NUREG/CP-0188

Transactions of the

2004 Nuclear Safety

Research Conference

To be held at

Marriott Hotel at Metro Center

Washington, DC

October 25-27, 2004

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Transactions prepared by

Brookhaven National Laboratory
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NUREG/CP-0188

Transactions of the

2004 Nuclear Safety

Research Conference

To be held at

Marriott Hotel at Metro Center

Washington, DC

October 25-27, 2004

Manuscript Completed: October 2004

Date Published: October 2004

Compiled by: S. Monteleone, Meeting Coordinator

S. Nesmith, NRC Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Transactions prepared by

Brookhaven National Laboratory
ABSTRACT

This report contains summaries of papers on reactor safety research to be

presented at the 2004 Nuclear Safety Research Conference (at the Marriott

Hotel at Metro Center in Washington, DC, October 25-27, 2004.

The summaries briefly describe the programs and results of nuclear safety

research sponsored by the U.S. Nuclear Regulatory Commission's Office of

Nuclear Regulatory Research. Also included are summaries of invited

papers concerning nuclear safety issues from U.S. government laboratories,

the electric utilities, the nuclear industry, and from foreign governments and

industry.

The summaries have been compiled here to provide a basis for meaningful

discussion of information exchanged during the course of the meeting, and

are in the order of their presentation on each day of the meeting.

iii
2004 Nuclear Safety Research Conference

October 25-27, 2004

Washington, DC, USA

CONTENTS

Page

Abstract .......................................................... iii

Monday, October 25

SESSION 1: MATERIALS AGING AND DEGRADATION RESEARCH

Chaired by: A. Hiser (NRC)

Objective: To discuss consideration of failures of passive systems and

components in risk assessments, and discuss two risk-informed initiatives

involving passive failures. Results will be presented from materials

degradation and aging research, emphasizing regulatory and safety insights.

Development of Technical Basis to Support Risk-Informed Revision

of the Pressurized Thermal Shock (PTS) Rule (10 CFR 50.61) ..... ........... 1

M. EricksonKirk, et al. (NRC)

LOCA Frequency Evaluation Using Expert Elicitation ....................... 3

R. Tregoning, L. Abramson (NRC), P. Scott (Battelle)

Review of the Reactor Coolant Pressure Boundary Leakage Data

and Requirements for U.S. Nuclear Power Plants .......................... 5

M. Srinivasan, D. Kupperman (NRC), W. Shack (ANL),

P. Krishnaswamy, G. Wilkowski (Engineering Mechanics Corp.)

Status and Results of NRC's Proactive Materials Degradation Assessment

Program ........................................................... 7

J. Muscara (NRC)

A Knowledge-Based Approach to Management of Materials Degradation .... .... 9

R. Dyle (Southem Nuclear), A. Marion (NEI)

V
SESSION 2: NEW REACTORS

Chaired by: M. Gamberoni (NRC)

Objective: To discuss ongoing recent or planned NRC and industry safety

research activities affecting new and advanced reactors and to discuss

infrastructure to support next-generation reactor technology.

Regulatory Structure for New Plant Licensing: Technology-Neutral Framework ... 11

M. Drouin, T. King (NRC), J. Lehner, T. Pratt, V. Mubayi (BNL),

D. Bley (Buttonwood Consulting)

Testing in the APEX Facility to Investigate Advanced Plant Thermal-Hydraulics ... 13

S. Bajorek (NRC)

NRC Nuclear Analysis Research for Reviewing the Advanced CANDU Reactor

ACR-700 .......................................................... 15

D. Carlson (NRC)

Pebble Bed Modular Reactor Research and Development

for an Innovative Small Reactor ........................................ 17

E. Wallace (PBMR Pty., South Africa)

ACR Research and Development ....................................... 19

A. White, D. Wren (AECL)

Advanced Nuclear Energy Research in the United States ...... .............. 21

C. Sink (DOE)

vi
Tuesday, October 26

SESSION 3A :SPENT FUEL RESEARCH

Chaired by: R. Meyer, M. Hodges (NRC)

Objective: Dry storage and transportation casks impose unique temperature

and pressure conditions on spent fuel rods, and the consequences of these

conditions are considered in cask licensing. Research results under these

conditions and their implications will be presented in this session.

High-Burnup Cladding Mechanical Performance during Cask Storage

and Post-Storage Handling and Transport ................................ 23

R. Daum, et al. (ANL)

Data Needs for the Transportation and Storage of High Burnup Fuel .... ....... 27

R. Einziger, et al. (NRC)

Perspective on Requirements for Spent Fuel Storage and Transportation ....... 29

A. Machiels (EPRI)

SESSION 3B: HIGH BURNUP FUEL UNDER LOCA CONDITIoNS

Chaired by: R. Meyer, F. Akstulewicz (NRC)

Objective: Risk-informed, performance-based rulemaking is planned to revise 10

CFR 50.46 and Appendix K for LOCA analysis. Results of recent research will be

presented, and these results will provide the technical basis for changing the

embrittlement criteria in the rule.

LOCA Integral Test Results for High-Burnup BWR Fuel ..................... 31

Y. Yan, T. Burtseva, M. Billone, (ANL)

LOCA Testing at Halden ............. ................................ 35

E. Kolstad, W. Wiesenack, V. Grismanovs (Halden)

Results from Studies on High-Burnup Fuel Behavior under LOCA Conditions .... 39

F. Nagase, T. Fuketa (JAERI)

Realistic Analysis of Fuel Rod Behavior under Large-Break LOCA Conditions .... 41

M. Nissley, C. Frepoli, K. Ohkawa (Westinghouse)

vii
Post-Quench Ductility of Advance Alloy Cladding .......................... 43

M. Billone, Y. Yan, T. Burtseva (ANL)

Overview of the CEA Data on the Influence of Hydrogen on the Metallurgical

and Thermal-Mechanical Behavior of Zircaloy-4 and M5Tm Alloys

under LOCA Conditions .............................................. 47

J-C. Brachet, et al. (CEA Saclay), J. Mardon (Framatome),

P. Jacques, A. Lesbros (EDF)

SESSION 4: DEVELOPMENT OF A PRA INFRASTRUCTURE

Chaired by: M. Drouin (NRC)

Objective: This session will consider insights on PRA Quality from SPAR and

ASP, phased approaches (e.g., Rg 1.200), and international activities. A panel

discussion on PRA methods & tools will debate our limitations, priorities, and

whether future risk applications might require additional PRA tools, methods and

guidance. Additionally, modeling organizational factors, integrating equipment

aging into the PRA model, and quantifying the risk of low-power shutdown will be

topics covered.

Insights on PRA Quality from the SPAR Model Development Program .... ...... 49

P. O'Reilly, M. Cheok (NRC)

Human Reliability Analysis (HRA) Good Practices .......................... 51

E. Lois (NRC), A. Kolaczkowski (SAIC), J. Forester (SNL)

The Joint U.S. NRC/EPRI Fire Risk Requantification Study ................... 53

J. Hyslop (NRC)

Phased Approach to Achieving an Appropriate Quality for PRAs

for Regulatory Decision Making ........................................ 55

G. Parry (NRC)

Guidance on the Treatment of Uncertainties and Alternative Approaches

used in PRAs . ..................................................... 57

M. Drouin (NRC)

viii
Wednesday, October 27

SESSION 5: RADIATION PROTECTION

Chaired by: C. Trottier (NRC)

Objective: To discuss current initiatives in radiation protection including new

recommendations from the International Council on Radiation Protection (ICRP),

the challenges of homeland security, and establishing a more robust materials

program.

More Robust Research Plan for Radiation Protection ....................... 59

S. Bush-Goddard (NRC)

2005 Recommendations of the ICRP .................................... 61

E. Holahan (NRC)

Challenges in Radiation Protection and Homeland Security .................. 63

C. Paperiello (NRC)

SESSION 6: CODES

Chaired by: S. Bajorek (NRC)

Objective: To explore issues in developing and applying computer codes to

analyze reactor systems. Topics will include the analysis, applications, and

maintenance of codes, the status of the TRACE project, the analysis of source

terrns, and the assessment and modernization of the MELCOR code.

TRACE: TRAC/RELAP Advanced Computation Engine ..................... 65

J. Staudenmeier (NRC)

MELCOR Development and Assessment ................................. 67

R. Gauntt (SNL)

Fuel Behavior Modeling in Accident Analysis .............................. 69

H. Scott (NRC)

The U.S. NRC Spatial Kinetics Code PARCS ............................. 71

T. Downar (Purdue University)

Use of Computer Codes in the Regulatory Review Process .................. 73

R. Landry (NRC)

ix
Recent Developments in Multidimensional Modeling of

Reactor Thermal Hydraulics using the NPHASE Code ...................... 75

M. Podowski, et al. (RPI)

SESSION 7: OPERA TiNG EXPERIENCE

Chaired by: J. Ibarra (NRC)

Objective: To present the efforts to enhance the NRC Operating Experience

Program utilizing the analysis conducted by RES. Prime concern is the impact

on this program from the recommendations of the Davis Besse Lessons Leamed

Task Force and the August 14, 2003 Northeast Blackout.

Overview of Approach to Reevaluate Station Blackout Risk ...... ............ 77

D. Rasmuson (NRC)

Generic Issues Program - Process and Screening Analysis ...... ............ 79

H. Vandermolen (NRC)

Operating Experience Implementation Task Force ....... .................. 81

T. Reis (NRC)

A Generalized Framework for Assessment of Safety Margins

in Nuclear Power Plants ............................................. 83

G. Lanik (NRC), J. Meyer (ISL)

Boric Acid Corrosion of Light Water Reactor Pressure Vessel Materials .... ..... 85

J-H. Park, et al. (NRC)

x
Development of Technical Basis to Support Risk-Informed Revision of the

Pressurized Thermal Shock (PTS) Rule (10 CFR 50.61)

Mark EricksonKirk, David Bessette, Mike Junge, Shah Malik,

Todd Mintz, and Roy Woods

United States Nuclear Regulatory Commission, Office of Nuclear Regulatory Research

Over the past 5 years, the Office of Nuclear Regulatory Research of the United States Nuclear

Regulatory Commission has conducted a project aimed at development of the technical basis for

a risk-informed revision of the pressurized thermal shock rule (10 CFR 50.61). Pressurized thermal

shock (PTS) includes any transient wherein severe cooling of the reactor pressure vessel (RPV)

occurs together with, or followed by, pressurization. A number of postulated accidents can

thermally shock the vessel. The tensile stresses produced in the vessel wall by the rapid cool-down

coupled with pressure induced stresses may, at high levels of neutron embrittlement, be sufficient

to drive a crack all the way through the reactor pressure vessel wall. To ensure that the likelihood

of vessel failure resulting from PTS is kept extremely low, 10 CFR 50.61 places statutory limits on

how embrittled the RPV wall may become before licensees are required to either submit to the

NRC a plant-specific rationale supporting the safety of continued operations beyond the statutory

embrittlement limit, or are obligated to take steps to keep embrittlement from exceeding the

statutory limit during the licensed life of the vessel.

As PWRs approach the end of their original 40-year operating licenses, and licensees consider

requesting a 20 year license renewal, compliance with 10 CFR 50.61 can become a factor that

limits the operational life of the plant. Addressing this issue on a plant-specific basis has consumed

considerable resources within both the NRC and industry. Additionally, it is now widely recognized

that state of knowledge and data limitations in the early 1980's necessitated a conservative

treatment of several key parameters and models that were used to establish the current statutory

embrittlement limit. These conservatisms suggest that re-evaluation of the technical basis for 10

CFR 50.61 based on a more up to date technological approach could justify reduction of

unnecessary regulatory burden while maintaining safety.

The re-evaluation includes probabilistic calculations to estimate the risk of vessel failure due to

PTS. This approach, which uses experimentally-benchmarked models, considers all of the factors

known to influence the risk of vessel failure during a PTS event while accounting for uncertainties

in these factors in a consistent manner across a breadth of technical disciplines. Two central

features of the approach are a focus on the use of realistic input values and models (wherever

possible), and an explicittreatment of uncertainties. This approach improves significantly upon that

employed to establish the 10 CFR 50.61 embrittlement limits, wherein intentional and un-quantified

conservatisms were included in many aspects of the analysis, and where uncertainties were treated

implicitlyby incorporating them into the models. Nevertheless, the current probabilistic model does

include a limited number of conservative factors. These conservatisms remain within an overall

model focused on producing realistic estimates of vessel failure risk due to PTS for one of two

reasons: (1) to help ensure the applicability of our findings to PWRs in general, or (2) due to the

lack of any more realistic models or parameter inputs.

I
The results of this research, which are summarized in this paper, demonstrate that the likelihood

of vessel failure due to PTS is extremely low (-10O4 lyear) even through the period of license

extension. These results provide a basis for potential relaxation of the 10 CFR 50.61 screening limit should rulemaking be undertaken.

2
LOCA Frequency Evaluation Using Expert Elicitation

Robert L. Tregoning

Lee R. Abramson

United States Nuclear Regulatory Commission

Washington, DC 20555 USA

Paul M. Scott

Battelle Memorial Institute

The emergency core cooling system (ECCS) requirements are contained in 10 CFR 50.46,

Appendix K to part 50, and GDC 35. Consideration of an instantaneous break with a flow rate

equivalent to a double-ended guillotine break (DEGB) of the largest primary system in the plant

generally provides the limiting condition in the required ECCS analysis. However, the DEGB is

widely recognized as an extremely unlikely event. Therefore, the NRC is developing a

risk-informed revision of the design-basis break size requirements for operating commercial nuclear

power plants. A central consideration in selecting a risk-informed design basis break size is an

understanding of the LOCA frequency as a function of break size.

LOCA frequency estimates have been developed using an expert elicitation process to consolidate

service history data and insights from PFM studies with knowledge of plant design, operation, and

material performance. This process is well-recognized for quantifying phenomenological

knowledge when data or modeling approaches are insufficient. Separate BWR and PWR piping

and non-piping passive system LOCA frequency estimates have been developed as a function of

effective break size and operating time through the end of license extension. The elicitation

focused solely on determining event frequencies that initiate by unisolable primary system side

failures that can be exacerbated by material degradation with age. The expert elicitation process

employed in this study is an adaptation of the formal expert judgment process used in

NUREG-1 150. This process included the decomposition of the complex technical issues which

impact LOCA frequencies into fundamental elements in order to more easily assess these

important contributing factors. The elicitation process required each member of the elicitation panel

to qualitatively and quantitatively assess these LOCA contributing factors and also indicate their

uncertainty in this assessment. This information was collected from each panelist in an individual

elicitation session.

The qualitative insights provided by the panel members are reasonably consistent. Most panelists

agreed that a complete break of a smaller pipe, or non-piping component, is more likely than an

equivalent size opening in a larger pipe, or component, because of the increased severity of

fabrication or service cracking. Many panelists thought that aging may have the greatest effect on

intermediate diameter (6 to 14-inch diameter) piping systems due to the large number of

components within this size range and the fact that this piping may receive less attention than

smaller or larger diameter piping. Frequency estimates are not expected to change dramatically

over the next fifteen years, or even the next thirty-five years. While aging will continue, the

3
consensus is that mitigation procedures are in place, or will be implemented in a timely manner,

to alleviate possible LOCA frequency increases.

The quantitative responses were analyzed separately for each panel member to develop individual

BWR and PWR total LOCA frequency estimates of the mean, median, 5th and 95th percentiles.

The LOCA frequencies for the individual panelists were then aggregated to obtain group LOCA

frequency estimates, along with measures of panel diversity. While there was general qualitative

agreement among the panelists about important technical issues and LOCA contributing factors,

the individual quantitative estimates are much more variable. Additionally, as the LOCA size

increased, the panel members generally expressed greater uncertainty in their predictions, and the

variability among individual panelists' estimates increased. Both trends are expected given the

underlying scientific uncertainty.

The elicitation LOCA frequency estimates are generally much less than the prior WASH-1400

estimates and more consistent with the NUREG/CR-5750 estimates. The small break (SB) LOCA

frequency estimates are similar once the steam generator tube rupture frequencies are added to

the NUREGICR-5750 PWR results. The elicitation medium break LOCA estimates are higher than

the NUREG/CR-5750 estimates by factors of 2.5 and 10 for BWR and PWR plant types,

respectively. The NUREG/CR-5750 LB LOCA frequency estimates, most comparable to the

elicitation LOCA Category 4, tend to be slightly higher (less than a factor of 3) than the current

results.

Sensitivity analyses were conducted to examine the robustness of the quantitative results to the

underlying analysis procedure. Sensitivity analyses investigated the aggregation method, the effect

of panelist overconfidence, percentile calculation, and panel diversity measurement. The bulk of

the sensitivity analyses resulted in insignificant changes to the LOCA frequency estimates.

However, the method used to aggregate the individual panelist estimates can result in large

differences compared to the baseline LOCA frequency estimates. Overconfidence adjustments

can also result in large, unsupportable increases in the frequency estimates and need to be done

with care given the extreme skewness of many of the individual distributions.

