IR 05000416/1986039

Insp Rept 50-416/86-39 on 861115-1219.Violation Noted: Failure to Follow Procedures for Intermediate Range Monitor Range 6 to Range 7 Correlation Test

ML20212R373
Person / Time
Site:	Grand Gulf
Issue date:	01/13/1987
From:	Butcher R, Dance H, Will Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20212R336	List: IR 05000416/1986039 ML20212R334 ML20212R347
References
50-416-86-39, NUDOCS 8702020619
Download: ML20212R373 (16)

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jp Hig UNITED STATES jo NUCLEAR REGULATORY COMMISSION

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,j REGION 11 101 MARIETTA STREET, *I c ATLANTA, GEORGI A 30323

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Report No.: 50-416/86-39

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Licensee: Systems. Energy Resources, In Jackson, MS 39205

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Docket No.: 50-416 License No.: NPF-29 Facility Name: Grand Gulf Nuclear Station Inspection Conducted: November 15 - December 19, 1986 Inspectors: ( &~~

R. C. Butcher, '$enior Resident Inspector

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/ D t/e Signed 6 ul - //3 6'7 W. F. Smith, P sident Inspector ' Daf Si ned Approved by: b Ms w H. C. Dance, 3ection Chief

/ /3 e77 hate / Signed Division of Reactor Projects SUMMARY

< Scope: This routine inspection was conducted by the resident inspectors at the site in the areas of Licensee Action on Previous Enforcement Matters, Operational Safety Verification, Maintenance Observation, Surveillance Observation, ESF System Walkdown, Reportable Occurrences, Operating Reactor Events, Inspector Followup and Unresolved Items, Design Changes and Facility Modifications, Plant Startup from Refueling, and a management meetin Results: One violation was identified: Failure to follow procedures for the Intermediate Range Monitor range 6 to range 7 correlation test (paragraph 13).

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Oy 870116 0 05000416 PDR

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REPORT DdTAILS Persons Contacted Licensee Employees

• D. Kingsley, Vice President, Nuclear Operatiens

• H. Cloninger, Vice President, Nuclear Engineering and Support

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+* E. Cross, GGNS Site Director

• F. Dale, Director, Nuclear Licensing and Safety

+*C. R. Hutchinson, GGNS General Manager

• F. W. Titus, Director, Plant Engineering

+*R. F. Rogers, Manager, Unit 1 Projects

• A. S. McCurdy, Manager, Plant Operations

• S. M. Feith, Director, Quality Assurance J. D. Bailey, Compliance Coordinator

• K. E. Beatty, Training Superintendent

+*M. J. Wright, Manager, Plant Support

• G. Cesare, Manager, Nuclear Licensing

+L. F. Daughtery, Compliance Superintendent

+*D. G. Cupstid, Technical Support Superintendent

• R. H. McAnulty, Electrical Superintendent

+R. V. Moomaw, Manager, Plant Maintenance

+ P. Harris, Compliance Coordinator

• J. L. Robertson, Licensing Superintendent

• L. G. Temple, I&C Superintendent

• J. H. Mueller, Mechanical Superintendent

• L. B. Moulder, Operations Superinterdent

• J. V. Parrish, Chemistry / Radiation Control Superintendent

+*J. W. Yelverton, Technical Assistar.' to Manager Operations

+H. D. Morgan, Manager, Unit 2 Constru.-tion

+E. G. Jung, Project Engineer

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+C. C. Hayes, Supervisor, QA

. +W. R. Patterson, Supervisor, Reactor Engineering

+J. E. Reeves, Manager, Nuclear Site QA Other licensee employees contacted included technicians, operators, security force members, and office personne Other Organizations R. Grummer, Fuel Representative, Exxon Corporation NRC Representatives

*L. A. Reyes, Deputy Director, Division of Reactor Projects, Region II

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• H. C. Dance, Section Chief, Division of Reactor Projects, Region II

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+ Attended exit interview (December 19,1986)

• Attended management meeting on December 15, 1986 l

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2 Exit Interview The inspection scope and findings were summarized on December 19, 1986, 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 inspection. The licensee had no comment on the following inspection findings:

r 416/86-39-01, Inspector Followup Item. Acceptability of airlock door seal test (paragraph 7).

416/86-39-02, Inspector- Followup Item. Discrepancies in ADS drawings and procedure /86-39-03, Inspector Followup Item. Correction of discrepancies in Bettis actuator seal replacement work completion records (paragraph 9).

,' 416/86-39-04, Unresolved Item. Unqualified electrical wiring used with Bettis and Hiller actuators for air operated valves (paragraph 10).

416/86-39-05, Inspector Followup Item. Remote shutdown panel rooms air conditioning evaluation test (paragraph 11).

