ML20040H200
| ML20040H200 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/12/1982 |
| From: | Mizuno G, Rothschild M NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML18023A006 | List: |
| References | |
| NUDOCS 8202170434 | |
| Download: ML20040H200 (44) | |
Text
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2/12/82 1
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD l
In the Matter of TEXAS UTILITIES GENERATING COMPANY, )
50-446 (Comanche Peak Steam Electric
)
Station, Units 1 and 2)
)
NRC STAFF'S ANSWER SUPPORTING APPLICANTS' MOTION FOR
SUMMARY
DISPOSITION OF CONTENTIONS 2 AND 7 I.
INTRODUCTION On January 26, 1982 Applicant:; filed, pursuant to 10 C.F.R. 5 2.749,
" Applicants' Motion for Sumary Disposition of CFUR's Contentions 2 and 7" (" Applicants' Motion")1/.
In support of the Motion, Applicants attach
-1/
Applicants' Motion was filed prior to CFUR's notification on February 8, 1982, to the parties and the Atomic Safety and Licensing Board (hereafter "the Licensing Board") that it was withdrawing all of its remaining contentions, including Contentions 2 and 7.
In view of CFUR's intentiens, on February 9,1982, the Licensing Board issued an " Order Cancelling Scheduled Evidentiary Session", since "the evidentiary sessions scheduled to begin on March 9,1982, were to address matters in controversy related to CFUR's contentions."
(Order, at 1). The Licensing Board indicated that it:
"...will await and consider CFUR's written filings on the matter of withdrawal before determining the appropriate disposition of CFUR's contentions" (Id_.)
Although CFUR has withdrawn Contentions 2 and 7, in order to assist the Licensing Board in its disposition of those contentions, the Staff is filing its answer to Applicants' Motion for summary disposi-l tion. The Staff's answer supports the Applicants' Motion, since there is no genuine issue of material fact to be determined with respect to both Contentions 2 and 7.
Accordingly, the Staff believes, as indicated infra, that the Applicants' motion should be granted and Contentions 2 and 7 should be dismissed.
j r202170434 820212 PDR ADOCK 05000445 0
a " Statement of Material Facts As To Which There Is No Genuine Issue Regarding Contention 2" and a " Statement of Material Facts As To Which There Is No Genuine Issue Regarding Contention 7" and affidavits. As stated below, the Staff submits that the attached Staff affidavits, together with the Staff's Safety Evaluation Report (SER)2/ and Supplement Nos. I and 2 thereto (.SSER),EI and Applicants' Motion and supporting documents demonstrate the absence of any genuine issue of material fact and that dismissal of Contentions 2 and 7 ie warranted as a matter of law.
Section II of this pleading discusses generally the law applicable to motions for sumary disposition.Section III of this' pleading sets forth the Staff's reasons for concluding that there is no genuine issue of material fact raised by Contentions 2 and 7.
II. GENERAL POINTS OF LAW _
Pursuant'to 10 C.F.R. 5 2.749 of the ConmiisIdn's Rules of Practice,
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sumary disposition is available to a party in NRC proceedings as to all or any part of the matters involved in the proceeding, as follows:
2/
" Safety Evaluation Report Related to the Operation of Comanche Peak Steam Electric Station, Units 1 and 2",(NUREG-0797, July 1981).
-3/
" Safety Evaluation Report Related to the Operation of Comanche Peak Steam Electric Station, Units 1 and 2" (NUREG-0797/ Supplement No. 1, October 1981) and " Safety Evaluation Report Related to the Operation of Comanche Peak Steam Electric Station, Units 1 and 2" (NUREG-0797/ Supplement No. 2, January 1982).
--g-w r.
(d) The presiding officer shall render the decision sought if the filings in the proceeding, depositions, answers to interregatories, and admissions on file, together with the statements of the parties and the affidavits, if any, show that there is no genuine issue as to any material fact and that the moving party is entitled to a decision as a matter of law....
The Comission's sumary disposition rule is analogous to Rule 56 of the Federal Rules of Civil Procedure, governing motions for summary judgment, and Federal court decisions interpreting Rule 56 may be relied upon in NRC proceedings for the interpretation of 10 C.F.R. 5 2.749.
See, e.g.,
' Alabama Power Co. (Joseph M. Farley Nuclear Plant, Units 1 and 2),
ALAB-182, 7 AEC 210, 217 (1974).
Under both Rule 56 and 10 C.F.R. 5 2.749, the party seeking sumary judgment has been held to have the burden of proof, viz., the burden of demonstrating the absence of a genuine issue as to any material fact.
Cleveland Electric Illuminating Co. (Perry Nuclear Power Plant, Units 1
~
and 2), ALAB-443, 6 NRC 741, 753 (1977), citing Adickes v. Kress & Co.,
398 U.S. 144, 157 (1970). The proponent of the motion for summary disposition must meet its burden of proof even if the party opposing the motion fails to present evidentiary material to the contrary.
- Perry, supra, 6 NRC at 754. On the other hand, where a properly supported motion for sumary disposition has been made, a party opposing the motion may not rest upon the mere allegations of its contention or answer.
10 C.F.R. 6 2.749(b); Virginia Electric and Power Co. (North Anna Nuclear Pow r Station, Units 1 and 2), ALAB-584,11 NRC 451, 453 (1980).
- Rather, pursuant to 10 C.F.R. 5 2.749(b), the party opposing summary disposition c
must present specific material facts showing there is no genuine issue to be heard:
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(b)... When a motion for sumary decision is made and supported..., a party opposing the motion may not rest upon the mere allegations or denials of his answer; his answer...
must set forth specific facts showing that there is a genuine issue of fact.
If no such answer is filed, the decision sought, if appropriate, shall be rendered.
Finally, in this regard, all material facts set forth in the statement filed by the moving party in support of its motion for sumary disposi-tion "will be deemed to be admitted unless controverted by the statement required to be served by the opposing party." 10C.F.R.I2.749(a).O In a recent Statement of Policy, the Comission emphasized the availability of sumary disposition in appropriate cases, as a means of expediting the hearing process.
In Statement of Policy on Conduct of Licensing Proceedings, 46 Fed. Reg. 28,533 (May 27, 1981), the Comission stated as follows:
In exercising its authority to regulate the course of a hearing, the boards should encourage the parties to invoke the sumary disposition procedure on issues where there is no genuine issue of material fact so that evidentiary hearing time is not unnecessarily devoted to such issues.
e 46 Fed. M. at 28,535. Similarly, the Appeal Board has stated that the sumary disposition rule provides "an efficacious means of avoiding unnecessary and possibly time consuming hearings on demonstrably insub-4/
Pursuant to 10 C.F.R. 6 2.749(a), answers to motions for summary disposition are to be filed within 20 days after service of the motion; and responses to "new facts and arguments presented in any statement filed in support of the motion" are to be filed within ten days after service thereof.
In deciding a motion for sumary disposition, "the record is to be viewed in the light most favorable to the party opposing the motion." Gulf States Utilities Co. (River c
Bend Station, Units 1 and 2), LBP-75-10, 1 NRC 246, 248 (1975);
Public Service Co. of New Hampshire (Seabrook Station, Units 1 and 2), LBP-74-36, 7 AEC 877, 879 (1974).
stantial issues." Houston Lighting and Power Co. (Allens Creek Nuclear Generating Station, Unit 1), ALAB-590,11 NRC 542, 550 (1980). As the Appeal Board noted recently, a hearing on each issue raised "is not inev-itable," but " wholly depends upon the ability of the intervenors to demonstrate the existence of a genuine issue of material fact...."
Philadelphia Electric Co. (Peach Bottom Atomic Power Station, Units 2 and 3), ALAB-654, 14 NRC 632, 634 (1981). A party cannot avoid sumary
" disposition "'on the mere hope that at trial he will be able to discredit
~
movant's evidence,'" nor may a party "'90 to trial on the vague supposi-tion that something may turn up.'"
Gulf States Utilities Co. (River Bend Station, Units 1and2),LBP-75-10,1NRC246,248(1975), quoting 6 Moore's Federal Practice 956.15[3]and[4].
III. ARGUMENT '
Contention 2 Contention 2, as admitted by the Licensing Board in its " Order Subsequent to the Prehearing Conference of April 30, 1980," dated June 16, 1980, 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 conclusions based on these computer codes are invalid.
The Staff believes that Contention 2 raises no genuine issue of E
material fact. After reviewing Applicants' Motion for Sumary Disposition and the accompanying Affidavits of Chun-Mong Jan, Arthur C.
