ML17250B199

Responds to NRC 900509 Ltr Re Violations Noted in Insp Rept 50-244/89-81.Corrective Actions:Improved Battery Load Profile Developed Incorporating Calculational Improvements Contained in Current Industry Std IEEE 485-1983

ML17250B199
Person / Time
Site:	Ginna
Issue date:	06/08/1990
From:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	Martin T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
NUDOCS 9006200487
Download: ML17250B199 (37)

Text

x REGULATOY INFORMATION DISTRIBUTZOYSTEM (RIDE) x~

x ACCESSION NBR:9006200487 DOC.DATE: 90/06/08 NOTARIZED: NO DOCKET

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50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION CREDY,R,C.

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RECIP.NAME

~ RECIPIENT AFFILIATION MARTINET.T.~ Region 1, Ofc of the Director

SUBJECT:

Responds to NRC 900509 ltr re violations noted in Insp Rept 50-244/89-81.

DISTRIBUTION CODE: IEOZD TITLE: General (50 COPIES RECEIVED:LTR 1 ENCL Dkt)-Insp Rept/Notice of Violation Response 3 SIZE:

NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 c RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-3 PD 1 1 JOHNSONFA 1 1 INTERNAL: AEOD 1 1 AEOD/DEIIB 1 1 AEOD/TPAD 1 1 DEDRO 1 1 NRR MORISSEAU,D 1 1 NRR SHANKMAN,S 1 1 NRR/DLPQ/LPEB10 1 1 NRR/DOEA DIR 11 1 1 NRR/DREP/PEPB9D 1 1 NRR/DRIS/DIR 1 1 NRR/DST/DIR 8E2 1 1 NRR/PMAS/ILRB12 1 1 NUDOCS-ABSTRACT 1 1 OE~I 1 1 OGC/HDS2 1 1 REG~ 02 1 1<

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TOTAL NUMBER OF COPIES REQUIRED: LTTR 22 ENCL 22

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55455 ROCHESTER GAS AND ELECTRIC CORPORATION ~ 89 EAST AVENUE, ROCHESTER, N.Y. 14849-PPPI June 8, 1990 TEEER<04C AREA CODE 7555 546 2700 Mr. Thomas T. Martin Regional Administrator Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406

Subject:

Response to Inspection Report 50-244/89-81 Safety System Functional Inspect'ion -- RHR System R. E. Ginna Nuclear Power Plant NRC Docket 50-244

Dear Mr. Martin:

This letter provides the initial 30-day response to the Safety System Functional Inspection (SSFI) of the Residual Heat Removal (RHR) System at the R. E. Ginna Nuclear Power Plant, conducted, between November 6 and December 8, 1989. The NRC letter of May 09, 1990 from Marvin W.

Hodges (NRC) to Robert C. Mecredy (RG&E) transmitted the report for that inspection. This letter provides the RG&E responses, pursuant to 10 CFR 2.201, to the two violations issued in conjunction with the SSFI report. In addition, we are providing schedule information concerning the unresolved issues, including the postulated flooding of the RHR room, identified in the inspection report. Additional information will be provided in the 120-day response to the SSFI report.

The nuclear industry is going through major upgrade efforts involving configuration management and design basis documents. RG&E is not alone in recognizing the benefits of these improvements and has been proceeding with these efforts. On March 6, 1990 RG&E made a formal presentation to NRC Region I staff and on March 27, 1990 made a presentation to NRR regarding our configuration management program. We have completed three pilot system design basis documents and are reviewing them to determine the optimal specification for the overall design basis document program for the remaining plant systems. In addition, RG&E has developed a separate program to provide further assurance that all design basis information and, commitments which may have been relied upon by the NRC are captured.

The objective of the NRC SSFI of the RHR systems was to assess the capability of that system to perform its design basis functions. As part of that inspection, the SSFI team assessed the overall design control program and other work processes used by RG&E. The review of these programmatic aspects was far broader than the RHR system.