416/86-39-06, Inspector Followup Ite Relocation of the ADS air supply makeup connection (paragraph 12).

416/86-39-07, Violatio Failure to follow procedures for the IRM j range 6 to range 7 correlation test (paragraph 13).

j 3. Licensee Action on Previous Enforcement Matters t

(Closed) Violation 416/83-35-0 Failure to implement Safety Review Committee (SRC) requirements by procedure. A Region II inspection was accomplished which reviewed the SRC and related TS requirements. Inspection

. Report 416/86-18 documented this inspection and concluded that the SRC

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satisfied applicable TS requirement . Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations or deviation One new unresolved item identified during this inspection is discussed in paragraph 1 .

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 staff. The inspectors made frequent visits to the control

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room such that it was visited at least daily when an inspector was on sit 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 control This inspection activity included numerous informal discussions with operators and their supervisors. Weekly, when the inspectors were onsite, selected Engineered Safety Features (ESF)

systems were confirmed operabl The confirmation is made by verifying the following: accessible valve flow path alignment, power supply breaker

, and fuse status, major component leakage, lubricant, cooling and general

, condition, and instrumentation. 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 verifications, shift turnover, sampling program, housekeeping and general plant conditions, fire protection equipment, control of activities in progress, radiation protection controls, physical i

security, problem identification systems, and containment isolatio The inspectors verified that the revised emergency procedures were available

in the control room and that the Safety Parameter Display System was opera-tional, as required by Attachment 1 to the GGNS license conditions.

l Maintenance Observation During the report period, the inspectors observed portions of the main-tenance review activities listed below. The observations included a review

of the work documents for adequacy, adherence to procedure, proper tagouts, j adherence to technical specifications, radiological controls, observation of all or part of the actual work and/or retesting in progress, specified

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retest requirements, and adherence to appropriate quality control MWO-ME5142, Conduct of Automatic Depressurization System (ADS) drop test in accordance with General Maintenance Instruction 07-S-14-324,

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Rev. O, ADS Air System Drop Test.

! MWO-M68123, Replacement of ten air start valves on Division 2 emergency

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diesel generator.

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Inspection of electrical wiring on air operated valve No violations or deviations were identifie . Surveillance Observation (61726)

The inspectors observed the performance of portions of the surveillance listed below. The observation included a review of the procedure for technical adequacy, conformance to technical specifications, verification of test instrument calibration, observation of all or part of the actual

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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 P-1P75-R-0004, Rev. 25, Standby Diesel Generator 12; 18 Month Functional Test, Tests III and I ME-1M23-R-0001, Rev. 26, Personnel Air Lock Door Seal Air System Leak Tes ME-1M61-V-0001, Rev. 28, Local Leak Rate Test of Valve E12-F050 C-1C11-M-0003, Rev. 20, Calibration Check of Scram Discharge Volume Level Switche OP-1C11-M-0012, Rev. 24, RPC Rod Block Functional Tes P-1P75-V-0012, Rev. 23, Standby Diesel Generator 11; Operability Verificatio .

06-0P-1P75-V-0013, Rev. 23, Standby Diesel Generator 12; Operability Verificatio On December 10, 1986, the inspectors witnessed an airlock door leak test conducted per 06-ME-1M23-V-0001, Rev. 25, with Temporary Change Notices (TCNs) 6 and 7, Containment and Drywell Airlock Seal Leak Test. The inspectors observed excessive amounts of what appeared to be silicone grease on the airlock door seals. TCN 6 added an acceptable lubricants list for seals and modified paragraph 5.3 which states, if desired a moderate application of suitable lubricant may be used to enhance inflatable seal performance. Additionally paragraph 5.3, which is accomplished prior to the seal leak test, requires the inflatable seals to be cleaned with dry rags, the airlock sealing surfaces to be cleaned with rags and cleaning fluid and then apply a moderate application of suitable lubrican This process destroys the as-found condition of the airlock seals and thus renders the leak test invalid. The quantity of lubricant on the seals appeared excessive and the inspectors could made furrows in the lubricant. Based on discussions with licensee management the licensee has agreed to conduct the as-found air lock seal leak tests. TCN 8 to 06-ME-1M23-V-0001 now requires the airlock seals be tested prior to cleaning and applying moderate amounts of lubrican This will document the acceptability of the airlock seals following use but prior to cleaning and applying lubricant to enhance seal performance. The inspectors did not consider the licensee's airlock door seal test acceptable when tests were performed as required by the noted procedure due to the lack of seal testing in the as-found condition and the application of lubricant to effect a satisfactory sea Region II management has evaluated the licensee's airlock door seal test and determined no further action is required at this tim This will be Inspector Followup Item 416/86-39-0 No violations or deviations were identifie .

