ML20045D247

Responds to NRC Ltr Re Violations Noted in Insp Rept 50-293/93-06 on 930314-0419.Corrective Actions:Procedures 2.1.7,Rev 27,2.4.24,Rev 8 & 1.3.37,Rev 37 Revised & Cooldown Simulation Being Evaluated for Improvement

ML20045D247
Person / Time
Site:	Pilgrim
Issue date:	06/18/1993
From:	Boulette E BOSTON EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BECO-LTR-93-77, NUDOCS 9306280129
Download: ML20045D247 (6)

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BOSTON EDISON l- Pilgrirn Nuclear Power Station i Rocky Hill Road Plymouth, Massachusetts 02360 ,

E. T. Boulette. FnD Senior Vice President-Nuclear  !

June 18, 1993  :

BECo Ltr. 93-77 U.S. Nuclear Regulatory Commission l Attn: Document Control Desk <

l Washington, D.C. 20555 l

i Docket No. 50-293 License No. DPR-35 r

SUBJECT:

REPLY TO NOTICES OF VIOLATION (REFERENCE NRC REGION' INSPECTION REPORT NO. 50-293/93-06)

Dear Sir:

i Enclosed is Boston Edison Company's reply to the Notices of Violation contained in the- .

subject inspection report. Enclosures 1 and 2 respond to Notices of Violation 'A' and i

'B', respectively. .

Please do not hesitate to contact me if there are any questions regarding the enclosed reply. }

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• E.T. Boule e, hD DWE/bal

Enclosure:

Reply to Notices of Violation  !

cc: Mr. Thomas T. Martin Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Rd.

King of Prussia, PA 19406 Mr. R. B. Eaton Div. of Reactor Projects I/II ,

Office of NRR - USNRC One White Flint North - Mail Stop 1401 11555 Rockville Pike Rockville, MD 20852 3

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Sr. NRC Resident Inspector - Pilgrim Station 9306280129 930618 "PDR ADOCK'05000293

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!. . ENCLOSURE 1 REPLY TO NOTICE OF VIOLATION 'A' Boston Edison Company Docket No. 50-293

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Pilgrim Nuclear Power Station License No. DPR-35 1

L During an NRC inspection conducted on March 14, 1993 - April,19, 1993, two violations ~of i NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10.CFR Part 2, Appendix C (1991), violation ' A' is listed below:

NOTICE OF VIOLATION 'A' Technical specification 3.6.A.2 states that critical core operation shall not be conducted.

unless the reactor vessel temperatures are above those defined by the appropriate curves on Figures 3.6.1, 3.6.2,'and 3.6.3._ At stated pressure, the reactor vessel bottom head may be maintained at temperatures below those temperatures corresponding to the: adjacent reactor vessel shell as shown in Figures 3.6.1 and 3.6.2.

Technical Specification 3.6.A.2 further states that in.the event this requirement is not met, achieve stable reactor conditions with reactor vessel temperature above that. defined by the appropriate curve and obtain an engineering evaluation to determine the appropriate course of action to take.

Contrary to the above, on March 13-14, 1993, during reactor depressurization, the reactor vessel bottom head pressure-temperature limits for subcritical cooldown as defined by Technical Specification Figure 3.6.2 were exceeded for approximately three hours and twenty minutes; however, an engineering evaluation to determine the appropriate course of action to take was not obtained until March 19, 1993 after:recctor restart and in response to NRC questioning.

This is a Severity Level IV violation (Supplement: 1).

REASON FOR THE VIOLATION The engineering evaluation. specified by Technical . Specification 3.6. A.2 was not conducted prior to reactor startup because the Reactor Vessel pressure-temperature (P-T) limit was not identified as having been exceeded until after the startup. The P-T' limit was exceeded during the March 13-14, 1993 cooldown. The condition was reported in LER 93-004-00. The cooldown occurred after the load rejection scram that occurred on March 13, 1993. The cooldow.n was complicated by a subsequent loss of preferred offsite power (345KV) that de-energized nonsafety-related components including the drive motors of the Recirculation System Loops ' A' and 'B' motor-generator (MG) sets / pumps. A loss of forced circulation in the' Reactor Vessel (RV) occurred as a-result of the de-energization of the MG sets / pumps 'A' and 'B'. A post trip review was conducted.

Licensed operator activities related to cooldown, loss of forced circulation in the RV, and post trip review are governed by approved procedures. The procedures include:

Procedure 2.1.7, " Vessel Heatup and Cooldown"; Procedure 2.4.24, " Reactor Vessel Cold Water Stratification"; and, Procedure 1.3.37, " Post-Trip Reviews". The administrative controls in these procedures, individually and/or collectively, were not sufficient-to-ensure the performance of an engineering evaluation if the P-T limit were exceeded.

