of licensee's

thermal- hydraulic analyses as well as for use

in analysis of safety issues. To avoid the loss

of legacy plant models that have been

developed for the existing codes, RELAP5,

TRAC-P, and TRAC-B, a Symbolic Nuclear

Analysis Package (SNAP) is being developed

which will be able to convert existing models,

with little or no modification, to TRACE

models. This graphical user interface is also

intended to enhance productivity by providing

a user-convenient environment for future

input model development, graphical display of

results, and input checking.

The consolidation of NRC's thermal-hydraulic

modeling into a single code was a decision

that was enthusiastically supported by the

ACRS. The ACRS understood this to be a

challenging undertaking. Indeed, the

consolidation has proved to be every bit as

challenging as anticipated. While

consolidation is nearing completion, many

"loose ends" remain to be resolved through

the assessment and improvement phases of

the project. The assessment and refinement

efforts constitute an important fraction of the

Thermal-Hydraulics Research Projects listed

in Table 14.

The TRACE code has been developed to the

point where it has provided useful validation

calculations for the ESBWR and AP1000

designs, as well as for the steam generator

hydrodynamic blowdown loads related to

Generic Safety Issue (GSI) 188, "Steam

GeneratorTube Leaks/ Ruptures Concurrent

with Containment Bypass." However, it is not

yet ready for routine regulatory application.

Therefore, the NRC will need to maintain the

RELAP5 code (Y6392). RELAP5 is being

used to carry out intensive thermal-hydraulic

analyses in support of the agency's initiative

in reevaluating the PTS screening criterion

and in supporting the certification of the

AP1000 advanced LWR design. The

maintenance of this older code should include

correction of deficiencies in RELAP5 when

necessary for satisfactory completion of these

analysis efforts.

In addition to the work described in the list of

sponsored projects, there have been

significant contributions from the RES staff

itself in the development and improvement of

the TRACE code. While the system codes

provide important thermal-hydraulic analysis

capability, it is encouraging to see that

computational fluid dynamic (CFD) methods

are being used by the RES staff for resolving

complex multidimensional thermal-hydraulic

issues. Ultimately, the coupling of CFD

methods with the system codes will add a

new dimension of confidence to thermalhydraulic analyses.

The development of new TRACE models and

the refinement of these models to include the

operating regimes of the advanced LWRs

have created the need for new empirical

data. To this end, the NRC is conducting

separate-effects experimental research at

universities. Electrically heated rod bundle

reflood experiments are being conducted at

Penn State University (Y6671). These tests

use advanced instrumentation to permit

greater insights into the reflood heat transfer

process so that peak clad temperatures can

be predicted with less uncertainty. Subcooled

boiling experiments and associated model

development are being conducted at UCLA

(W6749) to develop more accurate models

and reduce uncertainty associated with void

generation expected in the ESBWR.

Entrainment at "Tees" and in the upper

plenum of the AP1000 is being investigated

at Oregon State University (Y6507, Y6795) to

reduce uncertainty in prediction of the

AP1000 accident response. Experiments on

low pressure choked two-phase flow typical of

natural circulation operation and experimental

and modeling research on interfacial area

NUREG-1 635 58
transport for improving the accuracy of twophase models are being conducted at Purdue

University (W6698, Y6769). The motivations

for these efforts are the improvement of

break flow models at low pressure and

improvement of two-phase flow modeling

accuracy by adding transport equations for

the interfacial area. This is a generic effort to

improve the accuracy of two-phase flow

transient models.

To provide data needed for assessing the

applicability and accuracy of TRACE

simulations for the ESBWR and AP1000

designs, experiments are being conducted in

integral system test facilities. The PUMA and

APEX facilities are scaled models of the

ESBWR and AP1000 designs. The NRC also

participates in international programs such as

the CAMP Program for exchange of code

assessment and user experience and the

SETH Program in Germany and the PANDA

Program in Switzerland for exchange of

experimental data. The SETH Program will

contribute experience and data on boron

mixing. The PANDA Program will provide

natural circulation data for full-height facilities

for code assessment and verification of the

scaling methods used in the design of the

PUMA facility.

The code consolidation effort has not gone as

smoothly as was predicted at the time the

work was initiated. The effort is most

challenging and given the substantial

additional workload for thermal-hydraulic

research that has distracted from the effort,

delay is not a surprise. We note that the code

consolidation project involves many separate

organizations and is funded from a variety of

task orders having different NRC project

managers. One NRC project has about 17

separate task orders that address a variety of

thermal-hydraulic issues at separate

organizations. Surely, the coordination of

these efforts must be challenging.