Special emphasis was placed upon the engineering processes and their interfaces with other activities.

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2 The primary result of the SSFI was that no situations were identified that would prohibit the RHR system from performing its intended functions under normal and design basis accident conditions. As would be expected from an SSFI of any nuclear power plant, and in particular one of the early SEP plants, the SSFI identified. areas where improve-ment is warranted. Two Severity Level IV violations were cited., and ten specific unresolved items were documented.

The NRC letter of May 09, 1990 requires that the violations be addressed, pursuant to 10 CFR 2.201, within 30 days. The letter also requests that RG&E provide its evaluation of the specific unresolved items and planned actions, within 120 days. In addition, the NRC letter requests that RG&E also provide schedule information regarding the actions to address the unresolved items, within 30 days. The schedules requested are exclusive of unresolved item 89-81-11, Engineering Assurance, for which a response was requested in 120 days.

k Responses to two violations identified ar'e provided as Enclosures A &

B to this letter. The first violation involved. not maintaining an up-to-date load profile for the batteries. The actual capability of the batteries was not an issue, only the adequacy of the testing. RG&E had already reached a state of full compliance on this matter when the SSFI report was received.

The second violation cited had two parts. The first part involves having not already developed a periodic testing program for the molded case circuit breakers. The second part involves not having an explicit acceptance criterion in the test procedure for the setpoints of the dc undervoltage alarm relays. Although a generally accepted periodic test

>> method for molded case circuit breakers is not available in the industry today, we choose not to take issue with this violation. The industry is currently examining the need for and/or requirements for molded case circuit breaker testing. RG&E will implement, when .

available, those testing methods and requirements endorsed by the industry. With regard to the acceptance criterion for the undervoltage relay setpoints, we had already reached a state of full compliance on this matter when the SSFI report was received. In addition, on our own initiative, we have expanded this concern to include the ac undervoltage relays for the safety buses.

In addition to these violations, NRC also identified ten unresolved items. The identification of these items is contained in Enclosure C.

Several of these unresolved items have already been completed and several more are in process.

During the RG&E review of the SSFI report, management recognized that many of the unresolved items were examples of broader, underlying, programmatic concerns. Many of these concerns focused on engineering functions and, controls'. Because RG&E understands the importance of resolving the programmatic and management issues as well as the specific items cited by the NRC, we are developing a systematic approach to address both types of concerns. This approach is a two-part, parallel effort. The first part focuses on the management processes in a disciplined manner, while the second. part focuses on the resolution of the specific unresolved items.

P To begin the review of the broader concerns, we have re-reviewed the SSFI report and the cited issues, and have categorized them into general topical areas. For example, unresolved item 89-81-05 involves not having a mechanism to assure that design calculations are main-tained up-to-date. We see this specific item as being part of a more general area called design control. Enclosure D is a preliminary categorization of the unresolved items into the general topical areas.

In addition, RG&E is initiating a more detailed review of the work processes and their controls for each of the general areas which contain significant identified, weaknesses. This review will encompass identifying the cause of the violations, as well as the unresolved issues, identified by the SSFI report.

Enclosure E contains the schedular information as requested by the staff. We have separated this schedule information into two catego-ries: resolution completed and scheduled for resolution. RG&E has resolved items 89-81-04, 06, 07A, and 10 as identified in Enclosure C. In particular, RG&E has promptly resolved the issue regarding flooding of the RHR pump room. The UFSAR has been updated, and the EOPs and training documents have been revised. A detailed account of those actions taken to resolve the items identified above are con-tained in Enclosure E.

RG&E believes that the approach outlined in this letter assures proper and complete resolution of the specific issues identified as well as the more programmatic issues discussed.

