ML19282A098
| ML19282A098 | |
| Person / Time | |
|---|---|
| Issue date: | 02/26/1979 |
| From: | Donna Anderson, Hale C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML19282A087 | List: |
| References | |
| REF-QA-99900400 NUDOCS 7905100366 | |
| Download: ML19282A098 (10) | |
Text
8 i
U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION IV Report No.
99900400/79-01 Program No. 51100 Company:
The Babcock & Wilcox Company Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Inspection Conducted:
February 12-16, 1979 Inspector:
.cd M
EM.7/79 a-D. h. ?nderson, Principal Inspector, Vendor
' Date Inspection Branch Approved by:
(
C M h79 C. J. Hale, phief, Program Evaluation Section, Date Vendor hspection Branch Sumary Inspection on February 12-16, 1979 (99900400/79-01).
Areas Inspected:
Implementation of Title 10 CFR 50, Appendix B, and Topical Report BAW-10096A including design verification, Part 21 Report Follow-up, and action on previous inspection findings. The inspection involved thirty-two (32) inspector-hours on-site by one (1) USNRC inspector.
Results:
In the two (2) areas inspected, one (1) deviation from commitment and one (1) unresolved item was identified in one of the areas.
Deviation: Design Verification - The values for the high and low level / trip pressurizer setpoints identified in an equipment specification and a system description did not agree with the same values in the system requirements specification.
(See Notice of Deviation, Enclosure).
Unresolved Item: Values for pressurizer level trip setpoints in Table 7.2.1-2 of the Bellefonte FSAR are not in agreement with present calculated values.
(See Details Section paragraph D.3.).
'29051003/4 4 Details Section A.
Persons Contacted
- C. A. Armontrout, Lead Quality Assurance Engineer G. S. Carter, Senior Engineer R. A. Coe, Associate Engineer
- S. H. Klein, Manager, Quality Assurance Engineering
- A. L. MacKinney, Manager, Quality Assurance D. Mars, Licensing Engineer L. R. Seguin, Quality Assurance Engineer G. A. Shipman, Task Engineer R. M. Smith, Supei,isory Engineer C. C. Strempke, Task Engineer / Integration K. D. Tuley, Associate Engineer R. O. Vosburgh, Supervisory Engineer T. L. Wendland, Engineer
- Indicates attendance at the exit meeting.
B.
Action on Previous Insoection Findings (Closed) Unresolved Item (Report No. 78-02): An apparent discrepancy exists between fuel data presented in Table 4.2.3-1 of the Bellefonte FSAR and the data contained in the fuel specification.
Babcock and Wilcox has initiated action on this item by the issuance of DRN No.
86-2776-01, February 15, 1979, which documents the discrepancies identified in the Bellefonte FSAR. NRC/NRR has been advised of these discrepancies in the Bellefonte FSAR and any further action on this item will be conducted by NRC/NRR.
C.
Part 21 Report Follow-up-Reactor Vessel Insulation Loadina The inspector followed up on a headquarters request related to inspection of a design deficiency identified at North Anna Power Station, Units 3 and 4, and reported under the provisions of 10 CFR Part 21 by Virginia Electric and Power Company on June 30, 1978.
The loading effects of reactor vessel insulation on piping and nozzles had not been analyzed for a seismic event.
1.
Objectives The objective of this area of the inspection was to follow-up on a 10 CFR Part 21 report at the affected organization to verify that:
. a.
The report accurately describes the defect or failure to comply and satisfies the reporting requirements with respect to information provided and timing of submittal.
b.
The defect or failure to comply has been evaluated as required by 10 CFR 21 and reporting organizations procedures, stated safety hazard is a logical conclusion, factual and complete data has been used, generic implication has been assessed, and notification was in accordance with procedures.
c.
The stated corrective acti,n is appropriate, adequate, implemented or planned, and will prevent recurrence.
2.
Method of Accomolishment The preceding objectives were accomplished by an examination of:
a.