8. Engineered Safety Features System Walkdown (71710)

A complete walkdown was conducted on the accessible portions of the Automatic Depressurization System (ADS). The walkdown consisted of an inspection and verification, where possible, of the required system valve alignment, including valve power available and valve locking where required, instrumentation valved in and functioning; electrical and instrumentation cabinets free from debris, loose materials, jumpers and evidence of rodents, and system free from other degrading condition The inspectors noted that piping and instrument diagram (P&ID) M-1077C, Rev. 20 did not accurately reflect the piping between check valves B21-F130A and 021-F1308. The P&ID shows a run of 3/4 inch piping between two reducing couplings connected to the upstream side of the check valves. The as-built configuration consists of a 3/4 inch by 2 inch reducing tee. The area in question is where the the 3/4 inch piping from the booster compressors ties into the ADS air receivers shown in zone H-5 of the P&ID. This discrepancy should be corrected during a routine revision for other reasons and shall be tracked as Inspector Followup Item 416/86-39-0 The inspectors also noted test gauges attached to the closed drains on ADS air receivers A100A and A1008. The gauges were removed later as documented on surveillance procedure 06-0P-1821-C-0003, Revision 28, Nuclear Valve Operability. The inspectors identified a discrepancy in this procedure, unrelated to the gauges. Page 18 of Attachment I contained valves which do not exist in the plant, nor do they appear on the above P&ID. The valves are B21-F082A,B,C,E,G,K,L,M and N. Also, B21-F082T is missing from the sheet and 821-F081T,U,V and W appear twice. Upon reviewing the working copy of the procedure in the control room it was noted that the operators placed an "N/A" in the signature blank for the nonexistent valves and made a pen-and-ink change to B21-F091T making it B21-F082T, with a footnote that a change was requested for that entry. Although the document adequately certified the status of the intended valves, the procedure should have been revised to reflect what is te be certified in the first place. Correction of this deficiency shall be tracked as part of T.nspector Followup Item 416/86-39-02 noted abov No violations or deviations were identifie . Reportable Occurrences (90712 & 92700)

The below listed event reports were reviewed to determine if the information provided met the 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 event reports were reviewed using the guidance of the general policy and procedure for NRC enforcement action _

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6 The following Licensee Event Reports (LERs) are close *84-042 September 8, 1984 Inoperable Pipe Supports on SSW Piping

• 84-045 November 13, 1984 Reactor Scram Due to Low Water Level

• 85-008 March 18, 1985 Reactor Scram Due to Loss of Feedwater

• 85-043 November 7, 1985 General Failure of Terminal Strips Could Cause Loss of Hydrogen Analyzer

• 86-023 June 19, 1986 HPCS DG Design Error Could Cause Loss of Safety Function

The discrepancy of LER 86-23 is being tracked as IFI 416/86-20-0 The corrective action of LERs84-045 and 85-008 is being tracked to completion as IFI 416/85-09-0 (Closed) 10 CFR Partr 21, PRD 84-10, G. N. Bettis valve actuator EPR seals swelling. On August 3,1984, the licensee published a report to the NRC

that a deficiency concerning the possible swelling of the Ethylene Propylene Rubber (EPR) seals used on valve actuators manufactured by G. M. Bettis and installed on 17 safety-related valves by the Henry Pratt Company, and supplied to MP&L by Bechtel Power Corporation. G. M. Bettis determined that the Mobilgrease 28 used to lubricate the seals on the 17 valves at GGNS interacts with the EPR seals causing them to swell with the possible consequence of slowing valve stroke time. The licensee committed to monitor

the stroke time on the affected valves and by the end of the first refueling

outage have all seals replaced with a qualified, recommended lubricant

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that will not cause the swelling. The inspectors witnessed portions of

the seal replacement work during the outage and reviewed completed work i documents. The inspectors noted that the records which documented the seal i replacement of the above Bettis actuators contained a few discrepancies. Of i a minor nature, the inspectors found duplicate material tickets filed with incorrect packages. For instance, extra copies of tickets for valves P11-F130 and Z51-F003 were filed in the packages for P11-F131 and Z51-F010

respectivel The package for E61-F056 did not contain the as-found

! inspection data sheet and the final torque data sheet required by the valve maintenance procedure 97-5-14-317; however, there was sufficient objective

evidence that the job was done properly. One significant discrepancy was that the record for valve E61-F009 (Task Card No. ME0027) contained a

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material ticket showing receipt of an EPR seal kit for E61-F009 one day

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after the replacement was signed off as completed. This indicated possible rework. When the inspector questioned the absence of rework records, the

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licensee explained that the material ticket was erroneously charged against l E61-F009, when in fact the kit was utilized for E61-F007 (Task Card