Very truly yours, Robert C. Mec e y Division Manager Nuclear Production GAHN108 Enclosures xc: U.S. Nuclear Regulatory Commission (original)

Document Control Desk Washington, D.C. 20555 Allen R. Johnson (Mail Stop 14D1)

Project Directorate I-3 Washington, D.C. 20555 Ginna NRC Senior Resident Inspector

0 ENCLOSURE A Response to Notice of Violation 50-244/89-81 Violation 1

.Ins ection Re ort 44/88-81 VIOLATION 1:

STATEMENT OF VIOLATION-10 CFR 50, Appendix B, Criterion III, requires in part that measures be established to ensure that applicable regulatory requirements and design bases are translated into specifications and procedures.

These measures shall provide for verifying the adequacy of design by performance of design reviews.

Ginna Station Quality Assurance Manual, Section No. 11, "Test Con-trol," requires that engineering establish design test requirements and that testing be performed in accordance with approved procedures which incorporate the requirements and acceptance criteria contained in applicable Technical Specifications and regulatory requirements.

Contrary to the above, on November 15, 1989, the design reviews for Engineering Work Request (EWR) 3891 were inadequate in that the EWR did not establish the battery load, requirements thereby resulting in a battery load. profile used during the service test. not reflecting the design basis load requirements.

This is a Severity Level IV Violation (Supplement 1).

ACCEPTANCE OF VIOLATION:

RG&E agrees of EWR 3891.

that it did not update the battery load profile as part DISCUSSION:

The purpose of EWR 3891 was to replace the batteries because they were nearing the end of their service life and, while replacing them, to increase the capacity margin. 3891 did not include an updat-ing of the battery test profile EWR because it had been determined that no large loads had been added. to the battery since the original load profile had been developed.

The battery load profile was based upon the original Westinghouse design data. That information was consistent with industry practice at the time it was developed. Analytical techniques were not as sophisticated as those in use today. Rather than explicitly quanti-fying such factors as momentary loads and. the load starting currents, it was general practice to provide additional battery sizing based upon experience and engineering judgement. Today's standards (such as IEEE standard 485) suggest a more refined, more precisely quanti-fied analysis.

The actual battery capacity was sufficient to provide its safety functions. The battery has been shown to have adequate capacity as confirmed by a physical test.

Although there is no requirement for the Ginna Nuclear Power Plant to incorporate all newly-developed industry standards, we believe it prudent to use the current industry standards for developing revised battery load profiles, and have done so.

A-1

CORRECTIVE STEPS T A preliminary analysis, performed during the inspection, demonstrated that the battery size is adequate.

The revised battery size calculation had, been finalized subsequent to the NRC inspection and prior to the receipt of the inspection report, which confirms that the battery size is not a concern.

An improved battery load profile has been developed which incorpo-rates calculational improvements contained in current industry standard IEEE 485-1983.

The upgraded battery load profile (Design Analysis EWR 3341 "Sizing of Vital Batteries", dated March 12, 1990) has been transmitted by Engineering to the plant staff, and the battery testing and PT-10.3, Battery Service Tests) have been revised. The procedures'PT-10.2 batteries were tested during the recent outage using the revised procedures. The results demonstrated the adequacy of the battery capacity.

CORRECTIVE STEPS TO BE TAKEN TO PREVEBVP REClJRRENCE:

The applicability of this violation has been broadened by RG&E to assure that not only the important dc electrical loads are analyzed and tested, but also that the important ac electrical loads which may impact the operation of the plant emergency diesel generators are identified and tracked. We have implemented an electrical program as described under unresolved item 89-81-05.

load'rowth DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED:

Engineering established updated battery load requirements. The battery test procedures have been revised and the batteries have been tested using the new procedure. These actions were completed prior to the receipt of the NRC inspection report. RG&E is in full compli-ance.

A-2

ENCLOSURE B Response to Notice of Violation 50-244/89-81 Violation 2

I P

~

RG&E/Ginna Ins ecti Re ort 50-244/89-81 VIO N 2:

STATEMENT OF VIOLATION:

R. E. Ginna Technical Specifications Section 6.8.1 requires that written procedures be established and. implemented. for activities such as surveillance and testing activities of safety-related equipment.

