ML20040H235
| ML20040H235 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/12/1982 |
| From: | Diab S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18023A006 | List: |
| References | |
| NUDOCS 8202170480 | |
| Download: ML20040H235 (19) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of TEXAS UTILITIES GENERATING COMPANY,
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50-446
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(Comanche Peak Steam Electric Station,
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Units 1 and 2)
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AFFIDAVIT OF SAMMY S. DIAB I, Samy S. Diab, being duly sworn, do depose and state the following:
Q.1.
By whom are you employed, and what is the nature of the work you perform?
A.I.
I am a Nuclear Engineer employed by the U.S. Nuclear Regulatory Commission, Division of Systems Integration, Reactor Systems Branch ("RSB"). A copy of rny statement of qualifications is attached to this affidavit.
0.2.
What is the nature of the responsibilities you have regarding the Comanche Peak Steam Electric Station ("CPSES")?
A.2.
I was the Reactor Systems Branch lead reviewer for CPSES.
In this capacity, I was responsible for the safety review of the CPSES Final Safety Analysis Report ("FSAR") Sections 5.2.2, 5.4.7, 6.3, and 15.0, in accordance with the corresponding sections in the Standard Review Plan, NUREG-75/087.
8202170400'820212 PDR ADOCK 05000445 0
. 0.3.
Would you describe the scope of the subject matter addressed in your affidavit?
A.3.
My affidavit will eddress Contention 2, which alleges that:
One or more of the reports used in the construction of computer codes for the CPSES/FSAR have not been suitably verified and fonnally accepted; thus t
conclusion based upon these computer codes are invalid.
In particular, I have been asked to determine whether the facts presented in para 4raphs 2, 3, 4 and 5, of the Applicants'
" Statement of Material Facts As To Which There Is No Genuine Issue Regarding Contention C" (" Statement of Material Facts")
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are correct; and if the Staff supports the Applicants' position.
In addition, I have been asked to describe the review process 4~
for computer code topical reports.
Q.4.
Have you read, and do you agree with paragraph 2 of Applicants' Statenent of Material Facts with regard to NRC Staff accept-ance of Report 5, Report 24, Report 25, Report 26, Report 27,
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Report 29, Report 30, Report 31, and Report 32?
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A.4.
I have read paragraph 2 of Applicants' Statement of Material j
Facts on Contention 2, and I agree with the Applicants' state-l ment in that paragraph with regard to NRC Staff acceptance of the above-listed reports, except with respect to Report 5.
Report 5 has been accepted for use only in ATWS analysis.
l Q.5.
Have you read, and do you agree with paragraph 3 of Applicants' l
Statement of Material Facts with regard to Staff acceptance 1
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-3 of:
(a) those portions of Report 3 which were referenced in the CPSES Safety Evaluation Report ("SER"), Section 5.2.2, and the CPSES Supplemental SER ("SSER"), Section 5.2.2; (b) those portions of Report 6 which were referenced in the CPSES SER Section15.1.2;(c)thoseportionsofReport13whichwere referenced in the CPSES SER Section 15.3.9, and SSER Section 15.3.9;(d) Report 14;and(e) Report 77
- A.S.
I have read paragraph 3 of Applicants' Statement of Material Facts on Contention 2, and I agree with the Applicants' state-ment in that paragraph with regard to Staff acceptance of Reports 3, 6 and 13, with the excepcion that Report 3 is not referenced in SSEP. Section 5.2.2, and' Report 13 is not referenced in SSER Section 15.3.9.
I disagree with Applicants' statement regarding Report 7, and state that Report 7 was not referenced in the CPSES FSAR Section 6.3, and was not relied upon by the Staff in reaching its safety conclusions in CPSES SER Section 6.3.
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Q.6.
Have you read, and do you agree with paragraph 4 of Applicants' Statement of Material Fact, which states that Report 2 and f
f Report 17 were not used by the Applicants' in perfoming
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safety-related analyses referenad in the CPSES FSAR?
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A.6.
I have read paragraph 4 of Applicarits' Statement of Material
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Facts on Contention'2, and I agree with Applicants' statement, N
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to the e(tent that Applicants state that these reports are not
,, '.. N1 referenced in the current CPSES FSAR.
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8 4-Q.7.
Ha e you read, and do you agree with paragraph 5 of the Appli-
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cants' Statement of Material Facts that Report 22 was subnitted for informational purposes only?
A.7.
I have read paragraph 5 of the Applicants' Statement of Material
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Facts on Contention 2, and I agree with the Applicants' state-ment,in that parag9th.
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.A Q.8.