No. ME5167). There was a material ticket found in the E61-F007 record, but ,

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the licensee explained that the kit in that record was not used because it was damage The E61-F007 record failed to show this. The inspectors expressed concern that unless the records accurately identify which parts went into each valve, it would be difficult to prove that the valves have the proper quality to perform their intended design function. Based upon a review of all the record packages for the 17 Bettis actuators involved, the inspectors were satisfied that the job was accomplished satisfactorily; however, if the records are not corrected, the licensee might experience some difficulty in determining which trace number seal kits went into each valve if called upon to provide it in the future. The licensee issued a deficiency report to document this problem. Correction of the records shall be tracked under Inspector Followup Item 416/86-39-0 No violations or deviations were identifie . Operating Reactor Events (93702)

The inspectors reviewed activities associated with the below listed reactor event 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, scram reports and had discussions with operations, maintenance and engineering support personnel as appropriat On November 18, 1986, at 9:16 a.m., the 115 kV offsite line from Port Gibson to ESF transformer 12 lost power which caused the ESF bus 16AB ta de-energiz The Division 2 diesel generator started and tied to bus 16AB to supply powe A Division 2 isolation was initiate All systems functioned normally although the Division 2 diesel generator had not completed post maintenance testing and had not yet been declared operational. Two 500 kV offsite lines were still available and were connected to the remaining two ESF busse The licensee determined that the cause of the loss of the 115 kV line was due to a timber cutting crew working in the Vicksburgs area cutting down a tree which fell into the 115 kV line. This event is similar to the event of December 18, 1985, described in LER 85-04 The licensee should address what actions will be taken to minimize the possibility of similar occurrences in their event report for this even On November 28, 1986, the licensee notified the residents of a potential problem associated with use of unqualified wiring used with Bettis and Hiller actuators associated with air operated valves (A0Vs). Following an inspection of Raychem seals per IEN 86-53 (see Inspection Report 416/86-37)

the licensee cut out some seals that did not appear to meet Raychem minimum dimensions. NPE examined the cut out samples for adhesion, etc., and identified one piece of wiring which was not on their qualified lis Based on this finding, the licensee did a 100% inspection of 99 potentially af fected valve actuators. Nineteen of the 99 valves had come unqualified wirin This wiring affected 10 limit switches and 10 solenoids valve All questionable wiring was replaced before restart of the plant was initiate The unqualified wiring was traced to Bechtel's replacement of solenoid valves and the relocation of limit switches during the 1982 timeframe. It

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does not appear to be a vendor generic problem but an inhouse problem during one specific jo NPE is still evaluating reportability. This will be Unresolved Item 416/86-39-0 The resident inspectors have been monitoring licensee activities related to the correction of Standby Service Water (SSW) system deficiencies which were previously identified in NRC Inspection Reports 416/86-26, 86-32, 86-36 and 86-37. Briefly on August 26, 1986, while the plant was in hot shutdown (operating condition 3), the licensee identified errors made in the calculation of SSW flow to EST switchgear room coolers which resulted in the less than the design requirement This led to the discovery of piping and cooler fouling in those areas jointly served by Plant Service Water. While conducting piping design reviews for improved flow, it was discovered that a seismic qualification deficiency existed on six ESF equipment room cooling coil inlet and outlet nozzles located on four cooling coils, and that post accident temperatures in some of the Auxiliary Building corridors may be higher than originally calculate Through detailed engineering design reviews, system cleaning, cooler cleaning, and previously scheduled SSW system improvements, the problems were reduced to the point where recalculated heat loads and equipment space temperatures could be held to required values by all branches during design basis accident conditions and during normal plant operation with only the following additional hardware alterations:

The 3/4 inch diameter Reactor Core Isolation Cooling (RCIC) room cooler piping has been replaced with larger 1 1/2 inch diameter piping to reduce the resistance to flo The coil of one Division 2 EST switchgear room cooler located at Elevation 119 feet East has been replaced with a larger capacity cooling coil and its fan speed was increase In addition, flush and drain connections have been installed in EST switchgear cooler piping to facilitate more routine cleaning operations in the future, and annubars were permanently installed in the same piping to permit monitoring of SSW flow to these loads with minimal operational impac The licensee issued interim Licensee Event Report (LER) 86-029-03 on November 27, 1986, which provided a detailed summary of SSW problems experienced by the licensee since August 26, 1986, and in which MP&L considered the matter to be resolved to permit startup and resumption of power operations following the refueling outage. In view of the above problems, the licensee has committed to conduct an additional, independent scope design review of the SSW system to confirm its adequacy in fulfilling required safety functions. The actual scope, schedule, and other program details were to be discussed and provided to the NRC at a later dat The inspectors reviewed the LER and additional flow test results and saw no problem in proceeding with startup. The licensee is still evaluating the safety significance of plant operation prior to this outage with the reduced flow and other noted discrepancie SSW problems are being cracked under Unresolved Item 416/86-32-0 . . . _

