IR 05000312/1988005

Insp Rept 50-312/88-05 on 880204-0329.Violations Noted.Major Areas Inspected:Operational Safety Verification,Maint, Surveillance,Security & Health Physics Routine Modules & Followup Items

ML20154J351
Person / Time
Site:	Rancho Seco
Issue date:	05/10/1988
From:	Ang B, Dangelo A, Miller L, Myers C, Qualls P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20154J321	List: IR 05000312/1988005 ML20154J317 ML20154J344
References
50-312-88-05, 50-312-88-5, NUDOCS 8805260311
Download: ML20154J351 (29)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION V

Report.No:- '50-312/88-05 Docket No. 50-312-License No. DPR-54 Licensee: Sacramento Municipal Utility District P. O. Box 15830 Sacramento, California 95813 Facility Name: Rancho Seco Unit 1 Inspection at: Herald, California (Rancho.Seco Site)

Inspection conducte / , s [

Inspectors: V/ / h/ VM\ / 0 -8i A.G. 4 Angelo, (5 of Resyd,ent Inspector Date Signed r / /&

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Accompanying Personnel: H.- erch,Seniorp(ctorInspector,RegionI Approved By:  %

Reactor ProjectsSection II b/0'N Date Signed L. g Mil'ler, Chi Summary:

Inspection between February 4 and March 29, 1988 (Report 50-312/88-05)

Areas Inspected: This routine inspection by the Resident Inspectors and in

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part by a Regional Inspector, involved the areas of operational safety verification, maintenance, surveillance, security and HP routine modules and followup items. During this inspection, Inspection Procedures 71707, 71709, 71881, 71710, 92720, 62703, 51726, 92701, 92702, 62700, 25571, 92700, 30702, and 30703 were use Results: No general conclusions regarding the adequacy, strength or weakness of the areas' inspected, nor any significant safety matters were identifie One violation was identified pertaining to the failure to document a nonconforming test result. One new unresolved item relating to environmental qualification of modified safety related motor operated valves was identifie PDR ADOCK 05000312

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DETAILS

~1. Persons Contacted j

a .- . Licensee Personnel G. C. Andognini, Chief Executive Officer, Nuclear

'd. Firlit, Assistant General Manager (AGM), Nuclear Power Production D. Keuter, Director, Nuclear Operations and Maintenance J. McColligan, Director, Plant Support D. Brock, Acting Nuclear Maintenance Manager

• B. Croley, AGM, Technical and Administrative Services

. G. Cranston, Nuclear Engineering Manager W. Kemper Nuclear Operations-Manager

• J. Shetler, Director, System Review and Test Program T. Tucker, Nuclear.0perations Superintendent P. Kagel, Nuclear Mechanical Maintenance Superintendent L. Fossom, Manager, Scheduling and Outage Management J. Field,. Plant Support Group Supervisor p L~. Conklin, Manager of Management Controls S. Crunk, Manager, Nuclear Licensing J. Vinquist, Director, Nuclear Quality T. Fetterman, Electrical Engineering Manager J. Irwin, Supervisor, I&C Maintenance T. Shewski, Quality Engineer

• V. Lewis, Supervising Civil Engineer J. Robertson, Licensing Engineer P. Bosakowski, Supervisor of Licensing, Technical Support

• J. Delezinsky, Supervisor, NRR Coordination, Licensing

• G. Legner, Licensing Engineer H. J. Sefick, Jr. , Nuclear Security Manager Other licensee employees contacted included technicians, operators, mechanics, security and office-personne * Attended the Exit Meeting on March 29,'198 . Operational Safety Verification (71707,71709,71881)

l The inspectors reviewed control room operations which included access control, staffing, observation of decay heat removal system alignment,

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and review of control room logs. Discussions with the shift supervisors E and operators indicated understanding by these personnel cf the reasons l for annunciator indications, abnormal plant conditione and maintenance I work in progress. The inspectors also verified, by observation of valve I and switch position indications, that emergency systens were properly-aligned as required by the technical specification for the plant

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conditions.

l l During this inspection period, the licensee changed reactor modes from

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cold shutdown to the heatup-cooldown mode on March 1, 1988 and heated the unit up to the hot shutdown mode. The inspectors verified the technical

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specification requirements were being met for operations in these mode Tours of the auxiliary, reactor, and turbine buildings, including exterior areas, were made to assess equipment conditions and plant conditions. Also, the tours were made to assess the effectiveness of radiological controls and adherence to regulatory requirements. Required posting and secured high radiation areas were found intact with radiation protection personnel at their station Exit portal radiation monitoring equipment appeared to be operating properly. The inspectors also observed plant housekeeping and cleanliness, looked for potential fire and safety hazards, and observed security and safeguards practice Security personnel and manning requirements were met within the protected are No violations or deviations were identifie . ESF System Walkdown (71710)

During the portion of this inspection period when the unit was in cold shutdown, the inspectors walked down and verified the operability of the Decay Heat Removal system. After the reactor coolant system temperature had been increased above 280 F, the inspectors: 1) verified the capability to remove heat by use of two steam generators, 2) verified that one atmospheric dump valve (ADV) per generator was operable, 3) verified that the licensee had the minimum 250,000 gallons of water in the condensate storage tank, 4) verified the ADV isolation valve positions, 5) verified the operability of the steam safety valves, 6) walked down the lineup for the emergency feedwater system (EFW)

systems and verified operability, and 7) walked down the backup instrument air bottle supply systems for the atmospheric dump valves (ADVs) and emergency feedwater (EFW) system valve No violations or deviations were identifie . Independent Inspection (92720)

Review of B&W Owner's Group Recommendations (BW0G)

l l The inspector reviewed the status of the licensee's implementation of the i BW0G recommendations for trip reduction and transient response l improvement. Recommendations were developed by the Safety and Performance Improvement Program (SPIP) and were prioritized with the most important and beneficial recomendations being identified as "Key" trip reduction (TR) item The inspector reviewed all TR items (both key and non-key) for implementation. Key TR's which were not being currently implemented or complete were discussed with the licensee and justification for deferring the key TR was reviewed. The justification method used by the licensee was contained in procedure RSAP-0215, "RSL/LRSL Development and Administration."

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The selection criteria for implementation of TR's prior to restart of the reactor were whether the TR would assure that: the plant remained in the post-trip window (essentially, the reactor coolant system pressure and temperature remained within required limits), the facility would comply with technical specification requirements, and operator action outside the control room within the first ten minutes of an event would not be neede Key and non-key TR's which were outside this criteria were deferred to the next refueling outag The licensee had delayed five key items for implementation until after restar The deferred key items pertained to non-safety related, non-accident mitigating equipment within the plant (improvements in the integrated control system (ICS), main feedwater pump and block valve controls, and the moisture separator reheater (MSR) drain tank controls).

