IR 05000416/1985039

Insp Rept 50-416/85-39 on 851011-1115.Violation Noted: Failure to Prepare,Review & Approve Procedure Changes Re Surveillance of safety-related Equipment

ML20138C300
Person / Time
Site:	Grand Gulf
Issue date:	11/29/1985
From:	Butcher R, Caldwell J, Panciera V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20138C275	List: IR 05000416/1985039 ML20138C272 ML20138C289
References
50-416-85-39, NUDOCS 8512120537
Download: ML20138C300 (7)

Text

UNITED STATES

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, NUCLEAR REGULATORY COMMISSION J 'i - /^ REGION 11 h, ,k *.,

101 MARIETTA STREET. *1N -.g ATLANTA, GEORGI A 30323

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o Report N /85-39 Licensee: Mississippi Power And Light Company Jackson, MS 39205 Docket No.: 50-416 License No.: NPF - 29 Facility Name: Grand Gulf Unit 1 Inspection Conduc ed: October 11 thru November 15, 1985 Inspectors:

R. ""

utcher, Senic @ es gent I spector

[Date Signed 7 I , ^'=

. Cald , esiVnt Id5pector YUf Date Signed

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Approved by: k in m a h f V.' W. 'Pancter'a, Criief, Project Section 2B

// 2-9!/5 Ddte Signed Division of Reactor Projects SUvMARY i

Scope: This routine inspection entailed 243 resident inspector-hours at the site in the areas of Operational Safety Verification, Maintenance Observation, Surveillance Observation, Reportable Occurrences, Operating Reactor Events, and Inspector Followup and Unresolved Item Results: Of the six areas inspected, no apparent violations or deviations were identified in five areas; one apparent violation was found in one are PDR ADOCK 05000416 G PDR

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REPORT DETAILS-1. -Licensee Employees Persons Contacted

• J. E. Cross, General Manager C. R. Hutchinson, Manager, Plant Maintenance

• R. F. Rogers, Technical Assistant J. D. Bailey, Compliance Coordinator M. J. Wright, Manager, Plant Operations

• L. F. Daughtery, Compliance Superintendent D. Cupstid, Acting Technical Superintendent R.. H. McAnulty, Electrical Superintendent R. V. Moomaw, I&C Superintendent

• B. Harris, Compliance Coordinator J. L. Robertson, Operations Superintendent Other licensee employees contacted included technicians, operators, security force members, and office personne * Attended exit interview Exit Interview The inspection scope and findings were summarized on November 15, 1985, with those persons indicated in paragraph 1 above. The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspectio The licensee had no comment on the following inspection findings: Violation 50-416/85-39-01; Failure to follow Administrative Procedure 01-S-02-2 when making changes to Technical Section Instruction

09-S-08-2. (Paragraph 7) Inspector Followup Item 50-416/85-39-02; Division 1 Diesel Generator overspeed failure of November 6, 1985. (Paragraph 9.b) Inspector Followup Item 50-416/85-39-03; Failure of electrical terminal strips in Hydrogen monitors. (Paragraph 9.c) Licensee Action on Previous Enforcement Matters (92702)

Not Inspecte . Unresolved Items Unresolved items were not identified during this inspectio . - _ _

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, 5. Operational Safety Verification (71707)

The inspectors kept themselves informed on a daily basis of the overall plant status.and any significant safety matters related to plant operation Daily discussions were held with plant management and various members of the plant operating staf The inspectors made freauent visits to the control room such that it was visited at least daily when an inspector was on site. Observations included instrument readings, setpoints, and recordings; status of operating systems; tags and clearances on equipment controls and switches; annunciator alarms; adherence to limiting conditions for operation;. temporary alterations in effect; daily journals and data sheet entries; control room manning; and access controls. This inspection - activity included numerous informal discussions with operators and their supervisor Weekly, when onsite, a selected ESF system is confirmed operable. The confirmation is made by verifying the following: accessible valve flow path alignment; power supply breaker and fuse status; major component leakage, lubrication, cooling and general condition; and instrumentatio General plant tours were conducted on at least a biweekly basis. Portions of the control building, turbine building, auxiliary building and outside areas were visited. Observations included safety related tagout verifi-cations; shift turnover; sampling program; housekeeping and general plant conditions; fire protection equipment; control of activities in progress; radiation protection controls; physical security; problem identification systems; and containment isolatio In the areas inspected, no violations or deviations were identifie . Maintenance Observation (62703)

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During the report period, the inspector observed selected maintenance activities. The cbservations included a review of the work documents for adequacy, adherence to procedure, proper tagouts, adherence to Technical Specifications, radiological controls, observation of all or part of the actual work and/or retesting in progress, specified retest requirements, and adherence to the appropriate quality' control In the~ areas inspected, no violations or deviations were identifie . Surveillance Testing Observation (61726)

The inspector observed the performance of selected surveillances. The observation included a review of the procedure for technical adeourcy, conformance to Technical Specifications, verification of test instre. ment -

calibration, observation of all or part of the actual surveillances, removal-from service and return to service of the system or components affected, and review of the data for acceptability based upon the acceptance criteri .

