ML20009G663
| ML20009G663 | |
| Person / Time | |
|---|---|
| Issue date: | 06/24/1981 |
| From: | Breaux D, Brickley R, Hale C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20009G658 | List: |
| References | |
| REF-QA-99900404 NUDOCS 8108040528 | |
| Download: ML20009G663 (4) | |
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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION IV Report No. 99900404/81-02 Company:. Westinghouse Electric Corporation Nuclear Technology Division Monroeville, Pennsylvania Inspection Conducted:
June 2-5, 1981 l
Inspectors: '/2 M,3 b
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R. H. Brickley, Cc(tfactor Inspector Date Reactor Systems Section Vendor Inspection Branch N06.32ral 6/2W81 D. G. Breaux, Inspector Date Reactor Systems Section Vendor Inspection Branch Approved by:
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~$f C. 4( Hfle'; Chief Date Reac ror Systems Section Vendor Inspection Branch Summary Inspection conducted on June 2-5, 1981 (Report No. 99900404/81-02)
Areas Inspected:
10 CFR Part 50, Appendix B in the areas of control of computer programs, design verification, and three Regional and Headquarters inspection requests.
The inspection involved 57 inspector-hours on site by two NRC inspectors.
Results:
There were no nonconformance or unresolved items identified.
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DETAILS SECTION I (Prepared by R. H. Brickley)
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A.
Persons Contacted S. T. Maher, Engineer l
- B. A. Maul, Senior Engineer R. W. Steitler, Manager, Reactor Protection Analysis l
- Denotes those in attandance at the exit interview.
E.
Control of Computer Programs 1.
Objectives The objectives of this area of the inspection were to determine that:
a.
Computer programs have been developed, verified, qualified, and are being used in accordance with procedures which have been prepared, reviewed, and approved by authorized management.
I b.
A computer program custodian has been designated and has the responsibility for maintaining the security of the program.
c.
Each computer program that has been authorized for use has been qualified, has an appropriate user's manual, and this manual (or another manual) provides a detailed description of the mathematical models, empirical data, assumptions used, and applicable references.
d.
The computer program has been verified / qualified and that docu-mentation exists which includes:
(1) A descriptice of the program version and options validated.
(2) A detailed description of the test problems, including boundary conditions, mathematical model, and all key parameters.
(3) A listing of the test problem input data checks and a reprint of the program input and output, or reference to the location where this is stored.
(4) The comparison of solutions, evaluation of the program validity, and an analysis of any identified errors.
e.
Technically qualified individuals have reviewed and approved the verification / qualification of each computer program prior to its use in safety-related applications.
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Revisions and modifications have been subjected to the same review and approval as the original version of the program, g.
Errors identified in computer programs are promptly corrected and app.opriately verified prior to use.
h.
Errors which could result in significant deficiencies in nuclear plant design are reported to the NRC under the requirements of 10 CFR Part 21, 10 CFR Part 50.55(e), and 10 CFR Part 50, Appendix K, as appropriate.
2.
Method of Accomplishment she preceding objectives were accomplished by an examination of the i
user's manuals, theoretical manuals, and verification reports for several versions of the computer programs LOFTRAN and MARVEL.
In addition, the inspector examined applicable procedures implementing the QA program.
3.
Findings a.
There were no nonconformance, unresolved, or follow up items identified.
b.
The LOFTRAN computer program simulates transient behavior in a multi-loop pressurized water reactor (PWR) system.
The program simulates neutron kinetics, thermal-hydraulic conditions, and pressurizer, steam generators, reactor coolant pumps, and control and protection system operation.
Up to four independent reactor coolant loops may be modeled.
c.
The MARVEL (Multi-Loop Analysis of PWR System Transient) computer program calculates multi-loop detailed transient behavior of PWR systems.
It simulates two reactor coolant loops including two steam generators and associated systems.
It also simulates reactor kinetics, reactor control and protection system, safe-guards system and other subsystems.
This program is the precursor to LOFTRAN.
d.
Westinghouse reports, No. WCAP-7878 (LOFTRAN Code Description) and WCAP-9235 (MARVEL-A Digital Computer Code for Transient Analysis of a Multi-Loop PWR System), were found to satisfy the
" User Manual" requirements.
e.
All of the documents examined were completed during the period 1972 to 1979 and were found to have been processed in accordance with procedures existing at that time.
Present procedures have incorporated those items committed by Westinghouse in response to a previous computer program inspection (Report No.
99900404/78-03).
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I 4
C.
Design Verification l
1.
Objectives The objectives of this area of the inspection were to determine that procedures have been established and are being implemented that:
a.