The inspector concluded the items appeared have been appropriately deferred for post restart. The licensee has committed (on the long range scope list (LRSL)) to conduct a study for implementation of each item and to submit a proposed schedule to the NRC by June 198 . Monthly Maintenance Observation (62703)

Maintenance activities for the systems and components listed below were observed and reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, industry codes or standards, and the Technical Specification The following items were considered during this review: The limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing or calibration was performed prior to returning components or systems to service; activities were accomplished by qualified personnel; radiological controls were implemented; and fire prevention controls were implemente ) The inspectors observed the disassembly of the Transamerica DeLaval Industries (TDI) diesel lube oil pressure regulating valve. The valve had not been functioning as required. The licensee found that the valve, upon disassembly, had the flow directing piston installed backwards. The licensee believed the installation error had occurred due to poor marking of the piston. Additional guidance on

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assembly of the piston was provided to prevent reoccurrenc ) One control rod drive breaker under voltage trip device (UVD) failed to open during testing. Licensee personnel began to troubleshoot the trip breaker (GE Model AK2-25-2) failure under a Shift Supervisor Emergency Maintenance (SSEM) work reques Work controlled under an SSEM was not required to be preplanned; the technician investigates the cause based on whatever written guidance was contained in the work request. The technician, however, was required to document all work performe . .- . -.- - - . _ - - . -- . .. ,, ..... - - - - - - . ---

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During the investigation conducted under the SSEM, the licensee stated the potential cause for the VVD armature to bind and not trip open the breaker was an interference observed between the UVD armature and UVD nameplat Subsequent discussion between Region V and the licensee resulted in a decision by licensee management that an SSEM may not have been appropriate for this event. The SSEM may not have been appropriate in that a troubleshooting plan was not developed to identify and preserve the as-found conditions, and to assure that the root cause of the failure had clearly been found. A stop work was issued by the Quality Department on the use of SSEM' Further investigation of the UVD failure under the controlled conditions of a preplanned work request using the reactor trip breaker surveillance procedure, revealed a mislocated button welded onto the UVD armature. The button is a cylindrical metal tab which was the contact surface between the UVD armature and the trip paddle. The button appeared to have been welded slightly too low,

such that when the button contacted the trip paddle, the contact j surface was at the circumference of the button. That mislocation resulted in a small contact area which appeared to have grooved the button, and possibly cause the binding observe The licensee reinspected all other reactor trip breakers to determine if they had a similar mislocated button. The other five breakers had acceptable button locations, lhe licensee has contacted the vendor, General Electric Company, to determine what corrective action should be taken to rework the UV The licensee has also been developing criteria for the use of SSEM's. The issue of appropriate use of SSEM's will remain an open item pending inspection of the SSEM criteria (50-312/88-05-01).

3) Maintenance activities on the A2 TDI diesel air control pneumatic system were observed involving routine air bottle replacement for the safety related backup control air system.

i 4) Corrective maintenance on the B2 TDI diesel to reduce vibrations and repair cracked welds was observed. The licensee's resolution to the i

weld cracking problem on the exhaust shroud was to remove the shroud entirel Engineering analysis was performed in connection with this modification which appeared to have adequately addressed engine and room temperature consideration ) Removal of Spare Limit Switch Assembly From Limitorque Motor Operators During this period, the licensee identified that a previous modification of Limitorque motor operators had not been adequately reviewed for its effect on the environmental qualification of the electrical equipmen The modification involved the removal of a spare two-rotor geared l

limit switch assembly from a geared limit switch assembly containing l four rotors in the motor operato The removal could cause

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lubricant to leak out of the gear cartridge which in time could affect the remaining limit switch assembly and potentially make the motor operator inoperable. The licensee identified 69 valve actuators which were affected, 50 of which were identified as important to safety per 10 CFR 50.49. The spare rotor assenolies had been removed during a period from 1983 through 198 The licensee evaluated the as-found condition of the actuators and determined that the removal of the spare rotor assembly did not affect the environmental qualification of the actuator. However, prior to restart, the licensee indicated that all of the missing spare limit switch assemblies in Class 1 valve actuators would be replace The inspector reviewed the licensee's corrective action request CAR-88-07 which identified the generic concern and corrective action for the removal of the limit switches. The inspector found that the licensee appeared to be adequately resolving the specific hardware problem and also addressing the weakness in the control of the equipment qualification. This item is unresolved pending further review, including the completion of the licensee's corrective action to prevent reoccurrence of such modifications (50-312/88-05-03).

Various other plant maintenance activities including motor operated valve testing, instrument and control system (I&C) maintenance on emergency feedwater initiation and control (EFIC), other TDI diesel work, and electrical maintenance activities were observed in par No violaticas or deviations were identifie . Monthly Suiveillance Observation and Review of System Review and Test Program (SRTP) Testing (61726)

System testing under the System Review and Test Program (SRTP) was observed and reviewed to ascertain that the testing was conducted in accordance with the requirements of the approved special test procedure (STP). The STP test requirements to demonstrate system functionality were contained in the System Status Reports (SSR). Inspections and observations were conducted during the performance of the STPs and maintenance activities in support of the restart effort to ascertain that they were conducted properl The following testing activities were observed and reviewed during this report period: STP.961 - Loss of Offsite Power Performance of the initial sections of this test was previously observed and reported in Inspection Report 50-312/87-44. That report discussed test control problems which resulted in an erroneous operation of the B2 TDI diesel generator in parallel with offsite powe In this period, while performing a subsequent portion of the test, the licensee inadvertently drained approximately 1100 gallons of

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slightly contaminated borated water from the borated water storage tank (BWST) through the two lower reactor building spray rings. A safety features actuation signal opened the reactor building spray system isolation valves as part of the test. As planned, the spray pumps did not start (their breakers were lined up in the test position). However, contrary to the test plan, a manual valve in one train of the spray system was open when it should have been shut for the test. This allowed borated water to drain from the BWST (+47 ft, elevation) to the two lower spray rings (-2 ft, and +38 ft, elevation). Operators detected this condition after about thirty minutes, while restoring the normal valve lineup for containment spra The licensee's response to this event appeared to have been carefully planned and implemented. Action plans were developed for each functional area (operations, startup testing departments) to establish accountability and clearly describe responsibility for personnel action and performance during the conduct of a test. The action plan had been developed by each functional group to document their own expectation of acceptable performance. The inspector concluded this action was adeouate for this even Following resumption of test ang, the "A" TDI failed to start when a start signal was received. The engine rolled on starting air momentarily, but did not run. After troubleshooting efforts using the normal pre-start checks found no cause for the failed start, a single restart was attempted and was successful. Testing continued after a safety evaluation of the start failure. A second start later in the test was also successful. A work request was initiated to replace or rebuild some pneumatic control components which were suspected to have malfunctioned resulting in a failure of the pneumatic logic to properly reset to allow starting of the engin During the performance of a subsequent section of the test, the A2 TDI diesel output breaker failed to close into the A2 bus as expected when the bus experienced a loss of power. Investigation of the breaker lineup identified a test disconnect switch which was erronecusly open and prevented the alternate power supply breaker from Startup Transformer #2 into the A2 bus from opening. An interlock circuit had, therefore, prevented a close signal from being applied to the diesel output breaker, preventing the diesel generator from paralleling to the A2 bus, picking up additional loads from Startup Transformer #2. The licensee found that the test switch had been opened during the performance of previous testing in November, 1987. A walkdown of all breaker test switches was performed to ensure that any switch out of normal position was controlled per the test in progress. The licensee concluded the cause of the problem to be inadequate control of breaker test switch position as part of the prescribed electrical alignment for the tes The procedure was revised to incorporate test switch position verification as part of the electrical lineu During a shutdown of the "A" TOI in a later section of the test, the engine unexpectedly restarted and the unit was emergency stopped at