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The Resident Inspectors reviewed the Integrated Leak Rate Test (ILRT)

Procedure, Surveillance Procedure 06-ME-1M10-0-0002, Revision 20, and the Local Leak Rate Test (LLRT) Procedure, Surveillance Procedure 06-ME-1M61-V-0001, Revision 25, prior to the licensee initiating the tes The inspectors' questioned paragraph 2.14 of 06-ME-1M10-0-0002 which stated containment pressure shall not fall outside the limits of 10.5 and 13.5 psig during the test, and during the verification test the test pressure may fall below 10.5 psig. Technical Specification 4.6.1.2 requires containment leakage rates be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N-45.4-1972, and defines Pa as 11.5 psi CFR 50 Appendix J and ANSI N-45.4-1972 do not permit the test pressure to fall below Pa (11.5 psig).

The licensee revised 06-ME-1M10-0-0002 to not allow test pressure to fall below 11.5 psig. The LLRT procedure, paragraph 2.11, states that valve line-ups for the LLRT must be in accordance with Technical Section Instruction (TSI) 09-S-08- TSI 09-S-08-2, paragraph 6.3, states, "This instruction does not require revision whenever valve line up tables or sketches on the P & ID's on which they are based are revised." The licensee

, has been making revisions to TSI 09-S-08-2 without issuing formal changes as directed in Administrative Procedure ( AP) 01-S-02-2. Technical Specifi-cation (TS) 6.8.1 requires written procedures be established, implemented and maintained covering surveillance of safety related equipmen TS 6.5.3.1.a requires each procedure change to procedures required by TS 6.8 be prepared, reviewed and approved. These TS requirements are incorporated by AP 01-S-02-2, - Control and Distribution of the GGNS Operations Manual .

Paragraph 6.3 of AP 01-S-02-2 defines the methods to be followed when changing procedures. The failure of TSI 09-S-08-2 to meet TS 6.5.3. and AP 01-S-02-2 is a violation (50-416/85-39-01). Reportable Occurrences (90712 & 92700)

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The below listed Licensee Event Reports (LERs) were reviewed to determine if the information provided met NRC reporting requirements. The determination included adequacy of event description and corrective action taken or planned, existence of potential generic problems and the relative safety significance of each event. Additional inplant reviews and discussions with plant personnel, as appropriate, were conducted for the reports indicated by an asterisk. The LERs were reviewed using the guidance of the general policy and procedure for NRC enforcement action The following LERs are close LER N Event Date Event

~~83-068 6-3-83 Discrepancy in Secondary Containment Ventilation Isolation Damper Closing Time *85-036 9-16-85 Reactor Scram Due To Low Condenser Vacuu ._ _ _ _ _ _

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.85-031 8-13-85 Technical Specification Time Limits Exceede LER 85-036 is associated with Scram No. 32 discussed in paragraph 9 belo In the areas inspected, no violations or deviations were identifie . Operating Reactor Events (93702)

The inspectors reviewed activities associated with the below listed reactor events. The review included determination of cause, safety significance,

performance of personnel and systems, and corrective action. The inspectors examined instrument recordings, computer printouts, operations journal entries, and scram reports, and had discussions with operations maintenance and engineering support personnel as appropriate, Scram No. 32 occurred September 16, 1985, at 8:13 a.m. , with the reactor operating at approximately 100% power. The initiating event was loss of lube water flow to the circulating water pumps. This loss of lube water tripped both circulating water pumps which resulted in a low condenser vacuum trip of the main turbine and subsequent trip of the reactor. During this and previous Scram reviews, it was noted that the" main steam bypass valves did not operate the same as they did during startup testing. The licensee compared the present bypass valve performance to the startup test criteria and even though the valves were not operating the same as before they still met the startup test criteria. The licensee has been investigating the cause of the change in bypass valve performance during the present outage and have

