ML20197B237

Insp Rept 99900404/86-01 on 860106-10.No Violations or Nonconformances Noted.Major Areas Inspected:Implementation of 10CFR21 Reporting Sys & Selected Items Being Reviewed by Safety Review Committee

ML20197B237
Person / Time
Issue date:	04/30/1986
From:	Jocelyn Craig, Milano P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
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ML20197B242	List: ML20197B237 ML20244D831
References
REF-QA-99900404 NUDOCS 8605120486
Download: ML20197B237 (10)

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CRGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION INSPECTION NO.- 99900404/86-01 DATE- 1/6 10/R6 JN-9fTF HOUPS- 91 CORRESPONDENCE ADDRESS: Westinghouse Electric Corporation Nuclear Technology Division ATTN: Mr. J. L. Gallagher General Manager Post Office Box 355 Pittsburgh, Pennsylvania 51230 ORGANIZATIONAL CONTACT: Mr. P. T. McManus, Quality Assurance To FpunNF Nf!MAFR- l .i l 7 4 97R.7QRR NUCLEAR INDUSTRY ACTIVITY: Westinghouse provides NSSS components, other safety and non-safety related components, and services.

ASSIGNED INSPECTOR: ~~P I) (7sst_ . _- 4 t (

P. D. Milano, Special Projects Inspection Section Date (SPIS)

OTHERINSPECTOR(S): R. L. Pettis, SPIS P. Prescott, SPIS W. Banister, &G APPROVED BY- . p ( r. Y id-dl

,dirJohh W. Craig, Chief, SPIS. Ven' dor Program Branch Date T

1 INSPECTION BASES AND SCOPE:

A. BASES: 10 CFR Part 21 and 10 CFR Part 50.

B. SCOPE: The purpose of this inspection was to review implementation of changes to Westinghouse's 10 CFR Part 21 reporting system and review technical t. sis for selected items being evaluated by the Westinghouse Safety Review Committee.

PLANT SITE APPLICABILITY: Wolf Creek, 50-482; South Texas Project, 50-498/499; Vogtle, 50 424; Point Beach 50-266/301.

8605120486 860508 PDR GA999 ENVWEST 99900404 PDR

ORGANIZATION: WESTINGH0USE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.- 99900404/86-01 RESULTS: PAGE 2 of 10 A. VIOLATIONS:

None.

B. NONCONFORMANCES:

None.

C. UNRESOLVED ITEMS:

Westinghouse had not completed an evaluation of a problem discovered with an Engineering Safeguard Actuation System (ESAS). On-line testing of the reactor trip interlock (P-4) could not be accomplished, and only the P-4 relay for the Safety Injection block / reset function was presently covered by a test procedure. This could constitute an unreviewed safety question as defined in 10 CFR 50.59 if failure of the other P-4 contacts caused the Technical Specification or design limits to be exceeded. Therefore, this evaluation will be reviewed during a future inspection. (86-01-01)

D. STATUS OF PREVIOUS INSPECTION FINDINGS:

1. (Closed) Violation A (84-02) - Westinghouse failed to include in their report to the NRC the number and location of all defective Barton transmitters in use at, supplied for, or being supplied for one or more facilities or activities subject to the regulations in this part.

Westinghouse has modified Policy WRD-0PR-19.0, " Identification and Reporting of Substantial Safety Hazards, Significant Deficiencies, and Unreviewed Safety Questions," to include a statement for those plants under construction. In these cases, reports will be made to the NRC of these deviations which could create a substantial safety hazard if the plant were to have gone into operation without the deviation having been corrected.

Westinghouse had informed all utilities through an NSID Technical Bulletin and all corrective actions at Construction facilities and was completed. Issuance of a change to the 10 CFR 21 report to include the construction stage plants at this time would serve no function. Based upon the preceding, this item is closed.

2. (0 pen) Violation B (84-02) - Procedures did not specify what was to be in the evaluation record or when the record was to be prepared.