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the direction of the Assistant Shift Supervisor. The unexpected restart occurred when the operator manually pushed a circuit test switch "TEST BYPASS" during the shutdown. Although the operator had not been directed per procedure to manipulate the test switch, operator training had indicated that the test switch operation should not affect the engine operation. However, operation of the test circuit switch blocked the stop signal to the engine before it had stopped a!1owing it to continue runnin The licensee investigated the circuit design and concluded -that the circuit was operating correctly per its design. Additional training was given to all operators to make them aware of the problem to preclude mainipulation of the test switch during a shutdown sequence. In addition, the test switch was danger tagged to prevent operation for the remainder of the tes The inspector observed several test briefings, and performance of sections of the test, and found that the testing was performed in accordance with the procedure and appeared to have been successfully complete The inspector reviewed the results of the testing after completion of all sections of the test to determine if all. test deficiencies noted during the testing had been properly resolved. The inspector identified two test deficiencies which did not appear to have been adequately addressed by the licensee: A test deficiency in the load sequence interval had been identified by the licensee on review of the strip chart recordings of the bus votage following the start of the

"A" Bruce GM diesel generator. The load sequence interval did not meet the acceptance criteria of the test which required voltage to be restored within 40% of each load sequence time interval. The inspector reviewed the disposition of NCR No. 7813 which had been initiated and found that the test deficiency had been determined to be acceptable based on an interpretation of the USAR. USAR Section 1.6.9 commits the licensee to Safety Guide 9, "Selection of Diesel Generator Set Capacity for Standby Power Supplies." The Safety Guide provided criteria to be used for determining when a diesel

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generator was sufficiently stable (based on output voltage and

' * frequency) to continue to accept the next load block. That criteria required that the engine / generator recover to minimum

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voltage and frequency within 40% of the time interval between

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applying the next load. In discussions with licensee

representatives, the inspector questioned the validity of the l USAR interpretation as an appropriate technical basis for acceptable disposition of the NCR. To address the inspectors'

concern, the licensee reviewed and revised the NCR disposition to clarify the technical basis for the acceptability of the load sequence time interval. In addition, the licensee initiated a ci,ange to the USAR to clarify their commitments I with regard to both the Bruce GM diesel generators and the TOI l diesel generators. The licensee's position was that their

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measured engine / generator response time for recovery was approximately 46% of the interval time instead of the 40%

specified in Safety Guide 9. However, the most recent revision of that criteria permits a maximum interval time of 60%. Since the new criteria permitted a longer recovery (60%), the licensee considered that the diesel generators were acceptable

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as is. The inspector found the licensee's response to be adequate to. resolve the concern.

2. A failure of the "A"-TDI diesel generator to automatically hot i restart on safety features actuation signal (SFAS) had been noted by the licensee as a test deficiency not affecting the acceptance criteria of that section of the test during the conduct of the loss of offsite power test, STP 96 The inspector questioned licensee representatives regarding why the failure of the TDI engine to start was not considered to be

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a test failure. The licensee explained that when the TDI did not respond to the SFAS actuation, the test director was able i to manually recycle the pneumatic reset logic of the diesel l control circuit which enabled the diesel to start within five

. minutes of having been shutdown. Although automatic SFAS I actuation was the prescribed method of starting the diesel per

! the procedure, successful automatic SFAS actuation was not part of the acceptance criteria for the test. As a result, the test director's intervention to enable the engine start was not considered to affect the acceptance criteria. An occurrence description report (0DR) had been initiated by the licensee identifying previously initiated pneumatic component replacement from two previous failures to start as corrective action in addition to interim procedural requirements to ensure proper reset of the pneumatic controls following shutdown of the diese Procedure 961, "Loss of Offsite Power", section 6.17, tested the "A" TDI diesel generator hot restart function. This test ran the engine with the generator producing approximately three megawatts for one hour. The engine was then to be shut down remotely from the control room, stopped for less than five minutes, and then restarted automatically following a manual SFAS initiation from the control room. However, for this test, the engine failed to restart automatically. Test personnel in the diesel room manipulated local manual valves on the engine without a procedure describing their actions. STP 961, Section 6.17.2.11 required the generator to reach rated speed and voltage within ten seconds of the start signa The inspector was concerned that the failure of the "A" TDI to start automatically might not be adequately resolved since it was not identified for engineering disposition through an NC Furthermore, the inspector was concerned that troubleshooting by simply replacing components rather than determining the failure mechanism from controlled systematic examination would not be effective in determining the root cause of the proble I t

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Procedure QAP-17, titled "Nonconforming Material Control",

defines, in paragraph 4.1, a nonconformance as;

"A deficiency in characteristic, documentation, or procedure which renders the quality of an item (materials, parts, components or system) unacceptable or indeterminate. Examples of nonconformance include:

Physical defects, test failure, incorrect or inadequate documentation, or deviation from prescribed processing, inspection or test procedures. Nonconformances are hardware related."

Paragraph 5.1.1 further states that;

"Systems, equipment and appurtenances, components, parts or material which do not meet the specified requirements of purchase orders, design drawings, operational / test documents, or construction documents shall be considered nonconforming. Refer to Attachment 3 for guidelines on when to write NCRs."

Licensee rersonnel stated that the rationale for not preparing an NCR was that the failure to start of the diesel engine was known and did not effect the SFAS start of the engine and therefore was not a serious deficiency. The licensee had also established a compensatory measure in that the pneumatic system on the diesel was visually inspected to assure it was properly reset. When the reset was accomplished, automatic starting of the diesel was enabled and the engine would have met the technical specification starting requirement In response to the inspector's concern, the licensee reviewed the documentation of the problem and initiated an NCR to ensure engineering review of the adequacy of the ongoing wor Following replacement of the suspect pneumatic components, the reset perfonnance of the pneumatic control was improved but inconclusive in identifying the root cause of the problem. A more extensive investigation by the licensee identified that excessive field run lengths of a pneumatic sensing line were introducing unexpected restrictions to the pneumatic component venting, preventing proper repositioning of a pneumatic valv After rerouting the affected pneumatic tubing, the reset performance of the pneumatic controls for the TOI diesel was demonstrated to be reliabl Following careful inspection and engineering review by the licensee, it was discovered that the failure to reset of the pneumatic system was attributable to two causes. First, the pneumatic valves were operating in a sluggish fashion; this cause was always suspected by the licensee. Second, the length of the sensing line was routed in such a way as to introduce higher than expected line function losses. This problem was

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discovered by testing and engineering review after the NCR was generate STP 961. stated in the "Test Objective" section (paragraph 1.6)

that the diesel generators auto start on a simulated accident signal and loads are sequenced by the load sequencer. The procedure further specifies in step 6.17.2.9 that the engine will be started from the control room by depressing the SFAS channel button and then verify that the diesel generator obtains rated speed and voltage within 10 seconds on steps 6.17.2.11.3 and 6.17.2.11.5/6. The procedure makes no-provisions for intervention and manual operation of the control air system to achieve an automatic SFAS hot restart of the engin The inspector's prime concern was that a formal problem identification system estaulished by the licensee (the NCR process) was not used to ensure that effective corrective action was performed. The licensee subsequently established a detailed program to identify all the root cause problems with the pneumatic system after the NCR was generated. The failure to write an NCR is an apparent violation. (88-05-02) Auxiliary Feedwater System Testing Review The inspector sampled the test deficiencies documented during previous testing of the auxiliary feedwater system. The inspector found all deficiencies noted to be appropriately dispositioned and properly resolve STP.664 - Loss of ICS/NNI This test was performed to duplicate the system failure which initiated the overcooling event of December 26, 1985. The purpose of the test was to verify the adequacy of the system modifications which have been made to maintain control of plant temperature and pressure following a loss of power to the integrated control system (ICS) or the non-nuclear instrumentation (NNI).