Another observation is that no research is

directed at development of probabilistic

methods for use with TRACE for establishing

margin of safety at a prescribed confidence

level. There is a growing need for

uncertainty estimates in thermal-hydraulic

analyses used in the regulatory process.

The time is approaching when the TRACE

effort must reach fruition and a consolidated

code be integrated into the regulatory

process. There may be ways to improve the

effectiveness of this effort, such as:

* Creating a TRACE code peer-review

group to provide greater focus on

technical issues confronting the

development and help achieve

consensus on the approach.

* Combining the code consolidation

efforts and thermal-hydraulic model

development experiments under one

NRC project manager to achieve

more effective integration.

* Benefitting from an international

effort leveraging NRC resources with

similar undertakings within the

CAMP community.

The NRC bears the substantial burden of

developing and maintaining experimental

facilities for thermal-hydraulic research. The

burden has been lessened by the use of wellscaled facilities that are less than full height

such as the PUMA and APEX facilities. Data

from these facilities together with results of

thermal-hydraulic simulations have proven

the correspondence between data from the

scaled facilities and the results that would be

expected from the full-height systems of

AP1000 and ESBWR. This conclusion is

strengthened by the data from full-height

international facilities such as PANDA and

ROSA-IV. The ACRS believes that the

PUMA and APEX facilities are of proven,

current utility and will help in the certification

59 NUREG-1 635
of the ESBWRand AP1000 advanced reactor

designs.

In the longer term, the NRC may want to

expand international cooperative research

agreements for maintaining thermal-hydraulic

research facilities. The successes being had

in such arrangements in fuel behavior

research and severe accident research may

serve as examples of the advantages of such

an international approach.

NUREG-1635 60
Table 14. Thermal-Hydraulics Research Activities

Project Title Comments

W6698 PUMA Integral Test Facility Experimental facility to be used

for ESBWR design certification.

Provides data on LOCA natural

Y6769 PMcirculation instabilities and low- PUMA Test Facility pressure choked flow data for

assessment of the TRACE code

Y6673 TRAC-M Development Assessment - Small Consolidated NRC thermalLOCA Processes hydraulics code.

Y6300 User Support for Consolidated TRAC Code

Y6525 TRAC-M Code Maintenance

Y6583 Advanced Reactor Development of TRAC-M

Code - Advanced Reactor Portion of Y6300

Y6830 TRAC-M Development and Assessment -

ROSA-IV Tests

Y6666 Advanced Numerical Methods in TRAC-M This work to add a droplet field

and advanced numerical

methods to TRACE should be

accelerated. The work needs

greater focus and could benefit

from peer review

Y6507 APEX Experimental Validation Conduct tests for AP1000 review

and do confirmatory analysis

Y6795 APEX AP1000 Test Program using TRACE. Conduct beyond- design-basis test for AP1 000

Y6667 SNAP Implementation Maintain and modify the

graphical user interface for

TRACE to provide ability to

import TRAC-P and RELAP5

plant models for TRACE.

61 NUREG-1 635
Table 14. Thermal-Hydraulics Research Activities

(Continued)

Project Title Comments

W6749 Thermal-Hydraulics Research This task is very diverse and

might benefit from prioritization

Task orders to support TRACE consolidation: and additional technical direction

* Interfacial area transport perhaps from a peer-review

* Modularization of TRACE group.

* Development assessment numerical techniques

* Phase separation in 'Tees" tests within this

* Subcooled boiling at low pressure Project should be closely

* SNAP runtime and output visualization coordinated with the TRACE

development code development.

* SNAP nodalization knowledge-based expert

system

* Improvements to PARCS

* BWR model development and assessment

* OECD/NRC benchmark for a BWR turbine trip

transient

* Multiphase CFD enhancements to nuclear reactor

safety analysis

* MELCOR assessment against SCDAP/RELAP5

* Automated code assessment program

* Two-phase CFD enhancements for NPHASE code

* Peach Bottom turbine trip analysis

K6987 Analytical Support for Serena Program Program without any specific objectives; should be closed if

not yielding products to crucial

thermal-hydraulic activities.