Describe' ths ' subject matter of each topical report, and state:
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(a) whether or not the topical report constructs, verifies corkputer codes for the purpose of obtaining NRC Staff evalua-
' tion and formal acceptance; (b) indicate if the topical report was not referenced by the Applicants in the CPSES FSAR and (c) state if the report was accepted by the Staff?
A.8.
Report 2, WCAP-7769 " Overpressure Protection for Westinghouse Pressurized Water Reactors", dated October 1971', is an earlier version of WCAP-7769, Revision 1 (Report 3), which provides a methodology and a set of assumptions to analyze a nuclear plant's provisions for protection against overpressurization.
This report is not referenced in the current version of the CPSES FSAR.
In addition, this report does not describe, construct or verify any computer code.
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Report 3, WCAP-7769, Revision 1 " Overpressure Protection for 9
Westinghouse Pressurized Water Reactors, Revision 1" (June 1972),
was submitted by Westinghouse Electric Corporation (" Westinghouse").
This report sets forth a methodology and set of assumptions for
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5-analyzing a Westinghouse pressurized water reactor's provisions for protection against overpressurization. This report was~
referenced by the Applicants in the CPSES FSAR Section 5.2.2.
This topical report does not describe, construct or verify any computer code.
Report 5, WCAP-7907, "LOFTRAN Code Description" (October 1972),
was subnitted by Westinghouse. This report describes, constructs and provides verification for the LOFTRAN computer code.
LOFTRAN is a transient analysis code which calculates pertinent plant paraneters, including reactor coolant temperature and pressure, and power levels throughout the transient, based on a simulation of neutron kinetics, reactor coolant system, control systen for reactivity, pressure and temperature, the steam generator, and other associated components. This report is referenced in the CPSES FSAR in various sections (e.g., 15.1,15.2,15.3,15.4).
Report 5 has been fonnally accepted by the NRC Staff for the analysis of Anticipated Transients Without Scram (ATWS).
Although this topical report has not been accepted for general transient analysis applications, the Staff believes that the Applicants' utilization cf the LOFTRAN code, as referenced in the CPSES FSAR, adequataly predicts plant behavior throughout c
the transients and accidents for which the Applicants utilize this code. This is based on substantial experimental and calculational evidence.
First, the results of Staff's calcu-lations of the loss of main feedwater ATWS in a Westinghouse
. PWR utilizing the RELAP3-B computer code came within acceptable limits of those results obtained by Westinghouse utilizing LOFTRAN. Therefore, the Staff has concluded that LOFTRAN is an acceptable methodology for ATWS applications.1/ Second, a I
paper presented at the American Nuclear Society meeting in Denver during 1971_/ concluded that the LOFTRAN results 2
compared favorably with some transients tests on a Westinghouse reactor in Japan. The tested transients included a 25% load increase, a 25% load decrease, and a 100% load rejection.
Third, the Staff, through Los Alamos National Laboratory, con-ducted a loss of normal feedwater transient analysis for a Westinghouse plant similar to the Comanche Peak plant, utilizing theNRC-sponsoredcodeTRAC-PD2.$/ The Staff has compared the results of the transient of the above analysis to those calcu-lated for CPSES using LOFTRAN. The TRAC-PD2 calculations resulted in a peak pressure of 2296 psia while the LOFTRAN cal-culations resulted in a peak pressure of 2350 psia. Therefore, the Staff concluded that LOFTRAN censervatively predicted the pressurization transient. Finally, Westinghouse presented a 1/ NUREG-0460, Vol. II (April 1978), pp. XVII-3.
2/ T.W.T. Burnett, J.M. Geets, and P.M. Ginsburg, " Operational Per-formance of Westinghouse PWR Control Systems (paper presented at ANS meeting in Denver,1971).
E E.S. Idar and J.F. Lime (Loss Alamos National Laboratory), TRAC-PD2,
" Calculation of Loss of Normal Feedwater in a Westinghouse Prcssurized Water Reactor" (November 1981).
. comparison of the results obtained by the LOFTRAN code, with the results obtained by the WFLASH, RELAP, and other NRC, Babcock and Wilcox, and Combustion Engineering computer codes.1/ Based on this evidence, the Staff concluded in Section 15.1.2 of the CPSES SER that the use of the LOFTRAN code, as used and referenced by the Applicants in the CPSES FSAR, to analyze certain plant variables was acceptable,- until the report is fomally approved.
Report 6, WCAP-7908, "FACTRAN - A FORTRAN-IV Code for Themal Transients in a U0 Fuel Rod" (June 1972), was submitted by 2
Westinghouse for NRC Staff approval. This report was referenced by the Applicants in Section 15.2.3.2 of the CPSES FSAR.