4
Review of the Reactor Coolant Pressure Boundary Leakage Data

and Requirements for U.S. Nuclear Power Plants

M. Srinivasan', D. S. Kupperman, W. J. Shack2, P. Krishnaswamy 3 and G. M. Wilkowski3

'Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington

D.C. 20555-0001, U.S.A; 2Argonne National Laboratory, Argonne, IL 60439,

3Engineering Mechanics Corporation, Columbus, OH 4322

In 2002, the discovery of Davis Besse (DB) reactor vessel head corrosion cavity led the NRC to

form a Lessons Learned Task Force (LLTF). The LLTF conducted an independent evaluation of

the NRC's regulatory processes to ensure reactor vessel head integrity and identified and

recommended areas of improvement. One of the recommendations of the task force was to

assess the reactor coolant system (RCS) leakage requirements for potential improvements that

could provide enhanced safety. Specifically, the requirements needed to be evaluated for

sufficiency to (a) provide the ability to discriminate between the RCS unidentified leakage and the

reactor coolant pressure boundary (RCPB) leakage; and (b) provide reasonable assurance that

plants are not operated at power with RCPB leakage. The NRC Regulatory Guide 1.45 describes

the RCPB leakage detection systems and states the regulatory position that the sensitivity and

response time of each leakage detection system employed for unidentified leakage should be

adequate to detect a leakage rate, or its equivalent, of one gallon per minute (gpm) in less than one

hour.

The NRC staff recognizes that simply improving leakage detection and lowering allowable leakage

may not be sufficient to provide increased assurance of RCPB integrity. Leakage monitoring

assumes that the pressure boundary will fail only under a leak-before-break (LBB) scenario. The

leak rates associated with tight stress corrosion (SC) cracks or cracks which may be partially

plugged in stress corrosion-susceptible RCPB components could be too low to be detected reliably

by existing systems under normal operating conditions. Therefore, the scope of this review also

included methods which may be capable of detecting crack initiation and monitoring crack growth

before a through-wall crack develops and leakage occurs. To evaluate the current requirements,

the NRC's Office of Nuclear Regulatory Research (RES) conducted barrier integrity research at

Argonne National Laboratory (ANL) as a follow-up and an update of similar research conducted

by the NRC in the latel98Os.

This research resulted in a database that identifies the source of leakage, the leak rate, and the

resulting actions to remedy RCS leaks in U.S. LWRs. For each leakage event, the database

provides information on what equipment detected the leakage, how it was determined that the

leakage was through the pressure boundary, and the cause of leakage. Many reported leakages

were of very small leakage rate (c 0.01 gpm), were detected visually and were reported as drips,

weeping, seepage, "very small," boric acid deposits, etc. Cracks were associated with leaks about

40% of the time with a wide range of leak rates (< 0.01 gpm to > 100 gpm).

5
This research examined the capabilities of different types of leakage detection systems, which were

assessed for their sensitivity, reliability, response time and accuracy. Significant improvements in

leak detection capability will require implementation of new systems. Such systems are not only

sensitive and accurate, but often can be used to determine leak locations and thus reduce the

likelihood of unnecessary shutdowns. Newer commercially available systems include: acoustic

emission monitoring, humidity sensors, and air particulate detectors.

This research also determined that acoustic emission (AE) crack monitoring systems can also

detect crack initiation and growth before a leak occurs. The same systems used for leak detection

can also be used to detect crack initiation and growth. The AE from crack growth and leaks can

be separated from noise by monitoring at a high enough frequency. Such technology has been

successfully used to monitor selected locations at Watts Bar and Limerick power reactors.

Correlation studies between leak rates and crack size were conducted using computer models. For

a given leak rate, it was found that the corresponding crack size could vary by factor of 5 to 10

depending on the applied load on the cracked cross section. More refinements in the model and

mechanistic understanding are needed for more effective predictions.

Based on the results of this research, the NRC staff is considering how the regulatory requirements

for RCS leak detection and monitoring could be improved as per the DB LLTF recommendation.

However, many challenges need to be addressed in recommending improvements, such as the

need for backfit analysis or a cost-benefit analysis, that would potentially enhance public safety

while reducing unnecessary regulatory burden on utilities. Implementation challenges also exist

such as changes in technical specifications, regulatory orders, or potential rulemaking.

The contents of this presentation are based on the authors' knowledge and research sponsored

by the NRC Office of Nuclear Regulatory Research. This presentation is an independent product

of the authors and does not necessarily reflect the views or regulatory position of the US NRC.

6
Status and Results of NRC's

Proactive Materials Degradation Assessment Program

Joseph Muscara

Materials Engineering Branch, Division of Engineering Technology'

Office of Nuclear Regulatory Research, USNRC

There is a growing recognition both in industry and at the US Nuclear Regulatory Commission

(NRC) that proactive approaches to the management of material degradation are needed

particularly as plants age and continue to operate into their 20-year license renewal period.

Therefore, the NRC Office of Nuclear Regulatory Research (RES) has initiated work to develop a

foundation for regulatory actions to prevent materials degradation from adversely impacting safety.

A two-step approach will be used for this work. The first step is to identify susceptible materials

and locations where degradation can reasonably be expected in LWRs in the future. The second

step is to cooperatively develop and implement an international research program for the

components and degradation of interest. This research program would address materials and

degradation mechanisms, mitigation, repair and replacement, and nondestructive evaluation.

As a starting point for the first step, identification of susceptible materials and locations, RES is

conducting a review using the Generic Aging Lessons Learned (GALL) report to identify

components that have historically experienced degradation. The effectiveness of in-service

inspection and leakage monitoring requirements and techniques for these components is then

evaluated with respect to timely detection of known degradation mechanisms. Recommendations

for improvements in requirements and techniques will be made as necessary. The review will be

conducted by a Nondestructive Examination and Leak Monitoring (NDELM) task group comprised

of NRC Office of Nuclear Reactor Regulation (NRR), RES, and national laboratory experts in

inspection and leak monitoring techniques and requirements.

In addition to using the GALL report, a concurrent effort is underway to identify LWR components that may be susceptible to material degradation in the future. To accomplish this task, RES is using

a modified Phenomena Identification and Ranking Table (PIRT) process to perform the Proactive

Materials Degradation Assessment (PMDA). The PIRT process has several qualities that make

it ideal for the PMDA: a) provides a structured approach to expert elicitation where background

information is provided to an Expert Panel who identify material degradation of interest, b) requires

Panel Experts to document the bases for their selections, c) contains a provision for documenting

knowledge levels associated with decisions, and d) allows for ongoing documentation over the

course of the process.

RES is currently sponsoring a program at Brookhaven National Laboratory (BNL) to collect

operational experience and stressor information for the PWR and BWR components to be

examined by the PIRT Expert Panel and to manage and document the PIRT Expert Panel and

process. The Panel Experts will use the operational experience and stressor information to identify

potential degradation mechanisms, if any, for each component. The Expert Panel is comprised of

eight world-class materials experts from industry, research laboratories, regulatory agencies, and

universities from the US, Canada, Japan, France, and Sweden.

7
The Expert Panel will meet eight times, with each meeting lasting one week, to examine all of the

LWR components of interest. Four of the meetings will be used to examine PWR plant

components, and the remaining four will be used to examine BWR plant components. The first

Expert Panel meeting to examine the first set of PWR components was held in August 2004 and

there will be approximately six weeks between meetings. After each meeting, the components

identified by the Expert Panel as being susceptible to future degradation will be provided to the

NDELM task group who will assess NDE and leak monitoring techniques and requirements and

make recommendations for improvements.

A final report detailing the results of the Expert Panel deliberations for PWRs will be intemationally

peer reviewed and ready for publication in June 2005. The final peer reviewed report for BWRs

will be available in December 2005.

For each of the components susceptible to degradation identified through the GALL review and

PMDA PIRT effort, the Conditional Core Damage Probabilities (CCDPs) will be determined.

The probabilities of failure and associated uncertainty estimates for the components identified in

the above efforts will also be determined. Probability of failure information will be collected where

it already exists from other programs and sources such as aging programs, the LBLOCA

redefinition effort, and in licensee degradation assessments to support Risk Informed In-service

Inspection (RI-ISI) programs. This information will be provided by June 2005. The information

collected from these sources is unlikely to contain probability of failure estimates for all of the

components identified by the NDELM task group or by the PI RT Expert Panel, nor will it generally

contain uncertainty estimates. A comprehensive effort to determine these probability of failure and

uncertainty estimates will be completed in FY06.

8
A Knowledge-Based Approach to Management of Materials Degradation

Robin Dyle, Southern Nuclear Company

Alex Marion, NEI

Members of the Materials Technical Advisory Group

As a result of several recent industry issues involving materials degradation, the NEI Nuclear

Strategic Issues Advisory Committee (NSIAC), in May 2003, unanimously approved the

ulndustry Initiative on Management of Materials Issues and its associated "Guideline for the

Management of Materials Issues," NEI 03-08. NEI 03-08 creates a Materials Executive

Oversight Group (MEOG) and a Materials Technical Advisory Group (MTAG) to provide policy

direction, oversight and support to the industry's materials programs. An important element of

the oversight and support role is the creation and maintenance of a high-level, industry-wide

materials management strategic plan that identifies the highest priority challenges/activities.

Additionally, an Integrated Materials Management Work Plan has been developed in support of

the Strategic Plan. The Work Plan will provide details on each strategic issue, identify which

Issue Program(s) is working on particular topics and identify knowledge/technology gaps.

The Initiative's policy-objective is to have a more forward-looking, coordinated industry-wide

program, through the effective coordinated operation of the individual Issue Programs, for

managing materials aging and degradation. The intent of this strategic approach is to have a

comprehensive identification of issues and gaps that can be used by the Issue Programs in

developing their specific strategic and annual work plans.

A systematic approach to managing materials issues is necessary to effectively implement

this Strategic Plan as well as the elements of an effective materials issues management

program. The approach must focus on communicating with stakeholders, and on providing

appropriate direction and guidance to the industry. The systematic approach will identify,

assess, mitigate and/or repair known and future degradation mechanisms. It will incorporate

operating experience, impact of operational changes (e.g. power uprate), relevant laboratory

work and input from stakeholders. The first step in that process is to develop fundamental

understanding of degradation phenomena/mechanisms, and determine materials (and locations) that are known, or can logically be assumed, to be, susceptible to aging/degradation

phenomena when exposed to the operating environment.

In using the systematic approach to develop this Plan, a Degradation Matrix (DM) was

developed that addressed material degradation in a top-down fashion. All materials within

the scope of the initiative were considered and listed in the DM. Expert elicitation,

laboratory studies and field experience were then utilized to identify potential mechanisms

by which materials within this broad population can degrade. The material-mechanism

combinations were sorted into three large groups. The first group is the set of degradation

mechanisms/issues that have been adequately addressed and, where necessary, have

ongoing management programs in place. The second group is for those mechanisms/issues

that are determined to have work in place for tools/methodologies needed for effective

management. The third group is for those mechanisms that are identified as potential

challenges for which there is no program or effective methodology for understanding or

managing the mechanism at a phenomenological level.

9
The DM will be maintained as a living document and updated annually. It is a key resource

document that will be used with other reference material for the effective management of

materials aging and degradation.

10
Regulatory Structure for New Plant Licensing: Technology-Neutral Framework

Mary Drouin, Tom King, NRC

John Lehner, Trevor Pratt, Vinod Mubayl, BNL

Dennis Bley, Buttonwood Consulting

A risk-informed regulatory structure that can be applied to license and regulate advanced (new)

reactors, regardless of their technology, could enhance the effectiveness, efficiency, and

predictability (i.e., stability) of new plant licensing. As such, this new process, if implemented, could

be available for use later in the decade. The need for development of a risk-informed regulatory

structure for new reactors is based on a number of considerations. While the NRC has over 30

years experience of licensing and regulating nuclear power plants, this experience (e.g.,

regulations, regulatory guidance, policies and practices) has been focused on current light watercooled reactors (LWRs) and may have limited applicability to new reactors. There will be design

and operational issues associated with the new reactors that may be distinctly different from current

LWR issues.

The regulatory structure for current LWRs has evolved over five decades, and the bulk of this

evolution occurred without the benefit of insights from probabilistic risk assessments (PRAs) and

severe accident research. It is expected that future applicants will rely on PRA and PRA insights

as an integral part of their license applications. In addition, it is further expected that the

regulations licensing these new reactors will be risk-informed. Both deterministic and probabilistic

results and insights will be used in the development of the regulations governing these reactors.

Consequently, a structured approach for a regulatory structure for new reactors that provides

guidance about how to use PRA results and insights will help ensure the safety of these reactors

by focusing the regulations on where the risk is most likely while maintaining basic safety principles,

such as defense-in-depth and safety margin.

The objective of the regulatory structure for new plant licensing is to provide a technology-neutral

risk-informed approach that would enhance the effectiveness and efficiency of new plant licensing

in the longer term (beyond the advanced designs currently in the preapplication stage). This

regulatory structure is comprised of four major parts:

(1) a technology-neutral risk-informed framework that will provide guidance and criteria to the

staff for the development of technology-neutral requirements,

(2) the content for a proposed set of technology-neutral risk-informed requirements that will be

based on the guidance and criteria established in the technology-neutral framework,

(3) a technology-specific framework that will provide guidance and criteria for the staff on how

to apply the technology-neutral framework and requirements on a technology-specific basis,

and

(4) technology-specific regulatory guides that will be derived from the implementation of the

technology-specific framework and will provide guidance to licensees on how to apply the

technology-neutral regulations -on a technology-specific basis.

I I
The key features of the framework are as follows:

* a hierarchal approach based on a safety philosophy that is consistent with the

Commission's expectations for new reactors

* protective strategies that provide the basis for the technical requirements needed to ensure

that the safety philosophy has been met

* a defense-in-depth approach to address uncertainties that is based on previous

Commission guidance, on recommendations from the ACRS and on cornerstones

developed as part of the Reactor Oversight Program

* a probabilistic (risk-informed) approach that identifies and selects design basis accidents

and that assigns safety classification

* a performance-based approach that will establish performance standards and acceptance

criteria for results in development of the requirements

* incorporation of Commission direction on security related matters for new plants

To determine that the overall objectives of the regulatory structure have been met (e.g., enhanced

effectiveness and efficiency), the staff identified key characteristics for the regulatory structure.

Examples include:

* Flexible-The technology-neutral and technology-specific frameworks are developed in

such manner that they allow for changes and modifications to occur, in an efficient and

effective manner, that are based on new information, knowledge, etc., and can be adapted

to any technology-specific reactor design.

* Risk-informed-Risk information and risk insights are integrated into the decision-making

process such that there is a blended approach using both probabilistic and deterministic

information.

* Performance-based-The guidance and criteria, when implemented, will produce a set of

safety requirements that will minimize prescriptive means for achieving its goals, and

therefore, will be performance oriented to the extent practical.

* Uncertainty-The guidance and criteria will include treatment of the different types of

uncertainties.

* Defense-in-depth-Defense-in-depth is maintained and is an integral part of the framework.

12
Testing in the APEX Facility to Investigate Advanced Plant Thermal-Hydraulics

Stephen M. Bajorek

Office of Nuclear Regulatory Research

INTRODUCTION

The Advanced Plant Experiment (APEX) is a unique thermal-hydraulic integral system test facility

used to assess the performance of passive safety systems for the Westinghouse AP600 and

AP1000 designs. Data from the APEX facility was used extensively as part of AP600 Design

Certification, providing approximately 75 tests for use in code assessment and qualification of

AP600 safety margin. To address performance specific to the AP1000 design, the APEX facility

underwent significant modifications in 2002 in order that it more accurately represent AP1 000.

Included in the facility modification were an increase in the maximum core power, a new pressurizer

and surge line, a larger core makeup tank (CMT), larger diameter fourth stage automatic

depressurization (ADS4) system piping, and decreased line resistances for the CMT, ADS-4, and

PRHR heat exchanger.

Testing in the BAPEX-AP1000" facility began in 2003, with several integral experiments sponsored

by the U. S. Department of Energy (DOE) to investigate performance of AP1000 passive safety

systems at design basis accident conditions. The NRC Office of Nuclear Regulatory Research

(RES) also conducted several integral tests to explore beyond design basis performance and to

provide confirmatory information on thermal-hydraulic processes for which data from the APEXAP600 series of tests may not have been adequate. In particular, entrainment in the hot leg and

upper plenum is of significant interest because of the higher core power in AP1000 than in the

AP600. These data are also of interest because they provide unique information on upper plenum

thermal-hydraulics that may be useful in understanding carryover from an upper plenum of a

conventional pressurized water reactor during long term cooling.

TESTS EVALUATED

The table below lists the NRC APEX-AP1000 tests conducted for which an evaluation was

completed. Most of the 2003 tests investigated the double-ended guillotine break of a direct vessel

injection (DVI) line.

Test Identifier I Description

Double-ended guillotine break of DVI line #1 with complete failure of

NRC-AP1000-01 ADS-112/3 system.

Double-ended guillotine break of DVI line #1 with 2/4 ADS-4 valves

NRC-APIOOO-03 available. Both ADS-4 valves on pressurizer side failed closed.

NRC-AP1000-04 One-inch break at the bottom of cold leg CL4 with degraded sump.

Double-ended guillotine break of DVI line #1 with 2/4 ADS-4 valves

NRC-AP1000-05 available. Both ADS-4 valves on non-pressurizer side failed closed.

Two-inch break at the bottom of cold leg CL4 with 2/4 ADS- 4 valve

NRC-AP1000-06 failure. Both ADS-4 valves on pressurizer side failed closed.

13
CONCLUSIONS BASED ON NRC's APEX-AP1000 TESTS

1. The APEX-AP1 000 tests confirm significant entrainment and carryover of water to the

ADS-4 system during and after ADS-4 actuation. As expected, flow quality in the ADS-4 is low

when the water level in the vessel is above the bottom of the hot leg. Carryover to the ADS-4 can

also be significant, however, after the vessel level decreases below the bottom of the hot leg. In

addition, even when the upper plenum collapsed level is low, a water level in the hot legs can still

persist. This suggests that the processes of entrainment and flow dynamics within the hot leg

remain important to carryover to the ADS-4 even when the water level in the vessel decreases

well below the bottom of the hot leg.