The inspector observed the pre-test briefings for the test personnel and operators and the performance of the test conducted from the

control room. The inspector found the test to have been conducted per the approved procedure with good coordination and communication

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among test participant The initial performance of the test was suspended following a failure of the alternate turbine bypass valve controller to control as require Licensee investigation identified a miswiring in the controller which resulted in a short to ground. The licensee concluded that this factory error had existed for about 1 year since l the controller was installed in the panel but had never before been detected.

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After rewiring the controller to correct the problem, testing wat l resumed. During the final section of the test, the turbine driven

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auxiliary feedwater pump failed to achieve its rated spee Licensee investigation of the failure identified a miswired electrical connection in the turbine governor which had recently been rewired by maintenance. Following correction of the problem, the testing was completed successfully, meeting all acceptance criteri At the exit meeting, the inspector discussed the apparent need for management attention to the adequacy of post-maintenance and post-modification testing to establish reasonable assurance of component level functionality prior to the conduct of integrated system function testin The inspector observed that the reliance on an integrated system functional test as a post-maintenance / modification could result in equipment failure which could cause an operational event, such as overcooling, if the turbine bypass valve went to the full open positio The licensee acknowledged the inspectors' concerns and indicated that they were reviewing the adequacy of the in-process checking of component installations for ensuring component functionality at the close of this inspection perio STP.113 - EFIC Hot Functional Testing This procedure performed an integrated functional test of the Emergency Feedwater Initiation and Control (EFIC) system under hot shutdown conditions to demonstrate (a) automatic control of Once Through Steam Generator (OTSG) pressure using the atmospheric dump valves (ADVs), (b) automatic auxiliary feedwater (AFW) initiation on low level in the OTSGs, (c) minimum design AFW flow delivery to the OTSG, and (d) reactor coolant system (RCS) temperature control from the remote shutdown pane The inspector observed the pre-test briefings and performance of the tests. The test was conducted in a well controlled fashion per the approved procedure with satisfactory results obtained, meeting the test acceptance criteria and quality program requirements.

, STP.1156 - Letdown Cooler Testing

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On March 15, 1988, with the reactor in hot shutdown at normal operating temperature, a leak of about 100 gpm occurred when a downstream pressure relief valve (PSV 22021) stuck open while opening the upstream isolation valve to the "A" letdown cooler (SFV 22005) for motor operated valve testing, i

L The leak started when SFV 22005 was opened, allowing RCS pressure (about 2200 psi) to be seen at the Cooler Relief Valve, PSV 2202 The relief setpoint of PSV 22021 was 2500 psi. The licensee shut SFV 22005 and isolated the leak. Due to the valve testing in progress, the leak was initially thought to be a packing leak from SFV 2200 Inspection subsequent to the event determined the leak

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l was from PSV 22021. Based on containment sump drainage, it was I estimated that about 650 gallons of water was released. An Unusual Event was declared at 10:49 pm and terminated at 11:17 pm PST.

h As a result of the relief valve lifting, the licensee instituted l

additional caution in conduct of MOV testing with the plant in hot

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shutdow However, when susequently attempting to cautiously initiate flow to the "C" letdown cooler, the licensee again observed brief lifting of the cooler relief valve. After additional engineering review, the licensee determined the problem to be due to the rapid initiation of RCS letdown flow into the depressurized letdown coolers which could contain some voids due to cooling when isolated. The licensee concluded that the relief valve lifting was due to water slug impact on the valve seat when flow was initiated to the cooler. The licensee postulated that the combination of the high differential pressure initially across the inlet valve, and flashing within the cooler, accelerated the initial inlet flow to an impact velocity which resulted in a transient load in excess of the relief valve lift pressure, causing the valve to lif To correct the problem, the licensee implemented a modification of the cooler inlet valve ooerator control circuit to allow jog opening of the valve to slowly itiate flow to the coolers. In addition tu replacing the stuck relie, valve, the licensee also removed the discs from the check valves in the outlet of the coolers to allow all three coolers to remain pressurized whenever any cooler was in servic Following system modification and operating procedure changes, tho licensee conducted testing of the letdown system. The inspector observed the conduct of the testing per the approved procedure and found it to be successful in demonstrating the effectiveness of the corrective actions to preclude repeated lifting of the cooler relief valves when initiating flow to the cooler Procedures were modified to ensure the coolers were always unisolated through the

"B/C" letdown cooler inlet isolation valve, a globe valve. The "A" cooler inlet isolation valve is a gate valve, and was considered more likely to produce a water hammer when the system was '

unisolate . Control of Activities Affecting the Inservice Inspection (ISI) Program (92701)

NRC Inspection Report 87-05 observed that there was no formal system for notifying the ISI Group of changes to drawings, modifications or systems that could affect the ISI progra In response to this finding, the licensee issued a revised procedure, AP.44 Revision 10, containing requirements to coordinate activities that could affect the ISI progra The effective date of this procedure was February 5, 1987. However, during the current inspection, it was observed that the licensee has had problems in the area of changes that affected the ISI program. CAR 88-003, dated January 14, 1988 documented cases of code boundaries that were different from the original code of construction boundaries. In the I

instance noted, the controlled drawing showed only the original code of

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construction boundaries, and not the present ISI boundaries. Other similar problems identified by the licensee are documented in CAR 88-00 During his review, the inspector found that AP.44 had been revised in 1987 subsequent to Revision 10, and that the controls for coordination of changes that could affect the ISI program were renoved. CAR 88-003 could have been avoided if AP.44 Revision 10 had not been revised and the controls taken out. The inspector also determined that the corrective actions that were being taken under CAR 88-003 should provide the proper controls to coordinate Engineering with IS Also, the inspector reviewed the implementation of procedure AP.44 Revision 10, and determined that implementation of the procedure should <

eliminate identified problems in coordination of ISI and engineering / modification activitie This inspection revealed continuing problems in coordinating activities that affect the ISI Program. The licensee has made several changes to improve the ISI Program. These include actions such as revising procedures NDEl-0901, "Nondestructive Examination Program" and QAIP-1804,

"Control of Special Processes". The issue of coordination of ISI inspections with modification activities will remain an open item to followup on the effectiveness of licensee's corrective actions on this issue (50-312/88-05-04). Followup on Previous NRC Inspection Items (92702,92701,62700,25571, 92700)

(Closed) Violation 86-07-01 "Civil Penalty Violations Resulting from Inspection of the December 26, 1985 Event" On October 22, 1986, the Nuclear Regulatory Commission issued a Notice of Violation and Proposed Imposition of Civil Penalties (86-07-01). On November 20, 1986, SMUD responded to the Notice of Violation. This item collectively addresses the eleven parts of the Notice of Violation

identified during Inspections 50-312/86-06 and 50-312/86-07.

! The Notice of Violation included the following violations of specific requirements: Reactor coolant system cooldown rates during the event violated Technical Specification Failure to correct identified design deficiencies leading to loss of integrated control system power, and failure to develop procedures which would have lessened the severity of the ensuing transien C.1 No written procedures existed for securing high pressure injection following safety features actuation, or for manual emergency operation of an auxiliary feedwater system control valv C.2 No written procedures existed requiring periodic maintenance of an auxiliary feedwater system valve.