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discovered several problems associated with the bypass valve hydraulic system which would cause the change in their performance. These problems are.being worked and should be resolved this outag During post maintenance testing on November .6,1985, the Division 1 Diesel Generator (DG) oversped causing high vibration. The operators attempted to shut the DG down manually and it appears to have taken 10 to 15 seconds for the DG to shut down. 'The DG is manufactured by Transamerica DeLaval, Inc. (TDI) and is model DSRV-16. The licensee

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has disassembled the DG and is investigating the caure of- the overspee CG damage includes several connecting rod bushings (crankshaft end), one rod box. link pin bushing, and some apparent main bearing wear. Normal speed for the DG is 450 rpm and the licensee is not sure how much of an overspeed condition occurred. The licensee's investigation is continuing at this time and no root cause has been c determined. This will be Inspector Followup Item 50-416/85-39-0 E c-. On November 8,1985, the licensee found that some electrical terminal-strips located in the containment and drywell hydrogen analyzers had

, become cracked and broken. It appears the deterioration is due to excessive temperatures. These terminal strips are used to terminate

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the connections to the analyzer cell and heater. The terminal strips

were functional but the potential for failure was readily Lapparen .

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Incident Report 85-11-7 was prepared and the NRC was notified on November 8, 1985 at 5:00 The containment and drywell hydrogen analyzers were declared inoperabl The analyzers are required in operational condition 1, 2 and 3. The unit is presently in operational condition 4 (cold shutdown) for a maintenance outage. The analyzers (Model No. K-III) are manufactured by Compsi-Delphi Manufacturing and the terminal strips (Model No. 3-140 and 6-141) are manufactured by Cinch-Jones. The licensee has not determined corrective action at this time. This will be Inspector Followup Item 50-416/85-39-0 . Inspector Followup And Unresolved Items (92701) (Closed) IFI 50-416/85-22-04 On June 17,1985, General Electric (GE) submitted a 10 CFR part 21 report on an unqualified test switch (Switch CR2940) used on the Standby Liquid Control (SLC) system. GE Developed a -failure scenario which postulated a failure of the switches during a Loss of Coolant Accident (LOCA) such that the emergency power supply becomes degraded enough to inhibit operation of Emergency Core Cooling System (ECCS)

equipmen Switch failure could start the valved out SLC system pump valve motor, leading to motor overheating and shorting of the emergency bus which is protected only by fault current protective devices. Both division 1 and 2 emergency power buses are affected. The licensee's Nuclear Plant Engineering (NPE) evaluated the GE 10 CFR Part 21 report and determined it did not apply to Grand Gulf (GG). At Grand Gulf, multiple failures

, of class 1E circuit protective devices would have to occu These protective devices, for division 1 and 2 respectively are:

(1) Motor overload protection device (2) Containment penetration fuses,15214F and 163129 (3) Circuit breakers,52-152114 and 52-16312 Considering an independent single active failure _(loss of one of the above protective devices) coincident with a LOCA and redundant failure of the SLC CR2940 switches, adequate class 1E circuit protection remains available to preclude the failure of the emergency power bu This is because the containment penetration fuses, which were added by the licensee to sati s fy Regulatory Guide 1.63 requirements for over-current protection in circuits that penetrate containment, were not considered in the failure analysis by GE. The resident inspectors-discussed the licensee's evaluation of the' part 21 report with the Region II Plant Systems Section (T. Conlon/M. Hunt) and the' licensee's evaluation is considered adequate. No further action is require (Closed)lIFI 50-416/85-28-04 The inspector reviewed TCN 2 to Operations Section Procedure 02-S-01-3 and Revision 15 to Administrative Pr,cedure 01-S-06-2 which clarified

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the definition of the operator-at-the-controls, the control room operator and assistant control room operator. This item is closed.

, (Closed) IFI 50-416/85-02-02 In a letter from NRR (Mr. T. M. Novak) to MP&L (Mr. J. Richard) dated

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August 27, 1985, the NRC concurred in MP&L's position that surveillance testing of fuses used for overcurrent protection of containment

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_ electrical penetrations was not require However, MP&L was directed to include in appropriate plant procedures, a means for periodically verifying that correct fuse sizes are installed. The licensee states in Maintenance Section Procedure 07-S-01-205, Conduct of Maintenance

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Activities, attachment II, the requirement that during replacement of fuses, the size of the fuse will be verified against the as built electrical drawin Also a note states that a fuse will not automatically be replaced by a fuse of the same size just because that

fuse was installed originall This item is closed.

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