B&W Nuclear Power Generation Division Quality Assurance Manual,19A-N.1., Section 15, Nonconformances, Subsection 4.3.1.,
Significant Deficiency.
Administrative Procedure NPG-1707-01, Processing of Safety Concerns.
b.
PSC 5-78, Preliminary Report of Safety Concerns, February 3, 1978, and the followi.;g documentation:
Memoranda:
Preliminary Report of Safety Concern, 5-78, February 7, 1978.
c.
PSC 5-78, Preliminary Report of safety Concerns, File 205 T4.4, June 20, 1978; and the following related documents:
Memoranda:
Preliminary Significant Deficiency Report, PSC 5-78, June 13, 1978; Evaluation of PSC 5-78, May 15, 1978; Specifications:
18-1391001, RCS Displacements and Response Spectra, February 28, 1978.
Reports: B&W 835, As Built Response Spectra Curves for Containc.ent Cuilding, Davis Besse 620/0014, lanur. y 6,1975; 5 tress Report No. 32-9824-00, Analysis of Mirror Insulation Support Systems for Reactor Vessel, Bellefonte 1 and 2, October 4, 1978; Stress Report No. 32-9823-00, Analysis of Mirror Insulation Support Systems for Reactor Vessel, Davis Besse, October 6, 1978.
. Drawings: B&W 51-42-001-00, Toledo Edison / Davis Besse Nuclear Power Station General Arrangement Outline, February 1, 1974.
Forms Processed:
BWNP-20208, Preliminary Report of Safety Concerns (PSC). Completed form as required by NPG-1707-01, with concurrence by both Manager, Quality Assurance and Manager, Plant Integration, dated June 13, 1978.
3.
Findings The report accurately identifies that insulation and insulation a.
supports around the reactor vessel incore nozzles may cause loading on the nozzles during a seismic event that has not been previously analyzed.
b.
The subsequent evaluation and timing of submittal by B&W meets the reporting requirements of 10 CFR 21 and the B&W implementing procedures identified in C.2.a. above.
c.
Five (5) plants were originally identified as having possible generic implications, but subsequent analyses have identified the following plants as being affected:
(1) TVA-Bellefonte Units 1 and 2.
(2) VEPCO-North Anna Units 3 and 4.
B&W notified these utilities of their responsibility for reporting under 10 CFR Part 21 and management of B&W indicated that written confirmation had been received from the affected utilities that this item had been reported.
d.
Corrective action has been proposed by B&W which would result in stregthening of the structural supports for the insulation around the reactor vessel. This corrective action appears to be logical and should correct the problem described in the reports and evaluation identified in C.2.b.
above.
There were no deviations or unresolved items identified in e.
this area of the inspection.
D.
Desian Verification
~
<re c 1.
Objectives The objectives of this area of the inspection were to detemine 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 docum' nted, with the identification of the verifier clearly indic ated, and filed so they are identifiable to the document rev ewed and can readily be retrieved.
c.
Require +.nat the extent of design verification take into consideration the importance to safety, complexity, degree of standardization, state of the art, similarity with previously proven designs, applicability of standardized or previously proven designs, known problems and their effects, and changes to previously verified designs.
d.
Identify and document the method by which design verification is to be performed.
e.
Identify the items to be considered during design verification by reviews including selection and incorporation of inputs, necessary assumptions, quality and QA requirements, codes, standards, regulations, construction and operating experience, interfaces, design method used, comparison of output with input, item application suitability, material compatibility, and maintenance features.
f.
Prescribe the requirements for perfoming design verification by alternate calculations which shall' include performance by a person or persons other than those who performed the original calculation, the review of appropriateness of assumptions, input data, and code or other calculation met.cd use. The selection of method shall provide results consistent with the original calculation.