WCAP-7908 is being fomally reviewed by the Core Perfomance Branch (CPB) of the Staff and has not yet been formally accepted by the Staff. However, the Reactor Systems Branch accepts the use of this topical report for transient analysis, since the Staff review of Report 6 has progressed to the point that the results of analyses using this code will not be appreciably altered by any change in methods that may be required upon the completion of the Staff review.
l Al Letter from E.P. Rahe, Jr., Westinghouse, to J.R. Miller, U.S.N.R.C.,
" Transmittal of Slides from the December 15, 1981 Westinghouse Meeting on the LOFTRAN (WCAP-7907) and MARVEL (WCAP-7909) Codes' (January 19,1982).
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' Report 7, WCAP-7909, " MARVEL, A Digital Computer Code for Transient Analysis of a Multiloop Pressurized Water Reactor System" (June 1972), is a. topical report submitted for Staff evaluation by Westinghouse. Thisreportdescribesandveri-fies the MARVEL computer code, which calculates the multi-loop transteilt behavior of a pressurized water reactor. The computer code simulates two reactor cooling loops, including the steam generators and associated systems.
Report 7 was not referenced in Section 6.3 of the FSAR for CPSES. The Staff did not rely on Report 7 in reaching the safety conclusions in Section 6.3 of the CPSES SER.
Report 13 WCAP-8330, " Westinghouse Anticipated Transients Without Trip Analysis" (August 1974), was submitted by Westing-house and presents a methodology and a set of assumptions for analyzing ATWS in Westinghouse pressurized water reactors.
These assumptions include bounding values for plant parameters, such as reactor coolant flow, liquid relief discharge rates from safety valves, moderator temperature coefficient, and Doppler coefficient.
This report is referenced by the Appli-cants in Section 15.8 of the CPSES FSAR. This topical report does not describe, construct or present verification for any computer code.
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Report 14, WCAP-8424, "An Evaluation of Loss of Flow Accidents Caused by Power System Frequency Transients in Westinghouse Pressurized Water Reactors, Revision 1", was referenced by the Applicants in Section 15.3 of the FSAR for CPSES. The Staff's evaluation and approval in Section 15.2.2 of the CPSES Safety Evaluation Report ("SER") of Applicants' analyses of decreased cooling flow transients did not rely on Report 14.
Report 17, WCAPs-9168 and 9169, "Westignhouse Emergency Core Cooling System Evaluation Model-Hodified October 1975 Version" (September 1977), are the proprietary and non-proprietary versions of a topical report which describes the Westinghouse Emergency Core Cooling System'("ECCS") evaluation model. This report was not referenced by the Applicants in the current CPSES FSAR; and does not describe, construct or present verification for any computer code.
Report 22, WCAP-8768, " Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Sumaries" (no date), is a status report submitted by Westinghouse which summarizes the status of Westinghouse safety-related research and development for its reactors. This report was referenced by Applicants in Section 1.6 of the CPSES FSAR for informational purposes. The report does not describe, construct or verify a computer code. Nor does this report present any methodology, or set of assumptions for Staff review and acceptance.
Report 24, WCAPs-8170 and 8171, " Calculational Model for Core Reflooding After a LOCA" (July 1974) are the proprietary and non-proprietary versions of a Westinghouse report which describes, constructs and presents verification for the WREFLOOD computer code, which simulates the reactor core reflood hydraulics behavior.
It calculates several parameters, including the core flooding rates, core coolant pressures, and core inlet tempera-tures.
Report 24 was referenced by the Applicants in Section 15.6.5 of the CPSES FSAR. This report was evaluated and accepted by the Staff on May 30, 1975 as part of its review of the Westing-house ECCS evaluation mode.
Report 25, WCAPs-8200 and 8261, "WFLASH, a FORTRAN-IV Computer Program for Simulation of Transients in a Multi-Loop PWR" (June 1974), are the proprietary and non-proprietary versions of a Westinghouse report which simulates the core thermal hydraulic behavior after a small break Loss of Coolant A.cident ("LOCA"),
through the use of the WFLASH code. The WFLASH code calculates pertinent plant parameters, such as the reactor coolant pressure and temperature.
It is referenced in Section 15.6.5 of the FSAR for CPSES. Report 25 describes, constructs, and presents verifi-cation for the WFLASH computer code. Report 25 was accepted by the Staff on May 30, 1975, as part of its review and acceptance of the Westinghouse ECCS evaluation model.