Spencer, and William R. Spezialetti, the Staff believes that the Applicants' Statement of Material Facts As To Which There Is No Genuine Issue Regarding Contention 2 is correct. 5/ The Staff accordingly supports Applicants' position that. summary disposition of Contention 2 should be granted.
CFUR has identified 28 topical reports _/ which it claims describe, 6
construct, and provide verification for computer codes referenced in the Final Safety Analysis Report ("FSAR") for the Comanche Peak Steam
- Electric Station ("CPSES"). CFUR also alleges that these 28 topical reports have not been " suitably verified" 7/; and " formally accepted" by the NRC Staff.
5/
See attached Staff affidavits. The Staff has not addressed para-graphs 7 and 10 of Applicants' Statement of Material Facts As To Which There Is No Genuine Issue Regarding Contention 2 (" Statement of Material Facts").
In addition, as stated in the attached Staff affidavits, there are certain statements with which the Staff does not completely agree. However, under the Staff's theory of the case, establishment of such facts as being beyond dispute is not necessary to a favorable decision upon the motion for sunnary disposition.
6/
See Supplement to Petition for Leave to Intervene Cy Citizens for TiTr Utility Regulation (CFUR) (May 7, 1979); Report of CFUR's PositiononEachContention(April 10,1980); CFUR's September 15, 1980 Answer to Interrogatory 11 of Applicants' First Set of Interrogatories to CFUR (August 13,1980), and CFUR's Supplemental Answers to Interrogatory 11 (May 8, 1981); CFUR's March 11, 1981 Answers to Interrogatory C2-2 of NRC Staff's First Set of Interrog-atories to CFUR (January 19,1981), and CFUR's Supplemental Answers to Staff Interrogatory C2-2 (May 22,1981 and September 1,1981).
-7/
According to CFUR, the identity of the party who " suitably verifies" the topical reports is not an issue in Contention 2.
See CFUR's March 11, 1981 Answers to Interrogatory C2-4 of NRC Staff's First Set of Interrogatories to CFUR (January 19,1981). CFUR is of the c
belief that the Applicant suitably verifies and the NRC Staff I
accepts the topical reports. As will be discussed, infra, CFUR is mistaken in this regard. Topical reports are verified by the party submitting the topical report for NRC Staff acceptance. See also Affidavit of Samray S. Diab, Answer 11.
l
_7-In addition, CFUR has identified three (3) computer codes, itRI/STARDYNE, SCONV, and SPECTRA, which it alleges are based on topical reports which were not " suitably verified" and "fomally accepted" by the Staff. 8_/
These topical reports and computer codes are identified, and listed below in the same order as Applicants' tabulation of the CFUR-identified topical reports and codesEl in Applicants' Notion for Summary Disposition.
8/
See footnote 2, supra; see also CFUR's fiarch 3,1981 Answer to Tiiterrogatory C2-4 of StiTf'TTirst Set of Interrogatories to CFUR
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(January 19,1981).
9/
On January 22, 1982, theStaffreceivedCFUR'sSeventh[ sic]Setof Interrogatories to NRC Staff, dated January 20, 1982. CFUR's Interrogatory 8 to the Staff sought information and supporting documentation on the " nature, status, and date of final resolution" for 25 listed topical reports, eleven (11) of which were identified for the first time in this interrogatory.
In addition, Interroga-tory 9 to the Staff asked whether the Staff had " suitably verified and fomally accepted" eleven (11) computer codes, eight (8) of which were identified by CFUR for the first time in this interroga-tory. The eight newly-identified computer codes, like the three earlier-identified codes (NRI/STARDYNE, SCONV, AND SPECTRA), are not the subject of any current or past topical report requesting Staff evaluation and acceptance of those codes.
CFUR did not state in its September 15, 1980 or May 8, 1981 Answers to Interrogatory 11 of Applicants' First Set of Interrogatories to CFUR(August 13,1980), that these eleven topical reports and eight computer codes must be " suitably verified" and " formally accepted."
That interrogatory specifically requested CFUR to identify the topical reports and computer codes which CFUR contends were not verified and accepted. CFUR also did not indicate that these eleven reports and eight computer codes were not verified and fomally accepted in their Answer (March 11,1981) and supplemental answers (May 22, 1981 and January 19, 1981, respectively) to Interrogatory C2-2 of NRC Staff's First Set of Interrogatories to CFUR. This interrogatory also requested CFUR to identify the topical reports q
and codes which CFUR claims must be verified and eccepted. CFUR was l
l (Continued)
. CFUR-IDENTIFIED TOPICAL REPORTS AND COMPUTER CODES Applicants' Identification Report i
Report 1 WCAP-7706 (Non-Proprietary), "An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients,"
't.
February 1973.
Report 2 WCAP-7769, " Overpressure Protection for Westinghouse Pressurized Water Reactors," October 1971.
Report 3 WCAP-7769, " Overpressure Protection for Westinghouse Pressurized Water Reactors," Revision 1, Jure 1972.
Report 4 WCAP-7832, " Evaluation of Steam Generator Tube, Tubesheet and Divider Plate Under Combined LOCA Plus SSE Conditions," December 1973.
Report 5 WCAP-7907, "LOFTRAN Code Description," October 1972.
Report 6 WCAP-7908, "FACTRAN - A FORTRAN-IV Code for Thennal Transients in a 00 Fuel Rod," June 1972.
2 Report 7 WCAP-7909, " MARVEL, A Digital Computer Code for Tran-sient Analysis of a Multiloop Pressurized Water Reactor System," June 1972.
9/
(Continued) under a continuing obligation to supplement its answers to Applicants' and Staff's interrogatories by identifying topical reports and com-puter codes which must be verified and accepted, as alleged by CFUR pursuant to Contention 2.
See Dairyland Power Cooperative (La Crosse Boiling Water Reactor), Docket No. 50-409 (FTOL Proceeding)
(unpublished opinion issued July 8, 1980).
Since CFUR did not supplement its answers or otherwise specifically inform Applicants and the Staff that the eleven topical reports and eight computer codes identified for the first time in CFUR's c
Interrogatory 8 to Staff must be " suitably verified" and " formally accepted," the Staff assumes that CFUR does not allege, as part of Contention 2, that these reports and codes must be so verified and l
accepted.
I
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Report 8 WCAP-7950, " Fuel Assembly Safety Analysis for Combined Seismic and Loss of Coolant Accident," July 1972.
Report 9 WCAP-8163, " Reactor Coolant Pump Integrity in a LOCA "
September 1973.
Report 10 WCAP-8236 (Procrietary), December 1973 and WCAP-8288 (Non-Proprietary), " Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," January 1974.
Report 11 WCAP-8252, " Documentation of Selected Westinghouse Structural Analysis Computer Codes," Revision 1, July 1977.
Report 12 WCAP-8278 (Proprietary) and WCAF-8279 (Non-Proprietary),
" Hydraulic Flow Test of the 17x17 Fuel Assembly," February 1974.
Report 13 WCAP-8330, " Westinghouse Anticipated Transients Without Trip Analysis," August 1974.
Report 14 WCAP-8424. "An Evaluation of Loss of Flow Accidents Caused by Power System Frequency Transients in Westinghouse PWRs," Revision 1. June 1975.
Report 15 WCAP-8691 (Proprietary) and WCAP-8692 (Non-Proprietary),
" Fuel Rod Bowing," December 1975.
Report 16 WCAP-8720 (Proprietary) and WCAP-8785 (Non-Proprietary),
" Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations," October 1976.
Report 17 WCAP-9168 (Proprietary) and WCAP-9150 (Non-Proprietary),
Model - Modified October 1975 Version,{ stem EvaluationSeptember 19
" Westinghouse Emergency Core Cooling S Report 18 WCAP-9179(Proprietary)andWCAP-9224(Non-Proprietary),
" Properties of Fuel and Core Component Materials,"
September 1977.
Report 19 "MRI/STARDYNE, Static and Dynamic Structural Analysis System, User's Information Manual," developed by and proprietary to Mechanics Research, Inc.
Report 20 "SCONV", ref. cited in CPSES FSAR 3.7B(A).9, c
p.3.7B(A)-12. Gibbs & Hill, Inc.
Report 21
" SPECTRA", ref. cited in CPSES FSAR 3.7B(A).10,
- p. 3.7B(A)-15. Gibbs and Hill.
f
' Report 22 WCAP-8768, " Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Sunnaries," (no date).