Ginna Station Quality Assurance Manual,Section II, "Test Control,"

establishes the requirements for establishing and implementing test programs to demonstrate that safety-related systems and components will perform satisfactorily. Furthermore, this section requires that testing shall be performed in accordance with written procedures which incorporate acceptance criteria.

Contrary to the above, on December 9, 1989, Class 1E 480V ac molded case circuit breakers have not been subjected to scheduled periodic testing. Furthermore, there is no established acceptance criteria for testing the dc undervoltage relay al'arms in Procedure PT-11, "60-Cell Battery Banks 'A'

'B'his is a Severity Level IV Violation (Supplement 1).

ACCEPTANCE OF VIOLATION:

RG&E agrees that the periodic testing program of safety-related equipment at the Ginna.Nuclear Power Plant does not currently include molded case circuit breakers. RG&E also agrees that the Ginna periodic test procedure PT-11 "60-Cell Battery Banks 'A' 'B'" did not specify an acceptance criterion for the setpoint of the dc undervoltage relay alarms.

This violation has two parts which are addressed separately below:

Part 1: Molded Case Circuit Breaker Testing DISCUSSION:

Molded case circuit breakers are designed for nuclear and non-nuclear applications. This type circuit breaker is sealed. and does not include design features to test all the capabilities of the breaker beyond functional tests.

RG&E realizes the importance of assuring proper operation of these breakers. RG&E has not been lax in its attention to the importance of testing molded case circuit breakers. This problem was self-identified by RG&E and was incorporated into the RCM program. On our own initiative, we developed and implemented receipt-inspection testing for all new molded case circuit breakers at Ginna. We have also performed testing on molded, case circuit breakers in an effort to determine their characteristics.

Three years ago, RG&E performed special testing of all of its exist-ing magnetic only, molded case circuit breakers at Ginna Station on a special one-time basis. Successful operation has indicated no known degradation.

B-1

While the functioniO~ of molded case circuit hkers is important to safety and while there is an NRC requirement for a test program to assure that safety-related structures, systems and. components will perform satisfactorily, there is no specific requirement to test periodically every piece of equipment. As stated. in Appendix B, Criterion XI, "The test program shall include, as appropriate, operational tests ... of structures, systems and components." The term "as appropriate" is applicable and includes the availability of appropriate test methods. Molded case circuit breakers are not designed for in situ testing and would require determination and retermination to perform the testing. The vendors of this equipment have also not made recommendations for periodic testing. Because of generic applicability, periodic testing for molded case circuit breakers has been an industry-wide issue and no generally accepted test method has been developed at this time.

The nuclear industry has responded to the NRC through NUMARC concern-ing molded case circuit breaker testing and RG&E is pursuing this in conjunction with this effort.

CORRECTIVE STEPS TAKEN:

RG&E is continuing to work toward developing appropriate test methods for molded case circuit breakers, as part of the Reliability Centered Maintenance (RCM) program. The Ginna Nuclear Power Plant is one of the two "pilot plants" in the nation for the EPRI sponsored RCM program.

CORRECTIVE STEPS TO BE TAKEN TO PR1DGQFZ RECURRENCE-The industry is currently examining the need for, and benefits of, molded case circuit breaker testing. RG&E will continue to work closely with the industry and EPRI to determine appropriate test methods and. requirements.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED:

Although RG&E does not consider this a compliance matter, RG&E will implement, when available, those testing methods and requirements endorsed by the industry.

Part 2: Undervoltage Relay Alarm Acceptance Criteria CORRECTIVE STEPS TAKEN:

The periodic test procedure PT-11 "60-Cell Battery Banks 'A' 'B'"

has been revised to explicitly define the acceptance band/criterion for the dc undervoltage alarm relays.