. Report 26, WCAPs-8301 and 8305, "LOCTA-IV Program; LOCA Tran-sient Analysis" (June 1974) are the proprietary and non-proprietary versions of a report sumitted by Westinghouse as part of the ECCS evaluation model that describes, constructs, and presents verification for the LOCTA-IV computer code. This code was referenced by the Applicants in Section 15.6.5 of the CPSES FSAR.
LOCTA-IV calculates several parameters, including the peak clad temperature of a fuel pin as a result of both small and large break LOCAs.
Report 26 was accepted by the Staff on May 30, 1975, as part of its review and acceptance of the Westinghouse ECCS evaluation model.
Report 27, WCAPs-8302 and 8306, " SATAN-IV Program, A Compre-hensive Space-Time Dependent Analysis of LOCA" (June 1974) are the proprietary and non-proprietary versions of a Westing-house report which describes, constructs, and presents veriff-cation for the SATAN-IV computer code. SATAN-IV simulates the reactor core thermal hydraulic behavior during the blowdown phase of a large break LOCA.
It calculates several parameters, inc7uding core inlet flow and enthalpy, core pressure, core power, and cross-flow parameters within the core. This topical report was accepted by the Staff on May 30, 1975, as part of the Staff's review and acceptance of the Westinghouse ECCS evaluation model.
Report 29, WCAP-8339, " Westinghouse ECCS Evaluation Model, A Sumnary" (July 1974), was submitted by Westinghouse. This topical report describes the LOCA modeling of the Westinghouse ECCS evaluation model methodology. The report also describes the appropriate application and limitstions of the ECCS model.
This report was referenced by the Applicants in CpSES FSAR Section 15.6.5.3.
This report does not describe, construct or present verification for any computer codes. This report was accepted by the Staff on May 30, 1975, as part of its evalua-tion and acceptance of the Westinghouse ECCS evaluation model.
Report 30, WCAPs-8340 and 8356, " Westinghouse ECCS Plant Sensitivity Studies" (July 1974) are the proprietary and non-
~I proprietary versions of the sensitivity studies conducted by Westinghouse to examine the sensitivity of the ECCS evaluation model to variations in several parameters. The sensitivity analysis examines the degree of change in calculated parameter results obtained from various parts of the ECCS model, which are the result of variations in input parameters such as nodalization, time step, break size and location, and the single failure, application. This report is referenced in Section 15.6.5 of the CPSES FSAR. The topical report does not describe, construct or verify computer codes presented for the first time in this report. However, it does present partial verification for the computer codes which are a part of the f
Westinghouse ECCS evaluation model. This report was accepted by the Staff on May 30, 1975, as part of its evaluation and acceptance of the Westinghouse ECCS evaluation model.
Report 31, WCAPs-8341 and 8342, " Westinghouse ECCS Evaluation Model Sensitivity Studies, (July 1974), are a continuation of the sensitivity studies in Report 30. Report 31 is referenced in the FSAR for CPSES at Section 15.6.5.
Report 31 does not describe, construct or provide verification for a computer code presented in this report. However, the topical report does present partial verification for the computer codes used in the Westinghouse ECCS evaluation model.
Report 31 was accepted by the Staff on May 30, 1975, as part of its review and acceptance of the Westinghouse ECCS evaluation model.
Report 32, WCAPs-8471 and 8472, " Westinghouse ECCS Evaluation Model - Supplementary Information" (January 1975) are the proprietary and non-proprietary versions of Westinghouse responses to NRC Staff questions regarding the Westinghouse ECCS evaluation model. This report was referenced by the Applicants in the CPSES FSAR at Section 15.6.5.3.
This report does not describe, construct or provide verification for any computer code in this topical report.
Report 37 was accepted by the Staff on May 30, 1975, as part of the Staff's review and acceptance of the Westinghouse ECCS evaluation model.
. Q.9.
What are the principal characteristic sections of topical reports which describe and construct computer ccdes?
A.9.
Generally the body of such topical reports are composed of three main parts: methodology and modeling; application; and verification.
Q.10.
What are the responsibilities of the organization submitting a topical report for NRC Staff evaluation and acceptance?
A.10.
The submitting organization :nust set out the purposes of the topical report, and define its scope and applicability.
If there are limitations or restrictions on the use of the reports, the submitting organization must clearly identify them. The submitting organization is responsible for the accuracy and
- I completeness of the information provided in support of the topical report.
Q.11.
Who is responsible for verification of the computer code presented in the topical report?
A.11.
The burden of verification of any computer code presented in a topical report is placed with the submitting organization.
The organization seeking approval and acceptance of a computer code must provide sufficient evidence to clearly demonstrate the validity and accuracy of the methodology contained in the code submitted for review.