Report 23 WCAP-7956, "THINC-IV, An Improved Program for Thermal-Hydraulic Analy-is af Rod Bundle Cores", (October 1973)
Report 24 WCAP-8170 (Proprietary), and WCAP-8171 (Non-Proprietary),
" Calculational Mode's for Core Reflooding After a LOCA" July, 1974 Report 25 WCAP-8200 (Proprietary), and WCAP-8261 (Nen-Proprietary),
"WFLASH, A FORTRAN-IV Computer Progran for Simulation of Transients in a Multi-Loop PWR", June 1974.
Report 26 WCAP-8301 (Proprietary), and WCAP-8305 (Non-Proprietary),
"LOCTA-IV Progran; LOCA Transient Analysis", June 1974.
Report 27 WCAP-8302 (Proprietary), and WCAP-8306 (Non-Proprietary),
" SATAN-IV Program, Comprehensive Space-Time Dependent Analysis of LOCA", June 1974.
Report 28 WCAP-8326 (Non-Proprietary) and WCAP-8327 (Proprietary)}
"ContainmentPressureAnalysisCode(C0CO)",(July 1974 Report 29 WCAP-8339, " Westinghouse ECCS Evaluation Mode - A Sunna ry",
(July 1974)
Report 30 WCAP-8340(Proprietary),andWCAP-8356(Non-Proprietary),
" Westinghouse ECCS Plant Sensitivity Studies", July 1974.
Report 31 WCAP-8341 (Proprietary), and WCAP-8342 (Non-Proprietary),
" Westinghouse ECCS Evaluation Model Sensitivity Studies",
July 1974.
Report 32 WCAP-8471 (Proprietary) and WCAP-8472 (Non-Proprietary),
" Westinghouse ECCS Evaluation Model--Supplementary Infomation", January 1975.
L l
CFUR's wording of Contention 2, in its Supplement to Petition for Leave to Intervene (May 7, 1979), and its Report of CFUR's Position on Each Contention (April 10,1980), did not specify by name or identifying number the topical reports which CFUR alleges describe, construct and present verification for computer codes, and therefore must be "fomally accepted" by the Staff. As a result, Contention 2, as admitted by the I
Licensing Board, does not identify by name or identifying number the topical reports constructing and verifying computer codes which CFUR contends must be accepted by the Staff.
However, in CFUR's discussion of Contention 2 in its May 7, 1979 Supplement to Petition to Intervene, and its April 10, 1980 Report of CFUR's Position, CFUR listed 28 topical reports which it claimed con-structed computer codes, and therefore must be verified and accepted.
- In addition, CFUR listed 3 computer codes in its April 10, 1980 Report, namely, MRI/STARDYNE, SCONV, and SPECTRA, which it contended must be
" suitably verified" and "fonnally accepted".
In fact, these computer codes are not the subject of any topical reports which construct and provide verification for these codes. Affidavit of Frank Rinaldi and John P. Matra, Answer 6.
Subsequently, the Staff submitted its First Set of Interrogatories to CFUR, dated January 19, 1981.
Interrogatory C2-1 of that set
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requested CFUR tn:
Specifically identify the " reports" and the " computer" codes referred to in... contention [2].
In its Answer of May 22, 1981 CFUR referred the Staff to the May 7, 1979 Supplement to Petition, and its April 10, 1980 Report, listings of topical reports and computer codes which CFUR contended must be verified i
and accepted.
In the same response, CFUR stated:
CFUR has not ruled out the possibility of including additional reports and/or deleting reports.
CFUR's listing of 3 computer codes which are not the subject of any topical report, its wording of Contention 2, and CFUR's response to the Staff's Interrogatory C2-2, together suggest that CFUR believes that all l
l
12
~. computer ' codes referenced in the CPSES/FSAR are the subject of a topical report wherein each code is individually verified by the submitting party, and evaluated and accepted by the Staff. This belief is erroneous.
The CPSES FSAR references computer codes which are described, con-
. ' l,,.
structed and verified in a topical report.E/ The FSAR for CPSES also references'ccmputer codes which are not described, constructed and veri-fiedinatopicalreport.N It is useful at this point to discuss what topical reports are, and c
the function that they serve. Topical reports are described in the Affidavits of John S. Berggren and Samrqy S. Diab. Topical reports are submitted by nucle'ar industry organizations to the NRC Staff for Staff evaluation and acceptance. The report is evaluated independently of any specific construction pemit ("CP") or operating license ("0L") proceeding.
1f a report is accepted, the report may subsequently be referenced in an applicant's FSAR in a particular CP or OL proceeding.
In effect, Staff acceptance of a topical report is " generic" in nature. There is no further needbyindividdalapplicantstore-verifythevalidityofthesubject matter of the topical report. Affidavit of Berggren, Answer 4.
A topical report requesting Staff acceptance of a computer code presents the purpose and description of the code, and provides the docu-mentation for the applicability and validity of the code. Verification Affidavit of Diab, is the burden of the party submitting the report.
-10/ See, e.., discussions of Reports 26 and 27 in Affidavit of Taniqy fab, Answer 8.
11/ See discussion of computer codes MRI/ SPECTRA, SCONN, and SPECTRA in IfTidavit of Frank Rinaldi and John P. Matra, Answer 6.
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Answer 11. The Staff then reviews and evaluates the report, and in the process may request the submitting party to provide additional data, experimental results, or other verification. The Staff may also conduct i
its own cross-checks by comparing the results of NRC-developed codes with the results obtained by the code being reviewed. When the Staff is satisfied that the computer code is sufficiently verified, the Staff i
eccepts the topical report containing the code. The Staff then sends the
- . submitting party an " acceptance letter", which infonns the party that the i
topical report has been accepted, and identifies the conditions and limi-tations for such Staff acceptance. Affidavits of Berggren, Answers 10 and 11, and Affidavit of Diab, Answers 14 and 15.
Certain topical reports in the CPSES FSAR which are not accepted by the Staff nonetheless are referenced by the Applicants and/or relied upon by the Staff in reaching its safety conclusions on CPSES, as set forth in the SER for CPSES. However, as Dr. Diab explains in his affi-davit, the Staff may accept the use of topical reports which have not been formally accepted for use in applications where the review of the topical report has proceeded to the point that any alterations to the report will not change the results obtained by using the report in its current status. Affidavit of Diab, Answer 16.
The Staff will discuss, in seriatim, each of the 28 topical reports and 3 computer codes identified by CFUR as requiring verification and NRC Staff acceptance, in light of the discussion above.
I t
. Report 1 (WCAP-7706)
Report 1, (WCAP-7706), "An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients" (February 1973), was' submitted by Westinghouse. The report provides a reliability ar.alysis of the Reactor
'l,.
Protection System in Westinghouse reactors. The Protection System utili-zes solid state logic. Report I uses fault tree techniques to assess the probability of systen failure due to random component failures. A quali-
~ ~ tative analysis of comon node failures in anticipated transients is also included in the report. Applicants referenced Report 1 in Sections 7.1.3.7, and 7.2.2.2.3 of the CPSES/FSAR. The report does not construct or validate any computer code. Affidavit of Thomas F. Dunning, at Answer 5.
Report I was not accepted by the NRC Staff. Affidavit of Thomas G. Dunner, at Answer 6.
However, as stated by Mr. Dunning, the Staff's evaluation of the adequacy of the solid state protection system for CPSES did not rely upon a consideration of Report 1.
Affidavit of Dunning, at Answer 7.
Reports 2 and 3 Report 2. WCAP-7769, " Overpressure Protection for Westinghouse Pressurized Water Reacters" (October 1971), was submitted by Westinghouse.
This topical report is an earlier version of Report 3 WCAP-7769, Revision 1, and has the sane subject matter as the current report. Report 2 is not I
referenced in the current FSAR for CPSES, and the Applicants do not rely on its results to support the CPSES/FSAR safety conclusions.
In addition, this report does not describe, construct, or verify any computer code.
Affidavit of Sansny S. Diab, Answer 8.
1
)
__ Report 3. WCAP-7769, Revision 1. " Overpressure Protection Westinghouse Pressurized Water Reactors" (June 1972) was submitted by Westinghouse and is a revised version of Report 2.
This report sets forth a methodology and set of assumptions for analyzing a Westinghouse pressurized water reactor's ("PWR") provisions for protection against cverpressurization of the cooling system. This report was referenced by the Applicants in the CPSES FSAR at Section 5.2.2.
This topical report does not describe, construct, or provide verification for any computer code. J_d., Answer 8.