- g. ~~ ~ Prescribe the requirements for performing design verification by qualification testing which shall include requirements:
(1) For the identification, documentation, a demonstration of the adequacy of performance under the most adverse conditions, and consideration of all pertinent operating modes. Where the test is only intended to verify a specific design feature, the other features of the design shall be verified by other means.
e,.
-. (2) That testing be performed in accordance with written test procedures which incorporate or reference the test requirements, acceptance criteria limits and include provisions fo. assuring that prerequisites for the given test have been met, adequate instrumentation of the required range and accuracy is used, and that necessary monitoring is performed.
(3) That test results be documented and evaluated by the responsible designer and, if test results indicate that modifications to the item are needed, these modifications shall be documented and the item modified, retested, or otherwise verified.
(4) That scaling laws be established and verified for tests performed on models or mock-ups and the test configurations clearly defined and documented.
(5) That the results of model test work be subject to error analysis, where applicable, prior to use in final design.
2.
Method of Accomolishment The preceding objectives were acccmolished by:
a.
A review of the Babcock and h icox Topical Report, BAW-10096A, Section 3.0, Design Control, which summarizes the methods used by Babcock and Wilcox, Nuclear Power Generation Division to describe their activities related to the design verification process. The design verification process is defined, implemented, and enforced according to the following procedure:
(1) Quality Assurance Manual 19A-N.1, Section 3.0, Design Control.
(2) NPG-0405-30, Design / Test Documentation and Verification; NPG-0403-11, Signatures on Engineering Documents; NPG-0414-12, Preparation of Contract and Generic Licensing NPG-0405-22, Design Review; NPG-0405-21, Change Inquiry / Authorization (CI/A); UPG-0402-01, Processing of NPGD Prepared Calculations; NPG-0411-02, Preparation and Revision of Performance Criteria Specifications; NPG-0902-07, Development, Verification, and Release of Standard Mct5ematical Software; NPG-0412-55, Preparation of Design Specifications /
Requirements Documents; NPG-0412-56, Processing of Design Specifications./ Requirements Documents; NPG-0412-57, Preparation of NPGD Detail Design Documents and Plant Arrangement Drawings; NPG-07: 2-58, Processing of NPGD Detail Design Documents and Plant Arrangement Drawings.
b.
Assuring that the design verification effort is clearly documented in the following documents:
(1)
Equipment Specifications:
No. 08-1018000002-09, Nuclear Instrumentation / Reactor Protection System, May 11, 1978; No. 08-1018000006-03, Nuclear Instrumenta-tion and Reactor Protection System, August 14, 1978.
(2) System Descriptions: No. 15-402300000-04, Reactor Protection System (Nuclear and Non-nuclear Instrumenta-tion and Reactor Trip Circuitry), August 23, 1976; No. 15-402300000-01, Reactor Protection System (Nuclear and Non-nuclear Instrumentation and Reactor Trip Circuitry), February 19, 1975.
(3) Change Inquiry / Authorizations related to revisions to these documents: 88-4032-00, June 13, 1977; 88-3447-00, December 1, 1976; 88-3359-00, July 13, 1976; 88-3310-00, July 22, 1976; 88-2459-00, November 11, 1975; 88-1821-01, April 30,1975; 88-2391-00, July 9,1975; 88-4646-00, May 25, 1978; 88-4719-00, July 24, 1978; 88-4740-00, August 3, 1978.
(4) Drawings; 1003874EO, Reactor Protection System, October 16, 1978; 1002006B0, Logic for Control Rod Withdrawal Inhibit and Source Range Detector Power Supply Turn Off, June 3,1977.
c.
Verifying the extent of design verification in the following documents:
(1) Contract Technical Requirement:
No. 24-1003957-00, March 28, 1978.
(2) Applicable Documents List: No. 21-1005803-01, November 30, 1978.
(3) Procurement Authorization: No. 83-760115-15, November 30, 1978.
~
. (4) Change Inquiry / Authorizations related to revisions to these documents: 88-4883-00, November 7, 1978; 88-4520-00, July 31, 1978; 884545-01, September 22, 1978; 88-4833-01, November 3, 1978.
d.