Identify individuals or groups who are authorized to perform design verification reviews.
b.
Require the results of the design verification effort to be clearly indicated, and filed.
c.
Identify and document the method by which design verification is to be performea.
d.
Identify the items to be considered during design verification by the design review method.
e.
Prescribe the requirements for performing design verification by the alternate calculations method.
f.
Prescribe the requirements for performing design verification by the qualification testing method.
2.
Method of Accomplishment The preceding objectives were accomplished by an examination of:
i a.
Fourteen calculations that were part of the verification package on the LOFTRAN and MARVEL computer programs to verify that they were performed in accordance with applicable QA Program require-ments.
L.
Applicable procedures and revisions thereto in effect during the time these documents were generated.
3.
Findings a.
There were no nonconformance or unresolved items identified.
One follow up item was identified (See c. below.).
b.
These calculations were completed during the period 1970 to 1978 and were found to have been processed in accordance with pro-cedures in existance at that time.
c.
Follow up Calculation No. CN-RPA-76-53, part of the verification package for cycle 2 (version) of computer program MARVEL, could not be located during this inspection.
The inspector will examine this document during a future inspection.
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5 D.
Exit Interview An exit interview was held with management representatives on June 5, 1981.
In addition to those individuals indicated by an asterisk in paragraph A.
of each Details Section, those in attendance were:
T. M. Anderson, Manager, Nuclear Safety H. H. Brunko, Manager, Product Integrity and Design Assurance E. J. Hampton, Manager, Design Assurance Operations M. H. Shannon, Senior Quality Engineer R. A. Wiesemann, Manager, Regulatory and Legislative Affiars The inspector discussed the scope and findings of the inspection.
Manage-ment comments were generally for clarification only, or acknosledgement of the statements of the inspector.
4 6
DETAILS SECTION II (Prepared by D. G. Breaux)
A.
Persons Contacted B. F. Barng, Senior Engineer, Product Integrity D. E. Boyle, Manager, Primary Equipment T. W. Coffield, Mechanical Engineer M. R. Gasparro, Engineer M. Gooding, Senior Engineer, Nuclear Safety C. Johnrawley, Principal Engineer, Primary Equipment
- P. T. McManus, Manager, Design Assurance Systcms D. K. Weisensee, Senior Engineer, Reactor Internals
- Denotes those in attendance at the exit interview.
B.
Control Rod Guide Tube Support Pin l
This item is a follow up to a Westinghouse report to the NRC concerning a preliminary identification of potential failure of control rod guide tube support pins due to stress corrosion.
The event that prompted Westinghouse to report this item was an inspection at a foreign plant i
where stress corrocion cracking of these pins was detected.
1.
Objectives The objectives of this area of the inspection were to:
a.
Examine the results of the evaluation of this item to deter-mine that a proper evaluation was performed.
b.
Determine whether this item is generic or plant unique.
c.
Determine if this item was properly reported to the NRC and subsequently all parties involved.
2.
Method of Accomplishment The preceding objectives were accomplished by an examination of:
a.
Westinghouse letter to NRC Office of Nuclear Reactor Regulation, NS-TMA-2254, dated June 10, 1980.
b.
Northeast Utility letter to NRC Office of Inspection and Enforcement, dated April 15, 1980.
c.
South Carolina Electric and Gas Company letter to NRC Region II, dated Septemiser 3,1980.
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Westinghouse Internal Memo NS-LAA-80-185, dated December 8, 1980.
e.
Virginia Electric and Power Company letter to NRC Region II, dated March 18, 1980.
f.
Westinghouse letter to NRC Division of Operating Reactor, NS-TMA-2099, dated June 11, 1979.
g.
Westinghouse letter to NRC Office of Inspection and Enforcement, NS-TMA-2214, dated March 14, 1980.
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h.
Westinghouse letter to NRC Office of Nuclear Reactor Regulation, NS-TMA-2235, dated April 23, 1980.
i.
Safety Review Committee File 80-157.
j.
Five drawings.
k.
Four specifications.
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3.
Findings The potential failure of Westinghouse supplied control rod guide tube support pins due to stress corrosion cracking was reported to the NRC under 10 CFR Part 21 and 50.55(e).
It was reported by Westinghouse because the failure of a pin in a Westinghouse Upper Head Injectior. (UHI) plant could jeopardize control rod operation.
After testing of sample pins, Westinghouse concluded there was a greater susceptibility for stress corrosion cracking on pins tgat j
had a solution heat treat hardening process of lower than 1800 F.