Y6392 Maintain, Apply, Assess, and Develop NRC Consolidate RELAP5 capabilities

Computer Codes into TRACE. Maintain RELAP5.

Analysis of advanced reactor

designs. TRACE assessments of

PWR steam and feedwater line

breaks. This is a necessary

activity until TRACE can be used

to do the analyses.

Y6428 Evaluation of Steam Generator Tube Rupture This is a major effort using

Performance SCDAP/RELAP to analyze

steam generator tube rupture

accidents.

Y6503 AP1000 analysis TRACE analyses of AP1000

large and small break loss of

Y6662 AP1000 Confirmatory Thermal-Hydraulics coolant accidents.

Analysis

NUREG-1 635 62
Table 14. Thermal-Hydraulics Research Activities

(Continued)

Y6772 AP1000 Confirmatory Research Measure entrainment for

AP1 000 certification

Needs better coordination with

TRACE development and

assessment.

Y6526 Administer CAMP Meeting Support for users of RELAP5;

This is a valuable international

collaboration.

Y6571 SETH Program - Test Facilities Boron mixing programs in the

PKL facility in Germany and the

PANDA facility in Switzerland.

Y6589 Thermal-Hydraulics Research

Five task orders to support various thermal-hydraulics Applicable parts of this work

research: need to be coordinated and

* two-phase CFD enhancement combined with the TRACE code

* MOX neutronics development. The longer-term * Interfacial area transport tasks need peer review to

* separate effects experiments for model establish relevance.

development

* MELCOR assessment and application

Y6598 RELAP5 Analyses for Pressurized Thermal Complete PTS analyses for four

Shock plants.

Y6671 Rod Bundle Reflood Experiments - Phase 2 Rod bundle heat transfer tests in

support of TRACE reflood model

development.

Needs to be more closely

coordinated with the TRACE

code development.

Y6804 ESBWR Containment Support Thermal-hydraulics support for

ESBWR design certification.

Y6806 Review of GE Scaling Report

63 NUREG-1 635
16 REFERENCES

1. U.S. Nuclear Regulatory Commission,

"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 (ACRS), NUREG-1635, Vol.

5. June 2003.

2. U.S. Nuclear Regulatory Commission,

"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 (ACRS), NUREG-1 635, Vol.

4, May 2001.

3. Report dated May 16, 2003, from Mario

V. Bonaca, Chairman, ACRS, to Nils J.

Diaz, Chairman, U. S. Nuclear

Regulatory Commission, Subject:

Improvement of the Quality of Risk

Information for Regulatory

Decisionmaking

4. U.S. Nuclear Regulatory Commission,

"Evaluation of Air-Operated Valves at

U.S. Light-Water Reactors," NUREG1275, Vol. 13, February 2000.

5. U.S. Nuclear Regulatory Commission,

"Operating Experience Feedback

Report, Assessment of Spent Fuel

Cooling," NUREG-1275, Vol. 12,

February 1997.

6. U.S. Nuclear Regulatory Commission,

"Operating Experience Assessment - Effects of Grid Events on Nuclear Power

Plant Performance," NUREG-1784,

December 2003.

7. Report dated September 30, 2003, from

Mario V. Bonaca, Chairman, ACRS, to

Nils J. Diaz, Chairman, U. S. Nuclear

Regulatory Commission, Subject: Draft

Final Revision 3 to Regulatory Guide

1.82, "Water Sources for' Long-Term

Recirculation Cooling Following a Lossof-Coolant Accident."

8. U.S. Nuclear Regulatory Commission,

"Perspectives Gained from the Individual

Plant Examination of External Events

(IPEEE) Program," NUREG-1742,

Volumes 1 and 2, April 2001.

9. LaChance, J.L., et.al., "Circuit Analysis - Failure Mode and Likelihood Analysis,"

Sandia National Laboratory,

NUREG/CR-6834, September 2003.

10. U.S. Nuclear Regulatory Commission,

"Severe Accident Risks: An Assessment

for Five U.S. Nuclear Power Plants,"

NUREG-1150,1990.

11. National Fire Protection Association

(NFPA), "Performance-Based Standard

for Fire Protection for Light-Water

Reactor Electric Generating Plants,"

NFPA Standard 805, January 2001.

12. Meyer, R.O., R.K. McCardell, H.M.

Chung, D.J. Diamond, and H.H. Scott,

"A Regulatory Assessment of Test Data

for Reactivity-Initiated Accidents,"

Nuclear Safety, Special Issue on

Reactivity- Initiated Accident, Vol. 37,

No. 4, pp. 271-288, October-December

1996.