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On December 12,1986, at 2:50 a.m. , with the plant in hot shutdown (Opera-tional Condition 3) and the turbine on the turning gear at approximately 114 3 rpm, turbine bearing metal temperature high alarm P680-9A-33 came in. In +

l accordance with Alarm Response Instructions (ARI), bearing temperatures were

! checked and all temperatures were less than 120F except bearing number 210

which ranged from 235F to 190F, Operators noticed the bearing temperatures

! decreasing and at 3:15 a.m., indicated turbine speed had decreased to 0 rp A local check verified the shaft was stopped. Investigation by the licensee revealed that a wrench socket had been left inside an oil cavity in the number 10 bearing housing which then impacted on the exposed portion (approximately 8 inches wide) of the shaft journal. This impact area became very rough which resulted in the stoppage of the turbine shaft. The licensee is evaluating proposed methods of repairing the number 10 bearing and journal surfaces. The reactor could be shut down another three to six weeks

, depending on the actions required to repair the damag No violations or deviations were identifie . Inspector Followup and Unresolved Items (92701)

< (Closed) Inspector Followup Item 416/86-37-06, Completion of the Contain-ment Pressure Instrument Test. In paragraph 12 of NRC Inspection Report 416/86-37, the inspectors discussed having witnessed an integrated contain-ment pressure instrument test pursuant to GGNS License Condition (LC)

i 2.C.(33)(b). During that period, the drywell pressure instruments were satisfactorily tested. The remaining containment pressure instrument tests were tracked to completion by this IFI. Subsequently, on November 14, 1986,

! the remaining instruments were satisfactorily teste The inspectors l reviewed the completed test documentation for completeness and although the information reflected a satisfactory test, the acceptance criteria called 1,

out paragraph numbers that did not exist in the procedure, and there were

other editorial errors which reflected inadequate attention to detail when the procedure was written, and when reviewed for approval. The licensee's representative involved with the test committed to correct the acceptance

criteria paragraphs prior to archiving the completed tes The other editorial errors were corrected in accordance with GGNS administrative i requirements before and during test commitments made by MP&L in the April 3,

, 1986 letter (AECM-86/0042) referenced in LC 2.C.(33)(b), thus satisfying the

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license condition for startup.

I (Closed) Inspector Followup Item 416/86-32-08, Design Change Package (DCP)

85/4061, provide heating and cooling to the remote shutdown panel (RSP)

room During the period covered by Inspection Report 416/86-32, the inspectors witnessed work in progress and reviewed documentation associated with DCP 85/4061. The closeout of the DCP was tracked to completion under this item. The inspectors inspected the completed work, reviewed the closed r

out DCP for completeness, and conducted followup reviews to verify the j required changes were made to related drawings and procedure There were i' no Technical Specification (TS) changes required; however, TS License Condition 2.C.(33)(a) states that prior to startup following the first

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refueling outage, MP&L shall demonstrate the ability to maintain an effective temperature condition of 85F or less in the RSP room for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with an ambient outdoor temperature of at least 95F. Since the post modification performance ' test could only be conducted with an outdoor ambient temperature of 73F, the licensee performed an evaluation

to determine the adequacy of the modificatio This satisfied License

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Condition 2.C.(33)(a) for startup, thus this item is closed. The licensee committed in letter AECM-86/0363, dated November 24, 1986, to perform an additional test when ambient temperatures are satisfactory to confirm the results of the evaluation. This test shall be tracked as Inspector Followup Item 416/86-39-0 (Closed) Inspector Followup Item 416/86-28-01, Design Change Package (DCP)

l 81/5018, Installation of Triaxial Acceleromete The inspectors reviewed the Design Change Implementation Package (DCIP) for completeness. The required post modification calibration was accomplished. Procedural changes

were identified but were not completely accurat The DCIP listed the procedures on the left below as requiring revision but the actual procedure number that was revised is listed on the right. This indicated a lack of attention to detai ,

As Listed in the DCP Actual Procedures Revised 06-IC-C85-0-1003 06-IC-SC85-0-1003 06-IC-C85-0-1004 06-IC-SC85-0-1004 06-IC-C85-0-1008 06-IC-SC85-SA-1008 Technical Specifications have been revised and the procedural changes were adequat The reinstallation of the triaxial accelerometers within one degree will be tracked with violation 416/86-37-04. License Condition (LC)

2.C.(7) is complet (Closed) Inspector Followup Item 416/85-45-08. The licensee was notified j of a potential 10 CFR Part 21 defect in TDI Diesel Generators (DGs). This