2. Design basis tests from the APEX-AP1 000 facility did not result in core uncovery and heat up.

This significant finding indicates margin to uncovery in the AP1000 design for small break loss

of coolant accidents.

3. Both design basis tests by DOE and beyond design basis tests by the NRC (NRC-AP1000-03

versus NRC-APIOOO-05) in APEX-AP1000 show that failure of ADS-4 valves on the nonpressurizer side of the system results in a greater delay in IRWST injection than failure of ADS-4

valves on the pressurizer side of the plant. The likely cause is higher two-phase resistance in the

pressurizer side ADS vent path by water draining from the pressurizer.

4. Failure of the two of four ADS-4 valves were found to produce a core uncovery (NRC-AP1000-

05 and NRC-AP1000-06). While this is a beyond design basis accident scenario, it identifies a

sensitivity ADS-4 valve performance.

In general, the NRC tests provided new data for code assessment and model development. The

tests helped to confirm robustness in the passive safety system design and operation, since no

core uncovery was found to occur unless multiple failures were imposed on the transient.

14
NRC Nuclear Analysis Research for Reviewing

the Advanced CANDU Reactor, ACR-700

Donald E. Carlson

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Since September 2002, the NRC staff has been conducting a preapplication review of the

Advanced CANDU (CANada Deuterium Uranium) Reactor, ACR-700, in response to plans by

Atomic Energy of Canada Limited (AECL) to submit an ACR-700 design certification application

in March 2005. The NRC Office of Nuclear Regulatory Research (RES) has collaborated with the

NRC Office of Nuclear Reactor Regulation (NRR) on the preapplication review activities and is

preparing to support the design certification review by adapting and validating NRC's audit analysis

codes for use in predicting how the ACR-700 system will respond under operating, transient, and

accident conditions. Included in the audit analysis code work for ACR-700 are RES efforts on

phenomena identification and ranking, code adaptation and modeling, experiment and test data

evaluation, and code bias and uncertainty assessment (i.e., code validation) involving the technical

areas of nuclear analysis, thermal hydraulic analysis, and severe accident analysis. This paper

describes these RES efforts in the area of nuclear analysis.

In its nuclear design and operating characteristics, the ACR-700 differs significantly both from

NRC-licensed light water reactors (LWRs) and from existing CANDU reactors. Important nuclear

design changes in relation to existing CANDUs include the following: (1) Light water replaces heavy

water as the fuel coolant within the pressure tubes. (2) The core diameter is reduced by decreasing

the spacing between the pressure-tube fuel channels. (3) The volume of heavy water moderator

between fuel channels is further reduced by enlarging the diameter of the gas-filled calandria tube

that encloses each pressure tube. (4) A 43-pin CANFLEXm fuel bundle design replaces the

previous CANDU-standard 37-pin bundle while maintaining essentially the same overall bundle

dimensions. (5) The natural uranium oxide (U02) pellet composition used in conventional CANDU

fuel pins is replaced in ACR-700 by pellet compositions of slightly enriched U02 in all but the central

CANFLEXm fuel pin whose pellets are composed of natural U02 doped with dysprosium burnable

poison. (6) The average fuel bundle bumup at discharge increases from typically 7.5 GWd/t in

existing CANDUs to 21 GWdIt in the preapplication reference design of ACR-700.

To enable the simulation of such ACR-700 design features as. on-line fueling, pressure-tube core

geometry, separate moderator and coolant systems, and control rods that pass vertically between

horizontal fuel channels, new capabilities are being added to NRC's nuclear analysis codes that

have never been needed for analyzing LWRs. New capabilities being added to the Triton lattice

physics modules of the NRC's Standardized ComputerAnalysis for Licensing Evaluation (SCALE)

code system include employing the existing Keno Monte Carlo module to provide the detailed 3-

dimensional (3-D) neutron flux solutions needed for the lattice physics treatments of transverse

control rods and fuel bundle end effects in ACR-700. The fuel-depletion and control-rod lattice

results from SCALE/Triton provide the collapsed and smeared nodal data needed by the NRC's

Purdue Advanced Reactor Core Simulator (PARCS) code, which, when coupled to a system

thermal hydraulics code model of ACR-700, will provide 3-D static and dynamic solutions of in-core

15
neutron flux and power distributions during reactor operations, transients, and accidents. Further

code and modeling developments in SCALE/Triton and PARCS are needed to address the

channel-to-channel neutron spectral interactions that govern coolant void reactivity (CVR) effects

during the initial "checkerboard" voiding of alternate fuel channels in ACR-700 large-break loss-ofcoolant accidents.

Evaluating calculation bias and uncertainty is especially important in this context because

confirmatory measurements of CVR in an operating conventional CANDU or ACR-700 core are

inherently difficult and have never been performed in existing CANDUs. Associated ACR-700 CVR

validation activities in RES therefore include adapting the existing sensitivity and uncertainty

analysis methods in SCALE's Tsunami code modules for use in assessing the applicability and

adequacy of AECL's existing and planned sets of semi-prototypic critical experiments and fuel

irradiations. The modified SCALE/Tsunami modules will also be used to estimate the bias and

uncertainty in code predictions of CVR and other safety-significant nuclear effects in the ACR-700

core based on the emerging sets of code-to-data benchmark results. While motivated mainly by

specific needs and analysis issues associated with the staffs upcoming ACR-700 design

certification review, these improvements to NRC's nuclear analysis code infrastructure will provide

the staff with enhanced audit analysis capabilities that will prove broadly applicable to

understanding the safety-related behavior of other nuclear systems.

16
Pebble Bed Modular Reactor

Research and Development for an Innovative Small Reactor

Edward G. Wallace

Senior General Manager - US Programs

PBMR Pty, LTD., Centurion, South Africa

The pebble bed modular reactor (PBMR) being developed in the Republic of South Africa is a

development derived from German technology first deployed in the mid-1960s at the AVR plant

in Jolich, Germany. The AVR was a 30 MWth, indirect cycle research reactor that

demonstrated the basic characteristics of pebble fuel reactors. This demonstration was

followed by the 300 MWe THTR indirect cycle reactor which operated in the mid-1 980s for over

two years, before being shut down for non-technical reasons. The PBMR design is founded on

the development work and insights gained from the research, construction and operations of

these two predecessors. The PBMR Demonstration Power Plant is a full scale, single module

demonstration of the follow-on commercial multi-module plant design.

The design of the PBMR is different than its predecessors in that it employs a direct cycle (i.e.,

Brayton cycle) arrangement and operates at slightly higher core exit temperatures of 9000C.

The PBMR project began in 1993 in South Africa. This presentation will cover the research and

development, past, present and future, in four phases; basic R&D for concept finalization; initial

development for component design; verification programs and advanced research and

development.

Basic R&D Programs

Initial test programs were developed for core design, primary loop simulation, high temperature

reactor support components and power conversion unit components. PBMR and its key

suppliers performed a series of tests on part-scale and full-scale equipment to confirm the

conceptual design approach. Significant among the testing were the development and

operation of a high temperature nitrogen gas, three-shaft, Brayton cycle representation (PBMR

Micro-Model or PBMM) of the PBMR thermo-dynamic cycle. This facility demonstrated the

basic controllability of the three-shaft Brayton cycle power conversion unit (PCU) and provided

valuable benchmark data for PBMR thermodynamic codes. With the completion of the PBMM

testing, the initial concept design, system requirements and arrangements were completed.

Component Development Testing Programs

In 1999-2001, PBMR began development of individual test programs for high temperature

helium components and other reactor auxiliary items, and laboratory-scale fuel equipment and

associated quality measurement process development that were critical to the conceptual

design or fuel manufacturing process. This included a wide variety of fuel handling system

components and development of critical PCU components. The results of these development

tests confirmed the adequacy of a large number of individual components for use in the plant,

but were not under full environmental conditions. In the case of the PCU, test results coupled

with basic design analysis made way for a rearrangement of the PCU from a vertical, threeshaft cycle to a single, horizontal PCU.

17
Verification Testing Programs

Beginning in 2002, PBMR has identified a number of additional test facilities that would be

useful to verify and validate design performance codes and to demonstrate full reactor and

auxiliary component operability in the design high temperature PBMR environment. PBMR will

be constructing a High Temperature Facility (HTF) that will demonstrate system and component

operability and reliability under full design temperature and pressure conditions. The HTF will

also be full height for reactor and auxiliary system testing, where operating system performance

could be influenced by the physical scale of the system. The HTF will be constructed in 2005-6

and begin testing in mid-2006. PBMR will be constructing a Heat Transfer Test Facility to

validate all heat transfer mechanisms and coefficients in a pebble core configuration. In 2004-5

PBMR is conducting testing on the NACOK facility in Germany to further explore natural

circulation conditions and air ingress phenomena. Finally, PBMR is constructing fuel production

facilities at the Pelindaba site for the Koeberg Demonstration Power Plant. This facility will

demonstrate production fuel manufacturing capability and also provide early fuel for the PBMR

fuel irradiation testing program (FITP). The FITP will confirm the quality of the PBMR

manufactured fuel for the PBMR design envelop conditions and add additional confirmation

data to the established German fuel data base.

Advanced R&D Programs

PBMR has identified several different testing needs for the future to extend the PBMR design

beyond the technology used in the basic PBMR commercial design. These programs are

focused on three primary objectives: to improve fuel performance, utilization and disposal

options for existing PBMR plants; to develop component, materials, and fuel for VHTR

conditions approaching 12000C; and to use PBMR-based plants in other applications such as

hydrogen co-generation, desalination, or other process heat applications.

18
ACR Research and Development

A.J. White and D.J. Wren

Atomic Energy of Canada Limited

A new advanced reactor design, the ACR-700, is currently being developed to meet the needs

of the electrical utilities for safe, reliable and economic power generation technologies. The

ACR-700 is an evolutionary design that is based on the design of the CANDU reactors in

operation around the world. A few key design features differentiate the ACR-700 design from

the reference CANDU design, notably the use of slightly-enriched uranium fuel and the use of

light water as a coolant. The ACR-700 design is currently being reviewed by the US-NRC

under a pre-application review program.

The development of the ACR-700 is supported by a research and development program that

addresses the safety analysis of the design and the performance of the safety systems. In

addition, the ACR-700 development program includes activities to improve the economic

reliability and performance of the plant. In both cases, the research programs are incremental

and the technologies involved are based on the results of decades of research and operational

experience.

A key element of the ACR-700 development program is the extension of the safety analysis

code suite used to analyze the safety performance of CANDU reactors for application to the

ACR-700 design. A comprehensive program is in place to extend the validation of the

computer codes used to model the core neutronics and the transient thermalhydraulics of the

reactor coolant system under accident conditions. The former includes experiments in the

critical lattice facility (ZED-2) at the Chalk River Laboratories, analyses of data from other

relevant core configurations, and comparison with other code calculations (including MCNP).

The thermalhydraulics studies extend the validation of the CATHENA code for the ACR-700

design using results from the large-scale RD-14M test loop at the Whiteshell Laboratories.

The ACR-700 design, like the reference CANDU design, has a number of inherent, passive

features that slow the progression of potential severe accidents, mitigate the consequences,

and provide an extended time period for operator action to further mitigate event

consequences. The ACR-700 research program in this area is strongly coupled to the research

program supporting the operating CANDU plants. Research is in progress to support

understanding and modelling of the core degradation processes that are unique to the ACR700/CANDU design.

19
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20
Advanced Nuclear Energy Research in the United States

Carl Sink

U.S. Department of Energy

Advanced reactor and fuel cycle efforts in the United States sponsored by the Department of

Energy are carried out under three programs: the Generation IV Nuclear Energy Systems

Initiative, the Nuclear Hydrogen Initiative, and the Advanced Fuel Cycle Initiative. These

programs all support an overall strategy to create a new generation of nuclear energy systems

which will be deployable no later than 2030 in both developed and developing countries, both

for generation of electricity and to support a future hydrogen economy.

A technology roadmap for Generation IV nuclear energy systems was developed to identify the

most promising nuclear energy systems (consisting of both a reactor and associated fuel cycle)

for meeting the challenges of safety, economics, waste, and proliferation resistance and

physical protection. More than 100 experts from 10 countries evaluated more than 100

systems proposed by the worldwide R&D community leading to the selection of six Generation

IV systems: the Gas-Cooled Fast Reactor (GFR), the Lead-Cooled Fast Reactor (LFR), the

Molten Salt Reactor (MSR), the Sodium-Cooled Fast Reactor (SFR), the Supercritical-WaterCooled Reactor System (SCWR), and the Very-High-Temperature Reactor (VHTR). Significant

R&D is necessary to overcome technical challenges to develop each of the six concepts.

Crosscutting R&D is also being conducted in the areas of fuel cycle, fuels and materials, energy

products, risk and safety, economics, proliferation resistance and physical protection.

The Nuclear Hydrogen Initiative has been established to develop the production technologies

that can be most effectively coupled to next generation nuclear reactors for hydrogen

production. Efficient processes that can utilize the high-energy output of a nuclear reactor

include thermo chemical splitting of water and high-temperature electrolysis. The Advanced

Fuel Cycle Initiative is working to provide advanced fuel cycle technologies ready to be

deployed with light water- and gas-cooled reactors that have the ability to achieve a significant

reduction in the amount of high-level spent fuel requiring geologic disposal, to reduce

significantly the amount of accumulated plutonium in civilian spent fuel, and to extract more

useful energy from spent nuclear fuel components.

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22
High-Bumup Cladding Mechanical Performance during Cask Storage and PostStorage Handling and Transportation

Robert S. Daum, Hanchung Tsai, Yung Liu, and Michael C. Billone

Argonne National Laboratory (ANL)

Argonne, Illinois, USA

The assessment of cladding performance during spent-nuclear-fuel (SNF) pre-storage

operations, dry-cask storage, and post-storage handling and transport, including hypothetical

accidents, is important in ensuring that sub-criticality is maintained, that radioactivity is

contained, that cask external dose rates are limited, and that SNF assemblies can be safely

retrieved, handled and transported at the end of dry-cask storage. Although cladding failure is

not prohibited by federal regulations (CFR 71 and 72), the failure mode and extent of cladding

failure have a significant effect on the possible reconfiguration of SNF within storage and

transport casks. High-bumup, SNF Zircaloy-4 (Zry-4) cladding from pressurized-water reactors

is more susceptible to failure and possible fuel dispersal than low-bumup Zry-4 due to higher

hydrogen pickup, higher internal pressure and higher corrosion level. Hydride precipitation,

corrosion (i.e., wall thinning), irradiation-induced defects, and possibly even post-reactor

thermal creep, will reduce cladding ductility and impact resistance. Also, transfer-and-drying

operations conducted from the SNF pool to the storage cask may result in thermo-mechanical

conditions that promote radial-hydride-induced degradation of cladding ductility and impact

resistance during severe loadings associated with hypothetical accidents. Such cladding would

be more susceptible to brittle failure at the end of dry storage when temperatures are expected

to be 150-250'C - below the ductile-to-brittle transition temperature for zirconium hydrides. A

testing program has been developed to investigate these effects in high-burnup Zry-4 cladding,

with particular emphasis on the conditions that promote radial-hydride precipitation and the

effects of these radial hydrides on cladding integrity. Preliminary experimental results are

presented from axial tensile and thermal creep tests, as well as conditions for radial-hydride

precipitation. The test plan, which includes ring-compression-ductility and ring-crush-impact

screening tests, is also described

The current regulations for storage and transportation of SNF are designed primarily to

maintain sub-criticality and to ensure that doses are less than regulatory limits, that the cask

provides adequate fuel confinement and containment, and that the fuel is retrievable. Interim

Staff Guidance No. 11, Revision 3 limits high-burnup cladding temperatures to •400°C during

short-term operations and normal storage conditions. Although this temperature limit is

intended to minimize radial-hydride precipitation in high-bumup cladding, license applications

for high-bumup storage, storage-and-transport, and transport casks are evaluated on a caseby-case basis due to the lack of pertinent data to assess high-bumup cladding behavior during

the complete cycle of pre-storage operations, storage and post-storage handling and transport.

The data generated within the ANL test program are to be used both by applicants for highbumup cask licenses (e.g., nuclear industry and DOE-RW) and by cask-license evaluators

(NRC-SFPO).

23
Experiments have concentrated on the microstructural characterization and mechanicalproperty testing of stress-relieved Zry-4 (15x15 design) irradiated in H.B. Robinson Unit No. 2 to

a rod-average bumup of 67 GWdIMTU and a fast neutron fluence (E > 1 MeV) of 14x 1 02 n/m2.

In Grid Spans 3 and 4 (fuel midplane to 0.7-m above the fuel midplane), destructive

examinations of the as-irradiated cladding have been performed. Circumferentially averaged (8

locations) corrosion layers of 70±10 to 100±10 pm, with no spalling, have been measured,

along with cross-section-averaged (90 segments) hydrogen contents of 550±80 to 750±90

wppm. Precipitation of circumferential hydrides shows varying distribution, density, and particle

size along the axial, azimuthal, and radial directions of the cladding. In a few cladding

locations, transverse metallography in an etched condition shows a significantly high hydride

density localized in roughly a 90 arc directly under the outer-surface corrosion layer and to a

radial depth of z100 pm, suggesting the presence of a hydride "lens." Such hydride

microstructures are known to reduce cladding ductility under tensile loading, but little is known

about the evolution of these microstructures under thermo-mechanical conditions associated

with SNF drying, transfer and storage.