The dc relays have subsequently been calibrated and tested. The relays have been verified to perform within the specified acceptance criterion.

B-2

~ 1 0

. CORRECTIVE STEPS TAKEN TO PREVIXT RE E:

The applicability of this violation has been broadened by RG&E to assure that not. only the test procedures for dc undervoltage alarm relays have explicit acceptance criteria, but also that the test procedures for the ac undervoltage relays for the safeguards buses have explicit acceptance criteria.

The test procedure, PT-11 for the dc undervoltage alarm relays has been revised and PT-9.1 for the 480V ac safeguards buses is being revised to provide explicit acceptance criterion.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED:

The test procedure for the dc undervoltage alarm relays has This been revised to provide an explicit acceptance band/criterion.

action was completed prior to the receipt of the NRC inspection report. RG&E is in full compliance.

B-3

0 t'

ENCLOSURE C Identification of Specific Unresolved Items Note:

The statements of issues have been directly extracted from the SSFI report. In a few instances the issues have been condensed and paraphrased.

r 0 0

e

. 89-81-01 Service r Single Failure Suscept lity Potential loss of cooling water [flow] to both emergency diesel generators during or following a seismic event. The cooling water lube oil heat exchanger for the water jacket heat exchanger and non-seismic discharges through a common non-safety 10-inch discharge pipe. The cooling water discharge pipe would have to fail [or has been postulated by the NRC SSFI team to fail] so as to prevent

[block/pinch off] the flow of the service water.

89-81-02 Resolution of Safety Concerns The licensee was unable to provide the team with a documented or verifiable process available at RG&E that addresses how safety concerns raised outside the normal engineering process are brought to the attention of the Nuclear Safety and Licensing staff and resolved.

89-81-03 RHR Pump NPSH A consultant independently evaluated the available NPSH during post-accident recirculation mode from the containment sump and, a prelimi-nary result indicates that there may be some modes of operation of the RHR pumps under which adequate NPSH is not available. Licensee is evaluating the validity of these modes and the probability of occurrence. Licensee is also evaluating the possibility that the consultant's analytical model was too conservative.

89-81-04 Class 1E Battery Testing Failure to test the batteries with a load profile which truly repre-sented the load demand on the battery is considered a violation of 10 CFR 50, Appendix B, Criterion III.

89-81-05 Electrical Load Growth Control Program RG&E does not have a mechanism to assure that plant calculations affected by modifications are updated to ensure that they are main-tained up-to-date and accurate. The design process provides guidance to engineers to review the system capacity and other attributes, but the guidance addresses only specific modifications as they are performed. There is no formal load tracking system to ensure that system capacity is reviewed for the integrated effect of several modifications instead of just one. The licensee stated that an on-line program to capture electrical load growth and update affected calculations would be developed.

89-81-06 Molded Case Circuit Breakers and, Undervoltage Relay Alarms Failure to periodically test the molded case circuit breakers and not establishing an acceptance criteria for the undervoltage relay alarms are a violation of facility Technical Specifications 6.8.1, which requires testing of safety-related. components in accordance with established procedures.

.89-81-07A Calibrate of Control Room Instrume The control room dc voltmeters are not calibrated on a periodic basis to ensure reliable system voltage indication to operators.

89-81-07B Control Room P&IDs Piping and Instrument Diagram (P&ID) updates and Design Change team.

Requests (DCRs) posted in the control room were reviewed by the It was noted that the RHR system P&ID (33013-1247) did not reflect the current valve position configuration for the RHR system. Also, the existing DCRs outstanding against this drawing could not be used.

to derive the correct valve positions in that DCRs 1247-4, and 1247-5 had not been approved by RG&E Engineering and did not reflect the current position of valve 822B.

Processing of DCRs does not always occur in a timely manner such that the control room P&IDs can be immediately updated. Plant operations organization makes permanent changes to system valve positions, there is not an immediate markup or annotation made on the effected draw-ings.