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l D.1 Emergency Operating Procedure E.05, Excessive Heat Transfer, was inadequately implemented during the even D.2 Alarm procedures for evacuation of personnel for high gaseous i activity monitor alarms were inadequately implemente j D.3 Procedures for evaluation of unplanned radioactive releases were not implemente E.1 Emergency Plan in-plant notification procedures were inadequately implemente E.2 Emergency Plan procedures for notification of state and local E.3 governments were inadequately implemente (Twoexamples)

E.4 Emergency Plan procedures for implementation of radiation monitor alarm setpoints were inadequately implemente Further, Section 2.E of Inspection Report 50-312/86-07 (Item 86-07-01)

noted that the items collectively indicated a serious breakdown in the management controls which should have established and implemented important procedure Inspections of licensee corrective actions for the individual items noted above had been performed aad were documented in Inspection Reports 50-312/86-37, 87-02, 87-06, 87-21, 87-30, and 87-47. In addition, Inspection Report 50-312/88-08 documented inspection of items D.2 and D.3. NUREG 1286, the Safety Evaluation Report (SER), related to the restart of Rancho Seco Nuclear Generating Station following the event of

, December 26, 1985, and Supplement 1 to the SER, provided NRC staff evaluations of various details of the licensee's corrective action SMUD's Licensing Department was tasked by the licensee to track and clc eout comitments related to the violations listed abov The Lio nsing Department compiled a documentation package which included all the violations listed above, the licensee's corrective action commitments for each item, and closecut documentation f(r each item. The licensee's Quality Assurance Department performed a review of the package and determined that all corrective actions had been completed. The Assistant General Manager, Technical and Administrative Services, performed a management review of the package and determined that commitments for the licensee's corrective action for the Notice of Violation had been complete On the basis of the licensee's corrective actions and reviews, the various NRC inspections of the individual details of the Notice of Violation, the NRC staff evaluation provided by NUREG 1286 and Supplement 1 of NUREG 1286, the violation was close . . . . . - . .- -_ _ - - . .-

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86-07-03 (Closed) "! DADS Indicated Erroneous Airflow from Essential HVAC Systems for the Control Room" This item resulted from an observation by the inspector, that the Interim Data Acquisition and Display System (IDADS) indicated an airflow of between 400 and 600 cfm in each train of the essential HVAC system when the trains were shutdown and inoperable. The concern was that inaccurate data from IDADS would misrepresent the HVAC operating conditions to the operator The licensee's response was that the IDADS may show very low flow when the unit is shutdown. The display was attributed to the digital nature of signal processing in IDADS. When the IDADS displayed 400 and 600 cfm in each train, there was a 4 ma input signal to the transmitter. The transmitter was recalibrated from a 4 ma input signal to a zero ma input signal. The subsequent alignment check on this instrument showed plus or minus 150 cfm on the digital display anytime the essential HVAC systems were off. The licensee's explanation was that this was due to inherent instrument response at no flow conditions. Since this was apparently a characteristic of the flow instruments, this item was close (Closed) "E0P Rule Training" This open item concerned lack of documented training of the licensed operators in the five Emergency Operating Procedure (EOP) rules and the unfamiliarity of the operators with the requirements on when to use full High Pressure Injection (HPI). The inspector verified by reviewing training documents that all crews were subsequently trained on the five rules. The inspector also interviewed the on-shift operators for one crew about the requirements for full flow HPI and subcooling margin. The operators discussed what the training had covered and were knowledgeable of the five rules and when full HPI was required. This item is close (Closed) "Short Stroking of Motor Operated Valves" This item deals with a potential misoperation of motor operated valves (M0Vs) which the inspector had previously observed to potentially result in excessive valve stem thrust load The inspector had observed that "short stroking" an M0V was a way to cause the motor operator to realign from its manual handwheel drive mode into its electric motor drive mode without having to stroke the valve full open then closed. After manually closing the valve with the handwheel, initiating an electrical close signal will cause the operator to briefly energize and check shut itself. However, the inspector had observed that when operated in this fashion under test conditions, the resultant valve stem thrust could exceed the value developed under full

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stroke operation. The inspector was concerned that excessive thrust could be potentially damaging to the valve and/or the operato The licensee investigated the inspector's concerns to determine if short stroke operation of MOVs was a common practice and whether operat hns and maintenance personnel were aware of this potential conditio . - - , . . _ . . - - . - _ _ - _ . - _ . - . . . - , - _ -

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The inspector reviewed the results of the licensee investigation which concluded that "short stroking" was not a technique used by Operations or Maintenance personnel when restoring MOVs from manual operatio In review of the licensee's investigation results, the inspector found, however, that the licensee considered the technique to be undesirable because it was considered to result in inadequate closing thrust, thereby jeopardizing the containment isolation function. In discussions with licensee representatives, the inspector, again, emphasized his previous observation that the short stroking techniques appeared to result in excessive closing thrust which could jeopardize the struchral integrity of the valve operator, the valve internals, or the pressure boundary itsel '

Furthermore, to determine the licensee's awareness of this potential condition, the inspector reviewed the design basis report for ECN-R-5912 which modified the controls for the Limitorque operator of two safety-related MOVs to allow jog-open control of the valves. The inspector noted that when operated in this fashion, the potential for short stroke closure of the valve existed if the valve received a close signal (either manual or SFAS) while cracked open under jog contro However, the inspector found that this condition was not evaluated under the design basis report. In discussion with licensee personnel, the inspector found that the licensee considered that the risk of short stroke occurrence was negligible since jog control was used to slowly open a valve, but the final position of the valve would be either full open or full close Based on his review, the inspector concluded that the licensee's investigation addressed the concern in that management direction was given along with procedural direction that no valves were to be short-stroked. If a need arose to open a valve manually, the control room would cycle the valve one full stroke to prevent short-strokin Labeling was also added to all M0V's instructing site personnel not to manually engage M0V's without proper work clearances. This item is close (Closed),"TroubleshootingProcedure" ,

This item follows up on the licensee's developement of an equipment troubleshooting program which the licensee had committed to establish prior to restart. The licensee had previously established an Incident '

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Analysis Group to investigate and determine the underlying root c uses of l plant transients. However, this review group was only triggered into action following significant plant transients and was not routinely l invoked for more routine equipment troubleshootin The inspector reviewed the licensee's maintenance administrative procedure MAP-17, Root Cause Determination, which the licensee had established to evaluate the root cause of equipment failures and maintenance concerns. The inspector found that the procedure did establish a root cause review process for use within the maintenance ,

organization. The inspector noted that the root cause review was to be conducted by a single maintenance engineer. Although the procedure