Report 4_
Report 4, WCAP-7832 " Evaluation of Steam Generator Tube Tube Sheet, and Divider Plate Under Combined LOCA Pius SSE Conditions" (December 1973), was submitted by Westinghouse. The topical report evaluates the structural adequacy of the steam generator tubes, tube sheet, and divider plates of the Westinghouse Series 51 and Model D steam generators when subjected to simultaneous Loss of Coolant Acci-dent ("LOCA")andSafeShutdownduringEarthquake("SSE") loadings.
Affidavit of Jai R. N. Rajan, Answer 4.
Report 2 was referenced by the App 11 ants in Section 5.4.2 of the f
CPSES FSAR, but does not describe, construct or provide verification 1
for any computer codes. Jd.
This report was also accepted by the Staff. Notification of Staff acceptance was sent to Westinghouse on March 2, 1978. Jd.,seealso Affidavit of John S. Berggren, Answer 3, and attachments accompanying l
his affidavit.
[
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~ Report 5 Report 5, WCAP-79D7, "LOFTRAN Code Description" (October 1972), was submitted by Westinghouse and describes, constructs, and provides veri-fication for the LOFTRAN computer code. This code calculates pertinent plant parameters, including reactor coolant temperature and pressure, and core power levels. Report 5 is referenced by the Applicants in Section 5.2.3.2 of the CPSES FSAR. Affidavit of Sammy S. Diab, Answer 8.
This topical report has not been formally accepted in its entirety by the Staff, but has been approved only for analyses of Anticipated TransientsWithoutScram("ATWS").
Id.
The Staff believes that the Applicants' utilization of the LOFTRAN code to analyze certain plant variables during ATWS as referenced in the CPSES/FSAR Section 5.2.3.2 will adequetely predict plant behavior for tran-sients and accidents for which the Applicants utilize this code. CPSES Final Safety Evaluation Report ("SER") Section 15.1.2.
Id. The Staff's conclusion regarding the acceptability of the LOFTRAN code for use at CPSES in ATWS analysis is based on several factors.
As Dr. Diab explains, the Staff's calculational results of the loss of main feedwater ATWS in a Westinghouse PWR utilizing the RELAP3-B computer code came within acceptable limits of results obtained by Westinghouse using the LOFTRAN code. The Staff also conducted a loss of normal feedwater 4
analysis for a Westinghouse plant similar to Comanche Peak utilizing the NRC-sponsored code TRAC-PD2. The Staff compared the TRAC-PD2 results with the LOFTRAN results, and concluded that LOFTRAN conservatively predicted the pressurization transient.
In addition, a paper presented at the American Nuclear Society meeting in Denver during 1971 concluded that the LOFTRAN c,
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,-m_-
- - results compound favorably with transient tests on a Westinghouse reactor in Japan. Finally, Dr. Dieb states that Westinghouse presented to the Staff a comparison of the LOFTRAN code results, with the results obtained using WFLASH, RELAP, and other NRC, Babcock and Wilcox, and Combustion Engineering
~'
computer codes. Affidavit of Diab, at Answer 8.
The Staff concluded that the use of LOFTRAN to calculate certain plant variables for CPSES was accept-able.
Id.
Report 6 Report 6, WCAP-7908, "FAr.TRAN-A FORTRAN-IV Code for Thermal Transients In a U0 Fuel Rod" (June 1972), was submitted by Westinghouse. This report 2
describes and constructs the FACTRAN computer code, which calculates the transient temperature distribution in a cross section of a metal clad U02 fuel rod, and the transient heat flux at the surface of the cladding during certain transients and accidents. Report 6 was referenced by the Applicants in Section 15.2.3.2 of tne CPSES FSAR. Affidavit of Joseph J. Hononich, Answer 5; Affidavit of Diab, Answer 8.
Mr. Holonich explains in his affidavit that the Staff has completed its review of Report 6, but has not yet fonnally accepted the report j
because the Staff has not yet issued a formal Safety Evaluation Report i
for this topical report. Affidavit of Holonich, Answer 6.
The Staff has concluded that FACTRAN may be used in transient and accident analysis, subject to certain restrictions. The Staff's approval of Report 6 is based on a review of the report, the proposed Appendix to the report, and additional information submitted by Westinghouse. Mr. Holonich states that the Staff believes that FACTRAN will result in realistic analyses l
because the modeling concepts associated with the code are considered fuel rods. M.
acceptable for analyzing themal transients in U02 The Applicants' utilization of Report 6 for transient analysis was approved by the Reactor Systems Branch, as indicated by Dr. Diab.
Affidavit of Diab, Answer 8.
Mr. Diab affims that the Applicants' use of Report 6 is acceptable because of the Staff's conclusion that the results of analyses using FACTRAN wil1 not be appreciably altered by any change in methods that may be required by the Staff as a result of the completion of its review of Report 6. M.
Report 7 Report 7. WCAP-7907, " MARVEL. A Digital Computer Code for Transient Analysis of a Multiloop PWR System" (June 1972), was submitted by Westing-house. This report describes and verifies the MARVEL computer code for calculating the multiloop transient behavior of pressurized water reactors.
MARVEL simulates two reactor cooling loops, including the steam generators and associated systems. Affidavit of Diab, at Answer 8.
This report was not referenced by the Applicants in either Sections 6.2 or 6.3 of the FSAR for CPSES. Affidavit of Shum, Answers 5 and 7; Affidavit of Diab, Answer 8.
Therefore, the Staff did not rely on Report 7 in developing its safety conclusions in the CPSES SER Sections 6.2 and 6.3.
Report 8 Report 8. WCAP-7950, " Fuel Assembly Safety Analysis for Combined Seismic and Loss of Coolant Accident" (July 1972), was submitted for Staff approval by Westinghouse Electric Corporation (Westinghouse).
l This topical report analyzes fuel assembly response to combined seismic and loss-of-coolant loads. Affidavit of Ralph 0. Meyer, Answer 7.
As stated by Dr. Meyer, the Staff has accepted this report for use in license applications. M.;seealsoAffidavitofJohnS.Berggren, Answer 3, and attachments thereto. However, Report 8 applies only to Westinghouse 15x15 fuel assemblies, which will not be used at CPSES.
Affidavit of Meyer, Answer 7.
Accordingly, this report is not used as
".a reference in Section 4.2 of the Applicants' FSAR. Rather, Report 10, which is listed below and is applicable to Comanche Peak, is cited in Section 4.2 of the FSAR. M. Since Report 8 was not referenced by Applicants in CPSES FSAR Section 4.2, the report is not necessary to support the Staff's safety conclusions in the SER.
Report 8 is listed in the references for Section 3.7N of the Appli-cants' FSAR. However, it is not cited in the text for this section and is not necessary to support the conclusions of that section. M.
Report 9 Topical Report WCAP-8163 describes the behavior of reactor coolant pumps under the condition of loss of coolant accident (LOCA), and was referenced by the Applicants in Section 5.4.1.1. of the FSAR for the Comanche Peak Steam Electric Station. Affidavit of Barry Elliot, Answer 5.
This topical report, as referenced by the Applicants in Section 4.5.1.1 does not construct or verify any computer code. M.atAnswer6. More-over, the safety analysis in this topical report which was referenced by the Applicants in CPSES FSAR Section 5.4.1.1 was independently veri-
' fled by the NRC Staff by an analytical stress analysis perforted by the Staff. Ld. at Answer 5.
Report 10
'[.,
Report 10. WCAPs-8236 (proprietary) and 8288 (non-proprietary),
" Safety Analysis of the 17x17 Fuel Aste.bly for Combined Seismic and Loss of Coolant Accident" (October 1973), was submitted by Westinghouse.
- This topical report analyzes the fuel mechanical response of Westing-house fuel assemblies to combined loss-of-coolant-accident (LOCA) and safe-shutdown-earthquake (SSE) loads. Affidavit of Meyer, at Answer 7.
Report 10 is cited in Section 4.2 of the Applicants' FSAR for CPSES.
According to Dr. Meyer, the Staff detemined that the methodology pre-sented in this report is acceptable for analyzing 17x17 fuel assemblies with eight spacer grids, as will be used at CPSES.
Id_.
Furthermore, this report was accepted by the NRC Staff, contrary to CFUR's position on this topical report. Westinghouse was notified of Staff acceptance of Report 10 on February 6,1979.
Id., see also Affidavit of Berggren, Answer 3, and attachments accompanying that affidavit.
Report 11 Report 11, WCAP-8252, " Documentation of Selected Westinghouse Structural Analysis Computer Codes, Revision 1" (July 1977), describes and presents representative verification of eleven Westinghouse structural analysis computer codes used in the analysis and design of mechanical systems and components. As Dr. Rajan states in his affi-
davit, representative computer codes described, constructed and veri-I fied in this report include a code to derive a pressure-time analysis of blowdown, and a code which evaluates structural elements using a dynamic response analysis.