As compared to non-irradiated Zry-4 (15x15 design), room-temperature axial tensile properties

of the high-bumup (690±40 wppm H) Zry-4 at a strain rate of 0.1%/s show an increase in yield

(600 - 770 MPa) and ultimate tensile (765 -3 950 MPa) strengths and a decrease in uniform (6

- 3%) and total (14 - 4%) elongations. The strength increase appears to be due mainly to

radiation-induced hardening, while the ductility decrease appears to be due to both radiationand hydride-induced embrittlement. Thermal annealing tests with high-bumup Zry-4 samples

show that strength properties (based on microhardness data) appear to recover by =75% and

circumferential hydrides tend to homogenize across the cladding radius after 72 hours at

420'C. These results suggest that SNF drying operations may partially anneal radiation-induced

hardening. The degree of ductility recovery with annealing remains to be demonstrated.

Two thermal creep tests (C14 and C15) have been completed using defueled high-bumup Zry-4

cladding specimens, which are top-welded to active internal-gas-pressurization systems in

order to maintain constant gas pressure inside the creep specimens. These tests were

conducted at 400'C for 101 days at a pressure of 29.5 MPa and an initial hoop stress of 190

MPa. The specimens were depressurized periodically prior to cooling to room-temperature for

diameter measurements. Both specimens remained intact, no localized bulging (precursor to

rupture) was observed, the average hoop creep strains were =3.6% and the peak hoop creep

strains were =5%. In order to induce radial-hydride precipitation, the C15 specimen was cooled

at =2.4 Cfs from 400'C under full pressure during the final shut-down. The sample

depressurized at 205'C and a midplane true hoop stress of =205 MPa due to failure in the

upper weld region. Post-test metallography at three axial locations showed significant radialhydride and negligible circumferential-hydride precipitation. However, post-test hydrogen

measurements indicated substantial loss of hydrogen from the C15 specimen (=670 - 320

wppm at the midplane) to the Zircadyne-702 end-fittings (12 - 210 wppm at the bottom endplug). Redesign of the thermal creep test train and furnace is in progress to minimize the

hydrogen loss and the axial temperature gradient from the specimen midplane (400'C) to 30

mm above the midplane (=390'C). However, it is interesting to note that the hydrogen solubility

of non-irradiated Zry-4 is =210 wppm at 400'C. The absence of visible circumferential hydrides

at the specimen midplane (with 320 wppm) suggests that high-bumup Zry-4 is capable of

24
trapping about 100 wppm of hydrogen, which most likely precipitates during rapid cooling as

sub-micron-size hydrides. It will be interesting to determine if this excess hydrogen precipitates

as visible radial hydrides under the slow cooling rates (<4'C/day) typical of drying-transferstorage.

The preliminary axial-tensile, thermal-creep, and hydride-reorientation results have been used

to develop a test plan to better understand the mechanical behavior of high-bumup Zry-4

cladding under drying-transfer, storage, and post-storage handling-transport conditions. In

additional to tensile and creep tests of pool-stored high-bumup Zry-4, sealed specimens will be

annealed for =3 days at 380-420'C and at hoop stresses of 0, 60, 90, 120, and 150 MPa and

slow-cooled at -3-C/day under decreasing pressure. Rings cut from these 100-mm-long

samples will be subjected to ductility (diametral compression at 0.1%Is and 100%/s) and crushimpact failure-energy screening tests. These tests will be conducted at room-temperature and

150'C. The decreases in ductility and failure-impact energy will be correlated to the extent of

radial hydride formation to map out cooling conditions - especially stress at 400'C - that are

detrimental to high-bumup Zry4 cladding integrity. Additional tests (e.g., fracture toughness)

may be conducted on cladding subjected to these detrimental cooling conditions.

25
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26
Data Needs for the Transportation and Storage of High Bumup Fuel

RE Einziger, CL Brown, CG Interrante and GP Hoomseth

US Nuclear Regulatory Commission

Spent Fuel Program Office

The storage and transportation of low to medium bumup spent nuclear fuel

(<45 GWdIMTU) are mature, ongoing operations with a strong safety record. In order to

improve nuclear reactor utilization, utilities are operating plants with fuel that is licensed for

extended bumup. Fuel with bumups up to 62 GWdIMTU and possibly beyond will need to be

stored and transported. In addition, events of recent years have raised the question of the

safety of transport and storage casks that might come under terrorist attacks.

The NRC has been recently focusing on the storage and transportation of the higher bumup

fuel because it's cladding may have degraded mechanical properties relative to lower bumup

fuel. The increased fluence and time-in-reactor have caused changes in the fuel pellet and

cladding characteristics, In addition, newer cladding alloys are in use that have been developed

to meet in-reactor performance standards.

During extended operation, the fuel is in the reactor for a longer time, hence more cladding

oxidation occurs. Approximately 15-20% of the hydrogen generated during the oxidation

process diffuses into the cladding, and for the most part accumulates at the outer cooler edge.

Depending on the alloy, the hydrogen concentration in this outer rim of the cladding can reach

above 600 wppm compared to 300 wppm or less in lower burnup fuel. As the temperature of

the fuel is raised in the drying process, after cask loading, much of this hydrogen will go back

into solution. Later cooling of the fuel will re-precipitate the hydrogen as hydrides as the

solubility limit decreases with decreasing temperature. If the applied hoop stress due to the

internal rod gas pressure is sufficient, radial hydrides will form. Radial hydrides may degrade

the mechanical properties of the cladding and possibly lead to rod breach during transportation

and storage conditions. At the current time, this critical stress is not well defined.

To meet the performance needs in reactor, the vendors have developed a number of new alloys

that have more corrosion resistance than Zircaloy at comparable bumups. The corrosion of

these alloys produces less hydrogen thus making the potential for hydride generation and

reorientation potentially less detrimental. A comparative database on the creep behavior,

fracture toughness, and other mechanical properties must be established by the licensee to

determine if these alloys fall under the same guidelines for storage and transportation as the

Zircaloy alloys.

At higher burnup, the fuel pellet forms a rim region, representing about 4-8% by volume of the

fuel. This rim retains fission gas under high pressure, restructures to a submicron grain size,

and has higher plutonium content than the body of the pellet. Very little is known about the

behavior of this rim region under impacts that might be characteristic of a severe drop or

27
terrorist attack. Since the fuel grains are already in the respirable size range, it is important to

know the relative fracture and dispersal behavior of this fine-grained material compared to

behavior of grains from lower bumup fuel that are 100 times larger.

To address the effects of these changes on the safe transportation and storage of spent fuel,

expansion of the current databases by the vendors and utilities is expected. The NRC has

instituted a confirmatory data gathering activity to allow the staff to evaluate the adequacy of

these databases when license applications for the storage and transportation of high bumup

fuels are submitted.

The NRC continues to seek data and analysis methods from the nuclear industry to support the

safe storage and transportation of high bumup fuel. This paper will discuss data that are

needed to evaluate the storage and transportation of high burnup spent nuclear fuel, and the

NRC confirmatory programs to obtain data. Furthermore, the implications of the uncertainties

on pending changes for Interim Staff Guidance ISG-1 1 Revision 3 will also be discussed.

28
Perspective on Requirements for Spent Fuel Storage and Transportation

Albert Machiels

Electric Power Research Institute, Inc.

A significant regulatory milestone was achieved when the Spent Fuel Project Office published

Interim Staff Guidance (ISG) 11, Revision 2 entitled: Cladding Considerations for the

Transportation and Storage of Spent Fuel. The acceptance criteria specified in Rev. 2 were

supplemented by an additional set of acceptance criteria for low burnup fuel in Rev. 3 published

in November 2003.

In both revisions, the acceptance criteria address dry storage only, but not transportation. This

can be considered unusual for the following two reasons:

1. Most systems to be used for high-bumup spent fuel are intended to be dual-purpose.

Clearly, the acceptance criteria for loading spent fuel in such dual-purpose systems

should envelop both storage and transportation applications.

2. The most limiting considerations that led to the acceptance criteria for dry storage were

actually based on transportation, i.e., on limiting the potential formation of radial

hydrides in the spent-fuel cladding during dry storage in order to minimize potential

degradation of the cladding mechanical properties used in analyzing transportation

accidents.

Presently, the U.S. regulations (Part 71) do not have specific criteria with regard to performance

of cladding under hypothetical accident conditions.' However, the configuration of the spent

fuel in the transportation package after an accident is an input to the shielding and criticality

evaluation, as well as possibly to the confinement evaluation. Clearly, if it can de demonstrated

that no significant damage occurs either to the spent-fuel itself (no re-configuration), or to the

package (no potential for moderator ingress), the criticality analysis, generally considered as a

key driver from a regulatory perspective, would be greatly simplified.

Confirmatory and new experimental work is being conducted at ANL with the participation and

funding of several organizations, including the US NRC, US DOE, EPRI, and the fuel vendors.

This work is expected to demonstrate the conservative, but appropriately realistic, technical

basis, which resulted in the acceptance criteria for storage contained in Rev. 213 of ISG-1 1, and

its applicability to transportation applications. Concurrently, modeling of spent-fuel performance

under impact loading conditions is a necessary activity for both guiding the experimental work

and getting the most value from its results.

These activities are, or will be, supplemented by other generic efforts:

1. Risk assessment of criticality event during transportation

2. Implementation of full-bumup credit

However, Part 71 contains cladding performance criteria for normal conditions of transport.

29
This three-prong approach (risk assessment, fuel cladding performance, and burnup credit) is

expected to support the contention that transportation risk minimization is a direct function of

the reduction in the number of shipments. If this is indeed the case, the use of high-capacity

packages should generally be the preferred implementation path for dual-purpose systems.

30
LOCA Integral Test Results for High-Burnup BWR Fuel

Yong Yan, Tatiana A. Burtseva, and Michael C. Billone

Argonne National Laboratory (ANL), Argonne, IL 60439

LOCA integral tests with high-bumup BWR fuel have been conducted at ANL to provide NRC

and industry with data for assessing the applicability of the LOCA embrittlement criteria in 10

CFR50.46 to high-bumup fuel rods. Similar tests are planned for high-burnup PWR fuel.

These criteria limit peak cladding temperature to 22000F (12040C) and maximum oxidation

(equivalent cladding reacted, ECR) to 17% during high temperature steam oxidation to ensure

residual ductility during ECCS quench (T 2 135tC) and during possible post-LOCA seismic

events (T = 100'C). Appendix K specifies the use of the Baker-Just (BJ) correlation for

calculating reaction rate and ECR. NRC Regulatory Guide 1.157 (1989) allows for the use of

best-estimate correlations (e.g. Cathcart-Pawel [CP]) for calculating reaction rate and ECR. At

1204'C, 17% BJ-ECR = 13% CP-ECR. In anticipation of the degrading effects of high-burnup

operation on the cladding, NRC Information Notice 98-29 (1998) specifies that ECR should be

based on total oxidation (corrosion plus transient steam oxidation). For PWR Zircaloy-4 (Zry-4)

cladding, high-bumup operation results in peak corrosion layers of -100 gim, corresponding to

8-10% ECR, and peak hydrogen concentrations of 600-800 wppm. The ANL LOCA integral

tests, which are conducted with fueled specimens, are designed to improve our understanding

of the behavior of high-bumup fuel exposed to a LOCA transient, as well as to provide data for

the assessment of the LOCA embrittlement criteria.

LOCA integral test results are reported for fueled high-bumup BWR specimens. These results

are compared to baseline data for zirconia-pellet-filled, nonirradiated Zry-2 cladding specimens

exposed to the same tests. Four LOCA integral tests have been conducted with specimens

from Limerick BWR fuel rods at 56 GWdIMTU. In the as-discharged condition, the Limerick

cladding (Zr-lined Zry-2 GE-1I 9x9 design) has a corrosion layer of Z10 PM and a hydrogen

content of =70 wppm. The specimens were internally pressurized with helium to a gauge

pressure of =8.3 MPa at 300'C. During heating in steam at 5 C/s, the internal pressure rose to

s9 MPa prior to burst at =750'C. The full LOCA sequence (Fig. 1) calls for heating in steam at

5 C/s to 1204'C, holding for s5 minutes at 1204'C (s20% CP-ECR), slow-cooling at 3C/s to

800'C and bottom-flooding to quench the cladding from 800 to 100'C.

The ICL#1 test specimen was ramped to burst in argon and slow cooled. The ICL#2 specimen

was exposed to the LOCA test sequence with the exception of quench. The nondestructive

results from these tests indicated more similarities than differences between high-burnup

specimens and non-irradiated specimens. The ICL#3 specimen achieved partial quench

(800'C to 470'C) before failure of the quartz chamber that surrounded the specimen. The full

LOCA sequence with quench was demonstrated in the ICL#4 test (see Fig. 1). Nondestructive

examinations included photography and profilometry for all 4 specimens and gamma scanning

for the ICL#3 and #4 specimens. Destructive examinations were performed on ICL#2 and #3

specimens to determine oxide layer thickness, fuel morphology, and axial profiles of hydrogen

and oxygen concentration. Oxide-layer thickness and oxygen-content results indicate two-sided

31
oxidation in the ballooned-and-burst region of both high-bumup and nonirradiated specimens.

The axial profile of hydrogen pickup for ICL#2 and #3 specimens is shown in Fig. 2 and

compared to the data for nonirradiated Zry-2 cladding. For nonirradiated specimens, the

hydrogen pickup was low in the burst region and very high at 70-90 mm above and below the

burst mid-plane. For high-bumup-fueled cladding, the hydrogen peak was towards the burst

mid-plane. Because of the large secondary hydriding from the cladding inner surface,

significant degradation of post-quench ductility (PQD) is expected for the ICL#4 ballooned

region, even at 100-135-C. A 5t hot-cell integral test, with a 2-minute hold time at 1204-C, is in

progress to determine if the 17% BJ-ECR criterion is sufficient to protect the ballooned region

from embrittlement due to steam oxidation and hydrogen pickup at 1204'C.

In the uniform bumup region (within grid spans 2-5), the high-bumup PWR cladding for the next

set of LOCA tests differs from BWR cladding in terms of corrosion layer thickness (=40 to 100

pm) and hydrogen content (=400 to 800 wppm). For test planning purposes, the separate

effects of hydrogen on diametraf-compression PQD have been investigated with prehydrided,

nonirradiated 15x1 5 Zry-4 cladding rings after oxidation at 1204'C and quench. For as-received

(=10 wppm H) cladding that was oxidized at 1204-C, the ductile-to-brittle-transition CP-ECR

was 8% at room'-temperature, 12% at 100IC, and 14% at 135'C. In contrast, cladding with 400-

to-800 wppm hydrogen exhibited significant embrittlement, even after moderate oxidation at

1204-C. Samples prehydrided to 400-800 wppm and oxidized at 1204'C to 8% CP-ECR

exhibited no ductility. With anticipated secondary hydrogen update from the cladding inner

surface, the embrittlement ECR is expected to be «<17% for high-bumup PWR specimens

subjected to LOCA integral tests at 1204 C, even if the ECR is determined by the sum of the

corrosion layer and the BJ-calculated transient ECR. The baseline data from prehydrided

cladding are being used to plan the PWR hold-times at 1204'C such that the embrittlement

ECR can be determined effectively in the ballooned and non-ballooned regions of the PWR

LOCA integral test specimens.

In-cell LOCA Test ICLX4 with Limerick Fuel Sample, 314/04

u -T : M yk TC a

1600- - - - - - - -.-- - - -.-

C6 c- 1200-

CalM TC. 773°C |

o.400- .

0 400 800 1200 1600

Time (s)

Fig 1. Temperature and pressure histories for LOCA integral test (ICL#4) with high-bumup

BWR fuel.

3-
5E ,000

2,500 1 0

2,000

*~1,500

k 1,: 0

-110 -90 -70 -50 -30 -10 10 30 50 70 90 110

Distance from the Burst Center, mm

Fig 2. Axial profile of hydrogen pickup in high-burnup BWR cladding used in LOCA integral

tests ICL#2 and ICL#3 (5 minutes at 1204-C); OCL#1 I results are for nonirradiated Zry2 cladding.

33
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34
LOCA Testing at Halden

E. Kolstad, W. Wiesenack, V. Grlsmanovs

(OECD Halden Reactor Project)

The safety criteria for loss-of-coolant accidents are defined to ensure that the core will remain

coolable. Since the time of LOCA experiments in the '70s, which were largely conducted with

fresh fuel, changes in fuel design, the introduction of new cladding materials and in particular

the move to high bumup have generated a need to re-examine these criteria and to verify their

continued validity. Hot cell programs concentrating on embrittlement and mechanical properties

of high bumup cladding have been initiated in some countries.

The Halden reactor is suited for integral in-pile tests on fuel behavior under LOCA conditions. It

is intended to utilize fuel rods irradiated in commercial reactors to bumup levels >50 MWdlkg

with a thorough characterization regarding the state of the cladding and the bonding with the

fuel. Participating organizations have supplied both PWR and BWR fuel with desired

characteristics. It is the intention to include medium bumup (40-45 MWdIkg) fuel in the test

series in order to assess the difference between medium bumup and very high bumup fuel (>60

MWd/kg.

The Halden experiments are single pin tests and will focus on effects that are different from

those studied in out-of-reactor tests. A prototypical bounding LOCA transient does not exist,

and it was recommended that the test conditions be selected to meet the following primary

objectives:

- to maximize the balloon size to promote fuel relocation, and to evaluate its possible effect

on cladding temperature and oxidation, and

- to investigate the extent (if any) of - "secondary transient hydriding" - on the inner side of

the cladding around the burst region.

Target peak clad temperatures (PCT) for the pre-irradiated rods have been set at 8000C and

11 000C for high and medium bumups.

The first LOCA trial runs were carried out in the Halden reactor in May 2003, using a fresh,

unpressurized PWR rod with Zr-4 cladding. The main objective was practicing, to determine

how to run the later experiments with pre-irradiated segments. PCTs in the range 8000C -

11 000C for the initial six transients were successfully achieved.

