ML20141A334
| ML20141A334 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/18/1997 |
| From: | Mcdonald D NRC (Affiliation Not Assigned) |
| To: | Carns N NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| NUDOCS 9706200231 | |
| Download: ML20141A334 (4) | |
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j NUCLEAR REGULATORY COMMISSION
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June 18, 1997 Mr. Neil S. Carns Senior Vice President and Chief Nuclear Officer Northeast Nuclear Energy Company c/o Ms. Patricia A. Loftus Director - Regulatory Affairs F.O. Box 128 Waterford, CT 06385
SUBJECT:
SIEMENS POWER CORPORATION LARGE BREAK LOSS-0F-COOLANT ACCIDENT EVALUATION MODEL
Dear Mr. Carns:
The NRC recently completed an inspection of Siemens Power Corporation -
Nuclear Division (SPC) in Richland, Washington. As a result of the inspection, technical issues were raised that will require action by SPC to resolve. The purposes of this letter are to (1) provide you information regarding the open technical issues, and (2) alert you to the need to compile and maintain specific information pursuant to regulatory requirements pertaining to the computer codes used to perform loss-of-coolant accident (LOCA) analyses for Millstone Unit 2.
The exit for the inspection was conducted with Siemens at NRC Headquarters on May 13, 1997.
Some of the open technical issues are discussed in the enclosed letter from the NRC to SPC, and concern testing and analysis performed by SPC to support the development of the EXEM PWR large-break LOCA evaluation model.
No immediate action is required by licensees using SPC's methodology. However, licensees will be responsible for the assessment and reporting of the impact of any changes to SPC's approved methodology that may result from the resolution of these issues.
In addition to the issues discussed in the enclosed letter from NRC to Siemens on the inspection, the NRC inspection team found significant deficiencies in SPC's documentation of its analysis codes and models, and also determined that SPC performed inadequate verification and validation (V&V) of approved codes after changes were made to them. Regulatory requirements for documentation and V&V of LOCA codes are contained in 10 CFR Part 50, Appendix K, Section II.
I These requirements are imposed on licensees, sinca licensees are ultimately
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responsible for ensuring that the LOCA analyses i their plants comply with applicable regulations.
In addition, general requ.ements with regard to j
quality assurance, which are related to licensee responsibilities in ensuring
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thG LOCA codes and analyses comply with applicable regulations, are discussed in 10 CFR Part 50, Appendix B; for example, Criterion III (Design Control),
Criterion V (Instructions, Procedures, and Drawings), Criterion VII (Control
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of Purchased Material, Equipment, and Services), Criterion X (Inspection),
Criterion XVII (Quality Assurance Records), and Criterion XVIII (Audits).
NRC RE CBE CDPV 9706200231 970618 PDR ADOCK 05000336 P
Neil S. Carns 4 Licensees must maintain documentation that provides adequate assurance that their LOCA analyses comply with the appropriate provisions of 10 CFR Part 50, Appendix B and Appendix K, and that they meet the acceptance criteria of 10 CFR 50.46. This documentation should consist of (but is not limited to) reports of internal or external reviews of LOCA methodologies and results; audit or inspection repor's; or independent confirmatory analyses. While the staff does not require at this time that licensees submit this information, the staff will review the documentation as part of its inspections of
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licensees, such as those conducted in support of the core performance action plan.
Failure of licensees to provide adequate oversight of their contractors or to maintain adequate documentation could result in enforcement action.
The staff further reminds licensees that it is their responsibility to conduct meaningful audits of their contractors.
This is discussed in the context of requirements related to LOCA analyses in NRC Information Notice 97-15; however, it is generally applicable to other contractor activities, as well.
If the staff determines that contractor technical and quality assurance l
deficiencies should have been discovered by licensees' audits, the staff will generally initiate enforcement action against the licensees. The staff is currently reviewing whether enforcement action is warranted as a result of the SPC inspection.
Sincerely, Daniel G. Mcdonald, Jr., Senior Project Manager j
Special Projects Office - Licensing Office of Nuclear Reactor Regulation Docket No. 50-336
Enclosure:
As stated i
cc w/ encl:
See next page
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Neil S. Carns June 18, 1997 Licensees must maintain documentation that provides adequate assurance that their LOCA analyses comply with the appropriate provisions of 10 CFR Part 50, Appendix B and Appendix K, and that they meet the acceptance criteria of 10 CFR 50.46. This documentation should consist of (but is not limited to) reports of internal or external review:; of LOCA methodologies and results; audit or inspection reports; or independent confirmatory analyses. While the staff does not require at this time that licensees submit this information, the staff will review the documentation as part of its inspections of licensees, such as those conducted in support of the core performance action plan.
