ML20134F311

Requests That Encl Draft of Info Notice Be Reviewed to Ensure Technical Info Is Accurate Re Nonconservative Errors & Changes in Siemens Power Corp Large Break LOCA Evaluation Model & Compliance w/10CFR50.46(a)(3)

ML20134F311
Person / Time
Site:	Framatome ANP Richland
Issue date:	02/05/1997
From:	Chaffee A NRC (Affiliation Not Assigned)
To:	Curet H SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
References
NUDOCS 9702070355
Download: ML20134F311 (5)

Text

_ _ . . ._. __ __ _ _ _

,, Feinnry 5,'_1997

. H. D. Curet. Manager l Product Licensing Siemens Power Corporation 2101 Horn Rapids Road 3 ~ / 78 P.O. Box 130 Richland. Washington 99352-0130  ;

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SUBJECT:

REQUEST FOR A TECHNICAL REVIEW OF A DRAFT INFORMATION NOTICE REGARDING NONCONSERVATIVE ERRORS AND CHANGES IN SIEMENS POWER CORPORATION LARGE-BREAK LOSS-OF-COOLANT ACCIDENT EVALUATION MODEL AND COMPLIANCE WITH 10 CFR 50.46(a)(3)

Dear Mr. Curet:

The U.S. Nuclear Regulatory Commission is planning to issue an information notice discussing recent staff findings related to the review of Siemens Power Corporation. (formerly Exxon Nuclear) large-break loss-of-coolant accident emergency core cooling system analysis evaluation model changes and also to  !

remind licensees and reactor fuel vendors of the requirements contained in '

Section 50.46(a)(3) of Title 10 of the Code of Federal Reculations [10 CFR -

50.46(a)(3)] concerning the reporting of ECCS cooling model changes and i errors. We ask that you review the enclosed draft of that information notice  ;

to ensure the technical information is accurate. Your cooperation in this '

matter is appreciated. Please return any comments you may have as soon as

)ossible. A copy of this request and your response will be placed in the  !

Jublic Document Room for review by the public. Your response should be mailed I to:

U.S. Nuclear Regulatory Commission Washington. DC 20555-0001 ATTN: Stephen Koenick. NRR/PECB MAIL STOP: 011E4 Please address any questions you may have on this matter to Stephen Koenick of my staff. Mr. Koenick may be reached by phone (301) 415-2841. If no comments are received by close of business February 14. 1997, we will assume the technical information in the notice is correct.

[ Original signed by Robert L. Dennig]

for Alfred E. Chaffee. Chief  ;

Events Assessment and i Generic Communications Branch g g g y Division of Reactor Program Management l l

Office of Nuclear Reactor Regulation  !

Enclosure:

Draft Information Notice DISTRIBUTION TTMartin DRPM R/F AEChaffee PECB R/F SSKoenick Central-Files PDR I)i# ' I 0FFEE PECB:NRR SC/PECB:NRR C/ d t18 0 PECB SKoenick N/L EGoodwin6@IM ANff[e Y DATE 02/4/97 02/f/97 02/([97 7' 9702070355 970205 PDR ADOCK 07001257 C- PDR-

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, UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 February XX,1997 NRC INFORMATION NOTICE 97-XX: NONCONSERVATIVE ERRORS AND CHANGES IN A LARGE-BREAK LOSS-OF-COOLANT ACCIDENT EVALUATION MODEL OF A FUEL VENDOR AND COMPLIANCE WITH 10 CFR 50.46(a)(3)

Addressees All holders of operating licenses or construction permits for nuclear power reactors and all reactor fuel vendors.

Puroose The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert addressees about the recent staff findings related to the review of Siemens Power Corporation (SPC, formerly Exxon Nuclear) large-break (LB) loss-of-coolant accident (LOCA) emergency core cooling system (ECCS) analysis evaluation model changes and also to remind licensees and reactor fuel vendors of the requirements contained in Section 50.46(a)(3) of Title 10 of the Code of Federal Reaulations [10 CFR 50.46(a)(3)] concoming the reporting of ECCS cooling model changes and errors. It is expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required.