Report 11 was referenced by the Applicants
'l,,
in Section 3.6N of the FSAR for CPSES. Affidavit of Jai Raj N. Rajan, Answer 4.
l This topical report was evaluated and fomally accepted by the
- Staff. Notification of NRC Staff occeptance was sent to Westinghouse on March 2, 1978. M.,seealsoAffidavitofJohnS.Berggren,and attachments to his affidavit.
Report 12 Report 12, WCAPs-8278 (proprietary) and 8279 (non-proprietary)
" Hydraulic Flow Test of the 17x17 Fuel Assembly" (February 1974), was submitted by Westinghouse. This topical report analyses hydraulic flow tests for fuel assembly fretting and wear in 17x17 assemblies and is cited in Sections 4.2 and 4.4 of the Applicants' FSAR. The report describes the experimental results of the hydraulic flow tests, and analyzes those results. Affidavit of Meyer, at Answer 7.
In his affi-davit, Dr. Meyer states that the Staff has detemined that this report, along with additional Westinghouse test results, demonstrate that fuel assembly wear and fuel rod fretting for 17x17 fael rod assemblies are within acceptable limites. The Staff did not rely on any computer codes referenced in this report in reaching its safety conclusions regarding this report. M.
I
(
. The Staff accepted Report 12 and a letter notifying Westinghouse of Staff approval was sent on May 30, 1979. Jd.;seealsoAffidavit of John S. Berggren, Answer 3, and attachments accompanying his affi-
'l,.
davit.
Report 13 Report 13, WCAP-8330 " Westinghouse Anticipated Transients Without
" Trip Analysis" (August 1974), was submitted by Westinghouse and presents a methodology and 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 dis-charge rates from safety valves, and moderator temperature coefficient.
Affidavit of Saniqy S. Diab, Answer 8.
This topical report does not describe, construct and present verifi-cation for any computer code. Jd.
Report 14 Report 14, WCAP-8424, "An Evaluation of Loss of Flow Accidents Caused By Power System Frequency Transients in Westinghouse Pressurized Water Reactors, Revision 1" (June 1975) was submitted by Westinghouse for NRC Staff approval. According to James E. Knight, who is a Section Leader for the Power Systems Branch of the Staff, this report addresses coolant flow transients caused by electrical system frequency disturbances, and develops a maximum permissible credible decay rate for a utility grid.
Affidavit of James E. Knight, Answers 1 and 4.
. The Applicants referenced Report 14 in Section 15.3 of the FSAR for CPSES. See Affidavit of Knight, Answer 4, and Affidavit of Diab, Answer 8.
However, the Staff's evaluation and approval of Applicants' analyses of decreased cooling flow transients in the CPSES SER, Section 15.2.2.6 did not rely on Report 14. Therefore, Report 14 was not necessary to the Staff's safety evaluation of CPSES.
Report 15 Report 15, WCAPs-8691 (proprietary) and 8692 (non-proprietary),
" Fuel Rod Bow Evaluation" (December 1975), was submitted by Westinghouse, and analyzes fuel rod bowing for 17x17 fuel rod designs. Report 15 is cited in Section 4.2 of the Applicants' FSAR, and does not describe, construct, or provide verification for any computer code. Affidavit of Meyer, Answer 7.
As Dr. Meyer states in his Affidavit, a revised version of this report has been submitted to the NRC and is currently under review.
In the meantime, the Staff has utilized an interim, approved method of calculating rod bowing which is applicable to CPSES. The analysis of rod t, swing effects at CPSES were perfonned using the interim method; therefore,theStafffoundtheanalysisacceptable(SER4.2.3).
Id.
Report 16 Report 16,~ WCAPs-8720 (proprietary) and 8785 (non-proprietary),
" Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations" (October 1976), was submitted by Westinghouse. According to Dr. Meyer, Report 16 describes several revisions to a computer code
. called PAD-3.3, which is used by Westinghouse to analyze fuel rod thermal performance in reactors with designs similar to Comanche Peak.
The topical report is cited in Section 4.2 of the Applicants' FSAR.
Affidavit of Meyer, Answer 7.
The Staff approved the use of Report 16 with several restrictions.
g.; see also Affidavit of John S. Berggren, Answer 3, and attachments accompanying his affidavit. In the case of CPSES, the Staff concluded
.that these restrictions were not significant for first-cycle operation at full power (SER 4.2.2).
Therefore, the use of Report 16 in Section 4.2 of the CPSES/FSAR is acceptable.
Report 17 Report 17, WCAP-9168 (proprietary) and 9169. (non-proprietary)
" Westinghouse Emergency Core Cooling System Evaluation Model-Modified October 1975 version" (September 1977), are the proprietary and non-proprietary versions of a description of the Westinghouse Emergency Core Cooling System ("ECCS') evaluation model. Contrary to CFUR's claims, Report 17 does not construct and preseric verification for any computer codes. Affidavit of Diab, Answer 8.
This report is not referenced by the Applicants in the current FSAR for CPSES, and the Applicants do not rely on its results to support the CPSES/FSAR's safety conclusions.
Id.
e Report 18 Report 18 WCAPs-9179 (proprietary) and 9224 (non-proprietary),
" Properties of Fuel and Core Component Materials" (September 1977),
- was submitted by Westinghouse. This topical report documents material properties values and correlations used in the design of fuel assembly and core components for Westinghouse pressurized water reactors. Affi-davit of Meyer, Answer 7.
This report does not describe, construct.
'J.,
or provide verification for any computer codes. Rather, it presents data and correlations used in other computer codes, such as Report 16 above.
Id.
~~
This topical report is not yet approved by the Staff. Where property value correlations are required in computer codes, the Staff has reviewed and approved the significant correlations during the review of those codes.
Where the property value correlations are not presented in other topical reports, as was the case for hafnium in Section 4.2 of the Applicants' FSAR, the Staff has reviewed and approved the us_e of those specific cor-relations on the applicable docket. Therefore, approvel of Report 18 is not required for the CPSES safety evaluation.
I d_.
Report 19
" Report 19", as Applicants have identified the MRI/STARDYNE computer code, is not a topical report. The MRI/STARDYNE code is not a subject of any topical report review procedure. Affidavit of Frank Rinaldi and John P. Matra, Answer 6.
As stated by Frank Rinalrii and John P. Matra, MRI/STARDYNE is a ground purpose computer program designed to analyze linear elastic structural models.
Id., at Answer 5.
The Staff evaluated MRI/STARDYNE at the CPSES Structural Audit conducted on the premises of Gibbs and Hill, Inc., from May 11 to 15,
- 1981.
Id... at Answer 7.
This code was evaluated by the Staff to deter-mine if it was acceptable in accordance with the criteria set forth in Section 3.8.1 of the Standard Review Plan, as presented in Answer 7 of the Affidavit of Rinaldi and Matra. The Staff t.oncluded that MRI/
STARDYNE is a public domain computer program and it has a sufficient history of use to justify its applicability and validity without further verification. M. On the basis of the history of verification
. described above, the Staff accepted the Applicants' analyses which used the STARDYNE code. M. at Answer 4.
Report 20
" Report 20", like " Report 19", is not a topical report, but is the Applicants' identification of the SCONY computer code. SCONY is also not a subject of any topical report review process. Affidavit of Rinaldi and Matra, Answer 6.
SCONV is an in-house program of Gibbs and Hill. The program uses convolution integration to solve general dynamic problems for linear elastic systems with the support time history excitation R*S(t), where R is the scaling factor, and S(t) is the time-dependent support accelera-tion.
M., Answer 5.
SCONV was evaluated by the Staff at the May 11 to 15, 1981 Structural Audit where MRI/STARDYNE was evaluated, using the same criteria set out in Answer 7 of the Affidavit of Rinaldi and Matra. The Staff concluded that SCONV was acceptable for use in CPSES, since the results of a series of test problems obtained from the SCONV code were substantially identical to results obtained from the CSMP computer code. CSMP is a public domain
__- program offered by IBM as part of their scientific subroutine programs package. M. On that basis, the Staff accepted the Applicants' analyses which were based on SCONV. M., at Answer 4.
Report 21
" Report 21", as the SPECTRA computer code is identified by the Appli-cants, is not a topical report. SPECTRA is a Gibbs and Hill pcst-processor program which interpolates or extrapolates stored data, yielding spectral accelerations which correspond to a set of specified pairs of frequencies and associated model damping values. Affidavit of Rinaldi and Matra, Answer 5.