Test results on samples annealed at higher temperatures showed i
superior resistance to stress corrosion cracking further confirm-
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ing the Westinghouse conclusion.
All Westinghouse domestic UHI plants are being retrofitted with pins of a higher heat treat temperature to lessen their sus-ceptibility to stress corrosion cracking.
Since no examples of pin cracking in Westinghouse domestic plants have been made known to Westinghouse or the NRC, absolute generic determination cannot be made.
Westinghouse Quality Control would not have detected such cracks during fabrication, because these stress corrosion cracks propogate during subjection to actual operating environment.
All reporting requirements were properly met by Westinghouse in informing the NRC and the involved clients.
There were no nonconformance, unresolved, or follow up items identified in this area of the inspection.
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8 C.
Hydrogen Recombiner Power Supply Transformer This item is a follow up to a Georgia Power Company 10 CFR Part 50.55(e) report to the NRC concerning unqualified Hydrogen Recombiner Power 1
Transformers delivered to Vogtle Units 1 and 2.
1.
Objectives The objectives of this area of the inspection were to:
i a.
Examine the results of the evaluation of this item to determine that a proper evaluation was performed.
l b.
Determine whether this item is generic or plant unique.
c.
Determine if this item was properly reported to the NRC and subsequently all parties involved.
2.
Method of Accomplishe.Ont The preceding objectives were accomplished by an examination of:
a.
Georgia Power Company letter to NRC Region II, dated December 3, 1980.
b.
Georgia Power Company letter to NRC Region II, dated October 23, 1980.
c.
Westinghouse letter to Southern Company Services, Inc., dated September 16, 1980.
d.
Safety Review Comi,ittee File 80-97.
4 e.
Bill of Material W 9552010 Sheet 1 and 2.
f.
Engineering Change Notice ECN-31843.
I g.
QPS-300-1, " Quality Procurement Specification for Baseline Control, Revision 0, dated May 1, 1979.
j h.
Purchase Order Change Notice 546-CM3-402090 CN No. 4, dated March 12, 1981.
i.
Administrative Specification for Procurement of NSSS Components No. 54823, Revision 6, dated September 19, 1978.
3.
Findings On April 25, 1980, a Westinghouse QA inspection of halimar Elec-tronics (Halimar) revealed that an unqualified power supply
9 transformer had been supplied to Halimar by their subvendor Westinghouse-Greenville.
An inspection at Halimar determined that only one unqualified transformer for the Hydrogen Recombiner Control Panel units had been shipped from Halimar to the Georgia Power Company.
Georgia Power Company initiated a 10 CFR 50.55(e) report to NRC Region II.
The transformer was replaced.
All un-shipped Hydrogen Rec <"nbiner Control Panel units at Halimar are being replaced with qualifted transformers.
Proper interface between Halimar and Westinghouse-Greenville would have alerted H&limar that a transformer substitution had been made.
Halimar's own receipt inspection of the transformers should have made them aware of the substitution. Westing-house defines in its procurement contracts that any change be submitted by the vendor to Westinghouse in the form of a Contract Deviation Request (CDR).
To assure compliance in the future, Westinghouse is in the process of reviewing all Baseline Design Documents to assure that they define the design basis and qualification requirements accurately.
Westing-house also has added extra surve:llance activities at Halimar to further verify compliance.
Reporting requirements were properly met by Westinghouse in informing the involved client of the unqualified transformer.
There were no nonconformance, unresolved, or follow up items identified in thn area of the inspection.
D.
Boron Dilution This is a continuation of our inspection that was begun during the last inspection (Inspection Report 81-01).
This item refered specifically to one operating plant (North Anna 1 and 2) that required analysis of boron dilution in the cold shutdown operating condition.
Review of the Westinghouse Safety Analysis Standards (SAS), which is a standard for analysis used for licensing purposes, referenced inputs only in the refueling, startup, power, and manual rod control operating modes.
The cold shutdown mode was not considered to be un operating condition subject to safety analysis for licensing purposes for earlier plants.
North Anna 1 and 2 was the first Westinghcuse plant to analize inadver-tant boron dilution in the cold shutdown mode.
During this analysis the volume of the reactor vessel hot and cold legs, and steam generator were included when the Residual Heat Removal System was operating.
- However, during this mode of operation the stream generators would be isolated.
No volume was defined in the Westinghouse SAS for the operation in the RHR mode (ct i shutdown); so the responsible engineer and independent reviewer of the boron dilution analysis did not have a defined volume with which to compare.
There were no nonconformance, unresolved, or follow up items identified in this area of the inspection.
The design omission described above does not appear to be the result of a failure to follow procedures or deficient procedures.
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