Description of Circumstances Recently identifed changes and errors in SPC and General Electric (GE) LBLOCA analysis models have led to a series of 30-day reports and 10 CFR 50.72 reports as required by 10 CFR 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."

SPC LBLOCA ECCS Evaluation Model Changes The SPC LBLOCA ECCS model, TOODEE2, was originally approved by the NRC staff to meet the requirements of 10 CFR 50.46 in a letter dated July 8,1986 [ Accession number 8607150319), from D. M. Crutchfield (NRC) to G. Ward (Exxon). In 1991, FPC had made 3

changes to the NRC-approved fuel cooling test facility (FCTF) reflood heat transfer coefficient l correlation used in TOODEE2.

During August 1995, the NRC met with SPC about the LBLOCA ECCS evaluation model. As l

a result of that meeting, the staff sent a letter to SPC, dated November 13,1995

[9511150211), that identifed problems concoming changes in the TOODEE2 computer code

IN 97-xx

- February 1997 Page 2 of 4 1 ,

i specifically rr!ated to the 1991 changes to the NRC-approved FCTF reflood heat transfer coefficient correlation and the significance of the code changes. The staff then requested in i a letter dated March 13,1996 [9603150002], that SPC formally submit to the staff for its +

{ review and approval all model revisions and corrections implemented in TOODEE2 since the

staffs approval of the code in July 1986.

h On June 2,1996, SPC submitted topical report XN-NF-82-20, "EXEM/PWR Large Break i LOCA ECCS TOODEE2 Updates," Revision 1, Supplement 5 [9606260239], which described I ,

the updates made in the TOODEE2 computer code between 1986 and 1991. TOODEE2 is 4

part of the evaluation model used by SPC for pressurized-water reactors. The staff has completed its review of this report and has concluded that the proposed LBLOCA-ECCS  ;

model (i.e., the 1991 model) is not acceptable and the previously approved model (i.e., the

l 1986 model) contains an unacceptable error. This information was formally communicated to '

l SPC in a safety evaluation enclosed in a letter dated November 29,1996 [9612040294].

l After concluding that the 1991 model was unacceptable, the staff met with SPC and those j

licensees using SPC's LBLOCA evaluation model on October 16,1996, to discuss the unacceptable error in the 1986 model. The staff also requested and received information i

from the licensees that demonstrated that they were in compliance with 10 CFR 50.46 (see j

^

meeting summary dated November 5,1996 [9611140318]).

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Public Service Electric & Gas (PSE&G) Audit of GE During a recent licensee-conducted quality assurance (QA) audit of the fuel vendor (GE -

Wilmington, North Carolina), PSE&G, the licensee of Hope Creek Nuclear Generating Station, identified a weakness in GE's tracking of errors and changes in the LOCA evaluation models.

! Between 1990 and 1995, information sent to the licensee indicated that there had been no l known impact on the calculated peak cladding temperature (PCT). Earlier in 1996, two impacts had been reported by GE to the licensee and when reviewing the handling of this information during the audit, three additionalimpacts not previously reported to the licensee

! were discovered, dating back to 1990,1992, and 1993. In addition, the audit determined that GE had not been tracking the cumulative impact of errors and changes on the PCT as expected by the licensee and as required by 10 CFR 50.46. The cumulative PCT impact was previously known to be 35 'F (19 *C); however, on the basis of the errors identifed during i

the audit, the value is now raised to 100 'F (56 *C) exceeding the 50 'F (28 *C) reporting threshold. The licensee's recalculated PCT still remains below the ECCS acceptance criteria of 2200 'F (1200 *C).

j' The licensee stated that GE had been submitting annual reports to the NRC on behalf of its boiling water reactor customers since 1990. Although the audit determined that the cumulative change to the PCT was in error for the Hope Creek licensee, it was not clear at this time whether the vendor reports to the NRC were either wholly or partially in error, or if other licensees were similarly affected.