The team noted that permanent changes to valve positions in system operating procedures are occurring without the prior concurrence of RG&E engineering.

UFSAR, sections 5.4.5.3.5 and 5.4.5.2, refers to two remotely operat-ed valves which can be utilized to isolate an RHR loop from outside the pump room. The system walkdown and the upgraded P&IDs indicate that there is no longer any method available to isolate an RHR loop remotely (i.e., via reach rods). Although this information has been removed from the RHR P&ID, there is no identified punchlist item to delete this information from the UFSAR.

The team noted that uncontrolled training material (Lesson Texts) have not been updated to reflect system changes accomplished during the last outage. There is no station requirement to maintain this training material current. The inspection team considers that making this type of information available to control room operators in such an uncontrolled manner represents a notable program weakness.

The lack of timely operating information updates for control room use is considered an unresolved. item.

89-81-08 Equipment Environmental Qualification Evaluation The NRC questioned. the basis for the assumption that RHR pump seal failure will occur after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The NRC requests RG&E to sub-if stantiate the method of detecting any leak in the RHR pump room the pump seal were to fail before the stated 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

C-2

The safety relief valve test procedures contain general and minimal test instructions for performing the relief setpoint test. Standard the practices are not always performed. or documented. As written, test procedure requires only one successful setpoint and. test. Data from relief valve testing has been recorded inaccurately inconsistent-that ly in some cases. The NRC data recording requirements.

concluded RG&E should formalize test procedures instructions and During the on-going procedure upgrade effort, RG&E should assure that valve test procedures incorporate all new (1986) ASME Code Section XI, IWV-3512, and ANSI/ASME OM-1-1981 requirements"pop for safety relief valves. In particular, more than one successful test" at the designated lift pressure should letely and accurately documented.

be performed Valve and setpoint the and results leak comp-testing should also be performed with the allowable specification listed in the procedure. Valve test results and data should accurately reflect

'he results of all test activities. RG&E should also consider tests the the as-found benefits of adding other periodic valve such as relief lift setpoint, valve accumulation, and. valve capacity.

89-81-10 Translation of FSAR Requirements into Operating Procedures The Ginna UFSAR contains "operational" information and data which the inspectors determined to be invalid and, without a supporting design basis., Specifically, Section 5.4.5.3.5 states that in the event of a 50 gpm RHR pump seal leak and loss of both pump room sump pumps, operators have 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to isolate the leak before the RHR pump motors become flooded. The team determined that a 50 gpm leak into the pump room, with two failed sump motors; cannot be sustained in the RHR pump room for four hours before flooding the pump motors.

originally It intended was just suggested .that the four hour allowance was to indicate a rough system margin for coping with gross leakage in the pump pit.

The team was unable to find any consideration of this in any of the available design documents associated, with the RHR system. It also could not be found in any of the system operating or emergency procedures. The alarm response procedure for the high sump level alarm requires control room operators to dispatch an auxiliary operator to investigate possible pump room flooding, however there is no reference to maximum time limit to isolate a leaking RHR train if necessary.

The team reviewed the instrumentation devices available to control, room operators which would indicate RHR leakage in the pump room.

The only known indication would be from a high level sump alarm.

However, the sump alarm instrument is not qualified for service in a harsh environment.

Operating procedures, emergency procedures, and operator. training The material do not reflect the limiting design basis of the system.

apparently unsupported 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> flooding limit is considered an un-resolved item pending verification of the value by the licensee or correction of the UFSAR.

C-3

89-81-11 Engineer'ssurance O The design control measures as implemented/practiced by the licensee's engineering department were weak, and did not favorably compare to good engineering assurance practices generally accepted in the industry. There was lack of consistency in the implementation of approved engineering procedures among the various departments and engineering management did not appear to be cognizant of this incon-sistency. There was a lack of formal design interface control, lack of control over external communication with design consultants, and a lack of control over design documents/modification packages during the development and implementation phase.