SPECTRA is not the subject of any topical report review process.
Id., at Answer 6.
24 T
The Staff ev61uated the SPECTRA code at Gibbs and Hill along with the MRI/STARDYNE and SCONY computer codes. The SPECTRA code was evaluated for conformance to the criteria set forth in Section 3.8.1 of the NRC Standard Review Plan, which is set forth in applicable part in Answer 7 of the Affidavit of Rinaldi and Matra.
The Staff concluded that SPECTRA had been suitably verified, by comparing the results of SPECTRA with results derived from Regulatory l
Guide 1.60 calculations. M.
Report 22 Report 22, WCAP-8768, " Safety-Related Research and Development for I
c Westinghouse Pressurized Water Reactors, Program Sumaries" (no date), is l
a status report periodically submitted by L'estinghouse. The topical report describes ongoing safety-related research and development for Westinghouse i
i ls
(
PWRs, and gives a status report on program progress. Affidavit of Diab, Answer 8.
Report 22 was referenced in Section 1.6 of the CPSES/FSAR for ir.formational purposes. CFUR believes that Report 22 constructs and presents verification for a computer code. This belief is in error.
Report 22 does not describe, construct, and provide verification for any computer codes. Affidavit of Diab, at 8.
Report 23 Report 23, WCAP-7956, "THINC-IV, An Improved Program for Themal,
Hydraulic Analysis of Rod Bundle Cores" (October 1973), was submitted by Westinghouse. The topical report describes, constructs, and pre-sents experimental verification data for the THINC-IV computer code.
THINC-IV is used for the study-state, thermal-hydraulic analysis of reactor cores. Affidavit of Holonich, Answer 5.
As Mr. Holonich indicates, this computer code was referenced throughout the CPSES FSAR, and was used as background information in his review of Section 4.4 of the FSAR for CPSES.
I d_.
Contrary to CFUR's assertion, Report 23 was formally accepted by the NRC Staff. Westinghouse was notified of the fomal acceptance of Report 23 on April 19, 1978.
Id., see also Affidavit of John S. Berggren, Answer 3, and attachments thereto.
Report 24 C
Report 24. WCAPs-8170 (proprietary) 8171 (non-proprietary, "Calcula-tional Model for Core Reflooding After a LOCA" (July 1974), were sub-mitted by Westinghouse. This report describes, constructs, and presents verification for the WREFLOOD camputer code. This code simulates the reactor core reflood hydraulics behavior after the occurrence of a LOCA.
WREFLOOD calculates several parameters, including the core coolant reflooding rates, core coolant pressures, and core inlet temperatures.
e-Report 24 was referenced in Section 15.6.5 of the FSAR for CPSES. Affi-davit of Diab, Answer 8.
Wnile CFUR states that Report 24 has not yet been " formally accepted,"
. in fact, the Staff has formally accepted this report, as part of the Staff approval of the ECCS model. Notification of NRC Staff acceptance was sent to Westinghouse on May 30, 1975.
Id_., see also Affidavit of John S. Berggren, Answer 3, and attachments thereto.
Report 25 Report 25, WCAPs-8200 (proprietary) and 8261 (non-proprietary),
"WFLASH, a FORTRAN-IV Computer Program for Simulation of Transients in a Multi-Loop PWR" (June 1974), was submitted by Westinghouse as part of its ECCS evaluation model. WFLASH simulates the core themal hydraulic behavior after a small break LOCA, and calculates pertinent plant parameters such as reactor coolant pressure and temperature.
Report 25 describes, constructs, and presents verification for the WFLASH computer code. Affidavit of Diab, Answer 8.
CFUR incorrectly asserts that this report has not been verified, and accepted by the NRC Staf f.
In fact, Report 25 was accepted by the Staff as part of its approval of the ECCS evaluation model. Notification of Staff acceptance was sent to Westinghouse on May 30, 1975, as part of the letter notifying Westinghouse of NRC Staff approval of the ECCS
model.
Id_., See also affidavit of John S. Berggren, Answer 3, and attachments thereto.
Report 26 Report 26, WCAPs-8301 (proprietary) and 8305 (non-proprietary),
"LOCTA-IV Program, LOCA Transient Analysis" (June 1974), was submitted by Westinghouse for approval as part of its Westinghouse ECCS evaluation
model. This topical report describes, constructs and presents verifi-cation for the LOCTA-IV code. The LOCTA-IV computer code calculates several parameters, including the peak clad temperature of a fuel pin as a result of small and large break LOCAs. Applicants referenced this report in CPSES FSAR Section 15.6.5. Affidavit of Diab, Answer 8.
CFUR states that this report has not been verified, and accepted by the Staff. However, Dr. Diab points out that Report 26 was accepted by the Staff as part of its review and approval of the Westinghouse ECCS l
evaluation model. Notification of NRC Staff acceptance was sent to Westinghouse on May 30, 1975.
,Id., see also Affidavit of John S.
Berggren, Answer 3, and attachmentI thereto.
Report 27 Report 27,WCAPs-8302(proprietary)and8306(non-proprietary),
" SATAN-IV Program, A Comprehensive Space-Time Dependent Analysis of l
LOCA" (June 1974) is a Westinghouse report submitted as part of the Westinghouse ECCS evaluation model.
In his affidavit, Dr. Diab explains that Topical Report 27 describes, constructs, and presents verification for the SATAN-IV computer code; this code simulates the reactor core i
31 themal hydraulic behavior during the blowdown phase of a large break LOCA. It calculates several parameters, including core inlet flow and j
entholpy core pressure, core power, and core cross-flow parameters.
.,i '
Affidav'it of Diab, Answer 8.
Contrary to CFUR's assertions in their Supplement to Petiton for l
l Leave to Intervene (May 7,1979), and their Report of CFUR's Position on Each Contention (April 10,1980), Report 27 was evaluated and accepted
. by the Staff as part of its approval of the Westinghouse ECCS evaluation model. Notification of Staff acceptance was sent to Westinghouse on May 30, 1975.
Id., see also Affidavit of John S. Berggren, Answer 3, and attachments accompanying his affidavit.
Report 28 Report 28, WCAP-8326, " Containment Pressure Analysis Code (C0CO)"
(July 1974), was submitted by Westinghouse as part of its Emergency Core Cooling System evaluation model. The COC0 computer code predicts the temperature and pressure response of a reactor building to a loss of coolantaccident("LOCA"). The code analyzes the effectiveness of the engineered safeguards system, including internal and external sprays, containment venting, and ventilation for coolers. This report was referenced in Section 15.6.5.3 of the CPSES/FSAR.
CFUR contends that Report 28 was not " suitably verified" and " formally accepted" by the NRC Staff. This contention is in error. Report 28 was accepted by the Staff as part of its acceptance of the ECCS evaluation model.
Notification of Staff acceptance of Report 28 was sent to Westinghouse on
~
s
' May 30, 1975.
M., see also Affidavit of John 5. Berggren, Answer 3, and attachments thereto.
Report 29 a.
Report 29. WCAP-8339, " Westinghouse ECCS Evaluation Model - A Semary" (July 1974), describes the LOCA modeling of the Westinghouse ECCS evaluation model methodology. The report also describes the appro-priate applications and limitations of the ECCS model. The report was
~
~
referenced by Applicants in Section 15.6.5.3 of the FSAR for CPSES.
Affidavit of Daib, at Answer 8.
Dr. Diab states in his affidavit that Report 29 does not describe, construct or present verification for any computer codes. M. He also points out that Report 29 has been accepted by the Staff as part of the Staff approval of the ECCS evaluation model. Westinghouse was notified of the NRC Staff acceptance of Report 29 in an acceptance letter dated May 30, 1975.
Id., see also Affidavit of John S. Berggren, Answer 3, and attachments accompanying his affidavit.
Report 30 and 31 Report 30, WCAPs-8340 (proprietary) and 8356 (non-proprietary),
" Westinghouse ECCS Plant Sensitivity Studies" (July 1974), and Report 31, WCAPs-8341 (proprietary) and 8342 (non-proprietary), " Westinghouse ECCS Evaluation Model Sensitivity Studies" (July 1974), are sensitivity studies conducted by Westinghouse to partially validate the Westinghouse ECCS evaluation model. As Dr. Diab explains in his affidavit, these sensitivity analyses examine the degree of change in calculated parameter results r
obtained from other computer codes which are a part of the ECCS model, which are due to variations in various other parameters, including nodali-zation, time steps, break size and location, and simple versus multiple failures in the coolant system. Affidavit of Diab, Answer 8.