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referenced consultation with other Nuclear Organization personnel in determining the root cause of the failures, the inspector noted that no interdiciplinary participation in the investigation or review of the investigation findings and the root cause determination was require The licensee representative acknowledged the inspector's commen ~The inspector reviewed the licensee's maintenanc'e administrative procedure MAP-6, Work Request Planning, Revision 2, and found that specific provisions had been incorporated 'to control the conduct of troubleshooting. These provisions addressed as-found and as-left equipment status and conditions, retention of parts,for future investigations and stop work instructions for unanticipated condition This item is close (Closed), "Preventative Maintenance Program" This item encompasses two concerns which were identified by the inspector during a previous review of the licensee's preventative maintenance progra . The first involved the licensee's practice of scheduling preventative maintenance (PM) tasks to be performed prior to predictive maintenance testing. The inspector had been concerned that preconditioning the equipment by performing the PM might preclude identification o,f equipment degradation in the subsequent testing which supplied trending dat During this report period, the inspector sampled predictive maintenance procedure M.159, Vibration Monitoring, and found that the predictive maintenance was scheduled on a more frequent basis in addition to coinciding with surveillance testing. In discussions with licensee maintenance planners, the inspector determined that the licensee's use of predictive maintenance techniques for diagnostic, trending and monitoring of equipment performance appeared to adequately encompass the range of in-service conditions which would be expected between preventative maintenanc , The second issue involved the lack of a post-maintenance testing procedure for use by maintenance planning. During this report period, the inspector reviewed the licensee maintenance administrative procedure MAP-6, Revision 2 and found it to include guidelines for maintenance verification testing for use by the maintenance planners in specifying appropriate post-maintenance :

testing. The procedure also incorporated a verification test summary form to document the specified testing. The inspector found the licensee's response to both of these previously identified r concerns to be adequate. This item is closed.

l l 87-03-01 (Closed), "Quality Assurance Surveillance Program for ISI" During Inspection 87-03, the inspectors identified several deficiencies i in the licensee's oversight of contracted ISI activities. Specifically, l

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QA surveillance did not effectively assess contractor ISI activities and QA personnel did not appear to have appropriate technical skills to perfonn surveillance of ISI activitie The inspector reviewed the following documentation related to this open item:

Quality Assurance Procedure QAP-9, "Special Processes."

Quality Assurance Implementing Procedure QA1P-1804, "Quality Assurance Surveillance Program," effective date December 4,198 *

Licensee response to open item 50-312/87-03-0 The inspector also interviewed the manager of QC surveillance, and the Visual Level III certified QC inspector assigned to perform ISI surveillances. The inspector determined that Procedure QA1P-1804 requires assigned surveillance personnel to be qualified in the area they are auditing. He also determined that the QC inspector assigned to perform ISI surveillances is adequately qualified to perform this function. The inspector considers this item close (Closed), "Ultrasonic (UT) Examination Results for the Pressurizer Support Lugs" The Babcock and Wilcox evaluation of the ultrasonic test data for the indications at the pressurizer seismic lugs indicated that any differences between the dimensions reported in 1985 and in 1983 were within the accuracy of the ultrasonic testing technique. The licensee's contractor evaluated the indications in accordance with the ASME Boiler and Pressure Vessel Code,Section XI, 1977 Edition through Summer 1978 Addenda including linear elastic fracture mechanic analyses where necessary. The calculation included a fatigue crack growth analysis which indicated that crack growth for the largest individual indication was insignificant. The licensee's contractor also performed a stress analysis assuming a net section for the lugs reduced by an area inches wide over the length of each of the lug members. The resulting stresses were within ASME Code specified stress allowable The pressurizer lugs are fabricated from 3.5 inch thick rolled plate sections with full penetration welds. The licensee's consultant stated '

in their October 20, 1987 evaluation report that the indications are most likely slag, porosity, lack of fusion and lamination that have existed since original fabrication. At the time of fabrication, the seismic lug to vessel weld received only a surface (magnetic particle) examinatio The indications were identified in subsequent volumetric (ultrasonic)

examinations performed as part of the ISI Progra The licensee is currently changing their ISI Program to meet the requirements of the 1986 edition of the ASME Code. This edition of the code requires only surface exams of the subject area. The inspector discussed this situation with the licensee and was informed that the

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licensee will continue to perform ultrasonic examination of these indications for three scheduled inspections to ensure no service induced-degradation is occurrin The inspector reviewed the following documents related to this open item:

Ultrasonic data reports for support lugs85-006, 86-007,85-010 and 85-01 *

Drawings 25478E Revision 7 and 135484 Revision 1 Fracture mechanics report to NRC by Rancho Seco ISI report RSR 84-03 dated January 25, 1984 and JEW 86-391 dated September 16, 198 Babcock and Wilcox Company letter dated October 20, 198 Based on this review the inspector considers this item close (Closed), "Licensee Not Reviewing Ultrasonic Data Submitted from ISI Contractor" The ultrasonic data submitted by the ISI contractor was not complete in that it was impossible to determine what volume of the weld had been examined. The inspector's review of the ultrasonic documentation revealed no weld profiles for which evaluation of weld root and counterbore conditions were considered necessar The inspector reviewed the following documentation related to this open item:

Procedure NDEl-0901, dated February 3, 198 Procedure QA1P-1804, dated December 4, 198 The inspector determined that the NDE Instruction Manual (Procedure NDEl-0901) Section 5.4.3 requires a plant NDE Level III certified individual to perform surveillances of in-process ISI activities and to review ISI reports to assure adherence to applicable procedures and appropriate documentation. The inspector also determined that the J reorganized plant performance department includes a welding /NDE Engineering Section that has an assigned qualified NDE Level III individua .

Based on this review the inspector considers this item close (Closed), "Main Steam Piping Drawing Weld Identifier Discrepancy" The inspector reviewed the following documents related to this violation:

The licensee's response to the Notice of Violation.

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Welding manual M.308 section 6.1 e y'*- "

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Nuclear Engineering Procedure NEPM 5207.3, Section Nuclear Engineering Procedure NEPM 5207.16, Section 5.2, for checking all piping drawing *

NCR No. S-633 *

THe inspector confimed through his review that the subject drawing had been revised and associated documentation corrected. This is considered to be an isolated cas Based on this review the inspector considers this item close Temporary Instructions (Closed) Temporary Instruction (TI) 2515/71 "Inspection of Licensee's Actions in Response to IE Bulletin (IEB) 82-02" ,

IEB 82-02, Degradation of Threaded Fasteners in the Reactor Coolant '

Pressure Boundary (RCPB) of PWR Plants, required licensees to perform certain actions and submit a report to the NRC regarding P.CPB fasteners and thread lubricants. The licensee's response and IEB 82-02 were previously inspected and closed in Inspection Report 50-312/84-0 Additional NRC inspection instructions were provided by TI 2515/71 to perform follow-up inspections on specific bulletin related subjects for specific plants. The TI 2515/71 inspections were previously performed and documented in Inspection Report 50-312/87-0 The TI remained open pending a determination of the material of the fasteners for the valves listed below, a detemination of the lubricants last used on those fasteners, and performance of any corrective actions that were require HV-20001 RCS-001 HV-20002 RCS-002 DHS-015 CFS-001 DHS-016 CFS-002 Subsequent to Inspection 87-08, the licensee issued Rancho Seco Administrative Procedure (RSAP) 0221, Chemical Control Program, effective January 29, 1988. The procedure required chemicals that were in direct contact with plant gsystems and that were to remain in contact after the system exceeds 200 F to be designated Category I chemicals. Category I chemicals were required to contain only a maximum of 200 ppm each of chlorides, fluorides, and sulfur, among other specified chemical compositio The concern involved decomposition of any Molykote-G lubricant that may have been used on the fasteners, and its interaction with high-strength low-alloy steels, and austenitic and martensitic stainless steel During this inspection, the licensee determined that valves HV-20001, HV-20002, DHS-015, and DHS-016 had carbon steel fasteners and were therefore not a concern for possible sulfide stress corrosion cracking

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induced by Molykote-G. Also, during the inspection, the licensee determined that valves RCS-001, RCS-002, CFS-001, and CFS-002 were L