C-4

ENCLOSURE . D Preliminary Categorization of Issues Note:

The categories contained in Enclosure D were selected topics in 10 CFR 50 Appendix B and other sources. To begin the review of the broader concerns, we have reviewed the SSFI report and the cited issues, and have categorized them into general topical areas. For example, unresolved item 89-81-05 involves not having a mechanism to assure that design calculations are maintained up-to-date. We see this specific item as being part of a more general area calledunre-design control. Enclosure D is a preliminary categorization of the solved items into the general topical areas. .It is currently planned to categorize all the concerns identified in the inspection report.

DESIGN CONTROL General Control of Design Inputs Control of Design Process SSFI URI 89-81-05: Electrical Load Growth Con-trol Program SSFI URI 89-81-08: Equipment Environmental Qual-ification Evaluation Control of Design Outputs SSFI URI 89-81-07B: Control Room P&IDs Control of Design Interfaces and. Coordination Control of Design Changes Design Reviews/Engineering Assurance SSFI URI 89-81-05: Electrical Load Growth Control Program SSFI URI 89-81-11: Engineering Assurance Specific Design Concerns SSFI URI 89-81-01: Service Water Single Failure Susceptibility SSFI URI 89-81-03: RHR Pump NPSH PROCEDURES SSFI URI 89-81-09: Safety Relief Valve Testing DOCUMENT CONTROL SSFI URI 89-81-07B: Control Room P&IDs

0 ORGANIZATIONAL . ACES SSFI URI 89-81-02: Resolution of Safety Concerns SSFI URI 89-81-07B: Control Room P&IDs SSFI URI 89-81"10: Translation of FSAR Require-ments into Operating Proce-dures HANDLING OF SAFETY CONCERNS SSFI URI 89-81-02: Resolution of Safety Concerns SURVEILLANCE TESTING MAINTENANCE SSFI URI 89-81-07A: Calibration of Control Room Instruments SSFI URI 89-81-09: Safety Relief Valve Testing D-2

ENCLOSURE E Resolution of Specific Issues Note:

We have separated. the schedule information contained in this enclo-sure into two categories: resolution completed, and schedule for resolution. Listed first are those items for which RG&E has complet-ed resolution. Those measures taken by RG&E are identified. Some of the unresolved items listed cannot be adequately resolved, without addressing the broader more programmatic issues such as design control and engineering assurance and require more time to itemsresolve than the specific items. The schedules provided for some may change as RG&E further identifies the underlying concerns. An updated schedule will be provided in the 120 day response.

0

~ Resolution Com lete

~ 89-81-04 Class 1E Battery Testing This item was resolved prior to receipt of the SSFI report. Please see Enclosure A for actions taken for resolution.

89-81-06 Undervoltage Relay Alarms and Molded Case Circuit Breakers Please see Enclosure B for actions taken for resolution.

89-81-07A Calibration of Contxol Room Instruments This item was resolved prior to receipt of the SSFI report. The actions taken to resolve this issue include:

1) Calibration of all control room dc bus voltmeters during the recent refueling outage (the voltmeters were found to be within the specified acceptance criteria).

2) All dc bus voltmeters are now calibrated per Calibration Proce-,

dure CP-514 on an annual basis.

3) All emergency diesel generator and various secondary system power meter calibrations have been added to the CP-500 series procedures, and the meters were calibrated during the 1990 refueling outage.

89-81-10 Translation of the FSAR Requirements into Operational Procedures This item was resolved promptly. The actions taken to resolve this issue include:

1) Performance of a reanalysis, during the SSFI inspection, which determined that operators have two hours to respond. (Design Analysis, 10CFR50.59 Safety Evaluation, NSL-0000-015, Rev. 0, dated December 8, 1989, Residual Heat Removal Leakage Provi-sions.) T

2) Update of UFSAR sections 5.4.5.3.5, 5.4.5.2 and 6.3.3.8, submit-ted as part of the UFSAR update on December 16, 1989.