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defect concerned the recently experienced isolated failures of the intake

and exhaust valve springs after an extensive operating experienc This Part 21 report and the licensee's actions were previously discussed in

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Inspection Report 416/85-45. Although replacement of the springs was not required, the licensee inspected all intake and exhaust valve springs and replaced all springs that had white stripe The white stipes identified the springs as being from the Betts Spring Company, which manufactured the 1 suspect spring (Closed) Inspector Followup Item 416/86-26-04, Potential Loss of Unit 1 1 Safety Systems Due to Control Room Water Leak. The licensee has experienced several water leaks associated with temporary construction water systems and had repaired them. Several leaks were repaired in the temporary copper 4 lines supplying water to temporary air conditioning units in the control building. One of the water leaks on the Unit 2 side of the control room resulted in water leakage into the Unit 2 control room floor which is located directly over the Balance of Plant (BOP) computer. The 80P computer

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became inoperable due to the water from the leak. Also located under the Unit I control room floor is safety related wiring to the control panel The copper water lines had deteriorated and the licensee replaced the existing copper lines with carbon steel pipe.

'~ It appears that the temporary water lines to the temporary air conditioning

units were installed prior to receiving an operating license in 1982, and no safety analysis of the installation had ever been accomplished. There ,

are 17 of these air conditioning units in the control building on various elevations, 14 units on the Unit 2 side and three units on the Unit 1 side.

. A chanr to replace the existing copper lines with 3/4 inch carbon steel pipe was made under Temporary Alteration 86-0036 for the Unit 1 portion of l the control room. A Safety Evaluation Applicability Review appears to have

been conducted. The work was accomplished by Unit 2 personnel. As evidenced by the previous water leaks on the Unit 2 side of the control room, the 1 operability of Unit 1 safety systems could potentially affect safety related j equipment in the Unit I side of the control room. The licensee was requested to show what measures have been taken to ensure the above TS requirement is

} met when Unit 2 work which can affect the safety of Unit 1 is planned and/or

! accomplished. The licensee indicated that the measures are in the process

, of being placed in effec This shall be tracked under violation j 416/86-17-03 corrective action.

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l (Closed) Inspector Followup Item 416/86-32-13, DCP 84/3029, incorporate

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logic in Division 1 and 2 diesel generator (DG) protective trips to trip i only on engine overspeed and generator differential current under accident -

r conditions. During this inspection period, the DCP was completed on both ,

divisions and retesting was satisfactorily accomplished. The inspectors

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j j reviewed the completed DCP records and verified that all documentation

! required for signoff in support of satisfactory completion of this DCP was

in order. TS and licensee procedure changes were implemented consistent

! with DCP requirements and the inspectors noted that the required revisions to drawings were in the control roo The requirements for TS License _

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Condition 2.C.(37)(c) are satisfied. This item is close !

j (Closed) Inspector Followup Item 416/86-34-01, Licensee's resolution of  :

1- the Standby Service Water (SSW) system flow deficiencies and of their i conformance to License Condition 2.C.(20). On November 27, 1986, the

licensee published Licensee Event Report (LER) 86-029-03 as an interim i report documenting the resolution of SSW system deficiencies to permit l resumption of power operations. Discussions on the status of all SSW

appears in paragraph 10 of this repor Final resolution of all SSW ,

i problems area being tracked under Unresolved Item 416/86-32-05. Prior to  !

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startup, DCPs 82/5020 and 82/5020-1, which provided Loops B and A SSW (respectively) increased flow in support of License Condition 2,C.(20),

I were completed and closed out. Thus the license condition was satisfied

! prior to startup from the first refueling outage and there is no need to j continue tracking under this ite Closure of DCPs 82/5020 and 82/5020-1 J also sati sfy License Condition 2.C.(6), soil structure interaction

modifications requiring closure of DCP 84/4080-1 which reported closed in i

Inspection Report 416/86-37, and the above DCPs 82/5020 and 82/5020-1. This j item is close .

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(0 pen) Inspector Followup Item 416/86-36-02. Adequacy of Plant Service Water (PSW) Flow to Containment Penetrations. While reviewing PSW system drawings, the inspectors noted that some containment high energy piping penetrations depend on PSW cooling. A questions was raised, based upon degraded flow conditions found on SSW system low flow piping served by PSW during con-accident conditions, as to the adequacy of flow through these penetration Subsequently, the penetrations were identified as being limited to three associated with the Reactor Water Clean Up (RWCU) syste The licensee conducted flow tests and determined that there was no evidence

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of possible flow problems in these lines. The 3 inch pipe supplying the three smaller branches cooling the penetrations had over 34 gallons per minutes, and each branch had indication of full flow. Based on RWCU system operating requirements, the licensee explained that only one penetration of the three, number 83, will experience high temperature fluid, it being the RWCU return from the regenerative heat exchanger. The licensee implemented a special instruction to MWO 168509 to monitor for excessive temperatures at this penetration during hot plant conditions, to confirm there is no proble This item shall remain open until the special instruction is satisfactorily