Failure of licensees to provide adequate oversight of their contractors or to maintain adequate documentation could result in enforcement action.
The staff further reminds licensees that it is their responsibility to conduct meaningful audits of their contractors. This is discussed in the context of requirements related to LOCA analyses in NRC Information Notice 97-15; however, it is generally applicable to other contractor activities, as well.
If the staff determines that contractor technical and quality assurance deficiencies should have been discovered by licensees' audits, the staff will generally initiate enforcement action against the licensees.
The staff is currently reviewing whether enforcement action is warranted as a result of the SPC inspection.
Sincerely, Original signed by:
Daniel G. Mcdonald, Jr., Senior Project Manager Special Projects Office - Licensing Office of Nuclear Reactor Regulation Docket No. 50-336
Enclosure:
As stated cc w/ encl:
See next pt.ge DISTRIBUTION:
Docket. Filer PUBLIC SP0-L Reading SP0 Reading WTravers PMcKee LBerry DMcDonald JDonohew Alevin 0GC ACRS JDurr, RI DOCUMENT NAME: G:\\ MCDONALD \\SPC.LTR To oco, e copy of this document,Indcate in the M:
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1 WBeckner PMclCe'e DATE 06/ t$97 UV 06/J7/97 06/b97 06/@/97 06/ /97
'i 0FFICIAL RECORD COPY
Northeast Nuclear Energy Company Millstone Nuclear Power Station Unit 2
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cc:
Lillian M. Cuoco, Esquire Mr. F. C. Rothen i
Senior Nuclear Counsel Vice President - Work Services Northeast Utilities Service Company Northeast Nuclear Energy Company P. O. Box 270 P. O. Box 128 Hartford, CT 06141-0270 Waterford, CT 06385 Mr. John Buckingham Ernest C. Hadley, Esquire
- Department of Public Utility Control 1040 B Main Street Electric Unit P.O. Box 549 10 Liberty Square West Wareham, MA 02576 New Britain, CT 06051 Mr. D. M. Goebel
- Mr. Kevin T. A. McCarthy, Director
.Vice President - Nuclear Oversight Monitoring and Radiation Division Northeast Nuclear Energy Company Department of Environmental Protection P. O. Box 128 79 Elm Street Waterford, CT 06385 Hartford, CT 06106-5127 Mr. M. L. Bowling, Jr.
Regional Administrator, Region 1 Recovery Officer - Millstone Unit 2 U.S. Nuclear Regulatory Commission Northeast Nuclear Energy Company 475 Allendale Road P. O. Box 128 King of Prussia, PA 19406 Waterford, CT 06385 First Selectmen Mr. J. K. Thayer Town cf Waterford Recovery Officer - Nuclear Engineering Hall of Records and Support 200 Boston Post Road Northeast Nuclear Energy Company Waterford, CT 06385 P. O. Box 128 Mr. Wayne D. Lanning Deputy Director of Inspections Mr. B. D. Kenyon Special projects Office President and Chief 475 Allendale Road Executive Officer King of Prussia, PA *19406-1415 Northeast Nuclear Energy Company P. O. Box 128 Charles Brinkman, Manager Waterford, Connecticut 06385 i
Washington Nuclear Operations
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ABB Combustion Engineering Mr. Allan Johanson, Assistant Director i
12300 Twinbrook Pkwy, Suite 330 Office of Policy and Management Rockville, MD 20852 Policy Development and Planning Division Senior Resident Inspector 450 Capitol Avenue - MS# 52ERN Millstone Nuclear Power Station P. O. Box 341441 c/o U.S. Nuclear Power Station Hartford, CT 06134-1441 P.O. Box 513 Niantic, CT 06357 0
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Northeast Nuclear Energy Ccmpany Millstone Nuclear Power Station Unit 2 cc:
4 Citizens Regulatory Commission ATTN: Ms. Susan Perry Luxton 180 Great Neck Road l
Waterford, Connecticut 06385 l
Deborah Katz, President Citizens Awareness Network P. O. Box 83 Shelburne Falls, MA 03170 The Honorable Terry Concannon Co-Chair Nuclear Energy Advisory Council
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Room 4035 Legislative Office Building Capitol Avenue Hartford, Connecticut 06106 Mr; Evan W. Woollacuct Cv-Chair Nuclear Energy Advisory Council 128 Terry's Plain Road j
Simsbury, Connecticut 06070 i
Little Harbor Consultants, Inc.