. Q.12.
What is the responsibility of the Staff in reviewing a topical report that describes a computer code?
A.12.
The NRC Staff reviews the main parts of the topical report to ascertain that, when using the computer code described in the topical report, an applicant will adequately and conser-vatively predict the behavior of the plant. The Staff must confim that the limitations on the use of the computer code
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are conservative for the full range of normal and abnormal conditions for which the code is to be used.
Q.13.
How does the Staff review topical reports?
A.13.
The methodology and modelling of the computer code presented in the topical report are reviewed by the NRC Staff against accepted engineering practice. The applications of the computer code are evaluated to ensure its proper use in accordance with model conditions.
In evaluating the verification for the code, the Staff reviews the evidence submitted, which may j
include comparisons to data from operating plants or test facili-ties.
In some instances, the Staff may conduct audit calcula-tions to ascertain the validity and accuracy of the computer code.
Q.14.
What is the course of review if the Staff finds the topical report under review unacceptable?
A.14.
If the Staff finds deficiencies in the topical report, its first course of action is to request additional information.
. i Requests for additional information about (or comments on) a topical report under review may be handled either by a letter to, or a meeting with, the submitting organization.
If the Staff concludes that the topical report does not adequately verify the subject computer code, the Staff may request as a condition for further review, that the submitting organization provide additional information in the form of experimental results or plant test data. The Staff may also request that
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the submitting organization perform sensitivity analyses or additional analyses utilizing bounding input parameters.
Q.15.
What methods does the Staff utilize to independently verify computer codes that are the subject of topical reports?
A.15.
In verifying computer codes the Staff may request computer runs of standard problems using the code under review, then compare the results obtained with relevant plant experimental data. Alternatively, the Staff may conduct audit calculations using NRC-developed codes and compare the results to those obtained by the code being reviewed.
Q.16.
How does the Staff justify the use of computer codes which are the subject of topical reports when the topical report has not been fonnally accepted?
A.16.
The Staff may accept the use of a topical report that has not yet been fonnally accepted if the Staff review of the topical report has progressed to the point where the Staff has ade-quate assurance that any alterations to the code as a result of the completion of the review process will not change the results obtained by using the code in its current status, and therefore will not alter the conclusions drawn about the safety of the plant. The Staff may gain adequate assurance about the validity of a computer code by the demonstrated ability of the code to predict simulated experiments or plant data during normal or abnonnal plant operation.
Furthermore, the Staff conditions the use of the computer so that if, upon the completion of the review process, substantial changes to the topical report are required, the applicant will be required to reanalyze the affected transients and accidents using acceptable methods.
The above statements and opinions are true and correct to the best of my personal knowledge and belief.
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CD my S. Diab Subscribed and sworn to before me d ay of February, 1982.
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' Notary Publig My Comission expires:
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STATEMENT OF PROFESSIONAL QUALIFICATIONS SAMMY S. DIAB I am a Nuclear Engineer in the Reactor Systems Branch of the U.S.
Nuclear Regulatory Commission (NRC).
In this position, I am responsible for the technical. analysis and evaluation of reactor systems, accidents and transients, and applications for nuclear reactor operating licenses.
I have been in my.:urrent position since 1980.
From 1978 to 1980, I was a reactor systems reviewer in the Reactor Safety Branch, Division of Operating Reactors of the NRC.
In that position my responsibilities included:
systems analyses, accident and transient analyses, and reload application reviews.
From 1977 to 1978, I was a Nuclear Engineer in the Engineering Methodology Standards Branch, Office of Standards of the NRC.
In that position -I was responsible for updating and revising the standard review plan.
I developed Regulatory Guide 1.139, " Residual Heat Removal Guidance",
and Regulatory Guide 1.141, " Containment Isolation Provisions for Fluid l
Systems".
From 1973 to 1977, I was a Nuclear Engineer with Bechtel Power Corporation, Gaithersburg Power Division, Maryland.
I was responsible for reactor containment pressure and temperature analyses following a l
spectrum of high energy line breaks, jet impingement calculations, and subcompartment transient behavior.
I developed and used couputer codes.
l I also modified existing computer codes.
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. From 1971 to 1973, I was a research assistant with the Nuclear Engineering Department of the Pennsylvania State University.
In 1974 I was awarded a M.S. degree in Nuclear Engineering from Pennsylvania State University.
From 1967 to 1971, I was a researcher with the Egyptian Atomic i
Energy Establishment.
I received my B.S. degree in Nuclear Engineering from the University of Alexandria, Egypt.
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