Consecuently, as of May 2,1984, records were not adequate to ensure compliance with 10 CFR Part 21.21(a)(1) as follows:

ORGANIZATION: WESTINGH0USE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.- 99900404/86-01 RESULTS: PAGE 3 of 10

a. The evaluation records for Identified Item (ID)82-200, concern-ing Westinghouse Type AR relays, did not support the determ-ination that the item was not reportable to the NRC.

b. The evaluation record for ID 82-198, concerning steam generator J-tube failures, did not support the determination that the item was not reportable to the NRC subseouent to the 1983 Surry Unit 2 failures.

c. The evaluation records for Potential Items (PI)82-162, concern-ing non-seismic panels, and PI 82-154, concerning valves not fully qualified, did not support the decision not to refer them to the Safety Review Comittee (SRC) or the decision not to report the items to the NRC, WRD-DPR-19.0, Rev. 2, contains a requirement that evaluations must be prepared and retained. However, the specific records that must be included during the evaluation process is a function of the departmental procedures. This item will remain open until such time that these procedures can be reviewed and the new processing evaluated.

3. (Closed) Nonconformance B (84-03) - Procedures were not available to review the effects of computer program and system errors on design.

Westinghouse procedures have been modified to include requirements that correct this deficiency. This item is closed.

4. (Closed) Nonconformance C (84-03) - A proposed design change was not reviewed and verified by all functions involved in the original design.

The design change where this deficiency was noted has now received the required review. This item is closed.

E. OTHER FINDINGS OR COMMENTS:

1. The Westinghouse Water Reactor Divisions (WRD) 10 CFR Part 21 Reporting System: The NRC inspectors reviewed W procedure WRD-0PR-19.0, Rev. 2, dated December 1,1985, for the identification and reporting of substantial safety hazards, significant deficiencies, and unreviewed safety questions. Westinghouse personnel provided a brief presentation on the implementation of this procedure. The procedure states, in part, that in the event the WRD SRC believes

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA t

REPORT INSPECTION NO - 99900404/86-01 RESULTS: PAGE 4 of 10 that a potential significant deficiency or unreviewed safety question should be reported to an affected customer (s) and/or the NRC, the committee will notify the general manager (s) of the affected division (s), with a recommendation for the action as appropriate. In this manner, all potential significart deficiencies and unreviewed safety questions will be reviewed by the SRC and the recommendations of the committee will be acted upon by the general nanager(s) of the affected WRD division (s) as a prerequisite for reporting action.

Westinghouse Procedure WRD-0PR-19.0 had been revised recently as part of a program to decentralize the system for implementation of procedural requirements. The specific details for implementing the requirements have now been placed in each department's instruction.

Thus, WRD-0PR-19.0 provides only the overall objectives of the program leaving the specifics to each division and/or department's discretion.

It should be noted that the emphasis of this inspection was on the evaluation process conducted by the Westinghouse SRC. The determ-ination process within each department as to whether a deficiency need be presented to the SRC as a Potential Item (PI) was not reviewed. Thus, the inspection centered on a review of these deficiencies that had been determined to be potentially safety concerns. In this area, the evaluations for reportability, justifications for continued operation during the review period, and the determination of affected plants were reviewed.

2. Review of Westinghouse WRD Potential Item (PI) Files: The NRC inspectors reviewed the WRD PI files reauired by Westinghouse Procedure WRD-0PR-19.0. The PI files contain meeting minutes and information utilized by the Westinghouse SRC in their evaluation for reportability. The PI files included, in part, records related to evaluations or notifications to the Commission, copies of customer notification letters, conformation notices that all affected customers have been notified, and copies of NRC notification letters issued.

The following are the PI files reviewed, with a brief description of the files content and any discrepancies noted by the NRC inspectors.

a. File No. PI-81-143 contained information concerning a potential problem with a study performed to determine the impact of a steam line break analysis on the chiculated peak containrent pressure used to evaluate containment integrity for Unit 1, owned by Furnas-Centrais Electricas S.A.