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complete (Closed) Inspector Followup Item 416/86-32-14, Review of DCP 84/5007 closeout documentation, procedural changes and drawing change The inspectors reviewed the closed out documentation for the DCP and verified that TS changes, procedure changes and required drawing changes were implemented. This DCP was implemented in support of TS License Condition 2.C.(15), which requires prior to startup following the first refueling outage, installation of redundant scram discharge volume vent and drain valves and diverse and redundant scram instrumentation for each instrumented scram discharge volume, including delta pressure sensors and plant sensor The license condition was therefore satisfied, and this item is close (Closed) Inspector Followup Item 416/86-32-09, Design Change Package (DCP)

85/3122, Relocation of instruments to a maximum height of 70 inche The inspectors reviewed the completed DCIP 85/3122 and reviewed the remote shutdown panels for accomplishment. No procedural changes were required for this modificatio (Closed) Inspector Followup Item 416/86-11-01, Installation of new design inflatable seals on drywell and containment personnel air lock On September 19, 1983, the W. J. Wooley Company informed the NRC in a 10 CFR 21 report that a Prespray personnel air lock seal had failed a 465F acci-dent environment qualification tes On October 24, 1983, the licensee issued letter AECM-83/0674 to the NRC stating that differences between this test criteria and GGNS test criteria supported the conclusion that interim operation with the existing seals was justified until the first refueling outage when new seals would be qualified and available for GGNS. During this refueling outage the seals were replaced with upgraded inflatable seals in accordance with DCP 85/4504. The inspectors witnessed portions of the replacement, and reviewed the closed out DCP for completeness. This item is close .

(Closed) Inspector Followup Item 416/86-32-12, Addition of keylock switch for LPCI injection valve on remote shutdown panel. The inspectors reviewed the closed out documentation for DCP 86/3008 and verified that required procedural and drawing changes were incorporate The inspectors also verified the keys to the keylock switches were available. The keys are locked in a cabinet located in the remote shutdown room and access is limited to operation The keys were verified to actuate the keylock switches. This modification was made as part of LC 2.C.(18).

12. Design Changes and Facility Modifications (37700 and 37701)

The inspectors have been conducting document reviews and hardware inspections to ascertain that design changes and facility modifications associated with TS License Conditions were in conformance with the requirements of the facility license, TS, and 10 CFR 50.59. Part of the reviews were conducted during the two previous reporting periods and are documented in NRC Inspection Reports 416/86-32 and 416/86-37 and are being tracked as Inspector Followup Items. Those reviews that were continued and/or completed during this reporting period are documented in paragraph 11 above. The following additional DCP was reviewed during this reporting perio (Closed) Design Change Package 84/4075. By letter dated August 15, 1985 (AECM-85/0245), the licensee addressed an NRC concern regarding long term post accident operability of the Automatic Depressurization System (ADS) as part of NUREG-0737 item II.K.3.28. The existing make-up connection at that time was located in a high radiation area and would have resulted in excessive personnel exposure in the event of an accident. The licensee committed to relocate the ADS air supply make-up connection prior to startup from the first refueling outage. DCP 84/4075 was issued to accomplish the relocation of the ADS air supply make-up connection to the 139 elevation within the auxiliary building as committed in the referenced correspondenc The inspectors reviewed the completed DCP and reviewed the actual installatio Affected drawings and procedures were verified to incorporate the design chang One minor discrepancy was note Procedure 04-1-01-P53-1, Instrument Air System, Attachment 1, pg 62 listed valve P53-FA001 in area 9 on elevation 166 while reality valve P53-FA001 is located in area 9 on elevation 139. Although this commitment is complete it wit' be identified as Inspector Followup Item 416/86-39-06 for tracking purpose . Plant Startup From Refueling (71711, 61707)

The objectives of this inspection were to ascertain whether selected systems disturbed or tested during the refueling outage were returned to an operable status before startup, and to verify that plant startup, heatup, approach criticality, and core physics tests following the outage were conducted in accordance with approved procedure The inspectors conducted tours of the drywell, containment, and other areas of the plant as the predicted startup date neared, to ensure that general cleanliness of the areas which will have access restrictions during power operations is adequate and that equipment condition was being restored where