The rig with the fuel rod was located in a high-pressure flask connected to an ex-reactor highpressure loop incorporating a blow-down system. The cladding temperature transients can be

controlled by rod power and an annular heater surrounding the rod. A spray system at the top of

the rig is used to supply steam for the oxidation and hydriding processes. The extensive rig/rod

instrumentation enables power calibration and neutron flux monitoring, and includes a fuel

centre thermocouple (first rod only), three cladding thermocouples (at two elevations), rod

pressure sensor, cladding extensometer, heater thermocouples etc. The geometry of the test

section is shown in Fig. 1.

35
The second trial LOCA test run was successfully carried out on 28 May this year. The test was

performed with a fresh, pressurized PWR rod and consisted of a blowdown phase, heat-up,

hold at target PCT and termination by reactor scram (Fig. 2). The main objective was to achieve

ballooning and cladding failure to find out how to run later experiments with pre-irradiated

rodlets.

The target cladding temperature of 10500C was achieved, and rod rupture occurred at 8000C,

as evidenced by rod pressure and elongation measurements (Fig. 3) as well as the gamma

monitor on the blowdown line to the dump tank. The hold time above 9000C was 390 s and the

average temperature increase rate between 600 and 800'C was -70C/s. The azimuthal

temperature variation was small prior to cladding failure, within ± 2-30C, and the tensile hoop

stress -55 MPa. The spray was applied intermittently during the high temperature period and

the test was terminated by a reactor scram. The rod with its capsule will undergo gammascanning at Halden before it is shipped to Kjeller hotcells for detailed PIE.

Pre-test calculations were carried out by VTT using the FRAPTRAN/GENFLO code. The code

predicted the maximum cladding temperature with good accuracy. Also the timing and

temperature of the rod failure was well predicted. Further calculations will be performed in

preparation of the next test (the first with a pre-irradiated PWR segment).

36
034 Flask Heater Cable

Fig. 1. The geometry of the test section

lo M . I . . . I * .. ... .. . I....

0

us .0

I

UJ

0t

90,

80

70

60*

50

40-

30

20-

10.

4 ~E

-P- PF1

-E- EC2

.5

o E

z

0 -.5 z

-1 §

Ul

_ w1 0 600 i200 1800 2400 3000 3600

TIME (sec)

Fig. 2. Claddieg, beaterand loop inlet and outlet

temperatures duing the LOCA test. Also fuel and

heater poweris shown. Nomenclature:A = fuel

power, Q =heaterpower, I = ilet loop

temperature, dashed 1-2 = beater temperature

n- -- -, - -- - . - V- _

_- 200 400 60 800 1000

CLADDING TEMPERATURE (

0

C)

Fig. 3. Measurements of rodpressure and cladding

elongation as function cladding temperature during

the LOCA test (Rod rupture at 00'C)

37
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38
Results from Studies on High Bun-Up Fuel Behavior

Under LOCA Conditions

Fumihisa NAGASE and Toyoshl FUKETA

Department of Reactor Safety Research

Japan Atomic Energy Research Institute

To promote a better understanding of high bumup fuel behavior under loss-of-coolant accidents

(LOCAs), a research program is being conducted at the Japan Atomic Energy Research

Institute (JAERI). The program consists of integral thermal shock test and other separate tests

for oxidation rate and mechanical property of fuel claddings.

1. Intearal thermal shock test

In the test, short rods are heated up, burst, oxidized in steam and quenched by flooding water

in order to evaluate fracture-bearing capability of oxidized fuel claddings under the simulated

LOCA condition. The Japanese LOCA criteria for cladding embrittlement are based on results

from the integral thermal shock tests. Tests were performed with non-irradiated cladding tubes,

which were mechanically thinned and pre-hydrided to evaluate the separate effects. Claddings fractured into two pieces during the quench, depending primarily on the oxidation. The

threshold of the fracture in terms of the oxidation decreases with the higher concentration of

initially absorbed hydrogen and with the larger load in axial constraint.

Two PWR fuel rods, irradiated to 39 and 44GWd/t (rod average) at Takahama unit-3 reactor,

are subjected subsequently to the tests. The cladding material is low-Sn (1.3%Sn) Zircaloy-4 and thickness of oxide layer formed in the reactor ranged 18 to 25 pm. Before the tests, fuel

pellets were removed from 190mm-long segments, and alumina dummy pellets were loaded in

the defueled claddings. Zircaloy end-plugs were welded at the both ends of the claddings and

the fabricated test rods were pressurized to about 5MPa with argon gas. Six test rods were

quenched after rupture at about 800 deg C and isothermal oxidation at 1030 to 1192 deg C.

Two claddings which were oxidized to about 26 to 30%ECR* fractured during quench. The

fracture of the irradiated claddings agrees with the failure criteria for non-irradiated claddings containing similar hydrogen concentrations. Four claddings oxidized to about 16 and 25%ECR

survived the quench. These indicate that fracture/no-fracture threshold is not reduced so

significantly by irradiation to the examined bumup level.

2. Mechanical tests of oxidized and quenched cladding

Besides the integral thermal shock tests, mechanical tests are performed to develop methodology for predicting cladding fracture on quenching as well as to examine mechanism of

cladding embrittlement.

Ring-tensile and ring-compression tests were performed on non-irradiated Zircaloy-4 claddings which were pre-hydrided to 400 and 800 ppm, oxidized at 1000 to 1250 deg C, and finally

quenched.

- Ductility reduction observed in the ring-tensile tests was not remarkable for the oxidation

between 10 and 20%. Uniform tensile stress in the circumferential direction is applied to

the cladding in the ring tensile test, and this stress state is quite different from that is

applied during quench. This suggests that test methods should be carefully selected in

order to estimate cladding embrittlement under LOCA condition.

39
- The ring-compression tests detected the sudden ductility drop above 15% oxidation for

claddings which were oxidized at 1200 deg C without hydriding. The significant ductility reduction occurred at lower oxidation level in the pre-hydride claddings (400 and 800

ppm). This indicates that pre-hydriding enhances cladding embrittlement of oxidized

cladding and agrees with the results of the integral thermal shock tests.

- It has been generally considered that slow cooling after high-temperature oxidation

enhances oxygen diffusion from oxide layer into metallic prior-O phase, and consequently microstructure and ductility of metallic prior-O phase changes depending on the cooling

rate. To confirm that, ring compression tests were performed with cladding s which were

cooled at different rates after oxidation at 1100 and 1200 deg C. It is shown that the

influence of slow cooling differs depending on oxidation temperature and oxidation

amount. The influence becomes greater in the cladding oxidized at 1200 deg C and at

the lower oxidation level.

Acknowledgment

The integral thermal shock test with irradiated PWR fuel claddings has been performed as a

corporative research program between JAERI and Japanese PWR utilities.

* ECR is estimated by the Baker-Just equation, taking account of double sided oxidation

and wall thinning by ballooning. The nitial cladding thickness used in the estimation is

metallic thickness after corrosion during the reactor operation.

40
Realistic Assessment of Fuel Rod Behavior Under Large-Break LOCA Conditions

M. E. Nissley, C. Frepoli, K. Ohkawa

Westinghouse Electric Company

Fuel clad swelling and rupture can occur during a loss of coolant accident (LOCA), depending

on the core heatup transient and the pressure differential across the cladding. Clad rupture will

lead to release of fission products from the fuel, and double-sided metal-water reaction

(oxidation) within the ballooned region. In order to simplify the radiological dose calculations, it

is typically assumed that 100% of the rods in the core fail. However, it is instructive to consider

what a realistic failure fraction might be under more representative conditions. The objective of

this study will be to assess the extent of failure and the consequences for the cladding oxidation

for the large break LOCA scenario, with and without detailed treatment of uncertainties.

The first assessment will use a deterministic calculation of a large break LOCA under normal

operating (baseload) conditions, using the realistic computer program WCOBRA/TRAC. In

order to ensure some cladding rupture, a full train of ECCS will be assumed lost (worst single

failure) and bounding rod power conditions in the lead fuel assembly will be used. Estimates of

the extent of rupture throughout the core will be made by considering peak cladding

temperature dependence on rod power, rupture temperature as a function of cladding pressure

differential, bumup effects on rod internal pressure, and a core-wide census of rod power and

bumup. A comparison of the maximum local oxidation within and away from the ballooned

region will also be made.

The second assessment will use the results from a best-estimate plus uncertainties analysis of

a large break LOCA, performed using methods consistent with US design basis LOCA

regulatory requirements. Uncertainties in thermal-hydraulic models, plant operating conditions,

and fuel rod models are accounted for in this method by simultaneously sampling from the

uncertainty distributions of each parameter for each transient case. The plant operating

conditions considered in the uncertainty analysis include transient power distributions, such that

more severe axial shapes and higher linear heat rates are considered than in the first

assessment. The extent of rupture within the uncertainty cases will be examined, and

conclusions drawn relative to the threshold for rupture. Maximum local oxidation within and

away from the ballooned region will be reviewed for the most limiting cases, and those results

will be assessed for their dependence on the related fuel rod uncertainty parameters (burst

strain, degree of fuel relocation, etc.).

The information presented in these assessments should be interpreted as illustrative and

representative. Extent of rupture and degree of oxidation are highly dependent on the transient

conditions, which are highly dependent on plant-specific parameters such as core power,

nuclear peaking factors, ECCS capacity and other factors.

41
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42
Post-Quench Ductility of Advance Alloy Cladding

Michael C. Billone, Yong Yan, and Tatiana A. Burtseva

Argonne National Laboratory (ANL), Argonne, IL 60439

Diametral (ring)-compression screening tests have been conducted to assess the ductility of

17x17 Zry-4, ZIRLO and M5 samples oxidized to 0-20% ECR at 1OOtC, I 100 C and 1200C.

The 25-mm-long samples were exposed individually to two-sided steam oxidation in the same

test apparatus for the same test times, slow cooled to 8OO'C and water-quenched. Test times

were calculated using the Cathcart-Pawel weight-gain correlation and a reference wall thickness

of 0.57-mm (Zry-4 and ZIRLO). Based on sample weight increase (normalized to the surface

area), weight gain was determined and compared to Cathcart-Pawel (CP) predictions. As

expected, good agreement was achieved among Zry4, ZIRLO and M5 and the CP predictions

for the I 00-C- and 1200'C-oxidized samples, while differences in weight gain vs. time were

observed for the alloys oxidized at 1000C. After =3400 s at 1000C, the weight gains of M5

and ZIRLO were =36% and =20% less than Zry-4, respectively. For lower test times, the M5

weight gain was consistently lower than Zry-4, while the ZIRLO and Zry4 weight gains were

about the same. The experimental weight gains, along with the sample thickness (0.61 mm for

M5), were used to determine experimental ECR values.

Similar tests were performed with El 10 tubing (0.71-mm) at 1000-C and 1 IOOC to characterize

the onset of breakaway oxidation, subsequent hydrogen pickup, and decrease of post-oxidation

ductility. Weight gain for as-received El 10 could not be determined accurately because of the

early (<300 s at 1000-C) breakaway oxidation resulting in oxide flaking and spalling. However,

polished and/or machined-and-polished (0.58-0.69 mm wall) El 10 exhibited stable oxide growth

for oxidation times up to =300 s at 1000C and >1000 s at IO0C. For these samples, the

El10 weight gains were similar to M5: lower than Zry-4 at 1 OOOC and about the same as Zry-4

at 1100-C.

Ring-compression samples (8-mm-tong) were cut from the oxidized samples and tested initially

at room temperature and 0.033 mm/s displacement rate (0.35%/s diametral strain rate). Loaddisplacement curves were analyzed by the traditional offset-displacement method to determine

plastic ductility. It was found that this method over-predicts plastic displacement, determined

directly from pre- and post-test diameter measurements along the loading direction, by •0.2 mm

(2% strain). For rings with offset strains < 3%, direct measurement of post-test diameter after

the first through-wall crack proved to be a more reliable measure of ductility. Rings with

permanent strains < 1% were classified as brittle.

Zry-4, ZIRLO and M5 exhibited ductile behavior (offset strains 2 3% and/or permanent strains 2

1%) after oxidation at 1000 C and 1 OO'C for CP-calculated ECR 2 17%. The 1000-C results

are interesting in that the all three alloys exhibit =3% offset-strain ductility after oxidation for the

same test time (=3400 s) at 1000'C, even though the measured ECR values were 22.4%,

18.0% and 13.3% for Zry-4, ZIRLO and M5, respectively. These results suggest that oxidation

time at 1 OOOC and CP-calculated ECR correlate better with ductility than ECR based on actual

weight gain, especially for M5. For as-received El 10, embrittlement occurs after oxidation at

1000C for =625 s, corresponding to an average hydrogen pickup of =300 wppm. The

hydrogen concentration in this oxidized sample was highly non-uniform (25-560 wppm) in the

circumferential and axial directions with the high hydrogen concentrations occurring under local

areas of breakaway oxidation. The results suggest that hydrogen entering El 10 through cracks

in the oxide layer is essentially 'frozen" in position during the course of the test. For polished

and machined-and-polished El 10 oxidized at 1100 C for • 1011 s, the material was ductile up to

43
hydrogen concentrations occurring under local areas of breakaway oxidation. The results

suggest that hydrogen entering El 10 through cracks in the oxide layer is essentially mfrozenn

in position during the course of the test. For polished and machined-and-polished El 10

oxidized at I 00-C for •1011 s, the material was ductile up to 19% CP-ECR based on the

machined-and-polished wall thickness of 0.58 mm. While surface polishing was found to

stabilize oxide growth on El10 surfaces, pre etching with solutions containing HF tended to

de-stabilize oxide growth at earlier test times than observed for as-received El 10 tubing.

For Zry-4, ZIRLO, M5 and El 10 (RRC-KIIRIAR data) samples oxidized at 1200'C, the roomtemperature offset strains decreased rather abruptly from 5 to 10% ECR. Based on

interpolation of the permanent strain data, the embrittlement ECR values at roomtemperature were: z10% for Zry-4 and =12% for ZIRLO and M5. Based on the RRCKI/RIAR offset strain data, the embrittlement ECR for El 10 oxidized at 1200-C was z8%.

Zry-4, ZIRLO and M5 alloys oxidized at 1200'C were retested at 135 C and 0.35%/s, as well as

3.50%/os for Zry-4. The enhancement in post-quench ductility with test temperature was

remarkable. As shown in Figs. 1 and 2, ZIRLO and M5 (extrapolated) retained significant

ductility for measured and CP-calculated ECR values >17%. Zry-4 also maintained postquench ductility for ECR > 17% under these conditions. At the higher strain rate of 3.5%/s, Zry4 was also ductile for ECR > 17%. The implication of these results is that as-fabricated Zry-4,

ZIRLO and M5 satisfy the LOCA embrittlement criteria during and shortly after quench.

Although testing of the 17x17 cladding alloys at 100'C has not yet been conducted, the results

from testing of 15x15 Zry-4 at RT, 100'C and 135'C (Y. Yan, this meeting) suggest that these

alloys will retain ductility at the longer-term post-quench temperatures of = 1OOC.

However, in-reactor corrosion results in hydrogen pickup. This hydrogen can enhance

embrittlement directly, as well as indirectly by increasing the oxygen solubility and embrittlement

in the prior-beta layer of the post-quench cladding alloy. Preliminary test results at 135-C with

prehydrided 17x17 and 15xl5 (Y. Yan, this meeting) Zry4 samples oxidized at 1200'C indicate

that =5%-ECR Zry-4 embrittles at =600 wppm H, that =8-9%-ECR Zry-4 embrittles at =350

wppm H and =1l%-ECR Zry-4 embrittles at =300 wppm H. Future work will focus on the postquench ductility at 100-135 C of prehydrided-nonirradiated ZIRLO and M5 and high-bumup Zry4, ZIRLO and M5.

50-

-40

so

i

0 20

10

10

4 S 8 10 12 14 16 13 20 22

Measured ECR, %

Fig. 1. Post-quench ductility of ZIRLO oxidized at 1200-C and ring-compressed at RT and

135-C.

44
so

0 2 4 £ * 10 12 14 16 1 20 22

Measured ECR, %

Fig. 2. Post-quench ductility of M5 oxidized at 1200'C and uing-compressed at RT and 1 35'C.

45
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46
Overview of the CEA Data on the Influence of Hydrogen on the Metallurgical and

Thermal-Mechanical Behavior of Zircaloy-4 and M5T "' Alloys Under LOCA

Conditions

J.C. Brachet(*)"), L.. Portiere1 ), V. Maillot!'), T. Forgeron'2), J.P. Mardon (a,

P. Jacques"), A. Lesbros(4)

(1) CEA-DEN, DMNISRMA, CEANSaclay, 91191 Gif-sur-Yvette Cedex, France

(2) CEA-DSNI-Comb., CEAISaclay, 91191 Gif-sur-Yvette Cedex, France

(3) FRAMATOME ANP Nuclear Fuel, 10 rue Juliette R6camier, 69456 Lyon Cedex 06,

France

(4) EDF, SEPTEN, 12-14 avenue Dutrievoz, 69628 Villeurbanne Cedex, France

(*) corresponding author: jean-christophe.brachet@cea.fr

A few years ago, within the framework of the CEA/EDF/Framatome-ANP R&D cooperative

program, we made the assumption that the bum-up influence on the thermal-mechanical

behavior of the fuel cladding tubes under LOCA conditions should be strongly linked to the

hydrogen up-take due to the in-service oxidation (see for example discussion pp. 276-277 in

[1]). Thus, since that time, an extensive experimental program has been conducted in CEA labs

on as-received and pre-hydrided Zy4 and M5Tm advanced alloys of Framatome-ANP to get a

better insight into the influence of the hydrogen on the thermal-mechanical cladding behavior

during the first phase of the LOCA transient (ballooning and rupture) and for post-quenched

conditions (residual ductilityltoughness...) [2] [3] [4].