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IN 97-xx F;bru ry ,1997

, eg%dd Page 3 of 4 Discussion i

Although the LOCA analyses are performed by the fuel vendors, licensees are responsible for compliance with the regulations related to the LOCA analysis, that is,10 CFR 50.46(a).

l Section 50.46(a)(1)(i) requires licensees to calculate ECCS cooling performance with an

acceptable evaluation model. The staffs recent interactions with the licensees using the SPC's LBLOCA methodology (the review experience of the SPC LOCA evaluation model changes) and the Hope Creek QA audit indicate that licensees may not be closely monitoring the work of their respective fuel vendors. When the error in the 1986 model was discovered and when SPC changed the TOODEE2 code in 1991, the resulting changes in the PCT were, in some cases, significant, and the responsible licensees were not aware of the significant changes. "Significant"is defined in 10 CFR 50.46(a)(3)(i) as follows

"a significant change or error is one which results in a calculated peak fuel cladding temperature different by more than 50*F from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50 *F."

Licensees may not be performing adequate assessments of errors when they are aware of them. Furthermore, licensees' audits of SPC's code changes appear to have been ineffective in identifying the technicalinadequacy of the code changes. It should be noted that 10 CFR 50.46 allows fuel vendors or licensees to make code changes without the staffs prior approval; however, the licensees are responsible for identifying any deficiencies in the change process and reporting them to the NRC staff accordingly. In addition, the licensee determines whether the changes are significant.

It also appears that licensees may not be monitoring the cumulative effect of the code changes. In a given year, the impact of the code change may be less than 50 *F (28 *C) and hence the change is not significant. But the impact of the code changes over several years together can exceed 50 *F (28 *C) and, therefore, will be reportable as significant.

Section 50.46 places the responsibility for the reporting of code changes on the licensees.

Some licensees have apparently considered that the annual reports sent by the fuel vendor are sufficient to meet the requirements under 10 CFR 50.46(a)(3)(ii). Specifically,"the applicant or licensee shall report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually as specified in $50.4. If the change or error is significant, the applicant or licensee shall provide this report within 30 days.. " The reports submitted by the fuel vendors will not satisfy these reporting requirements; however, licensees are allowed to refer to the vendor's annual reports. As stated in 10 CFR Part 50, Appendix B, Section Vil, "The effectiveness of the control of quality by contractors and subcontractors shall be assessed by the applicant or designee at intervals consistent with the importance, complexity, and quantity of the product or services."

in summary, licensees are reminded that to meet the ECCS acceptance criteria their

! responsibilities include:

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- i i DRAFT iN e7-xx Fcbruary ,1997 Page 4 of 4 (1) Section 50.46(a)(1)(i) requires licensees to calculate ECCS cooling performance with an acceptable evaluation model.

(2) Section 50.46(a)(3)(ii) requires licensees to report changes and/or errors and their estimated effects on the limiting ECCS analysis to the Commission at least annually, and if the change or error is significant, the licensee shall provide this report within 30 days.

(3) Individual licensees are responsible to assess effectiveness of quality of ECCS evaluation models provided by the vendors as required by Part 50, Appendix B.

Meaningful technical audits may be necessary to meet Appendix B.

This information notice requires no specific action or written response. If you have any questions about the information in this notice, please contact one of the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.

Thomas T. Martin, Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contacts: George Thomas, NRR Joseph L. Staudenmeier, NRR (301) 415-1814 (301) 415-2869 E-mail: gxt@nrc. gov E-mail: jls4@nrc. gov Stephen Koenick, NRR (301) 415-2841 j E-mail: ssk2@nrc. gov l

Attachment:

List of Recently issued NRC Information Notices 4