These a*
reports do not construct and present verification for computer codes which are presented in these reports. However, Reports 30 and 31 present partial verification for computer codes described in other topical reports
. submitted as part of the Westinghouse ECCS evaluation model.
Id.
Reports 30 and 31 were accepted as part of the NRC Steff approval of the ECCS model, contrary to CFUR's claim. Westinghouse was notified of the NRC Staff acceptance of Reports 30 and 31 in an acceptance letter dated May 30, 1975.
M., see also Affidavit of John S. Berggren, Answer 3, and accompanying attachments thereto.
Report 32 Report 23,WCAPs-8471(proprietary)and8472(non-proprietary),
" Westinghouse ECCS Evaluation Model - Supplementary Information" (January 1976), was submitted by Westinghouse and presents Westinghouse responses to NRC Staff questions regarding the ECCS evaluation model.
Affidavit of Diab, Answer 8.
As Dr. Diab states in this affidavit, Answer 14, the NRC Staff may request additional informtion to enable it to properly evaluate a topical report.
In this case, this informa-tion was provided in Report 32. Affidavit of Diab, Answer 8.
Report 32 was referenced by the Applicants in Section 15.6.5.3 of the FSAR for CPSES. This report does not describe, construct or present verification for a computer code presented in this topical report. Id.
Report 32 was evaluated and accepted by the Staff as part of its review and approval of the ECCS model. Notification of Staff approval of Report 32 was sent to Westinghouse on hay 30, 1975.
Id., see also Affidavit of John S. Berggren, Answer 3, and attachments accompanying his affidavit.
Conclusion Regarding Contention 2 In view of the discussion of the 28 topical reports and 3 computer codes identified by CFUR in its May 7,1979 Supplement for Leave to Intervene and its April 10, 1980 Report of CFUR's Position on Each Con-tention, it is apparent that CFUR has identified topical. reports which have in fact been approved by the Staff; reports which do not construct or verify computer codes; reports which were not referenced by the
/
Applicants in the CPSES FSAR; and reports which were not relied upon by the Staff in reaching its safety conclusions for CPSES, as indicated in the CPSES SER. For the topical reports not yet approved, the Staff concluded that their use in CPSES was adequately justified. Finally, CFUR identified three computer codes which are not the subject of topical reports.
The Staff believes that the Applicants' Motion and documents filed in support thereof, together with the affidavits and documents submitted by the Staff, establish that there is no genuine issue of material fact as to Contention 2.
The Staff believes that the Applicants have met the burden of obtaining sumary disposition on Contention 2 as a matter of law and accordingly, the Staff believes that summary disposition of Contention 2 is appropriate.
i j
Contention 7 The Staff believes that Contention 7 raises no genuine issue of material fact. After reviewing the Applicants' Motion and the accom-panying documents, inculuding the supporting affidavits, the Staff believes that the " Statement of Material Facts As To Which There Is No Genuine Issue Regarding Contention 7" (hereafter " Statement of Material Facts") is correct (Affidavit of Robert C. Stewart at 2 and Affidavit of Owen Thompson at 3), and the Staff supports Applicants' position that summary disposition of Contention 7 should be granted.
Contention 7, as admitted by the Licensing Board in its June 16, 1980 Order, supra, states as follows:
" Applicants have failed to adequately evaluate whether the rock 'overbreak' and subsequent fissure repair using concrete grout have impaired the ability of Category I structures to withstand seismic disturbances.
As the basis for this contention, CFUR cites "I.E. Inspection Report 75-05," and states that:
" Evidence of rock 'overbreak' around the perimeter of the excavations for the containment building due to blasting wrs reported as a deficiency in I.E. Inspection Report 75-05."
he " Supplement to Petition for Leave to Intervene By Citi-zens For Fair Utility Regulation (CFUR)," May 7,1979 (at 22).
CFUR contends that the specifications or procedures used in repairing the rock overbreak are inadequate, stating that:
"CFUR has reason to believe that loose rock material was thrown into the over-excavation prior to the pouring of I
concrete." See "CFUR Answers to NRC Staff's First Set of Interrogatories to and Request to Produce From CFUR,"
March 11,1981, at 8 (Answer to Staff Interrogatory C7-12).
CFUR also:
... suspects that to save money throu5h the use of less concrete, Applicants incorporated loose rock and other material into 1
1
. the foundation of CPSES." (Emphasisadded). See "CFUR's Supplement to Answers to Applicants' First Set of Interrogatories to CFUR and Requests To Produce " May 8. 1981, at 6 (Answer to Applicants' Interrogatory 112e).
As the basis for its suspicions, CFUR has cited interviews which it says it conducted in May 1980 with George Clancy and C. A. Thetford, "ex-employees of Brown and Root." See CFUR's Answers to NRC Staff's First SetofInterrogatories, supra,at8(AnswertoStaffInterrogatoryC7-13)
and CFUR's Response to NRC Staff Second Set of Interrogator'es, May 22, 1981, at 12 (Answer to Staff Interrogatory C7-23).
A.
The Rock Overbreak and Fissures Were Froperly Repaired NRC Staff inspections indicate that during initial blasting of previously undisturbed rock in excavating for the Unit I and Unit 2 Containment Buildings, overbreak occurred in adjacent rock. (Affidavit ofRobertC.Stewartat3). Surface rcck, varying in depth from one (1) to six (6) feet, was shattered and fissured on the perimeter of both the Unit 1 and Unit 2 Containment Buildings and the areas of the Safe-guards, Fuel, and Auxiliary Buildings. The Applicants reported this matter in accordance with 10 C.F.R. I 50.55(e). (Id.)
The incidents of rock overbreak described above were repaired in accordance with Gibbs & Hill, Inc, specifications and Brown & Root, Inc.
procedures and procedures developed by Mason-Johnston Associates, Inc.
(H. )
During independent inspections perfonned from 1975-1978, the cogni-zant NRC inspector (Mr. Robert C. Stewart) periodically observed activi-ties involved in the repair of the rock overbreak, namely, cleaning of rock surfaces, placement of concrete, and grouting.
(H. ) Such observa-
tions generally involved discussions with the cognizant Mason-Johnston Associates, Inc. field geologist, Herbert C. Crowder, as well as a review of the applicable Gibbs & Hill, Inc. specifications, the Brown and Root, Inc. and Mason-Johnston procedures and Brown and Root, Inc. inspection records, all of which were maintained in the on-site QA/QC document control vault.
(,I_d. ) Mr. Stewart's personal observation and evaluation indicated that the reapir of the rock overbreak, including the placement
. of concrete and grouting, were made in accordance with the applicable specifications and procedures.
(M.,at4).
The overexcavation which caused the rock overbreak also resulted in fractures in the rock on which or against which the foundations of the Unit 1 and Unit 2 Containment Buildings, Safeguards Buildings and the Fuel Building were pourad.
(M.) Such fractures were properly identified, mapped for grouting and repaired in accordance with Gibbs l
& Hill, Inc. specifications and Brown and Root, Inc. and Mason-Johnston Associates, Inc. procedures.
CFUR has implied that " loose rock and other foreign material" resultedinthe"formationofairpocketsintheCPSES(ComanchePeak SteamElectricStation) foundation." See CFUR's Supplement, supra, at 7 (Answer to Interrogatory 115).
CFUR has not presented any evidence to confirm its suspicions.
CFUR has merely stated that an interview with "ex-Brown and Root employees," George Clancy and C. A, Thetford, provides the basis for its suspicions. Although CFUR's interview with George Clancy may have provided the basis for CFUR's suspicions, Mr. Clancy failed to provide any infomation to NRC Staff inspectors about loose rock
- material being through into the overexcavation, when he was inter-viewed as part ')f an NRC Staff investigation of allegations made by:
Mr. Clancy concerning other deficiencies in Comanche Peak construction.
(M.,at6-7). Moreover, none of the NRC Staff's inspections of Comanche Peak construction lend credence to CFUR's suspicions that any loose rock was thrown into th'e excavation prior to the placement of concrete, resulting in air pockets in the Comanche Peak foundations.
~
(StewartAffidavit,at5-7). These inspections revealeif only that a chainfall was left in the Unit I containment base mat (M., at 5) and that there were two voids in the foundations of seismic Category I structures.