On the one hand, one of the main assumptions here was that the microstructural defects, and

the resultant hardening produced under heavy neutron irradiation within the zirconium matrix,

are annealed early upon the first phase of the LOCA transient (i.e. first thermal ramp) and thus,

that the main effects of high bum-up should come from the hydrogen uptake. To assess this

hypothesis, specific thermal-mechanical tests have been performed on virgin, pre-hydrided and

irradiated cladding tubes. This confirmed that the effect of hydrogen uptake dominates over that

of irradiation on the thermal-mechanical response of the materials.

So, in a first part of the presentation, we will briefly summarize the main results obtained here

and, from the metallurgical point of view, we will illustrate the strong influence of hydrogen on

the decrease of the alpha-to-beta phase transformation temperatures of the zirconium alloys

studied.

On the other hand, studies have been performed on the post-quench mechanical behavior of

as-received and pre-hydrided cladding tubes after single-face oxidation at 1000-12000C and

quenching. In parallel with these mechanical tests, in-depth metallurgical investigations have

been developed 151, to be able to quantify the resultant phase thickness (that is, ZrO2, Alpha(O)

and Ex-Beta phase layers) and their specific chemical composition - especially their oxygen

content which is known to influence strongly the residual mechanical properties. Also,

fractograph analysis has been applied on failed samples to get a better knowledge of the failure

47
mechanism as a function of the materials and of the hydrogen concentration, for different

oxidation conditions.

So, in the second and major part of the presentation, we will focus on LOCA post-quenched

behavior of as-received and prehydrided Zy-4 and M5"1 cladding tubes for typical hydrogen

contents ranging from -200 up to -600 wt-ppm depending on the alloy. Ring compression,

impact, and bending tests at Room Temperature have been performed for different oxidation

conditions. The mechanical results will be presented and briefly discussed, taking into account

the metallurgical analysis (resultant phase morphology and thickness, chemical composition -

oxygen contents, failure mode,...).

References

[11 T. FORGERON, J.C. BRACHET, et al., a Experiment and modelling of advanced fuel rod

behaviour under LOCA conditions a:>,B phase transformation kinetics and EDGAR

methodology v, Zirconium in the Nuclear Industry: 12t. Int. Symposium, June 1998, Toronto,

Canada, ASTM STP 1354, (2000), pp. 256-278

[2] J.C. BRACHET, L. PORTIER, et al., 'Influence of hydrogen content on the aw*p phase

transformation temperatures and on the thermal-mechanical behavior of Zy-4, M4 (ZrSnFeV)

and M5ThA (ZrNbO) alloys during the first phase of LOCA transient," Zirconium in the Nuclear

Industry: 13uh. Int Symposium, June 10-14 2001, Annecy, France, ASTM STP 1423, (2002), pp.

673-701

[3] L. PORTIER, T. BREDEL, J.C. BRACHET, V. MAILLOT, J.P. MARDON, A. LESBROS,

'Influence of Long Service Exposures on the Thermal-Mechanical Behavior of Zy-4 and M5"1

Alloys in LOCA Conditions", Zirconium in the Nuclear Industry: 14m. Int. Symposium, June 13-17

2004, Stockholm, Sweden,, to be published in ASTM-STP

[4] J.-C. BRACHET, J. PELCHAT, D. HAMON, R. MAURY, P. JACQUES, J.-P. MARDON,

(( Mechanical behavior at Room Temperature and Metallurgical study of Low-Tin Zy-4 and M5T"

(Zr-NbO) alloys after oxidation at 11 OOC and quenching", Proceeding of TCM on "Fuel behavior

under transient and LOCA conditions' Sept. 10-14, 2001, organised by IAEA, Halden, Norway

(5] J.C. BRACHET et al., "Quantification of the a(O) and Prior-P phase fractions and their

oxygen contents in high temperature (HT) oxidised Zr alloys (Zy4, M5rm)", Proceeding of

"SEGFSM Topical Meeting on LOCA' 25-27t. of May 2004, ANL, Chicago, USA

48
Insights on PRA Quality from the SPAR Model Development Program

Patrick O'Relily, Michael Cheok

U.S. Nuclear Regulatory Commission

The purpose of the NRC's Standardized Plant Analysis Risk (SPAR) model development

program is to provide NRC staff with readily available and easy-to-use analytical tools for

performing plant-specific risk assessments. The Office of Research has completed SPAR

models for all 72 plant sites for internal initiating events at full power operation. In addition,

models for the low power/shutdown modes of operation, for external initiating events, and for

calculating large early release frequencies are being developed. SPAR models are used to

evaluate risk significance of inspection findings as part of the Reactor Oversight Process, to

evaluate risk associated with operating events as part of the Accident Sequence Precursor

program, to perform analyses in support of generictsafety issue resolution, to perform analyses

in support of the staff's risk-informed review of license amendments, and to independently verify performance indicators as part of the Mitigating Systems Performance Index (MSPI).

Modeling assumptions and methodology in the SPAR models are "standardized for consistency

and ability to understand differences in risk results. The initial Rev 3 internal-events SPAR

models has been completed and benchmarked against licensee PRAs. This paper would

provide the results of the comparison, discuss major factors that influence differences in risk

(CDF) results and discuss some insights and possible implications with regard to the goal of

achieving PRA quality.

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50
Human Reliability Analysis (HRA) Good Practices

Erasmia Lois', Alan Kolaczkowski,2 John Forester 3

'U.S. Nuclear Regulatory Commission, Washington, DC, USA

2Science Applications International Corp., Eugene, OR, USA

3Sandia National Laboratories, Albuquerque, NM, USA

The U.S. Nuclear Regulatory Commission is establishing "good practices" for human

reliability Analysis (HRA), documented in NUREG-1792, Good Practices for

Implementing Humam Reliability Analysis (HRA), Draft Report for Comment, July 2004.

NUREG-1792 -provides a reference guide to HRA practitioners for performing HRAs

(whether for the first time or when analyzing a change to current plant practices) and for

reviewing HRAs to assess the quality of an analysis.

The HRA good practices are developed as part of the NRC's initiatives to address

probabilistic risk assessment (PRA) quality issues and supports the implementation of

Regulatory Guide (RG) 1.200 entitled: OAn Approach for Determining the Technical

Adequacy of Probabilistic Risk Assessment Results For Risk-informed Activities." Its

elements are directly linked to RG 1.200 which reflects and endorses, with certain

clarifications and substitutions, the American Society of Mechanical Engineers (ASME)

Standard RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power

Plant Applications," and Revision A3 of the Nuclear Energy Institute (NEI) "Probabilistic

Risk (PRA) Peer Review Process Guidance," (NEI-00-02).

The HRA good practices reflect NRC and contractor experience from developing HRA

methods (e.g., THERP and ATHEANA), performing HRAs (e.g., NUREG-1150), and

reviewing HRAs, particularly the Individual Plant Examinations. 'Good Practices' means

those processes and individual analysis tasks and judgments that would be expected of

a HRA (considering current knowledge and state-of-the-art) in order for the HRA results

to sufficiently represent the anticipated operator performance when making risk-informed

decisions. The document is principally focused on the process for performing HRAs.

Therefore, are of generic nature, that is, are not tied to any specific methods or tools that

could be employed to perform an HRA. They are written in the context of a risk

assessment for commercial nuclear power plant operations occurring nominally at full

power, however, many are also applicable to low power and shutdown operations.

Similarly, are addressing internal initiating event analysis, however, should generally

also be appropriate to external initiating events.

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52
The Joint U.S. NRCIEPRI

Fire Risk Requantification Study

J. S. Hyslop

U.S. Nuclear Regulatory Commission

Fire risk analysis (FRA) methods have been used in the Individual Plant Examinations of

External Event (IPEEE) program to facilitate a nuclear power plant examination for

vulnerabilities. However, in order to make finer, more realistic decisions for riskinformed regulation, FRA methods need to be improved. Licensee applications and U.S.

Nuclear Regulatory Commission (NRC) review guidance with respect to many regulatory

activities such as the risk-informed, performance-based fire protection rulemaking

(endorsing National Fire Protection Association Standard 805)[2] will benefit from more

robust methods. In order to address the need for improved methods, the NRC Office of

Nuclear Regulatory Research (RES) and Electric Power Research Institute (EPRI) are

participating in a program to develop state-of-art methods in FRA.

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54
Phased Approach to Achieving an Appropriate Quality for PRAs

for Regulatory Decision Making

Gareth W. Parry

Division of Systems Safety and Analysis

Office of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission

The Commission, in Staff Requirements Memorandum (SRM) COMNJD-03-0002,

"Stabiliing the PRA Quality Expectations and Requirements", introduced the concept of

a four phase approach to achieving an appropriate quality for PRAs used in regulatory

decision making., This phased approach is needed because not all the guidance

documents defining PRA quality are available for all the risk contributors. Currently, the

staff guidance on PRA quality in RG 1.200 addresses only internal events at full power,

and includes a limited level 2 PRA (LERF) analysis. Guidance for external initiating

events, internal fires, and low power and shutdown modes of operation will follow

development of PRA standards for these risk contributors. The approach lays out a

path, in a phased manner, for how risk-informed applications can be implemented while

the guidance documents needed to define PRA quality for all risk contributors are

developed. In the SRM, the Commission directed the staff to develop an action plan that

would define a practical strategy for its implementation. In addition, the SRM directed

the staff to discuss the resolution of technical issues, such as model uncertainty,

treatment of seismic and other external events, and human performance issues.

An implementation plan has been developed jointly by the Offices of Nuclear Reactor

Regulation (NRR) and Nuclear Regulatory Research (RES). In this plan, PRA quality is

measured, as in Regulatory Guide (RG) 1.174 and RG 1.200, in terms of its

appropriateness with respect to scope in terms of risk contributors as discussed above,

level of detail, and technical adequacy. Inherent in this definition is that a PRA of

sufficient quality to support an application need only have the scope and level of detail

sufficient to support that application, but it must alvays be technically adequate.

The plan covers the first three phases defined in the SRM, and identifies the activities

required to support implementation. As directed by the SRM, the feasibility of the fourth

phase will be assessed following achievement of Phase 3. The phases are achieved for

specific risk-informed activities when guidance documents are available to support those

activities, and in particular to address the issue of the quality of PRA necessary to

support the activities.

Phase 1 represents the current situation, where guidance on PRA quality is general, and

staff review of the base PRA supporting the activity is performed on a case-by-case

basis. Phase 2 takes advantage of the work that has been performed to develop PRA

standards. Phase 2 occurs when there are PRA standards and the associated

regulatory guides in place to address those PRA scope items that are significant to the

55
decision. To be in Phase 2 for an application, the licensee's submittal is expected to be

in conformance with the published standards. The PRA standards are being developed

on different schedules. As a result, the risk-informed activities will transition to Phase 2

on different schedules according to which scope items are significant to the decision.

Phase 3 provides a regulatory framework for the development of a PRA that will be of

sufficient quality to support all current and anticipated applications.

In implementing the phased approach, the plan calls for specific application types to be

defined, and the necessary guidance documents identified. Additional guidance

documents will be developed on a schedule that is a function of when the standards are

developed. The implementation schedule has a built-in grace period to allow licensees

to implement the new guidance. To implement the phased approach, a process will be

developed for prioritizing and scheduling submittal reviews. This prioritization process is

necessary to balance the need to use staff resources effectively and efficiently and the

need to provide incentives for licensees to develop more complete PRA models. The

staff plans on working closely with industry in the development of the guidance

documents, and will develop the necessary standards not developed by a Standards

Developing Organization (e.g. ASME, ANS).ln addition to providing a strategy for the

implementation of the phased approach, the plan addresses the identification and

resolution of technical issues (e.g., model uncertainty, the treatment of seismic and other

external events, and human performance).

56
Guidance on the Treatment of Uncertainties and Alternative Approaches

Used in PRAs

Mary Drouln

Probabilistic Risk Analysis Branch

Division of Risk Analysis and Applications

Office of Nuclear Reactor Research

The NRC is extensively using information from probabilistic risk assessments (PRAs) in

its regulatory decision making. To streamline staff review of licensee applications using

risk insights, professional societies and the industry undertook the following initiatives to

establish consensus standards and guidance on the use of PRA in regulatory decision

making:

The American Society of Mechanical Engineers (ASME) has developed a

standard for a Level 1 analyses (i.e., estimation of core damage frequency

(CDF)) and a limited Level 2 analysis (i.e., estimation of large early release

(LERF)) covering internal events (transients, loss of coolant accidents, and

internal flood) at full power.

* The Nuclear Energy Institute (NEI) has developed a "PSA Peer Review

Guidance," (NEI-00-02) covering internal events at full power-Level 1 and

simplified Level 2.

* The American Nuclear Society (ANS) has developed PRA standards for external

hazards (December 2003) and is developing PRA standards for: - low power and shutdown with a tentative publication date of December 2004 - internal fires with a tentative publication date of December 2004

It is expected that licensees will use the PRA standards and industry guidance to help

demonstrate and document the adequacy of their PRAs for a variety of risk-informed

regulatory applications. Therefore, the staff should document its position on the

adequacy of the standards and industry guidance to support regulatory applications.

Such documentation will indicate in which areas staff review can be minimized and

where additional review may be expected. To accomplish this, the staff has developed Regulatory Guide 1.200 to provide an approach for assessing the adequacy of PRA

results used in support of regulatory applications and an accompanying Standard

Review Plan (SRP) chapter.

A major challenge in development of PRA standards is identification and resolution of

the technical issues. One important issue is that of model uncertainty. For example,

there are elements of the PRA where there is uncertainty regarding the appropriate

model to use (e.g., human reliability analysis). When a consensus cannot be reached

as to whether there is a clearly preferred approach, a decision must be made taking into

account the impact of adopting different models. Another challenge in the development

57
of PRA standards is that the analyst may use alternative approaches than those

provided in the standard to account for scope items or technical requirements that are

not addressed in its PRA.

Guidance is not available on acceptable approaches of how to address, in detail,

uncertainties in the decision making process, nor is their guidance on the acceptability of

alternative approaches. The staff has initiated efforts to develop guidance on:

* Acceptable approaches to supplement a PRA that is not of a full scope, or has

deficiencies in some elements which includes the appropriate use of bounding

analyses, screening methods, or qualitative approaches

* Identification and performance of sensitivity studies

* How to use the results from the "uncertainty analyses" in the decision making

process, including the role and definition of defense-in-depth.

58
More Robust Research Plan for Radiation Protection

Stephanie Bush-Goddard, Ph.D.

Radiation Protection, Environmental Risk,

and Waste Management Branch

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

The Nuclear Regulatory Commission's (NRC's) standards for protection against ionizing

radiation, 10CFR Part 20, continue to serve as a coherent and logical framework for the

variety of radiation sources and exposures under NRC jurisdiction. However, in 2003,

the Commission directed the NRC staff to develop a more robust materials research

program in the Office of Research that would take into account developments in

environmental and individual radiation protection may have significant implications for

NRC's regulatory framework. In addition, revisions in dosimetry results stemming from

the Hiroshima/Nagasaki assessments being undertaken by the Radiation Effects

Research Foundation, and to be addressed in the upcoming report by the National

Academy of Sciences Committee on Biological Effects of Ionizing Radiation, will warrant

an assessment of the potential impact on our current regulations

As a result, a research plan was developed and provided to the Commission in

February, 2004, outlining a research program plan addressing radiation protection and

health effects issues. The Commission paper containing this research plan can be found

in ADAMS under accession number m1040360736. This paper will discuss the key

components and goals of this research program.

To summarize, the goals of the research plan are as follows:

Goal No. 1: Maintain and Improve NRC's Knowledge of Radiation Health Effects

Evaluate data and information on health effects attributed to radiation exposure in order

to serve as a technical information resource. Ensure that NRC has the most up-to-date

information in order to make realistic estimates of radiation health effects.

Goal No. 2: Support Development of Radiation Protection Standards and

Implementation

Develop and enhance methodologies for radiation detection, measurement, monitoring,

and dose assessment to provide technical information with greater accuracy and less

uncertainty in order to enhance the effectiveness and efficiency of regulatory decisions.

Promote consistency in the development of radiation protection standards by evaluating

the impact and relevance of new technical and scientific developments on the NRC

regulatory program and actively participating in the development of radiation protection

recommendations to ensure that they, are based on sound and logical scientific

principles.

59
Goal No. 3: Support Radiation Protection Rationales and Technical Bases

Provide models that support realistic decision making in radiation protection. The NRC

needs up-to-date information to accurately assess the impact on public health and safety

in developing new regulations and reviewing licensed activities.

Goal No. 4: Develop Technical Basis for Risk-Informing Materials Applications

through Analysis of Operational Experience

Develop information to support regulatory decision-making on risks associated with the

use of radioactive materials. By accurately assessing the risks associated with expected

or accidental exposures to radioactive sources, the staff will be better able to risk-inform

NRC regulations on byproduct and source material.

60
2005 Recommendations of the ICRP

E. V. Holahan, Ph.D.

Radiation Protection, Environmental Risk,

and Waste Management Branch

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

The Intemational Commission on Radiological Protection (ICRP) was established to

advance for the public benefit the science of radiological protection, in particular by

providing recommendations and guidance on all aspects of radiation protection. The

aim of the recommendations is to provide an appropriate standard of protection for

mankind from sources of ionizing radiation, without unduly limiting beneficial practices

that give rise to exposure to radiation. In preparing its recommendations, the ICRP

considers the fundamental principles and quantitative bases on which appropriate

radiation protection measures can be established, while leaving to various national

protection bodies the responsibility of formulating the specific advice, codes of practice,

or regulations that are best suited to the needs of their individual countries.