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appropriat The inspectors noted that the drywell in particular was restored to an excellent level of cleanlines There were no apparent equipment condition deficiencie The Control Rod Drive Hydraulic (CRDH) system and the Automatic Depressurization System (ADS) were selected at random for a walk through to determine operational readiness. The equipment was inspected for the presence of anomalies and the inspectors verified that the systems were being placed back into an operable status in accordance with the Technical Specifications (TSs) and approved plant operating procedures and lineup The ADS was inspected in greater detail in accordance with NRC inspection procedure 71710, which is an Emergency Core Cooling System walkdown. The results are documented in paragraph 8 abov '

In addition, the inspectors had opportunities to witness the proper restoration to service of Division 1, 2 and 3 emergency diesel generators as they witnessed portions of the retests and operational surveillances performed following DCP work and in the case of Divisions 1 and 2, the Quality Review / Design Revalidation work which involved extensive teardown, inspection, and reassembl Throughout the refueling outage, the inspectors verified that plant procedures were revised to properly reflect and support the numerous hardware changes made in accordance with the DCPs inspected as discussed in paragraphs 11 and 12 of this report. These procedure changes were approved as required by the TS On November 29, 1986, the refueling outage was terminated as the licensee placed the plant in operational condition 2, startup. Within about an hour, the inspectors witnessed the commencing of control rod withdrawal to achieve criticality. As control rods were withdrawn, the appropriate TS surveillances were conducted. The withdrawal sequence was conducted without incident except that several control rod drive mechanisms had to be vented, which created minor delays. Prior to startup, all control rod drive mechanisms and piping had been cycled, timed and vented; however, the licensee explained

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that additional air could have been drawn from condensate reject water after

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the venting procedures were performed. The reactor became critical at 4:48 a.m., on November 30, 1985. The inspectors then observed the source range nuclear instrument responses to control rod movement, source range monitor to intermediate range monitor (IRM) overlap, and the IRM Range 6 to Range 7 correlation test in accordance with Technical Section Procedure 09-S-02-20, Revision 0, Neutron Monitoring System Performanc During performance of the IRM Range 6 to Range 7 correlation test, the inspectors noted that the prerequisite plant conditions were not being met,

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1.e., step 6.4.1 of the above procedure requires power to be adjusted, if required, until the particular IRM channel being tested indicates 50 to 100 on Range 6. Channel B was initially tested at 35 on Range 6 and 3.0 on Range 7 which did not meet the acceptance criteria. As power drifted down due to the slight heatup in progress, the test engineer tested Channel B again at 20 on Range 6, 2.0 on Range 7, recorded the data, and signed it off

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as satisfactory. When the inspectors challenged the results, a discussion ensued regarding the ambiguities of the procedure. The test engineers did not appear to understand the intent of the test with regard to maintaining power level above 50 on Range 6. The inspectors pointed out that the previous shift had problems understanding the procedure, and that a change was made to clarify the requirements. The change apparently did not accomplish this to the extent required. It was apparent to the inspectors that supervision did not adequately consider clarifying the procedure such that all the people conducting the test could understand and follow the procedure. After two subsequent attempts to clarify the procedure, the test was resumed and satisfactorily complete TS 6.8.1 requires written procedures to be established, implemented and maintained covering test activities of safety related equipment. Contrary to this requirement, licensee personnel conducting the IRM Range 6/7 correlation test prescribed in step 6.4 of Technical Section procedure 09-S-02-20 failed to meet the required plant conditions, i.e., power leve This is violation 416/86-39-0 The resident inspectors witnessed portions of the startup in conjunction with a Region II inspector visiting the site for that purpose. The results of the joint inspection activities and additional startup inspections conducted by the visiting inspector are documented in NRC Inspection Report 416/86-3 . Management Meeting (30702)

On December 15, 1986, Region II management and the resident inspectors met with licensee management at the GGNS. Attendees are noted in paragraph The licensee presented an overview of the operating history of the plant, the first refueling outage and a preliminary summary of lessons learned from the first refueling outage. A brief summary of the status of major problem areas was discussed. The major problem areas discussed were the excessive number of reactor trips and the scram reduction program; emergency planning deficiencies and licensee initiatives; standby service water system problems and corrective actions taken or planned; quality program implementation improvement initiatives; drawing control corrective actions; Raychem splices and motor operated valve discrepancies and corrective actions; licensing activities and the updated Final Safety Analysis Report review progra Also discussed was the INPO accreditation status which noted an INP0 accreditation team vicit is scheduled for January 19, 1987, which will complete GGNS accreditation if successful. Region II management noted the licensee's actions on major problems areas was comprehensive but also noted that NRC inspection reports indicate a continuing problem in the timeliness and effectiveness of corrective action This is still one area of NRC concern. A short familiarization tour of the control room, turbine floor, auxiliary building and the Energy Service Center was conducted. A licensed operator awards banquet was held that evening in Vicksburg, MS, and Mr. Luis Reyes presented NRC license certificates to recently licensed Reactor Operators and Senior Reactor Operators.