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allowed by the vendor drawing to have stainless steel fasteners. CFS-001 and CFS-002 were confirmed by the licensee to have stainless steel fasteners. The licensee was unable to confirm the fastener material for RCS-001 and RCS-002 during the inspection. Furthermore, the licensee was unable to determine from maintenance records the lubricant last used on the fasteners for RCS-001, RCS-002, CFS-001, and CFS-00 IEB 82-02 stated that molybdenum disulfide (MoS) contains a significant level of sulfide content. The Bulletin further stated that experience suggested that MoS_ has a pronounced tendency to decompose in the presence of i:'3 1 temperature and prolonged exposure to moisture conditions to rrieasa sulfides, which are known promoters of stress corrosion crackin The licensee's evaluation of the noted condition for valves RCS-001, RCS-002, CFS-001, and CFS-002 indicated that there was no evidence of leakage that would expose the valves fasteners to moisture even if Molykote-G nad been used on the fasteners. In addition, the licensee stated that the valves and fasteners were covered with insulation. The licensee considered that if a concern existed, it would be a long-term concern related to breakdown of sulfides, if Molykote-G had been used due to the lack of indication for a moisture concern. The licensee committed to perform the following actions before the end of the next refueling outage for valves RCS-001, RCS-002, CFS-001, and CFS-002: Disauemble and inspect all pressure boundary fasteners of the subject valve Perform appropriate corrective actions for any discrepancies noted by the inspectio Clean and reinstall fasteners using RSAP 0221 Category I lubrican Enter the above commitments in the computerized and QA audited coninitment tracking ' syste The licensee evaluation appeared to be reasonable and the commitments appeared to provide reasonable assurance that potential problems related to fastener degradation for valves RCS-001, RCS-002, CFS-001, and CFS-002 would be orevented. On the basis of the licensee's commitments, the TI was close No violations or deviations were identifie Part 21 Reports Master Tracking Item 85-13-P g Master Tracking Item (MTI) 85-13-P was comprised of the following Part 21 reports dealing with the TOI diesel generators: 85-04-P, 85-13-P, 85-21-P, 85-22-P, 86-01-P, 86-04-P, 86-08-P, 86-10 P and 86-16- P was closed in Inspection Report 50-312/86-34. All of the

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remaining issues are addressed in the following subparagraphs. This MTI

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a. 85-13-P (Closed)

The Part 21 report concerned the potential to overheat some components in the TDI aiesel control cabinets. This was documented by a letter dated May 17, 1985 from Transamerica Delaval to the NR The licensee made the panel modifications suggested by the letter, which included installing temperature alarm and fans in the control cabinets. The effectiveness of the modification was rated during the diesel 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> endurance operations. The temperature remained well within the requirements. This item is close b. 85-21-P (0 pen)

The Part 21 report concerned potential problems with exhaust valve springs for the TDI diesel which could result in engine non-availability. The licensee inspected the TDI engine exhaust springs and found no failed springs. Due.to the nature of the failures which appeared to occur after extensive operations (5000 to 7000 hours0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br /> of operation), the licensee has committed 'to replace

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these springs prior to the completion of the first refueling outage after startup. This commitment is listed in and being tracked by the licensee's coordinated commitment tracking system. This item will remain open pending installation of these spring c. 85-22-P (Closed)

This Part 21 report listed a problem at a facility in which the intake air cylinder end caps had not been welded. The report requested that licensees inspect their installations to ensure that the welds were in place. The inspection at Rancho Seco found that the welds were properly installed. This item is closed, d. 86-01-P (Closed)

This Part 21 report concerned a problem in which the starting air valve assembly capscrews, which hold the assembly to the engine cylinder, were too long. If the capscrew was too long, it would

"bottom out" on installation resulting in improper torquing into place. The licensee inspected their capscrews and verified that they were the proper length. This item is close e. 86-04-P(Closed)

This Part 21 report concerned a fuel line which broke due to excessive vibration and shut down the engine. The licensee responded to the concerns by performing the corrective actions listed in the repor They installed fuel line guards, spray shields and a shrouded (dual-wall) fuel line. This item is close f. 86-08-P (Closed)

This Part 21 report concerned a potential problem with inadequate turbocharger thrust bearing lubrication. The licensee performed modifications as approved by the manufacturer. The lubrication appeared adequate during the diesel testing program. This item is closed.

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23 P (Closed)

This Part 21 report concerned a potential problem with lube oil check valves. The valves had a liner seat material made of ethylene propylene rubber (EPR). EPR is not suitable for use with petroleum base material Rancho Seco noted that the seat material was deteriorating. The licensee replaced the seat material with a material suitable for this use. This item is closed.

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This Part 21 report concerned a potential defect with the lube oil sump tank foot valv This item was not applicable to Rancho Sec This item is close Licensee Event Reports LER 86-17-L0 (Closed) "DG Loading During LBLOCA did-not Account for AFW Pump Runout" This LER reported a design deficiency in which the licensee discovered that an emergency diesel generator (DG) could be overloaded. In 1983 the

"A" AFW pump autostart was added to the emergency electrical bus. The design did not account for the full run out loading of 2955 Kw and a continuous rating on the "A" Bruce-GM DG of 2750 Kw. The licensee, prior to plant restart, shifted some of the emergency loads required to the output of the TDI diesels, which are now operable. This resolves the potential to overload the DGs. The licensee is also retraining the senior design engineers to prevent a recurrance of this type proble This item is close LER87-22-L0(Closed)"DieselGeneratorCoolantLeak" This LER. involved the discovery of excessive seal leakage from the jacket coolant water pump during surveillance testing of the "A" Bruce GM emergency diesel generator resulting in the diesel generator being declared inoperable. At the time, the "B" emergency diesel generator was also out of service. This resulted in a condition in which both loops of decay heat removal were considered to be inoperable due to the loss of ooth trains of emergency power. The licensee declared an Unusual Event per the Emergency Plan and made appropriate notifications. Following replacement of the jacket coolant water pump and satisfactory performance of the surveillance test, the licensee secured from the Unusual Even The inspector had initially followed up on this event as reported in Inspection Report 87-24 and had identified a concern regarding the qualification of the spare pump used to replace the leaking pump.

l Followup of the licensee's actions to address weaknesses in their program l

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for dedication of commercial grade equipment was identified as Open Item 87-24-0 No other concerns were identified by the inspector in his review of this

even This LER is close .