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l l . An overview of the procedures is provided as follows:

• Procedure 2.1.7 is the principal method of monitoring RV metal temperature and pressure during a heatup or cooldown. The' procedure referenced Technical Specifications 3.6.A.1/4.6.A.1 and 3.6.A.2/4.6.A.2. The procedure included Attachment 1 for logging temperatures and pressures and also included Technical ,

Specifications Figures 3.6.1, 3.6.2, and 3.6.2. Attachment 1, however, did not !

specify a check or comparison of the recorded pressure and pressure to the P-T  !

limit.

• Procedure 2.4.24 did not direct the operators to review Technical Specification Figure 3.6.2 if a loss of recirculation flow occurred during a heatup or cooldown or Figure 3.6.3 if a loss of recirculation flow occurred during power operation.

The loss of preferred offsite power that occurred after the scram on March 13, 1993, resulted in a loss of forced circulation in the RV. The operators did not check or compare the resulting or subsequent RV pressure-temperature condition to Figure 3.6.2 because Procedure 2.4.24 did not direct this action.

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• Procedure 1.3.37 did not specifically require an evaluation or comparison of j transitory parameters governed by Technical Specifications to Technical Specifications including Figure 3.6.2. The RV pressures and temperatures recorded during the cooldown on March 13-14, 1993, were not compared to the P-T limits of Figure 3.6.2. Therefore, the P-T limits were not identified during the post trip review as being exceeded during the cooldown. Consequently, an engineering evaluation of the P-T condition experienced during the cooldown was not requested to determine the appropriate course of action to take as specified by Technical Specification 3.6. A.1.

The licensed operator training program includes simulator training. This training includes simulated cooldowns. The cooldown simulation, however, assumed an uninterrupted cooldown. Consequently, operator simulator training for a cooldown did not include operator actions if a RV pressurization or P-T condition requiring action occurred during a cooldown.

An INP0 document (SER 5-93) describing two industry events in 1992 involving RV cooldown and heatup rates and P-T limits being exceeded due to the loss of recirculation flow l following a scram was issued on February 24, 1993. The document was processed as part of l the Operating Experience Review Program (0ERP). The document was assigned for disposition

, by the Operations Section (comments 1 and 3) and Nuclear Training Department (comment 2) l on March 8, 1993. The Operations Section disposition was that comments 1 and 3 were addressed by procedure 2.4.24. The comments were dispositioned on March 10, 1993.

Comment 2 was dispositioned by the Nuclear Training Department on May 4, 1993.

Essentially, the disposition was to incorporate SER 5-93 into the Licensed Operator j Requalification Training Program and the training module for Shift Technical Advisors and i Senior Reactor Operator candidates.

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. 00'RRECTIVE ACTION TAKEN AND RESULTS ACHIEVED 1 l

P'rocedures 2.1.7 (then Rev. 27), 2.4.24 (then Rev. 8), and 1.3.37 (then Rev. 7) have been j revised. Essentially, the purpose and focus of the changes were to specifically reference and/or incorporate Technical Specifications Figures 3.6.1, 3.6.2, and 3.6.3, as applicable. The changes to procedure 2.1.7 included additional steps that compare RV P-T conditions to the P-T limits and initiate actions, including an engineering evaluation, if a P-T limit is exceeded. The changes to procedure 2.4.24 included operator direction to refer to procedure 2.1.7 and determine that RV pressure-temperature is in compliance with-Technical Specifications limits. The changes to procedure 1.3.37 included a check of selected transitory parameters governed by Technical Specifications, and preliminary assessment of selected systems' responses.

Licensed operators received training as part of corrective action. The training included discussion and review of LER 93-004-00, the INP0 document SER 5-93, and the changes to procedures 2.1.7 (Rev. 29) and 2.4.24 (Rev. 9). The training was completed during the recent refueling outage (RF0 9) prior to.startup.  ;

The effects of exceeding the P-T limit during the cooldown 'were evaluated. The evaluation l concluded the RV did not exceed ASME Section III structural limits nor did the RV exceed -

ASME Section XI fracture toughness limits. ]

Reports submitted to the NRC since 1972 and previous scram reports were reviewed. The review focused on events involving RV pressurization with no forced circulation. The review identified no previous event or condition involving a pressurization with no forced circulation in the RV.

