ML20059H855
| ML20059H855 | |
| Person / Time | |
|---|---|
| Issue date: | 10/28/1993 |
| From: | Borchardt R Office of Nuclear Reactor Regulation |
| To: | Marriott P GENERAL ELECTRIC CO. |
| Shared Package | |
| ML20059H858 | List: |
| References | |
| REF-QA-99900403 NUDOCS 9311100247 | |
| Download: ML20059H855 (5) | |
Text
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UNITED STATES j, t. f.
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WASHINGTON. D.C. 20566 0001
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October 28, 1993
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Docket Nos.52-001 and 99900403 Mr. Patrick W. Marriott, Manager Licensing & Consulting Services GE Nuclear Energy 175 Curtner Avenue San Jose, California 95125
Dear Mr. Marriott:
SUBJECT:
NOTICE OF NONCONFORMANCE i
(NRC INSPECTION REPORT NO. 99900403/93-02)
This letter addresses the inspection of your facility at San Jose, California, conducted by Richard P. McIntyre, Robert L. Pettis, and Billy H. Rogers of the Nuclear Regulatory Commission's (NRC's) Vendor Inspection Branch, George Thomas of the Reactor Systems Branch, Joseph L. Staudenmeier of the Analytical Support Group, Robert A. Gramm of the Performance and Quality Evaluation Branch, and Sampath K. Malur of the Special Inspection Branch on September 7 through 10, 1993. The details of the inspection were discussed with Mr.
Robert Berglund, General Manager, Advanced Reactor Programs, and other members of your staff during the inspection and at the exit meeting on September 10, 1993.
The purpose of the inspection was to determine if quality activities performed as part of the design of the Advanced Boiling Water Reactor (ABWR) project were conducted under the appropriate provisions of the.GE-NE 10 CFR Part 50, Appendix B, quality assurance (QA) program, as implemented by the Quality Assurance Program Description topical report (NED0-11209-04A) that has been approved by the NRC.
The scope of the inspection included the review of the Design Record Files (DRFs) for computer code input modeling and independent design verification for the computer codes ODYNA, REDYA, and SAFER; the review of the implementation of the QA controls in place for activities performed as part of ABWR Full Integral Simulation Test (FIST) used to qualify SAFER; the review of certain Standard Safety Analysis Report (SSAR) Chapter 6 and 15 safety analyses calculations that will not be reanalyzed if the combined operating license (COL) uses the certified design reference core fuel design; and the review of residual heat removal (RHR) and reactor building cooling water (RCW) system calculations for which GE-NE and the Japanese technical associates (Toshiba and Hitachi) have the design lead.
The inspection results indicate that some of the design, testing, and verification activities that support the ABWR Standard Safety Analysis Report (SSAR) and the certified design were conducted without fully implementing the QA commitments contained in the ABWR SSAR, the GE-NE topical report, and as-wg g$% q"' gjidibd
. required by Appendix B to 10 CFR Part 50.
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Chapter 17 of the ABWR SSAR states that common engineering design documents are reviewed and approved by GE-NE and that GE-NE is responsible for the design and supporting calculations and records for the ABWR project. The inspection team reviewed the DRF for the reactor building cooling water system (RCW), for which a Japanese technical associate had the design lead. While the common engineering documents were reviewed by GE-NE, the inspection team found no evidence that the supporting engineering calculations performed by the technical associates for RCW had been similarly reviewed. Additionally, GE-NE's annual audits of the technical associates did not examine the technical adequacy of the supporting calculations.
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The review of input modeling and verification of selected computer codes identified significant QA deficiencies. The FIST, which was used to qualify the SAFER computer code, was apparently conducted as a non-safety-related activity, although it is used to support the ABWR SSAR accident analysis. The following records and documentation, which are required to be contained or referenced in the DRF, were missing: (1) final as-built facility drawings, (2) reference to original data tapes from the tests, (3) records of disposition of all test anomalies, (4) test log and QA forms for each test, and (5) documentation of analytical or experimental verification of engineering calculations.
In addition, the FIST rod thermocouples were purchased from an unapproved commercial supplier and their calibration was not verified by GE-NE.
The DRFs for the computer codes reviewed did not document adequate independent verification of code changes and GE-NE's procedures did not require that such documentation be maintained.