3) Revision of Training System Description RGE-25 during the inspection.

4) Revision of EOPs prior to receipt of the inspection report.

(Procedure E-1, Loss of Reactor or Secondary Coolant, Step 18 was added and ES-1.3, Transfer to Cold Leg Recirculation, a note before Step 9 was added.)

.Schedule for Resolu . n 89-81-01 Service Water Single Failure Susceptibility As noted in the inspection report, the failure of the 10 inch dis-charge line in a manner which would stop service water flow to the diesel generators is a low probability event. This event is also beyond the design and licensing basis of the plant. Nevertheless, RG&E plans to further evaluate the potential risk of this scenario during the PRA/IPE effort. Our IPE is currently scheduled to be submitted in the third quarter of 1991.

89-81-02 Resolution of Safety Concerns An interim process for handling safety concerns is under-development and will be discussed in our 120 day response.

89-81-03 RHR Pump NPSH Documentation of the analysis findings is scheduled to be completed by December 31, 1990. In addition, RG&E plans to consider this matter in the PRA/IPE.

89-81-05 Electrical Load Growth Program RG&E has implemented an interim process for all modifications to perform the following actions:

Current system loadings for the dc batteries have been estab-lished in Design Analysis, EWR 3341, Sizing of Vital Batteries, and for the diesel generator loads in Design Analysis, EWR 4136, Diesel Generator Loading.

2) An Electrical Engineering Design Guide, Electrical Interface Checklist EDG-15D, Rev. 0, is being implemented on all modifica-tions which requires identification of load changes to the dc batteries and the diesel generator ac loads.

3) A process controlled by Electrical Design Guide, Design Verifi-cation Model EDG-15B, Rev. 0, has been established within the Electrical Engineering Design Verification Group which updates the loading data for the impacted power supply and determines the remaining capacity margin for ac and dc loads.

We are taking actions to integrate this process into the appropriate Engineering (QE) procedures. We anticipate completion of these actions by the date of our 120 day response.-

89-81-07B Control Room P&IDs RG&E has considered the examples identified by the staff which resulted in the staff's conclusion that information updates forhas control room use are not implemented in a timely manner. RG&E resolved several of the examples identified. These include:

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.1) RG&E has impleilhted improved controls in Z" Drawing Change Request ( DCR ) process . RG&E has assigned Station Engineer with responsibi lity for tracking and processing all DCRs .

2) The UFSAR has been reviewed to assure that the appropriate information with regard to the isolation of the RHR pump seal is

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3) RGGE has revised the lesson text to reflect the revised RHR 'pump seal leakage time limitation of two hours.

An interim process for enhancing the update process for control room information is currently under review and will be discussed, in the 120 day response.

89-81-08 Equipment Environmental Qualification Evaluation The passive failure 'of a RHR pump seal is assumed to occur at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, consistent with SRP 15.6.5. The consequences of this assumed passive failure, concurrent with the assumed design basis LOCA, was evaluated, by the NRC during the review of SEP Topic XV-19 and found to be acceptable. Nevertheless, RGGE plans to further evaluate this scenario during the PRA/IPE effort with its attendant requirement to perform an internal flooding analysis. Our IPE is currently sched-uled to be submitted in the third quarter of 1991. The results of this evaluation will determine if the upgrade of the sump level switches to a safety-related status is recommended.

89-81-09 Safety Relief Valve Testing and, Documentation RGGE has commit/ed to incorporate ASME Code Section ZI-IWV-3512 (1986) and implement ANSI/ASME OM-1-1987 as part of the IST Program Upgrade. Procedure changes to incorporate these requirements were completed. prior to receipt of the SSFI report. RGfiE will have completed. all testings under these new requirements by December 31, 1994.

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