ML20028D020
| ML20028D020 | |
| Person / Time | |
|---|---|
| Issue date: | 11/22/1982 |
| From: | Brickley R, Hale C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20028D014 | List: |
| References | |
| REF-QA-99900529 NUDOCS 8301140405 | |
| Download: ML20028D020 (12) | |
Text
{{#Wiki_filter:* ORGANIZATION: REACTOR CONTROLS, INC. SAN JOSE, CALIFORNIA REPORT INSPECTION INSPECTION NO.: 99900529/82-01 DATE(S) 10/12-15/82 ON-SITE HOURS: 23 CORRESPONDENCE ADDRESS: Reactor Controls, Inc. ATTN: Mr. J. C. Millett Vice President 1245 S. Winchester Boulevard San Jose, CA 94070 ORGANIZATIONAL CONTACT: Mr. R. K. Crum, QA Manager TELEPHONE NUMBER: (408) 246-3801 PRINCIPAL PRODUCT: Control rod drive hydraulic systems and installation of reactor pressure vessel internals. NUCLEAR INDUSTRY ACTIVITY: Reactor Controls, Inc., (RCI) consists of about 130 employees, 110 of which are located in the San Jose office. RCI is currently performing work on Grand Gulf, Rancho Seco, Shoreham, La Salle, Phipps Bend, Clinton, Nine Mile 2, and Hope Creek. it/22/F2. ASSIGNED INSPECTOR: 7. Id.I3 R. H. Bricklegteactor Systenis Section (RSS) Dste OTHER INSPECTOR (S): I I b M V APPROVED BY: M g ale, Chief, RSS Date INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 21 and 10 CFR Part 50, Appendix B. B. SCOPE: Inspection made in response to the following: (1) a construction deficiency report from Region II concerning supports installed and accepted by i l RCI that are not in accordance with approved design; (2) a potential 10 CFR Part 21 submitted by RCI concerning hydrodynamic loads on the control rod drive hydraulic system (CRDHS); and (3) allegations concerning analysis of. - the CRDHS. p ~~ / PLANT SITE APPLICABILITY: p Docket Nos. 50-416, 50-417, 50-461, 50-462, 50-410. f ,yr#x + 8301140405 821126 PDR GA999 EMVREACT 99900529 PDR
ORGANIZATION: REACTOR CONTROLS, INC. SAN JOSE, CALIFORNIA REPORT INSPECTION NO.- 99900529/82-01 RESULTS: PAGE 2 of 5 A. VIOLATIONS: None B. NONCONFORMANCES: None C. UNRESOLVED ITEMS: None D. OTHER FINDINGS OR COMMENTS: 1. Control Rod Drive Hydraulic System (CRDHS) Supports - This item concerns a 10 CFR Part 50.55(e) report submitted to Region II by Mississippi Power & Light Company (Grand Gulf, Units 1 and 2) regarding pipe support installations which were not in accordance with design requirements. The NRC inspector examined the records concerning this item consisting of internal and external correspondence, QC Hold Information Reports, and the related drawings and engineering change notices (ECN). The examination of RCI records disclosed that the nonconforming supports were identified by RCI site QC personnel during the performance of reinspections of previously accepted supports. All QC holds had been dispositioned to repair the deficiencies and/or a stress reanalysis performed, and ECN's were issued and incorporated into applicable drawings. In addition, site QC personnel have received additional training on the necessity to follow the requirements of their QA procedure for inspections. There were no nonconformances or unresolved items identified in this area of the inspection. 2. CRDHS Hydrodynamic Loads - This item concerns a potential 10 CFR Part 21 report made by RCI to the Office of Inspection and Enforcement regarding the results of a CRDHS piping stress' analysis for the startup scram condition at Grand Gulf, which indicated that the piping is overstressed by factors as high as six times ASME Code allowables due to pressure peaks and high displacement effects.