CORRECTIVE ACTION TO PRECLUDE RECURRENCE The cooldown simulation is being evaluated for improvement. The focus of the evaluation is to consider an interruption during a cooldown and/or operator actions if a RV pressurization or P-T condition requiring action occurs during a cooldown.

The changes to Procedure 1.3.37 will be reviewed as part of the regularly scheduled training program for senior reactor operators.

Technical Specification 3.6.A will be evaluated for possible improvement. The focus for improvement is to specify and/or clarify the plant conditions in the same manner identified on Figures 3.6.1, 3.6.2, and 3.6.3. This additional action complements the changes made to Procedures 2.1.7, 2.4.24, and 1.3.37.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved when the engineering evaluation of the P-T condition was completed. The evaluation was approved at the end of March 1993, and a copy was provided to the NRC Resident Inspector.

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. ENCLOSURE 2 REPLY TO NOTICE OF VIOLATION 'B' Boston Edison Company Docket No. 50-293 Pilgrim Nuclear Power Station License No. DPR-35 During an NRC inspection conducted on March 14, 1993 - April 19, 1993, two violations of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (1991), violation 'B' is listed below:

NOTICE OF VIOLATION 'B' 10 CFR 50 Appendix B, Criterion III, " Design Control" states that design changes shall be subject to design control measures commensurate with those applied to the original design.

Boston Edison Quality Assurance Manual Volume II, Section 3.3.2.8 requires that methods for verifying design changes, such as design reviews, alternative calculations, and .

' qualification testing be properly chosen and followed; the most adverse design conditions be specified for test programs used to verify the adequacy of designs.  ;

Contrary to the above, the licensee failed to establish adequate design controls to ensure that the trip setpoints for the main input breakers to the voltage regulating i

transformers, which were installed via plant design change during the 1992 mid cycle outage to service the safety related 120 VAC Y-3 and Y-4 busses, were proper.

REASON FOR THE VIOLATION The main input circuit breakers of the soltage regulating transformers X55 and X56 tripped open and 120 VAC Safeguards Buses 'A' and 'B' consequently became de-energized because the ,

trip settings of the main input breakers were too low. The event was reported in LER 93-004-00. A root cause analysis concluded the settings were set too low because of an unauthorized change to the trip settings. The analysis included review of the design package (PDC 91-59A), dedication plan test instructions, receipt inspection, and pre-operational test procedure (TP 92-58). The supplier test documents indicated the trip ,

settings were left at the correct setting of '5'. The receipt inspection activities included documentation, physical damage, identification and/or markings, protective covers and seals, cleanliness, and electrical tests, but did not include a requirement to check or verify the trip settings. After installation, transformers X55 and X56 were pre-operationally tested (TP 92-58). The testing included voltage regulation, input breaker contact resistance, current leakage, initial startup and energization, transformer ratio, relay and alarm functional tests. The testing did not include a requirement to check or verify the trip settings since there were no installation or testing activities that would have caused the settings to be changed. The analysis could not determine when the change to the trip settings occurred.

CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED The trip settings of the main input breakers for X55 and X56 were increased. The change was implemented via a modification change (FRN 93-02-03) on March 15, 1993. The new trip settings included additional margin to preclude a recurrence. The original design trip setting of '5' was sufficient to prevent an unnecessary trip of the input breakers.

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i i l . C0'RRECTIVE ACTION TAKEN TO PRECLUDE RECURRENCE i Visual irispections of selected electrical equipment were conducted while shut down during i RF0'9. The purpose of the inspections was to provide additional assurance the i unauthorized change of the input breakers' trip setting was an isolated occurrence. The inspections were completed with satisfactory results. The inspection results indicated the unauthorized change of the input breakers' trip setting was an isolated occurrence.

Appropriate nuclear engineering organization personnel were made aware of the root cause of the trip of the main input breakers. Interim corrective action consisted of reminding engineering personnel to include in-process verification of adjustable trip settings, when appropriate. Long term corrective action consists of revising NED procedure 3.02' (currently Rev. 30), " Preparation, Review Verification, Approval and Revision of Design Documents for Plant Changes". The change is focused on including verification of the setting of a device having an adjustable setting. When appropriate, the verification would be specified as part of post installation or preoperational testing. The change is expected to be approved by August 1993.

Preventive action supplementing the above noted corrective action was also assessed. The focus for preventive action was the root cause analysis conclusion regarding the unauthorized change to the trip setting of the main input breakers. The preventive action will be to include a summary of the response to this violation in the weekly nuclear  ;

organization Division Managers' meeting for discussion with their personnel. <

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved when the trip setting of the input breakers of transformers X55 and X56 was changed via FRN 93-02-03 on March 15, 1993. j i

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