13. Diaz, N.J., "To License and Regulate - Sharping the Edges," Presented at the

INPO 4t Annual CEO Conference,

Atlanta, Georgia, November 6-7, 2003.

65 NUREG-1 635
14. Poucet, A., "The European Benchmark

Exercise on Human Reliability Analysis,

" Presented at the American Nuclear

Society, International Topical Meeting

on Probability, Reliability, and Safety

Assessment, PSA 89, Pittsburgh, PA,

April 2-7, 1989.

15. Committee on the Biological Effects of

Ionizing Radiation (BEIR), National

Research Council, "Health Effects of

Exposure to Low Levels of Ionizing

Radiation: BEIR V," National Academy

Press, Washington, D.C. 1990.

16. International Commission on

Radiological Protection (ICRP),

"Recommendations of the International

Commission on Radiological Protection,

ICRP Publication 60," Ann. ICRP, 1991.

17. U.S. Nuclear Regulatory Commission,

"Radiological Assessment for Clearance

of Materials from Nuclear Facilities,"

NUREG-1 640, Vols. 1-4, June 2003.

18. Senior Seismic Hazard Analysis

C o m m i t t e e ( S S H A C ) ,

"Recommendations for Probabilistic

Seismic Hazard Analysis: Guidance on

Uncertainty and Use of Experts,"

Lawrence Livermore National

Laboratory, NUREG/CR-6372, April

1997.

19. Code of Federal Regulations, Titlelo,

Part 72 (10 CFR 72), "Licensing

Requirements for the Independent

Storage of Spent Nuclear Fuel, HighLevel Radioactive Waste, and ReactorRelated Greater than Class C Waste,"

U.S. Government Printing Office,

Washington, D.C., 2003.

20. Code of Federal Regulations, Title 10,

Part 100 (10 CFR 100), "Reactor Site

Criteria," U.S. Government Printing

Office, Washington, D.C., 2003.

NUREG-1635 66
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(289) (Assigned by NRC, Add Vol. Supp, Rev.,

NRCM 1102. BIBLIOGRAPHIC DATA SHEET and Addendum Numbers, If any.)

3201.320ILIGAHI2AA HE

(See Instructions on the reverse)

2. TITLE AND SUTrrLE NUREG-1635, Vol. 6

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

Report to the U.S. Nuclear Regulatory Commission 3. DATE REPORT PUBLISHED

MONTH YEAR

March 2004

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

Technical Report

7. PERIOD COVERED (Incdusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (IfNRG-parodeD sg tOfliceorRegon. U.S. NudearRegdatoryComnvnssisn. andmaing address;fcontractor.

pnonide name and mnalkn address.)

Advisory Committee on Reactor Safeguards

U. S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (t NRC type Same as above contraor, provide NRC Dysion. Oft rReof ,n U.S. Nuear Regutatory Cornmitssn.

and making address.)

Same as above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 ivtds or less)

This report 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). This

report focuses on that portion of the NRC research program dealing with the safety of existing nuclear reactors and advanced

light water reactor designs, APIOOO, and ESBWR, submitted for certification. In Its review of the NRC research activities, the

ACRS considered the programmatic justification for the research as well as the technical approach and progress of the work.

This review attempts to identify research crucial to the NRC mission. It also attempts to Identify research activities that have

made valuable contributions to the agency mission In the past, but now have reached the point where additional research Is

not needed for efficient and effective safety regulation. The review also attempts to Identify areas where greater International

cooperation in research useful to the NRC could leverage resources of partners In the research and yield superior technical

products. This report does not address research on the vulnerability of existing nuclear power plants to acts of sabotage and

terrorism.

12. KEY WORDSIDESCRIPToRS (List wordsorphrases thatfwassstresearers hb cg the report.) 13. AVAILABILITY STATEMENT

observations and recommendations unlimited

operational data radiation protection4. sEcuRf CLSSIFICATION

containment systems safety research (MIS Pfte)

digital Instrumentation and control systems seismic research unclassified

fire safety research severe accident research (This Report)

reactor fuel research thermal-hydraulics research unclassified

neutronics and criticality safety

human factors and human reliability research 15. NUMBER OF PAGES

materials and metallurgy

probabilistic risk assessment 16. PRICE

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