Assuring that the following documents were reviewed for identification and documentation of the method by which design verification is to be performed:
Specifications: 08-1000196-02, Reactor Trip Switch Station NSS205 Standard Plant Design, October 12, 1976; 08-109700000-03, The Integrated Control System, March 14, 1978, e.
Identifying the items to be considered during design
. verification:
(1) System Requirements Specification: 35-6031000004-01, Reactor Protection System (Nuclear and Non-nuclear Process Equipment and Reactor Trip Circuitry), October 17, 1978.
(E) Document Release Notice, January 29, 1979.
f.
Identifying the requirements prescribed for performing design verification by alternate calculations:
(1) Calculations: 32-4602-01, Bases for RP!-II Functional Requirements, September 29, 1978; 32-4409-00, Maximum Ramp Rates for RPS Sensor Qualification, July 18, 1978; 32-4352-02, RPS Setpoints for Accident Analysis, October 11, 1978.
(2) FSAR: Bellefonte Nuclear Final Safety Analysis Report, Volume 6, RPS Trip Setting Limits, Table 7.2.1-2.
g.
Identifying the requirements prescribed for performing design verification by qualification testing:
(1) Specification:
11-1305000002-01, Seismic Design Criteria and Qualification Standards of Class IE Control and Instrumentation Equipment, July 23, 1976.
(2) Seismic Test Rerarts:
BAW 58-0451-00, Signal Converter Module, June 7, 1978; BAW 58-0512-00, 885 Reactor Trip Module, September 6,1978; BAW 58-0491-00,
I i
i
_g.
Detector Power Supply Module, August 16, 1978; BAW 58-0405-00, 885 Pressure Level Test Module, Maren 21, 1978.
3.
Findings In this area of the inspection, one (1) deviation from commitment was identified (See Notice of Deviation, Enclosure). The following unresolved item was also identified during this inspection:
During a review of the RPS Trip Setting Limits in Table 7.2.1-2 of the Bellefonte FSAR, the inspector noted that the pressurizer high and low level trip points of 380" and 125" respectively did not agree with a recene snalysis:
Calculation No. 32-4352-02, October 11,197E, which specified the pressurizer high and low level trip points as 310" and 80" respectively.
10 CFR 50.b.(4) requires that the Final Safety Analysis Reports "A final analysis i
and evaluation of the design and performance of structures, systems, and components with the objective stated in (a)(4)* of this section and taking into account any pertinent information developed since the submittal of the preliminary safety analysis report."
- (a)(4) refers to preparation of preliminary analyses.
The inference abo've to " final analysis" indicates that information should not be submitted to NRC/NRR which is " preliminary" in nature.
This item will be forwarded to NRC/NRR for resolution of what constitutes " final analysis" for purposes of FSAR submittals.
Also, management of Babcock.and Wilcox indicated that as yet a CI/A has not been prepared on these new values of the high and low level pressurizer trip setpoints. The inspector will follow-up on this item at the next inspection.
E.
Exit Meetina An e.,it meeting was conducted with Babcock and Wilcox management personnel at the conclusion of the inspection on February 16, 1979.
Those individuals indicated by an asterisk in Section A above were in attendance.
In addition, the following were present:
J. W. Donnell, Manager, Quality Assurance Audits C. M. Fletcher, Manager, Quality Control Surveillance B. A. Karrasch, Manager, Plant Integration Unit
. J. H. MacMillan, Vice President P. J. Motiska, Quality Assurance Internal Auditor D. H. Roy, Manager, Engineering G. R. Skillman, Engineering Staff Assistant B. W. Whitaker, Manager, General Services E. A. Womack, Jr., Manager, Plant Design The inspector discussed the scope of this inspection and the details of the findings identified during this inspection.
Management representatives of Babcock and Wilcox acknowledged the statements by the inspector with respect to the one (1) deviation presented.
O