The ICRP regularly examines the status of its recommendations and reviews the

increasing knowledge of the effects of exposure to ionizing radiation to decide whether

new recommendations are needed. Since the publication of the last set of

recommendations in 1990, the ICRP has published ten reports that provide additional

guidance for the control of radiation exposures. The draft 2005 Recommendations are

intended to consolidate into a single document and within a unified scheme a number of

recommendations that have been made in individual publications since the 1990

Recommendations, Publication 60, was issued.

The full text of the ICRP 2005 Recommendations is available for review on the ICRP

website. The Main Commission has stated that the 2005 Recommendations are not

intended to change the fundamental basis for radiological protection nor are they

intended to prompt significant changes other than updating the Interagency and

European Basic Safety Standards. The draft recommendations maintain the public and

occupational dose limits originally recommended in 1990, update tissue and radiation

weighting factors, recommend dose constraints from single sources, emphasize that

patient dose should be commensurate with the clinical benefit, and include a policy on

radiological protection of non-human species.

A presentation of the 2005 ICRP Recommendations will highlight the evolution that has

occurred relative to recommendations previously issued by ICRP in Publications 26

(1977) and 60 (1990) and the Standards for Radiation Protection as contained in 10

CFR Part 20.

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62
Challenges in Radiation Protection and Homeland Security

Carl Paperiello

Director, Office of Nuclear Regulatory Research

US Nuclear Regulatory Commission

Radiation protection issues play a large variety of roles in homeland security. Most

recognize a variety of activities in responding to radiological dispersal devices(RDD's)

and similar malevolent events. These activities correspond in many ways to existing

radiological emergency planning and response responsibilities. Additional radiation

protection needs have emerged in the prevention of terrorist and malevolent activities.

These include radioactive materials security, border protection, and search capabilities

for radioisotopes including special nuclear material (SNM). The use of ionizing radiation

and radioactive sources in homeland security activities raises new issues with respect to

public exposure, a significant increase in the number of radiation workers and their

training, and the security of radioactive material in new devices. The role of NRC

research activities in this area is still uncertain as much of the research appears to be

the responsibility of organizations other than the NRC.

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64
TRACE: TRACIRELAP Advanced Computation Engine

J. L. Staudenmeler

Office of Nuclear Regulatory Research

Safety Margins and Systems Analysis Branch

Overview

The TRACE project was started so that the NRC could replace their existing suite of

reactor system safety analysis codes (RELAP5 , TRAC-PWR, TRAC-BWR, and

RAMONA) by a single code in order to reduce code maintenance and user training

costs. The previous code architectures were designed in the 1970's and were subject to

constraints imposed by both computer hardware and the FORTRAN 77 language.

TRACE was built with a modem code architecture that is easy to maintain and extend

with new models to address future safety problems. TRACE is also integrated with the

Symbolic Nuclear Analysis Package (SNAP) graphical NRC analysis environment.

TRACE can also be run in a coarse grained parallel processing mode and has a built in

mechanism to communicate with other computer codes.

Code Capabilities

TRACE is a two-fluid compressible flow code with I 2, and 3 dimensional flow

geometry and is coupled to the PARCS 3 dimensional reactor kinetics code. The code

also has the capability to model heat structures and control systems that interact with

the fluid solution. TRACE will be able to perform any type of reactor analyses that were

performed by the predecessor codes (RELAP5 , TRAC-PWR, TRAC-BWR, and

RAMONA) and has component models and mesh connectivity that allow a full reactor

and containment system to be easily modeled. The current models and correlations are

the TRAC-PF1/MOD2 models. The NRC recognizes that users have a large investment

in previously developed input decks. TRACE will be able to run input models from

RELAP5, TRAC-PWR, and TRAC-BWR with minimal or no modification. This will allow

for a gradual migration to TRACE through the transition period of the next few years

during which both RELAP5 and TRACE are still supported. TRACE also supports

coarse grained parallel processing in which the user can decide how to divide a model

so that different pieces run on different processors.

User Interface

SNAP was developed to be an analysis environment for NRC safety analysis codes that

will aid the user and make them more productive. SNAP currently supports TRACE and

RELAP5. It will support CONTAIN and MELCOR soon. A user can develop and modify

input models and then run and analyze the results within SNAP. SNAP also has the

interactive calculation and mask animation capabilities of the Nuclear Plant Analyzer

(NPA). SNAP will soon have the capability to automate parametric studies of a given

65
input parameter over a range specified by the user. In the future SNAP will also allow

custom user defined macro components that are comprised of multiple TRACE

components.

Code Architecture

The TRACE code architecture was designed to take advantage of the capabilities of

modem computer hardware and compilers. Features of Fortran 95 such as dynamic

memory allocation and derived data types make the large container array and the

associated error prone memory management techniques used in TRAC and RELAP

obsolete. The large memory capacity of current desktop computers makes the use of

memory saving techniques of bit packing and temporary memory overlays unnecessary.

The TRACE code architecture was designed to be portable, easily maintained and

extended. Gains in developer productivity have already been realized while developing

a new BWR channel component that has the ability to model current BWR fuel bundle

geometries. TRACE also has the built in capability to communicate with other codes or

other TRACE runs locally or over a network though the Exterior Communications

Interface ( ECI ). TRACE is currently being developed and run on multiple operating

systems including Windows NT/2000/XP, Linux, and Mac OSX.

Exterior Communications Interface ( ECI )

The ECI is a request driven interface that allows TRACE to communicate with any code

that implements the ECI without any modifications to TRACE. The ECI uses a standard

BSD sockets interface at the operating system level. The external code can be written in

any computer language that can call BSD sockets. The ECI has allowed TRACE to be

easily coupled to codes such as SNAP, CONTAIN, and MATLAB. The interface should

allow TRACE to be coupled to CFD or other special purpose codes in the future.

Current Status and Future Development

TRACE has undergone limited assessment and is being used for some applications

within the NRC. TRACE is currently undergoing a more complete code assessment to

identify deficiencies in its current physical models and correlations. As deficiencies are

identified new (or existing physical models from other predecessor codes ) will replace

the original TRAC-PWR models. New reflood and condensation models are currently

being implemented in the code. Work is also underway to add a droplet field to the code

and to make the numerical methods more implicit to improve code robustness and

runtime performance. The NRC will also couple its other reactor analysis codes such as

FRAPCON, FRAPTRAN, and MELCOR to TRACE in the future. TRACE should provide

a robust and extensible platform for safety analyses well into the future. The long range

objective of TRACE development is to produce a tool capable of simulating both large

and small break loss of coolant accidents in conventional and advanced nuclear power

plants.

66
MELCOR Development and Assessment

R.O. Gauntt

Sandia National Laboratories

MELCOR is a fully integrated, engineering-level computer code that models the

progression of severe accidents in light water reactor nuclear power plants. MELCOR is

being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory

Commission as a second-generation plant risk assessment tool and the successor to the

Source Term Code Package. A broad spectrum of severe accident phenomena in both

boiling and pressurized water reactors is treated in MELCOR in a unified framework.

These include thermal-hydraulic response in the reactor coolant system, reactor cavity,

containment, and confinement buildings; core heatup, degradation, and relocation; coreconcrete attack; hydrogen production, transport, and combustion; fission product release

and transport behavior.

Initially, the MELCOR code was envisioned as being predominantly parametric with

respect to modeling complicated physical processes (in the interest of quick code

execution time and a general lack of understanding of reactor accident physics).

However, over the years as phenomenological uncertainties have been reduced and

user expectations and demands from MELCOR have increased, the models

implemented into MELCOR have become increasingly best estimate in nature. The

increased speed (and decreased cost) of modem computers (including PCs) has eased

many of the perceived constraints on MELCOR code development. Today, most

MELCOR models are mechanistic, with capabilities approaching those of the most

detailed codes of a few years ago.

The latest version, MELCOR 1.8.6 is scheduled for release in early 2005 and will include

many new and improved phenomenological models, especially in the area of core melt

progression. Included in the next code release will be significant improvements in the

treatment of in-core and lower head molten pool and crust modeling, allowing for

analysis of a TMI-like core degradation where a core-region convecting molten pool can

be contained within confining perimeter crust and side wall release of the molten pool

upon failure of the crust can be predicted. In the lower head region, improved molten

pool models can account for internal pool natural convection and phase partitioning of

the pool contents into disctinct metallic and ceramic regions. Other code modeling

improvements include new models for release of Ag-In-Cd aerosol, for degradation and

oxidation of PWR-type boron carbide control rods, for improved treatment of

reflood/quench behavior, and many code extensions and enhancements. Additionally,

significant advances in modeling parameters and properties have been gained based on

insights derived from the international research efforts such as Phebus (IRSN),

QUENCH (FzK) and MASCA (lbrae), with the aim of implementing best estimate

treatment of severe accident phenomena into the MELCOR code.

67
Today MELCOR is finding numerous applications both within the USNRC in both the

support of licensing activities as well as best estimate source term characterization for

severe accidents. Examples include the support of AP-1000 design certification, the

upcoming ACR-700 design certification request by AECL, and new best estimate

analyses of accidents in spent fuel pools. Emphasized in contemporary MELCOR

analyses is a best estimate treatment of phenomena combined with true uncertainty

analysis, as opposed to bounding deterministic analyses and sensitivity studies in order

to provide risk-informed assessments of safety issues. In support of this philosophy, new

capabilities have been developed for the MELCOR code suite to perform Monte Carlo

uncertainty studies with MELCOR to supplement traditional deterministic applications of

the MELCOR code.

68
Fuel Behavior Modeling In Accident Analysis

Harold Scott

U.S. Nuclear Regulatory Commission

Fuel design limits are used to prevent fuel damage during normal operation and

anticipated operational occurrences. Accident analysis determines the extent of fuel

damage (to ensure coolable geometry) during hypothetical transients and accidents.

From the standpoint of fuel behavior, these transients and accidents may be broadly

considered as either undercooling or overpower events. Regulatory limits and technical

specifications are used to ensure that events of either type do not progress into core

melt scenarios

Three postulated accident types are under investigation by NRC to determine the

behavior of high-bumup fuel. For undercooling events, the design-basis loss-of-coolant

accident (LOCA) is considered. To represent the broad class of overpower events, the

PWR control-rod-ejection accident, a reactivity initiated accident, is considered. One

beyond-design-basis event is considered in which both overpower and undercooling

occur, i.e., powr oscillations in a BWR when failure to scram is assumed.

Loss-of-coolant accident (LOCA)

A number of fuel-related phenomena, such as cladding oxidation and hydrogen uptake

during normal operation, and cladding ballooning and rupture during the transient, may

affect core coolability during small-break and large-break LOCAs. Further, fuel pellet

cracking during the bumup process leads to the possibility of fuel relocation into

ballooned sections of fuel rods, thus altering the axial distribution of decay heat.

Ballooning size and cladding oxidation may also be affected by bumup. Substantial new

data are being generated to address these effects (see Billone et. al. and Grismanovs

et. al. in these Proceedings), but analysis is needed to estimate the impact of these

effects on core behavior during a LOCA, such as coolant flow around ballooned regions in the fuel bundle. Therefore, high-bumup needs to be considered for both damage

criteria (10 CFR 50.46) and evaluation models (10 CFR 50 Appendix K). The need for

detailed fuel behavior modeling in system thermal hydraulic codes will be discussed.

Control-rod-ejection accidents In PWRs

As discussed previously, fuel rod cladding accumulates hydrogen during normal

operation and this may lead to loss of ductility. As a consequence, high-bumup fuel

rods may not withstand the rapid thermal expansion of the fuel pellets that occurs during

a power pulse. Since NRC's damage criteria in Regulatory Guide 1.77 was derived for

fresh fuel, these criteria need to be revised. Previous and more recent data from

several different test reactors was analyzed to obtain a realistic cladding failure

69
threshold correlation at high bumup. This work was completed last spring and results

were issued as a Research Information Letter, No. 0401, on March 31, 2004.

Power oscillations associated with anticipated transients without scram (ATWS) in

BWRs

BWR ATWS scenarios in which power oscillations might occur were reviewed several

years ago (NUREG/CR-6743). Cladding temperature is expected to rise and eventually

the cladding may dry out without subsequent rewetting. To analyze this accident, the

FRAPTRAN single-rod code was coupled with the Finnish GENFLO hot-channel

hydraulics code. Calculations with these coupled codes were described at the 2001

NSRC (NUREG/CP-0176, page 381). Aspects of further analysis and perceived data

needs will be discussed.

70
The U.S. NRC Spatial Kinetics Code PARCS

Thomas J. Downar

Purdue University West Lafayette, Indiana

PARCS is a three-dimensional reactor core neutronics simulator that solves the steady-state

and time-dependent muttigroup neutron diffusion or SP3 transport equations to predict the

steady-state and dynamic response of the reactor to reactivity perturbations such as control rod

movements, boron concentration or changes in the temperature/fluid conditions in the reactor

core. The code is applicable to both PWR and BWR cores loaded with either rectangular or

hexagonal fuel assemblies. The PARCS code was chosen by the U.S. Nuclear Regulatory

Commission as its best estimate core neutronics code [Downar, 19981, and PARCS is coupled

directly to the U.S. NRC thermal-hydraulics systems codes TRACE and RELAP5. As of version

2.6, PARCS is coupled to TRACE as a static link library, whereas PARCS is coupled to

RELAP5 using Parallel Virtual Machine (PVM) message passing interface. A depletion

capability is also available in PARCS which includes a cross section interface capability to

lattice physics codes such as HELlOS or SCALE/TRITON.

PARCS solves the multigroup time dependent diffusion and SP3 equations with either assembly

or pin-by-pin spatial discretization. The nodal options include both the analytic nodal method

(ANM) and the nodal expansion method (NEM). The pin-by-pin finite difference option (FMFD)

accepts pin homogenized cross sections in an arbitrary number of energy groups. The

multigroup SP3 transport option is available in both the NEM and FMFD kernels. The various

geometry types and solution kernels available in PARCS are summarized in Table 1.

Table I Solution Kernels Available In PARCS

Geometry Kernel Solution Energy Angle Comment

Type Name Method Treatment Treatment Comments

CMFD FD 2G Diffusion

Cartesian ANM nodal 2G Diffusion

3D FMFD FD MG SP3 _

NEMMG nodal MG SP _

Hexagonal CMFD FD 2G Diffusion

3D TPEN nodal MG Diffusion

Cylindrical CMFD FD 2G Diffusion To be released in v2 7

3D FMFD FD MG Diffusion/ To be released in v2.7 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _S P 3 _ _ _ _ _ _ _ _ _

The problem size which can be treated in PARCS is restricted by available memory. The finest

level of group constant homogenization is at the pin-cell level, but there is no restriction on the

number of energy groups for the NEM and FMFD options. Computational times are minimized

by the use of advanced numerical methods such as a BILU3D preconditioned Krylov solver

(BICGSTAB) to solve the coarse mesh finite difference problem in PARCS. Execution times

will vary depending on the solution kernel chosen, however, 2-group nodal diffusion solutions

for typical LWR transients (e.g. control rod ejection) are on the order of minutes on a current

generation PC. Each new version of PARCS is tested on a series of Operating Systems (e.g.

Windows, LINUX, Solaris) and Fortran compilers (e.g. DVF 6.6, Intel 7, NAGWare 4.2, Sun F90

71
6.2). A comprehensive assessment of the PARCS code for a wide range of reactor

benchmarks has been reported previously, to include fast reactor transients such as control rod

ejections [Joo, 1996], MOX fuel applications [Kozlowski, 2002], and coupled code transients for

both PWRs [Kozlowski, 20001 and BWRs [Lee, 2002]. The most recent focus of BWR

assessment for TRACE/PARCS is the OECD Ringhalls Stability Benchmark.

A fuel depletion capability is available in PARCS, which employs a macroscopic depletion

method [Xu, 2001]. During each "bumup step" the power in each fuel node is used to time

advance the bumup of the node, and the assembly cross sections derivatives and other

neutronics parameters are computed at the new bumup for the corresponding fuel composition.

Altemately, a given core "bumup" distribution can be specified by the user and the depletion

module in PARCS will generate the fuel assembly neutronics data at the specified bumup. Thus

the code provides the functionality to meet the two primary fuel cycle problems of the analyst, to

deplete a core to find a typical "equilibrium" condition or to accept the bumup distribution for a

specific core and to find the corresponding cross sections and steady-state condition. The fuel

macroscopic cross sections and other nuclear data (e.g. kinetics parameters, detector

information, discontinuity factors, form functions, Xe/Sm data, etc) at the appropriate fuel

bumup conditions are prepared using lattice physics code such as HELIOS, TRITON, CASMO,

etc. Because the output format of the each lattice code is unique, the cross section processing

program, GENPMAXS, was written to process the output of the lattice codes and prepare a

common cross section format file, PMAXS, which is read by PARCS. A separate manual is

available which describes in detail the cross section processing treatment in GENPMAIXS.

Near term development work on PARCS will address methods upgrades for analysis of

advanced core designs such as the ACR-700. Other near term work will include the

incorporation of PARCS into the TRACE/SNAP interface software. Specific questions can be

addressed to the PARCS users group which is available via the PARCS website which also

provides more detailed information on the PARCS code:

www.enpineerina. Durdue.edu/larcs

[Downar, 1998] PARCS: Purdue Advanced Reactor Core Simulator, User Manual, PU/NE-98-26, NRCV1.00, T. Downar, H. Joo, D. Barber, 1998.