Millstone - ITPOP Project Office P. O. Box 0630 Niantic, Connecticut 06357-0630 L
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May 7, 1997 4
i Mr. H. D. Curet, Manager i
Product Licensing Siemens Power Corporation 2101 Horn Rapids Road P. O. Box 130 l
Richland, WA 99352-0130
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ((RAI) REGARDING REFLOOD 1
HEAT TRANSFER I
Dear Mr. Curet:
The inspection team identified two open issues concerning reflood heat transfer during the NRC's inspection at Siemens Power Corporation (SPC) from March 17-21, 1997.
Those issues are:
(1) assessment of the scaling methodology used to apply the FCTF 17x17 PWR reflood heat transfer correlation to non-17x17 geometries (Inspection Issue 13) and (2) validity of the FCTF reflood heat transfer coefficients derived from the original 17x17 testing (Inspection Issue 17).
Both of these open issues were discussed with you as part of the inspection.
Regarding the first issue, in the NRC staff's Safety Evaluation Report (SER) on Exxon Nuclear's (ENC, now SPC) LBLOCA Evaluation Model, a commitment was confirmed to perform reflood heat transfer testing in 15x15 bundle geometry to validate ENC's proposed scaling methodology.
In correspondence in October 1992 and August 1993, SPC requested and the NRC agreed to a new approach, in which existing data from other tests could be used instead of data from FCTF 15x15 tests.
While the staff did not request that Siemens submit the scaling evaluation for review in its August 1993 letter, the expectation was that the assessment would be performed and would be available for the staff to au(it at its discretion.
It appears from the results of the inspection that the assessment was not actually performed until very recently, and the team's initial review of the information provided during the week of March 17, 1997, indicated that the documentation did not meet the staff's expectations.
With respect to the second issue, the team reviewed internal Advanced Nuclear Fuels (ANF, now SPC) correspondence from the period April-June 1987, indicating that concerns had been raised about the correct programming of data analysis algorithms in the XRAD/ LEPER code used to derive heat transfer coefficients from the FCTF 17x17 reflood heat transfer tests.
Those heat transfer coefficients were used to develop the original PWR 17x17 reflood heat transfer correlation. A subsequent internal memorandum, from March 1989, indicates that some evaluation of the original FCTF data was, in fact, performed using a " controlled" version of the XRAD/ LEPER algorithms, and that agreement between the original and "new" analysis was very good.
However, the team has found no evidence-that a systematic re-analysis of the 17x17 data was ever performed.
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Enclosure
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Mr. Curet 2
9 In order to help resolve these two issues, the staff requests the following:
1.
That SPC complete the assessment of the applicability of the reflood heat transfer scaling methodology to geometries other than 17x17, as t
discussed during the inspection, and submit the results for staff review i
as part of the overall assessment of the reflood heat transfer correlation; and i
1 2.
That SPC provide the staff, as soon as possible, a schedule for a complete reanalysis of temperature data from the original 17x17 tests used to develop the FCTF reflood heat transfer correlation.
The
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reanalysis should be performed with a computer code that has been l
1 verified, validated, and controlled according to the requirenents of SPC's quality assurhnte program.
Upon completion of the reanalysis, the 4
results are to be submitted to the NRC for staff review.
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For both (I) and (2), SPC should provide sufficient evidence to demonstrate that the PWR reflood heat transfer coefficient correlation, as currently implemented in the T00DEE2 code, produces results that are conservative when compared to test data.
Two RAls on these topics are being provided to SPC i
under separate cover with additional details about what the staff will need to
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proceed with the review of the revised FCTF correlation and the scaling methodology.
j The staff also reminds SPC that it may have responsibilities with regard to j
one or both of these issues, pursuant to the requirements of Part 21 of Title 10 of the Code of Federal Regulctions (10 CFR 21).
i A copy of this letter is being provided to all licensees that could be 1
affected by either of these issuas.
I Sincerely, Original Signed By{
James E. Lyons, Acting Chief i
Reactor Systems Branch Division of Systems Safety and Analysis Office of Nuclear Reactor Regulation i
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