(M. ) NRC inspectors monitored and observed corrective action takeri by the Applicants to repair incidents of foreign material and voids. '(M.,
~
at6). NRC inspections do not confim the existence of any foreign material, s
loose rock or voids which would have an kffect on the static, dynamic and-R
~
o
,y engineering properties of these structures. (M.,at6:
l B.
The Rock Overbreak and Subsequent Repair Using Concrete-and Grout Did Not Impair the Ability of Se,jsnic. Category I Structures to Withstand Seismic Disturbances--
/
-~.
The CFUR also alleges that:
"Since loose rock, air pockets and other foreign material have different structural properties than.dcas solid con-crete, their incorporation and fomation in the foundation of CPSES will affect the static,' dynamic <andfengineerir;g properties of the structure.", CFUR's Supplement, ju_prg, '
at 7 (Answer to Interrogatory 116).
l
/
/
First of all, as shown above, there is no basis for CFUR's sssurrption that there is loose rock, air' pockets and-other ' foreign materials in"
- ,J
. c Comanche Peak foundations. The NRC Staff has evaluated the stability.
of subsurface materials and foundations of the Comanche Peak plan,t.
See s.
f' s
/, '
F
Section 2.5.4 of the Staff's " Safety Evaluation Report Related to the Operation of Comanche Peak Steam Electric Station, Units 1 and 2,"
(NUREG-0797, July 1981),andAffidavitofOwenThompson,at3-4. Based on this evaluation, the Staff concluded that 1) the Applicants have adequately determined the static and dynamic properties of the materials underlying the site and the response of these materiais to dynamic loading; 2) the rock supporting the plant foundations will not modify (i.e.neitherattenuatenoramplify)thedesignearthquakemotionbecause
~
such modification would occur only if the foundations were soil-supported; and 3) there are no areas of actual or potential surface or subsurface subsidence, uplif t or collapse resulting from man's activities.
(M.,
at 3-4).
The Staff also evaluated the effects on static and dynamic founda-tien stability. of replacing undisturbed limestone and claystone foundation rock with concrete having a compressive strength of 2500 pounds per squareinch'(psi),attwenty-eightdays,asprescribedfortherepairof areas where 'there was' rock overbreak.
(M., at 4 ).
~
Iq this regard, the procedures for repair of the rock overbreak require'd removal of broken rock and its replacement with concrete having a minimum compressive strength of at least 2500 psi, at twenty-eight days.
(M.). By comparison, the compressive strengths of the e
undisturbed limestone and claystone foundation materials are about 200 tons per square foot (TSF) and 20 TSF (i.e., 2770 psi and 277 psi),
respectively. (M. ) Much higher safety factors arc obtained if the foundation bearing-materials are assumed to be limestone.
(M. ) There-m fore, replacing foundation rock with concrete does not compromise static 4
r -'
-i e..
2.' or dynamic fhundation stability because the replacement concrete has a
/
much higher strength than the material, assumed in the design.
(i.e.,
claystone).
(Id.).
~
The'5taff als'o evaluated the effects on static and dynamic founda-
. l,,,
tion stability of' cracks and fissures resulting from blasting and of the grouting. procedures specifked for siibsequent repairs of cracks and fiss'bris. (Ijd.at5).
X ' According to classical foundation engineering bearing capacity
~ theory, there is a stressed elastic zone, enclosed by the foundation
~
base and ima'ginary planes extending downward from the foundation edges, atanangleof6with.thehorizontal,sothatthesurfa:esmeetbelow Qd.)
(The~ angle $ is the angle of dlthecenterofthe1 foundation. internal friction'for the foundation bearing material and, for the 7.- Comancheheaksite,isinexcessof40 degrees). Qd.) The' cracks and fissures.beneath safety related structures at the r Comanche Peak site that have been identified by the Applicants are i ~~ c early all within this zone. (H.) Consequently, these cracks and fissures will be closed by the imposed stresses and have no adverse influence on static foundation stability or on the elastic properties F of.the foundation materials in close proximity to the base of the struc-V tures. (Id.) The procedures specified for the repair of overbreak areas con-taining cracks'and fissures required cle'aning, inspecting and backfilling y with concrete, followed by grouting. (M.) These procedures assure that any cracks, voids and fissures of any significant size have been filled with concrete. (Id.,at5-6). The subsequent grouting beneath ,~ d ,\\ f e*- g
- the concrete provides added confirmation that no significant voids or cracks remain. (M., at 6). The specified procedures for crack and fissure repair are adequate to assure that no significant voids or unfilled cracks remain in the foundation bearing material. (Id.) Therefore, the cracks and fissures resulting from blasting and the subsequent repair of the cracks and fissures will have no adverse impact on foundation static or dynamic stability because the treated foundation bearing materials in the elastic zone are not susceptible to displacement. In sum, the specified procedures or disturbed for repair of the rock overbreak and cracking required replacement of disturbed material by concrete having a strength greater than the design strength of the undisturbed foundation bearing material and the procedures required for filling fissures, cracks and joints, as well as the str ss conditions ir. these areas, assure that no significant fissures, cracks or joints remain. (M.,at7). Therefore, the ability of the repaired foundation bearing materials to withstand seismic disturbances will not be impaired. (M.). C. Conclusion Regarding Contention 7 In contrast, nowhere in CFUR's filings to date or in its answers to the Staff's and Applicants' interrogatories, has CFUR provided any factual basis for the allegations in Contention 7. CFUR has not pre-sented any evidence disputing that the repair of rock overbreak and fissures was properly perfomed in accordance with appropriate speciff-C cations and procedures. All CFUR has to offer is its suspicions that such repairs were not properly performed. Similarly, CFUR has not con-troverted the evidence i. hat the rock overbreak and fissure repair will not impair the ability of seismic Cateogry I structures to withstand seismic disturbances. The Staff believes that the Applicants' Motion and documents filed in support thereof, along with the affidavits submitted by the Staff, establish that there is no genuine issue of material fact as to Contention 7. The Staff believes that the Appli-cants have met the burden of obtaining sumary disposition on Contention 7 as a matter of law and accordingly, the Staff believes that summary disposition of Contention 7 is appropriate. CONCLUSION For the reasons stated above, the Staff supports Applicants' conclusion that no genuine issues of material fact remain to be resolved with respect to Contentions 2 and 7. Therefore, Applicants' motion for summary disposition of Contentions 2 and 7 should be granted and those contentions should be dismissed as matters in con-troversy in this proceeding. Respectfully submitted, uken /k% cme / l l Marjorie Ulman Rothschild Counsel for NRC Staff l l N Geary S. Mizuno I Counsel for NRC Staff Dated at Bethesda, Maryland this 12th day of February,1982 j,
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD _ ) In the Matter of Docket Nos. 50-445 TEXAS UTILITIES GENERATING COMPANY, ET AL. 50-446 ) ~~- ) (Comanche Peak Steam Electric Station, ) Units 1 and 2) CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF ANSWER SUPPORTING APP MOTION FOR
SUMMARY
DISPOSITION OF CONTENTIONS 2 A captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk through deposit in the Nuclear Regulatory Commission's internal mail system, this 12th day of February, 1982: Mrs. Juanita Ellis Marshall E. Miller, Esq., Chairman
- President, CASE r
Administrative Judge 1426 South Polk Street Atomic Safety and Licensing Board Dallas, TX 75224 U.S. Nuclear Regulatory Comission Washington, DC 20555 David J. Preister, Esq. Assistant. Attorney General Dr. Kenneth A. McCollom Environmental Protection Division Administrative Judge P. O. Box 12548, Capital Station Dean, Division of Engineering, Austin, TX 78711 Architecture and Technology Oklahoma State University Mr. Richard Fouke Stillwater, OK 74074 1668-B Carter Drive Arlington, TX 76010 Dr. Richard Cole, Administrative Judge
- Atomic Safety and Licensing Board Nicholas S. Reynolds, Esq.
U.S. Nuclear Regulatory Comission Debevoise & Liberman Washington, DC 20555 1200 17th Street, N.W. Washington, DC 20036 J. Marshall Gilmore, Esq. 1060 W. Pipeline Road Hurst, TX 76053 P i p.
l c - Docketing and Service Section (1)* Atomic Safety and Licensing Board Office of the Secretary Panel
- U.S. Nuclear Regulatory Comission U.S. Nuclear Regulatory Comission Washington, DC 20555 Washington, DC 20555 Atomic Safety and Licensing Appeal Panel (5)*
U.S. Nuclear Regulatory Comission Washington, DC 20555 i
- w ha~ &Nacu Marjorie Ulman Rothschild Counsel for NRC Staff Geafy 5. Mizuno Counsel for NRC Staff C
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