[Joo, 1996] Methods and Performance of the Neutron Kinetics Code PARCS, H. Joo, D. Barber, T.

Downar, PHYSOR-96, International Conference on the Physics of Reactors, Mito, Japan, September,

1996.

[Kozlowski, 2002] Benchmarking of the Multigroup, Fine Mesh, SP3 Methods in PARCS with the VENUS2 MOX Critical Experiments, T. Kozlowski, C. Lee, T. Downar, School of Nuclear Engineering Purdue

University, PHYSOR-2004, Seoul Korea, October, 2002.

[Xu, 2001] Fuel Cycle Depletion Capability for the U.S. NRC Neutornics Code PARCS, Y. Xu, T. Kim, D.

Tinkler, T. Downar, M&C 2001 Salt Lake City, Utah, USA, September 2001

[Lee, 2002] Analysis of the OECD Peach Bottom Turbine Trip 2 Transient Benchmark with Coupled

Neutronic and Thermal-Hydraulics Code TRAC-M/PARCS, D. J. Lee, T. Downar, T. Ulses, K. lvanov, B.

Akdeniz, PHYSOR-2004, Seoul Korea, October, 2002.

[Kozlowski, 2000] Analysis of the OECD MSLB Benchmark with the Coupled Neutronic and ThermalHydraulics Code RELAP5/PARCS, T. Kozlowski, R. Miller, T. Downar, D. Ebert, PHYSOR-2000,

Pittsburgh, 2000.

72
Use of Computer Codes in the Regulatory Review Process

Ralph R. Landry

U.S. Nuclear Regulatory Commission

Traditional regulatory reviews of computer codes have focused on the documentation of the code

and results of the application of the code. Codes developed for generic application to numerous

nuclear power plants and plants of differing designs have necessitated application calculations that

are non-plant specific. That is, the calculated plant response is for a fictitious plant that is assumed

to be "representative" of plants to which the code will be applied for licensing purposes. The

ensuing code review must also make broad statements about the code and its applicability and the

manner in which it behaves.

The Office of Nuclear Reactor Regulation has, in recent years, further exercised its regulatory

authority in so far as the applicants for approval of thermal hydraulic systems codes have been

asked to submit the code itself with a sample input deck in addition to the usual documentation and

analytic results. To remove computer hardware, operating system, and compiler effects from use

of the individual code, the NRC has acquired numerous computer workstations so that the codes

can be run on the same platform as used by their respective developers.

Review and approval of codes such as RETRAN-3D, S-RELAP5, and TRACG was based in part

on extensive code execution by the staff of the NRC.

A significant part of the analytical process followed since the development of the Code Scaling,

Applicability and Uncertainty evaluation methodology is development of a Phenomena Identification

and Ranking Table, or PIRT. The PIRT and the process leading to is development provides

significant insight into those processes and phenomena having the most influence on the

progression and results of a nuclear plant accident ortransient. Understanding which phenomena,

and ultimately which models, are key to an event's progression, permits code developers to focus

their efforts where they would expect to need to have the greatest accuracy in modeling. From the

perspective of the regulator, the PIRT permits the focusing of review activities on those areas

having the most importance thereby using available resources more efficiently.

Code utilization has taken several forms during the review process. Receipt of the subject code

in both a precompiled executable and source form permits execution of the code precisely as would

be done by the applicant. Of course, results of a calculation are expected to match exactly with

the results supplied with the code documentation. In addition, availability of the source code

permits the staff to make modifications to specific models and correlations, compile a new

executable code and perform test analyses. Through these test cases, the staff is able to further

understand the level of importance of the various models and correlations within the code, as well

as confirm the level of importance identified through the PIRT process.

73
As a further step in the review process, the staff also performs parallel calculations of the specific

applicant code transient using the NRC's computer codes. In several reviews it has been obvious

that there are benefits to both codes in this process. Reviews of applicant codes in the area of the

reactor kinetic models, by comparison with the diffusion code NESTLE and a transport code, found

that while some codes produced virtually identical results, others produced dramatically different

results. That work was done prior to the availability of PARCS.

Parallel applicant/NRC thermal hydraulic calculations have been very successful in that they have

identified both code to code shortcomings, as well as errors in input modeling that might have gone

otherwise undetected. Refining the modeling and upgrading the codes' capabilities ultimately

resulted in more accurate and more reliable nuclear plant calculations. From the perspective of

the NRC, we have an improved independent tool, while from the perspective of the regulator, we

have more confidence in the results from an approved code that will be submitted for future review.

Including execution of the applicant's computer code as well as the NRC's independent computer

codes in the code review and approval process has enhanced our understanding of the applicants'

codes as well as. improved our confidence in the results of their use of the codes. An additional

benefit has been improvement of our own codes by identifying those areas were models were in

need of greater accuracy.

74
Recent Developments in Multidimensional Modeling of

Reactor Thermal Hydraulics Using the NPHASE Code

M. Z. Podowski, S. P. Antal, B.Z. Wierzbicki and P. Tlwan

Center for Multiphase Research, Rensselaer Polytechnic Institute

Troy, New York 12080-3590

Recent progress in both the understanding of the basic physics of two-phase flows and the

computational capabilities using CFD methods clearly shows the great potential of fordetailedthree

dimensional mechanistically-based numerical simulations of two-phase flows in boiling channels

and nuclear reactor systems. These simulations can give significant insight into the hydraulic

characteristics of next generation reactor systems. It is not surprising that most advancements to

date are related to dispersed flows. Indeed, several closure laws based on ensemble-averaging

the instantaneous equations governing the interacting phases have already been developed for

both bubbly and particle flows and successfully applied to study various multiphase/multicomponent

flow phenomena. In order to extend the existing predictive capabilities to other flow regimes such

as slug flow, chum flow, and annular flow, where a phenomenological approach has been mainly

used until now, new closure laws must be developed and incorporated in an appropriate

multidimensional computational framework.

The purpose of this presentation is to summarize the results of recent work on the development

of mechanistic models of two- and multiphase flow for the NPHASE code. Specifically, closure

laws consistent with the multifield modeling and computational concepts used by NPHASE are

discussed and selected results of NPHASE simulations are shown. Furthermore, new results are

presented, concerning coupling a new Level-Set-based interface tracking computational model with

the NPHASE solver, and the use of the combined code to simulate the effect of bubble deformation

and channel geometry on gas-liquid interaction.

The accuracy of all CFD predictions depends strongly on the quality of closure models for local twophase flow phenomena mass and momentum transfer. Typically, CFD codes use simplified

closure laws, actually applicable only to dispersed bubbly/particle flows. Recently, several generic

closure laws (defined by specific mathematical operators and formulae) have been developed for,

and/or implemented into the solution algorithm of, the NPHASE code, that are applicable to various

flow regimes, including bubbly, slug, chum turbulent, and annular flows. Also, work was has been

performed on the development of mechanistic local models of both general and flow-regimespecific local interfacial phenomena, and several new modeling and computational issues have

been resolved that are critical for the consistency of the overall CFD model. A key effort here has

been on the implementation of the new models with the underlying numerical treatment, to

75
assure a tight coupling between the numerics and models that is needed to improve the robustness

and convergence characteristics of an advanced multiphase CFD code. In addition, a fully coupled

mass/momentum solver has been developed for the NPHASE code. This advanced numerical

method allows the solution of both the mass conservation and momentum conservation equations

simultaneously.

The ability to predict the shape of the gas/liquid/solid interfaces is important for various multiphase

flow and heat transfer applications. Specific issues of interest to nuclear reactor thermalhydraulics, include the evolution of the shape of bubbles attached to solid surfaces during

nucleation, bubble-surface interactions in complex geometries, etc. Additional problems, making

the overall task even more complicated, are associated with the effect of material properties that

may be significantly altered by the addition of minute amounts of impurities, such as surfactants

or nanoparticles. An important practical issue is concerned with the development of an innovative

approach to model time-dependent shape of gas/liquid interfaces in the presence of solid walls.

The present approach combines a modified level-set method with the NPHASE CFD solver. The

coupled numerical solver that is being currently developed will be capable of simulating the

evolution of gas/liquid interfaces in two-phase flows for a variety of geometries and flow conditions,

from individual bubbles to free surfaces (stratified flows).

As a result of the present work, a consistent first major step has been made toward the

development and encoding the models of individual flow regimes into a closed-form model allowing

to predict varying flow conditions in heated reactor channels using multidimensional computer

simulations. Interestingly, the present work also provides new insight into the modeling issues

encountered in reactor system codes.

76
Overview of Approach to Reevaluate Station Blackout Risk

Dale M. Rasmuson

Operating Experience Risk analysis Branch

Division of Risk Analysis and Applications

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

301-415-7571

dmr&nrc.aov

NUREG-1032 presented the risk basis for the Station Blackout (SBO) Rule. It provided loss of

offsite power (LOOP) frequencies, probability estimates for not recovering offsite powerto a safety

bus, and core damage frequency estimates of LOOP and SBO accident scenarios. NUREG-1032

used LOOP event information from 1968 through 1985. NUREG/CR-5496 updated the LOOP

frequencies and nonrecovery probabilities using LOOP event information from 1980 through 1996.

On August 14, 2003, large portions of the Midwest and Northeast United States and Ontario,

Canada, experienced an electric power blackout. The outage affected an estimated 50 million

people and 61,800 MW of electric load. Nine U.S. nuclear power plants experienced reactor trips

as a consequence of the power outage. Nine plants used their emergency diesel generators to

power their safety-related buses during the power outage. Offsite power was restored to at least

one safety bus after a period of time ranging from about two hours to about 14 hours, with an

average time of about seven hours. These LOOP events were longer in duration than previously

considered under the SBO rule.

As part NRC's grid action plan, RES has or is performing the following tasks:

1. Provide short-term risk insights of the impact on nuclear power plants based on the

observed changes in grid reliability to allow the agency to evaluate if action is needed

before the Summer of 2004.

2. Provide preliminary accident sequence precursor analyses of each affected plant for

technical review.

3. Evaluate SBO implications. Using data from recent LOOP events, update the SBO LOOP

frequency and duration.

4. Evaluate SBO risk. Calculate SBO risk (core damage frequency) with updated Standardized

Plant Analysis Risk models for a spectrum of plants.

This paper will present a summary of the ASP risk analyses completed to date and an overview of

the methods being used to update LOOP frequencies and nonrecovery probabilities. It will also

provide an overview of the steps to be used in the reevaluation of core damage frequency

associated with LOOP and SBO.

77
References

1. U.S. Nuclear Regulatory Commission, Evaluation of Station Blackout Accidents at Nuclear

Power Plants, NUREG-1 032, June 1988.

2. U.S. Nuclear Regulatory Commission, Evaluation of Loss of Offsite Power Events at

Nuclear Power Plants: 1980- 1996, NUREG/CR-5496, November 1998.

78
Generic Issues Program - Process and Screening Analysis

Harold Vandermolen

U.S. Nuclear Regulatory Commission

The Generic Issues Program first began formally in response to a Commission directive in October

of 1976. In 1983, it became one of the first programs to make successful use of probabilistic risk

information to aid in agency decisionmaking. In the 20 years since the program became

quantitative, approximately 840 issues have been processed. Although there is far less reactor

licensing activity than in the 1 970s, new issues continue to be identified from research programs

and operational experience, and the generic issue program remains very active.

Generic issues tend to involve rather difficult questions of safety and regulation, and an efficient

and effective means of addressing these issues is very important if regulatory effectiveness is to

be achieved. If an issue proves to pose a genuine, significant safety question, then swift, effective,

enforceable, and cost-effective action needs to be taken to eliminate an accident before it happens.

Conversely, if an issue is of little safety significance, the issue should be dismissed in an

expeditious manner, avoiding unnecessary expenditure of resources and regulatory uncertainty.

The generic issues process will be discussed. The screening analysis of a generic issue arising

from operating experience will be presented.

79
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80
Operating Experience Implementation Task Force

Terrence Reis

U.S. Nuclear Regulatory Commission, (NRR)Task Force Lead

When the Office for Analysis and Evaluation of Operational Data was dissolved in 1999 the

operating experience (OpE) program functions were transferred to the Office of Nuclear Reactor

Regulation (NRR), Office of Nuclear Regulatory Research (RES), Office of Administration, and the

Incident Response Organization. Since that time, some difficulties have been encountered in the

OpE program. NRR and RES began discussions in 2001 to correct some of the program

shortcomings. In 2002, the Davis Besse Lesson Learned Task Force, also recommended that the

OpE Program be re-assessed. An interoffice Reactor Operating Experience Task Force conducted

an assessment of the OpE Program and provided 23 recommendations to correct the OpE

Program shortcomings. This presentation will briefly discuss the activities being conducted to

implement a new OpE Program.

81
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82
A Generalized Framework for Assessment of Safety Margins

in Nuclear Power Plants

George Lanik, NRCIRES; Jim Meyer, ISL

A generalized framework for assessment of safety margins is proposed with the goal of using both

probabilistic and deterministic analysis to generate an overall safety margin index. Through this

technique, the proposed framework merges elements of safety regulation such as defense-indepth, traditional safety margins, and probabilistic risk. Separate safety margin metrics are

proposed for fuel performance, primary pressure boundary performance, and containment

performance, with the possibility of combining these into an overall safety metric.

A primary goal of this work is to develop a method to assess changes at plants due to power uprates, longer cycles, extended fuel bum-up, and aging processes. The process could assess other

changes as well, such as operating procedures and technical specifications. Such an approach

might be adapted to assess proposed changes such as those currently addressed by Regulatory

Guide 1.174, 'An Approach for Using Probabilistic Risk Assessment in Risk-informed Decisions

On Plant-Specific Changes to the Licensing Basis."

At the current state of development, the focus has been on the status of the fuel during accidents

and transients with performance metrics related to peak clad temperature and peak enthalpy

deposition. Preliminary results have been generated from analysis of medium sized Loss-ofCoolant Accident sequences.

83
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84
Boric Acid Corrosion of Light Water Reactor Pressure Vessel Materials

J.-H. Park, 0. K. Chopra, K. Natesan, W. J. Shack and W. H. Cullen

U.S. Nuclear Regulatory Commission

This report presents experimental data on corrosion rates of the materials found in or near typical

reactor pressure vessel penetrations, in boric acid solutions of varying concentrations, at

temperatures of 95-316*C (203-600*F). Tests were conducted in (a) high-temperature,

high-pressure aqueous solutions with a range of boric acid concentrations, (b) high-temperature

(150-C-316-C) H-1-O solutions at ambient pressure, wetted and dry, and (c) low-temperature

(- 95'C) Saturated, aqueous, boric acid solutions. The results indicate significant corrosion only for

the low-alloy steel, and no corrosion for Alloy 600 or 308 stainless steel cladding. Corrosion rates

were most significant for two conditions: (a) a boric acid-saturated, aqueous solution at 97.50C

exhibited corrosion rates up to 100 mm/year, and (b) H-B-0 solutions exhibited a corrosion rate of

about 125 mm/year at a temperature of 150'C. No material loss was observed in boric acid melts

or deposits in the absence of moisture. Tests were also conducted with alloys of various Cr

contents in the range of 0.2-25 wt.% to evaluate the effect of Cr loss from Type 308SS weldment

on the corrosion rate in a room temperature-saturated boric acid solution at temperatures between

150' and 316tC at a pressure of 12.4 MPa. The corrosion test results from this study are

compared with the existing corrosion/wastage data in the literature.

85
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(9-204) (Assigned by NRC, Add Vol., Supp., Rev,

NRCMD S.7 and Addendum Numbers, ff any.)

BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse) NUREG/CP- 0188

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED

Transactions of the 2004 Nuclear Safety Research Conference MONTH YEAR

October 2004

4. FIN OR GRANT NUMBER

A3988

S. AUTHOR(S) 6. TYPE OF REPORT

Summaries of Conference Papers by various authors; Compiled by Susan Monteleone, BNL Conference summaries

7. PERIOD COVERED (lncdusie Dates)

October 25-27, 2004

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (if NRC, provide Division. Ofice or Region. US. Nucdear Regulatory Commission, and ma ding address; if contractor.

provide name and mailing address.)

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (if NRC, he 'Same as above' if contractor, provide NRC Division, Office or Region, U.S. Ndear Regulatory Commission,

and mahig address.)

Same as Item B above.

10. SUPPLEMENTARY NOTES

S. Nesmith, NRC Proiect Manager, Transactions prepared by Brookhaven National Laboratory

11. ABSTRACT (200 words or less)

This report contains summaries of papers on reactor safety research to be presented at the 2004 Nuclear Safety Research

Conference (formerly titled the Water Reactor Safety Information Meeting) at the Marriott Hotel at Metro Center in Washington,

DC, October 25-27, 2004.

The summaries briefly describe the programs and results of nuclear safety research sponsored by the U.S. Nuclear Regulatory Commission's Office of Nuclear Regulatory Research. Also included are summaries of invited papers concerning nuclear

safety issues from U.S. government laboratories, the electric utitlities, the nuclear industry, and from foreign governments and

industry.

The summaries have been compiled here to provide a basis for meaningful discussion of information exchanged during the

course of the meeting, and are in the order of their presentation on each day of the meeting.

12. KEY WORDS/DESCRIPTORS (Lit words orphrases that wilassist researchers in locating the repot.) t3. AVAILABILITY STATEMENT

reactor safety research unlimited

nuclear safety research 14. SECURITY CLASSIFICATION

pus Page)

unclassified

(This Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (9-2004) PRINTED ON RECYCLED PAPER
NUREG/CP-0188 TRANSACTIONS OF THE 2004 NUCLEAR SAFETY

RESEARCH CONFERENCE

OCTOBER 2004

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

OFFICIAL BUSINESS