For the REDYA, ODYNA, and SAFER codes, independent verification was accomplished through a design review process.
This process relied on a design review team's evaluation of a description of the code models and of the results of test cases selected by the code developer. However, this process did not include an independent verification of implementation of the changes in the models described to the design review team, nor a quantitative evaluation of the accuracy of the results of test cases.
Examples of failure to perform quantitative assessments of results of code changes include changes to REDYA and ODYNA for the internal recirculation pumps and to SAFER for the isolation condensers, which did not compare calculational results to measured data. The test cases only verified that the code results were qualitatively correct.
Several ABWR calculation notebooks were poorly maintained and lacked sufficient information.
For example, the internal pump flow area in the ABWR SAFER model was based on a combination of an undimensioned drawing assumed to be to scale and an individual's memory of the pump shaft diameter.
In l
general, the notebooks lacked sufficient information to enable an independent review by a technically qualified person without the assistance of the analyst or system designer.
In two instancer the inspection team identified technical inconsistencies between the ASWR design and information contained in the ABWR SSAR.
We are concerned that the design control process has not ensured the accurate
Mr. Patrick W. Marriott October 28, 1993 translation of information into the SSAR which is relied upon for our design certification safety judgement.
The above findings do not appear to have major safety significance for the ABWR design at this point in time. This conclusion is based on the fact that there is-a significant thermal hydraulic design margin built into the ABWR codes ODYNA, REDYA, and SAFER, that 'sas demonstrated through thermal hydraulic computer analyses and the FIST test program. There is significant safety-margin, especially in SAFER for LOCA analyses where Appendix K analysis requirements are fairly conservative.
In addition, the calculated peak cladding temperature (1116' F) is considerably less than the allowable value (2200* F).
However, based on the recent SBWR and ABWR QA inspections, there is a lack of attention to QA by GE-NE in many of the activities related to the Advanced Light Water Reactor Program (ALWR).
If not properly addressed.by GE-NE, it could have the potential to cause significant safety concerns.
The lack of significant immediate safety concerns does not relieve GE-NE of the responsibility for implementing adequate design and test control.
Please provide us within 30 days from the date of this letter a written statement in accordance with the instructions in the enclosed Notice of Nonconformance. We will consider extending the response time if you can show good cause for us to do so.
In addition, your response should specifically address (1) the safety significance of the concerns identified in this report, (2) the general integrity of the GE-NE QA program implementation for the ALWR program, (3) Unresolved Items (93-02-08) and (93-02-09), and (4) actions planned to be taken to rectify the adverse conditions identified.
The responses requested by this letter and the enclosed Notice of Nonconformance are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, Public Law No.96-511.
In accordance with 10 CFR Part 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC's Public Document Room.
Mr. Patrick W. Harriott-October 28, 1993 Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely, Rfh9OorcNNtfbheckoEY Standardization Project Directorate,,
Associate Directorate for Advanced Reactors and License Renewal.
Office of Nuclear Reactor Regulation cc: See next page DISTRIBUTION:
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A Mr. Patrick W. Marriott Docket No.52-001 General Electric Company 99900403 cc:
Mr. Joseph Quirk Mr. Raymond Ng GE Nuclear Energy 1776 Eye Street, N.W.
General Electric Company Suite 300 i
175 Curtner Avenue, Mail Code 782 Washington, D.C.
20086 San Jose, California 95125 Mr. Victor G. Snell, Director Mr. L. Gifford, Program Manager Safety and Licensing Regulatory Programs AECL Technologies GE Nuclear Energy 9210 Corporate Boulevard 12300 Twinbrook Parkway Suite 410 Suite 315 Rockville, Maryland 20850 Rockville, Maryland 20852 Director, Criteria & Standards Division Office of Radiation Programs U.S. Environmental Protection Agency 401 M Street, S.W.
Washington, D.C.
20460 Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C.
20585 Marcus A. Rowden, Esq.
Fried, Frank, Harris, Shriver & Jacobson 1001 Pennsylvania Avenue, N.W.
Suite 800 Washington, D.C.
20004 Jay M. Gutierrez, Esq.
Newman & Holtzinger, P.C.
1615 L Street, N.W.
Suite 1000 Washington, D.C.
20036 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.
Room 8002 Washington, D.C.
20503 Mr. Frank A. Ross U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874