ORGANIZATION: REACTOR CONTROLS, INC. SAN JOSE, CALIFORNIA REPORT INSPECTION NO.: 99900529/82-01 RESULTS: PAGE 3 of 5 The NRC inspector examined the records concerning this item consisting of internal and external correspondence, QA instructions, and specifications. The examination of RCI records disclosed that they had not previously considered hydrodynamic loads resulting from the scram function in their analysis of the CRDHS piping because RCI had not been provided the correct scram valve opening time. RCI had been using a 350 millisecond (msec) opening time for the scram inlet valve, which reportedly produced negli-gible hydrodynamic loads. This value was obtained from General Electric Specification No. 21A8781 (Control Rod Design and Performance Require-ments, Revision 2, dated May 20, 1975) which in Section 5.2.10 stated, " Scram valve time from pilot valve voltage interruption: (see 6.2.3) 100% open - 0.35 sec. max." In performing the stress analysis for Grand Gulf (BWR-6), RCI was specifically requested to consider the hydrodynamic loads of this system for the first time; therefore, requesting confirma-tion of the valve opening time. The response indicated an opening time of 20 msec; and, upon further inquiry, that this value also applied to earlier units of the BWR. When the hydrodynamic loads using the new value were considered in the analysis, the resultant overstress condition previously mentioned was identified. Recent data from tests conducted at Grand Gulf during preoperational scram checks indicate that the previous analysis gave results that were approxi-mately 20% higher than the measured peak pressures. Currently, RCI is conducting a reanalysis of the CRDHS on those plants with active contracts l consisting of Grand Gulf, Nine Mile Point, Clinton, River Bend, and Hope Creek. There were no violations, nonconformances, or unresolved items identified in this area of the inspection. 3. Analysis of the CRDHS - This item concerns allegations by an anonymous individual to Region V alleging that: (1) pool swell loads were not analyzed in the manner required by General Electric Containment Report-4 (CLR-4); (2) the control rod insert and withdraw lines were not analyzed for the pipe break condition; (3) " brow beating" analysts to sign computer printouts which represented system problems; (4) stress problems and load reports have no QA standards or controls; and (5) seismic category lines are not being analyzed in the proper fashion. The NRC inspector exarsined the records concerning these items consisting of QA instructions, two engineering control checklists, seven analysis i { l
j l ORGANIZATION: REACTOR CONTROLS, INC. SAN JOSE, CALIFORNIA REPORT INSPECTION NO.: 99900529/82-01 RESULTS: PAGE 4 of 5 files, two design analysis outlines, two computer program verification files, and two specifications. In addition, the NRC inspector privately interviewed five randomly-selected individuals involved in the performance / checking of CRDHS analyses. a. Allegation No. 1 "The pool swell loads have been made static equivalent loads for analysis, instead of being done correctly per CLR-4, which says they must be analyzed by time history method." This allegation could not be substantiated. The examination of ~ records and discussions with RCI representatives did not disclose any reference to or knowledge of document CLR-4 or the requirement to use the time history method in the analysis. Subsequent to the inspection, the inspector identified CLR-4 as a General Electric document titled, " Containment Load Report," Revision 4, dated January 1980. A review of Revision 3 to this docu-ment and discussions with GE representatives failed to disclose any requirement that pool swell loads had to be analyzed by the time history method or that the time history method was a more conserva-tive or correct analytical method. b. Allegation No. 2 " Pipe break for the high pressure control rod insert and withdraw lines were ignored even though they were covered under ' Design Analysis Outline' from Sargent & Lundy for high energy criteria." This allegation could no't be substantiated. The examination of the design analysis outline for Clinton, Unit 1 (SA-463-DAO), that had been reviewed and accepted by Sargent & Lundy, disclosed that piping and branch connections equal to or less than a nominal pipe size (NPS) of 1 " were excluded from high energy pipe rupture analysis. The same exclusion was found in Bechtel Specification 9645-M-316.0 and the FSAR for Grand Gulf. The insert line is 1-\\" and the with-draw line is 1" NPS; and, thus, are excluded from the analysis, c. Allegation No. 3 "The brow beating of analysts to sign computer prict-auts, which represent system problems. . Any checker not rubl6r stamping printout, ' Problems,' are either layed-off or trans-ferreu to some other work." This allegation could not be substantiated. The private interviews with five randomly-selected individuals involved in the performance / checking of CRDHS analysis did not disclose any feeling of " brow beating" or that they felt threatened in any way by management.
ORGANIZATION: REACTOR CONTROLS, INC. SAN J03E, CALIFORNIA REPORT INSPECTION NO.: 99900529/82-01 RESULTS: PAGE 5 of 5 d. Allegation No. 4 "The stress problems have no quality assurance standards incorporated in the organization of information, to print-out, to report. There are no titles, table of contents, check-lists, signoff sheets, assumptions, stress or load summaries or any attempt to organize analysis into any coherent definition of problems... No QA controls of reports, contents, organization, or procedures exist, or is apparent." This allegation could not be substantiated. The examination of QA instructions, project instructions, and seven analysis files disclosed that the alleged items were adequately addressed and were being implemented. e. Allegation No. 5 "The seismic category 1 over 2 lines are not being analysed (sic), or even being considered in the proper fashion." This allegation could not be substantiated. The examination of client specifications and analysis files did not disclose any deficient areas. l l t l l
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