ORGANIZATION: WESTINGH0USE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.- 99900404/86-01 RESULTS: PAGE 5 of 10 An analysis was completed based on up-to-date containment and auxiliary feedwater data supplied by the Architect-Engineer (A/E) Gibbs & Hill. The results were also based on a current analytical method which showed that the peak containment pressure remained below the design limit of 46 psig, including the effects of leaking MSIVs. The potential problem thus was considered closed.

During the review of this file one (1) discrepancy was noted. I&E Information Notice No. 82-11, concerning potential inaccuracies in wide range pressure instruments used in Westinghouse plants was included in the file. This notice did not pertain to the original analysis and appeared to be misplaced in this file. This fact was brought to the attention of Westinghouse.

b. File No. PI-81-143 was established to evaluate a problem discovered with an Engineered Safeguard Actuation System (ESAS) modification which was to allow on-line testing of the Reactor Trip Interlock (P-4). This interlock in the Engineered Safety Features Actuation System utilizes contacts in the reactor trip breakers to verify reactor trip. In a letter to the NRC dated November 7, 1979, Westinghouse notified the NRC of the potential problem. On March 22, 1985, Westinghouse notified the NRC that the design modification to allow on-line testing still allowed the possibility for undetectable failures. The letter also stated that a previously issued test procedure would be utilized in the performance of the required tests.

However, this notification dealt only with the P-4 contacts that permit an operator to block safety injection actuation.

Westinghouse internal letter, NS-I&CSL-85-101, dated March 15, 1985, stated that the reported P-4 problem was not complete.

In addition to the P-4 contacts of SI block and reset, none of the other auxiliary switch P-4 contacts were being tested.

This letter stated that because of this failure to adequately test P-4 contacts, Westinghouse would have to test them or justify why their failure would not prevent completion of a safety function.

The March 15, 1985 letter also indicated that the NRC Electrical

& Instrumentation and Control Systems Branch had, in a telephone conversation to Westinghouse, asked about the tests of the other P-4 contacts. The Westinghouse reponse was that the item was

i ORGANIZATION: WESTINGH0USE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.- 99900a04/86-01 RESULTS: PAGE 6 of 10 being reviewed at that time. Finally, the letter discussed setting up a meeting to decide which contacts (functions) needed on-line testing.

Another Westinghouse internal letter, from Control System Design and Analysis to Nuclear Safety Department, dated June 4, 1985, indicated that a meeting was held within Westinghouse on May 16, 1985 to discuss the cnneern over inability for on-line testing. The letter stated that an evaluation would be required to determine the consequences of either inadvertently actuating the P-4 permissive, or else not clearing the P-4 permissive when restarting af ter a trip. The letter also provided an evaluation of the effects of a P-4 permissive failure on the steam dump control system for plants that have the turbine trip without reactor trip option.

Westinghouse internal letter, NS-RAT-PTA-85-220, dated July 17, 1985, provided as an attachment the safety evaluation of the consequences of a P-4 failure during a turbine trip on reactor trip and on feedwater isolation on low reactor coolant system temperature.

While the preliminary conclusion for a turbine trip failure was that it would probably be bounded by the analysis for a steam-line break, the report also concluded that "the basis of Technical Specification and design limits such as heatup and cooldown curves or thermal cycling limits may be violated." If correct, this would constitute an unreviewed safety question as defined by 10 CFR 50.59, since the margins of safety as defined in the basis for the technical specification would be reduced.

Based on this, Westinghouse Plant Transient Analysis Group recommended in this letter that a procedure be developed to periodically test the P-4 contacts for operability.