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87-24-01 (Closed) "Procurement of Diesel Generator Coolant Pump" This open item deals with a concern regarding the qualification of a spare jacket cooling water pump which the licensee had installed on the

"B" Bruce-GM emergency diesel generator. The inspector had found the licensee's program to be inadequate for controlling the dedication of commercial grade components for service in safety related application This concern was addressed in a subsequent special Procurement and Warehousing inspection by the NRC Vendor Inspection Branch in January, 198 Based on the followup through the special inspection, this item is considered to be close No violations or deviations were identifie . Followup on Employee Concerns and Allegations Allegation RV-88-A-0002 Characterization A quality control inspector's signature was falsified on an NCR (nonconformance report) #S-753 . NCR #7330 was improperly revise . NCR #7330 was improperly voide . The engineering disposition of the NCR was inadequat . A potential "cover-up" may have existed of inadequate previous QC inspections on the high point vent pipin Implied Significance to Design, Construction or Operation Potential administrative and engineering disposition problems with the NCR program may impair the effectiveness of the quality assurance program to identify and cause corrective action on conditions adverse to safety or qualit Assessment of Safety Significance NCR S-7539 contains a "Prepared By" block on the NCR form which contains a quality control inspector's name written in scrip The individual whose name appears in the prepared by block

, stated that he did not sign his name in that bloc The concerned individual had also 9 % n bis same concerns to the licensee for followup. The lic mseeh investigation had identified a person involved with th: ln ;essing of NCRs who admitted to writing, in script form, other individual's name The licensee's investigation concluded that the person writing other persons names in script form was not aware of a potential problem with signing for other individuals on the NCR for +* *

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In the licensee investigator's opinion, there was no intent by the individual to forge signature .' In October 1987, NCR #7330 was written to document potential quality problems with the high point vent piping. The NCR identified potential problems involving welding deficiencies such as arc strikes, slag, under-cut and pipe support configuration problems such as structural member sizes being incorrect, clearances between members and piping being incorrec This NCR number was deleted from the NCR log book by

"whiting-out" the entry in the log book. Quality supervision had intended to separate the items contained in the initial NCR into another NCR (#7530) and work requests and engineering action request items. Iicensee personnel stated that their intent was to separate out items from the original NCR which were not required to be documented on ar. NC The licensee had stated that all items which were originally ,

identified on NCR #7330 were reinspected by Quality Control i with all deficiencies being documented on an NCR if the deficiency met the NCR procedure criteria. The inspector sampled approximately 25% of the items originally documented on NCR #7330 and found the items to have been addressed on NCR #7530 per procedure. NCR #7680 had been issued on 1/20/88 to document the technical issues concerning the high point vent pipin . The licensee had been made aware of the voiding of the NCR ;

(7330) by the concerned individual and had investigated the concern. Licensee investigation had documented the need to I correct their process to revise quality department document As part of their corrective action to previously identified problems with the NCR procedure (QAP 17) contained in '

inspection report (50-312/87-44) the licensee implemented a new ,

problem identification and corrective action system called i Potential Deviation From Quality (PDQ). The PDQ described in procedure RSAP-1308 would appear to prevent reoccurence of this !

problem by controlling signatures on the original PDQ for i The concern involves the reactor coolant system high point vent system pipe support design. Specifically, the concern is that the current pipe support design did not provide sufficient flexibility to accomodate thermal movement of the piping syste Licensee management in the quality area were made aware of the concern by the alleger and had investigated the support configuration. Licensee engineering had walked down the high point vent system for compliance with the as-built drawings and had also reviewed the stress analysis for the high point vent syste Their conclusion was that the design was acceptable

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and piping stress levels were within code allowables. The 1 inspector had reviewed the analysis, and had not identified any violation or deviations from requirement . Licensee quality management had conducted an investigation of a potential cover-up of prior Quality Control inspections and documentation. The investigation was started after the licensee was made aware of the concern by the allege The licensee's investigation had concluded that there was not a cover-up as such. However possible contributing effects were identified that may have caused the individual involved to believe a cover-up had taken plac The licensee's investigation found that an apparent lack of sensitivity by some Quality Control supervisors to quality control inspectors' concerns and finP'igs existed. Also the originator had not been kept informed of NCR status during the review cycle. Little communicaticn existed between the individuals who were reinspecting the high point vent and the originator of the NCR. Methods to be used by the quality -

control office to transfer information between forms and the

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NCR logbook changing of NCR numbers was not clear and did not have precise instructions within the previous procedure for handling NCRs (QAP-17).

The inspector spoke with three quality control inspectors, four welding craft personnel, two quality control supervisors and the quality director to ascertain if there was a reluctance by quality people in the field to identify conditions adverse to quality. All licensee personnel interviewed by the inspector indicated they had no reluctance to either identify or cause to be generated a PDQ to document deficient condition Conclusion

. The allegation was partially substantiated in that improper steps were taken by quality personnel to revise and void (by whitir1-out an entry in the NCR log) an NC The issues were known by the licensee prior to the inspector's involvement during the inspection and appear to have been satisfactorily investigated by the license The corrective actions taken by the licensee appear to be adequate to resolve the concern. The allegation is considered close . Facility Events (93702) On February 7, 1988, the licensee inadvertantly drained about 1100 gallons of slightly contaminated borated water from the Borated Water Storage Tank (BWST) through the two lower rings of the containment spray header inside the reactor building. The incident occurred when a safety features actuation signal opened the reactor building spray system isolation valves as a planned part of the loss of Offsite Power (LOOP) test. As planned, the spray pumps did not i

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start (their breakers were lined up in the test position). However, contrary to the test plan, a manual valve (CBS-009) in one train of the spray system was open when it should have been shut for the tes This allowed borated water to drain from the BWST (+47 f elevation) to the two lower spray rings (-2 ft. and +38 ft, elevation). Operators detected this condition after about thirty minutes, when restoring the normal valve lineup for containment spra The licensee's investigation showed that positive control of these valves had been lost. As a result of this incident, the licensee suspended testing, verified plant status and had in place a positive means to control plant status prior to resumption of testin . On February 8, 1988 the licensee responded to a fire in the "B" side annunciator panel. The panel is located in the west 480 volt electrical equipment room. As a result of the panel fire, the licensee deenergized both "A" and "B" annunciator panels resulting in a loss of all control room annunciator window The licensee response to the fire resulted in fire damage being limited to the one panel (the annunciator system is contained in four cabinets all side by side) with no damage discovered to other plant equioment or electrical parels. Fire brigade response was observed by the inspector and appetred to be timely and well controlled. Fire personnel were dispatched to 480 volt room quickly and necessary equipment including self contained breathing apparatus was available. Security had established boundaries to control

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personnel access to the area in a timely fashio Operator response to the event appeared to be controlled and orderly. Discovery of the fire was accomplished by operations personnel before smoke detection devices had alarmed. Operations personnel had begun an investigation of the annunciator panel when a number of control room windows appeared to indicate in an unstable fashion. During followup investigation by the licensee, operations personnel discovered smoke beginning to exit the annunciator panel and also smoke detecting devices began to alarm. Local fire suppression efforts by licensee personnel were successful in containing the fire.

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Trouble-shooting by the licensee was performed in an orderly manner and carefully established which annunciator panels were effected by -

the fir Plant conditions were maintained without inciden The licensee's investigation revealed that the apparent cause of the fire was overheating of a resistor on the flasher circuit within one of the panels. Possible contributors to the fire, as found by the

. licensee were: the circuit boards and holders were flammable, the boards were operated at excessive voltages (125 VDC), components showed indication of operating at maximum ratings, and heat emitting I components were mounted in contact with flammable circuit board ,

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,' 28 Corrective action was undertaken to restore the annunciator to service, with all windows in service at the end of this report perio Reinspection and rework was completed of all circuit cards in all annunciator panels. with substitution of components on the caras made to less flannable electrical _ components. The licensee has undertaken a long term engineering study to consider potential replacement of the panels with a different desig Licensee action to deal with the annunciator fire and restoration appeared to be controlled and_ completed in a tinely fashio No violations or deviations were identifie . Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations or deviation . Exit Meeting (30702,30703)

The inspector met with licensee representatives (noted in Paragraph 1) at various times during the report period and formally on March 29, 198 The scope and findings of the inspection activities described in this report were summarized at the meeting. Licensee representatives

! acknowledged the inspector's findings and the violation identified.

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