From the evaluation conducted for a P 4 failure and its effect on feedwater isolation, Westinghouse concluded that the transient would be mitigated by other safety functions. Normally the presence of the P-4 signal closes the main feedwater regulating below setpoint. The P-4 signal also permits valvesonT@locktoisolatefeedwateronlowcompensated the P-15 in T and steamline and feedwater isolation on low-low c8hn,satedTcold (excessive cooldown protection).

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.- 99900404/86-01 RESULTS: PAGE 7 of 10 Based on an interview with a Westinghouse engineer, it was stated that the safety analysis to support the scenarios would be completed in February 1986. Also, if it is determined by Westinghouse that en-line tests are needed for these other P-4 contacts, approximitely twelve different tests would be required to cover the var;ous Westinghouse designs.

The PI file did not contain any information nor was Westinghouse able to provide evidence that the NRC was aware of the test-ability issues discussec in the March 15, 1985 letter.

c. This is Unresolved Item 86-01-01. File No. PI-85-011 was opened on May 23, 1985 after Westinghouse was informed by Rosemount that the Westinghouse design specification requirerent for accuracy at elevated temperature of 130 F with less than a 0.5% change in reference accuracy from 80 F could not be met by Rosemount Model 1153 transmitters. This was stated in an NSID internal memo, EQ/EI (85) - 150, dated April 3, 1985.

Westinghouse internal letter, NS-I&CSL-85-117, dated May 17, 1985, stated that setpoint studies were being done to assess the impact on the affected channels. Evaluations performed up to that date indicated that the allowance for overall channel accuracy could accommodate the shift. Also, the letter stated that a quantitative study, including a safety analysis impact, was needed prior to Wolf Creek initial criticality. However, there was no information in the file which indicated that the final results of a study for Wolf Creek or the South Texas Project, the affected units, had been conducted.

d. File No. PI-83-193 contained information concerning a potential problem item identified on July 6,1982 with equipment start on low voltage coincident with a seismic event for the Vogtle plant. Af ter about eight months of consideration, on February 16, 1983, a PI number was assigned to the problem.

By June of that year (1983) the technical and risk assessment technology organizations had completed their analysis and recommendations for the problem. On January 24, 1984 the Nuclear Safety Department reiterated their position on the PI problem and recommended closure. On February 9, 1985 the PI was determined to be not reportable and was closed. The file did not contain any information which addressed whether or not operating plants should continue operation while reporta-bility was being determined. Though not identified by plant l

l t

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION No.- 90000404/86-01 RESUtTS: PAGE 8 of 10 name, early correspondence implied wide generic implications, but this was changed after document WCAP10598 "Justificatwn for Reduced Voltage Operating During a Seismic Event" was issued.

e. File No. PI 83-206 was opened following a Westinghouse evaluation for Wisconsin Electric (Point Beach) of a postulated reactor vessel head drop accident as discussed in NUREG-0612

" Control of Heavy loads at Nuclear Power Plants." An excerpt from the Westinghcuse analysis (Letter No. PT-SSD-527) submitted to Point Beach in November 1982 stated that the analysis indicated that although the reactor vessel primary nozzles would not be stressed above allowable limits, the vessel supports may' collapse and cannot be relied upon to prevent severe damage to the reactor coolant loop piping.

While the plant is on RHR during a refueling outage, permanent damage to the reactor coolant pump and steam generator supports may result from the accident and include primary coolant piping damage. This may affect loop nozzles used for decay heat removal systems. The PI file did not contain any letters raising this issue to the Safety Review Committee. Also there was no evidence presented to show that Wisconsin Electric Power Company requested additional information from Westinghouse in regard to the conclusion reached on the collapse of the vessel support legs.

This PI has been open since 1982 with a possible formal closeout set for February 1986,

3. NSD Safety Evaluation Checklist Westinghouse Policy WRD-0PR-2.3, Rev. O, dated December 1, 1985, requires that written safety evaluations to be prepared by NTD Nuclear Safety after a request is received from a Westinghouse organization. These safety evaluations are performed primarily to support design changes for operating plants. However, as a secondary function, the NSD can use the safety evaluations to document possible safety-related questions. This function while described in NSD Instruction NS-0PLS-IG-3 is not defined in Procedure WRD-0PR-2.3.

After the actual safety evaluation has been completed, a Safety Evaluation Checklist is completed. This checklist provides a means for documenting necessary FSAR and/or technical specification changes. Additionally, the results of the safety evaluation and reference to the actual safety evaluation is documented on this form.

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.- 99900404/P6-01 RESULTS: PAGE 9 of 10 Since the Nuclear Safety Cepartment performs safety evaluations to support the evaluation of potential safety concerns, there may be Potential Item (PI) files which are opened based on the outcome of these evaluations. It is not a requirement to file these checklists within the PI file or to cross-reference the unique checklist number of the PI file number. Thus, it is difficult to correlate those checklists which support PI files.

The departmental implementing instructions for NSD are in the form of internal letters rather than the typical procedure format. These letters provide instructions obtaining checklist (SECL) log numbers and the information on the checklists. Within the Operating Plant Licensing Support group, the instructions of the use of the checklist is in Instruction and Guidance NS-0PLS-IG-10.

While reviewing several completed SECLs, several administrative deficiencies were noted. The block to be utilized for recording the customer or Westinghouse references is not being used. Also, the " Justification" section is not always complete in its description of the evaluation results. Finally, the reference to the actual safety evaluation report is not always provided in the appropriate block.

Safety Evaluation Checklist SECL 85-097, 7300 System NTC Card Podification, was prepared on December 11, 1984 to support the implementation of a temporary modification to the Temperature Test Card (NTC) in the Model 7300 Process Protection System. This problem was reported to the NPC on January 1, 1983 in Westinghouse letter, NS-EPR-2774. The problem (PI 82-79) was that the contact bounce of the mercury wetted test relay during seismic events could cause up to a 20 second delay in the performance of the over temperature delta temperature and over power delta temperature trip function. The SECL stated that bypassing the relay eliminated the source of concern. This SECL did not contain a reference to either a customer or Westinghouse document to indicate why the evaluation was being performed.

SECL-85-286, dated July 18, 1985, duplicated the same information as SECL-85-098 while adding more affected units. This SECL did reference a Field Change Notice FCN-6258. However, neither of the SECLs discussed the fact that on-line testing could no longer be l

performed using the modified card. This fact was stated in FCN-6258 along with a means to replace the card during testing. Thus, neither of the SECLs described the complete scope of the problem or the outcore of the intended fix.

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ORGANIZATION: WESTINGHOUSE ELECTRIC COPP0 RATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.- 09900404/86-01 RESULTS: PAGE 10 of 10

4. Time of Response During the review of PI files the time utilized by Westinghouse in processing Pls was noted. The following table reflects the length of time from when the problem was first noted and the PI file was opened until the time for PI review to closure. It was also noted that the elapsed time from the identification of a potential problem item to the assigning of a PI number has increased in 1985.

PI Number Status Problem to PIA Pld to Closed 82-151 Closed 2 day 3 yr 83-193 Closed 8 mth 2 yr 83-216 Closed 1 mth 23 months84-268 Closed 1 wk 8 wk 84-247 Closed 2 wk 5 mon 83-209 Open 5 wk (2iyr){hs84-240 Open I wk (22 months)y 85-020 Open 6 mth (3 months) y 85-011 Open 7 wk (8 months) y 85-027 Open 2 wk (2 months) y

5. Westinchouse Radar Reports During the inspection, the inspectors requested copies of the Westinghouse Radar Response Forms. These forms were in the style of information letters to utilities and described problems encountered with equipment. The Radar system, however, no longer exists. Upon the request by the inspectors, Westinghouse committed to providing copies of the Response Forms. As of the date of this report, the forms have not been received.

(1) PI still open. Time indicates period from opening of PI file to time of NRC inspection.

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