IR 05000312/1988016

Insp Rept 50-312/88-16 on 880601-0708.No Violations or Deviations Noted.Major Areas Inspected:Operational Safety Verification,Health Physics & Security Observations,Esf Sys Walkdown,Maint & Surveillance & Testing

ML20207D734
Person / Time
Site:	Rancho Seco
Issue date:	07/27/1988
From:	Dangelo A, Miller L, Myers C, Qualls P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20207D732	List: IR 05000312/1988016 ML20207D727
References
50-312-88-16, NUDOCS 8808150450
Download: ML20207D734 (17)

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U. S.- NUCLEAR REGULATORY COMISSION

REGION V

Report No: 50-312/88-16 Docket N License No. DPR-54 Licensee: Sacramento Mur.icipal Utility District P. O. Box 15830 Sacramento, California 95813 Facility Name: Rancho Seco Unit 1 Inspection at: Herald, California (Rancho Seco Site)

Inspection conducted: June 1 through July 8, 1988 Inspectors: LO M M t/ MT * II A. J. D' Angelo, Sen'ior Resident Inspector Date Signed

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C. J. Myers, R4sident Inspector

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Ni th Date Signed W h W 7474%

P. M. Qualls, R'esident Inspector Date Signed Approved By: 'l for M1*I8I L. F. Miller, Chief, Reactor ProjectsSection II Date Signed Summary:

Inspection between June 1 and July 8,1988 (Report 50-312/88-16)

Areas Inspected: This routine inspection by the Resident Inspectors involved the areas of operational safety verification, health physics and security observations, engineered safety features system walkdown, maintenance, surveillance and testing, review of core power distribution limits and reactor shutdown margin determinations, review of nuclear instrumentation calibrations, and follow-up item During this inspection, Inspection Procedures 71707, 71709, 71881, 71710, 61726, 62703, 93702, 61702, 61705, 61707, 72700, 92702, 92701, 92700 and 30703 were use Results: A significant strength observed was the operators response to the main feedwater block valve closure. A weakness was observed in that the manner in which troubleshooting maintenance was to be performed was not documented in the preplanned section of the work reques No violations or deviations were identifie PDR ADOCK 05000312 O PNU

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DETAILS Persons Contacted Licensee Personnel J. F. Firlit, Chief Executive Officer, Nuclear D. Keuter, Acting Assistant General Manager (AGM),

Nuclear Power Production

. B. Croley, AGM, Technical and Administrative Services

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J. Vinquist Director, Nuclear Quality

• D. Brock, Acting Nuclear Maintenance Manager G. Cranston, Nuclear Engineering Manager '

• W. Kemper Nuclear Operations Manager

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• J. Shetler, Director, System Review and Test Program T. Tucker, Nuclear Operations Superintendent P. Kagel, Nuclear Mechanical Maintenance Superintendent

. L. Fossom, Manager, Scheduling and Outage Management J. Field, Plant Support Group Supervisor L. Conklin, Manager of Management Controls

• S. Crunk, Manager, Nuclear Licensing T. Fetterman, Electrical Engineering Manager J. Irwin, Supervisor, I&C Maintenance

• Coleman, Quality Engineering Supervisor T. Shewski, Quality Engineer V. Lewis, Supervising Civil Engineer

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J. Robertson, Licensing Engineer P. Bosakowski, Supervisor of Licensing, Technical Support J. Delezinsky, Supervisor, NRR Coordination, Licensing

• G. Legner, Licensing Engineer S. Rutter, Supervisor, Incident Investigation Review Group (IIRG), Licensing

• P. Johnson, Manager, Engineering Support D. Schuman, Incident Analyst, Licensing /IIRG L. Houghtby, Nuclear Security Manager Other licensee employees contacted included technicians, operators, i mechanics, security and office personne * Attended the Exit Meeting on July 8, 1988.

. Operational Safety Verification (71707, 71709, 71881)

During this inspection period, Rancho Seco was critical and was conducting power ascension program testing at various power levels between 24 and to 70 percent reactor power. The inspectors reviewed control room operations which included access control, staffing, observation of decay heat removal system alignment for low pressure injection, and review of control room log Discussions with the shift -

supervisors and operators indicated understanding by these personnel of the reasons for annunciator indications, abnormal plant conditions and

maintenance work in progress. The inspectors also verified, by 3

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observation of valve and switch position indications, that emergency systems were properly aligned as required by technical specifications for plant condition 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. The inspectors also observed plant housekeeping and cleanliness looked for potential fire and safety hazards, and observed security and safeguards practice Through discussions with various licensee _ personnel, the inspectors verified that the employees are aware of and when required involved in the ALARA program Radiation Work Permit (RWP) review revealed that the RWP did include:

job description, radiation levels, contamination, airborne radioactivity (if expected), respiratory equipment, protective clothing, dosimetry, special equipment, expiration, health physics (HP) coverage and signatures. The RWP radiation and contamination surveys were kept curren Employees understood the RWP requirement The inspectors observed personnel in controlled areas were wearing the ;

proper dosimetry, and personnel exiting the controlled areas were using the monitors properly. Labeling of observed containers appeared

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appropriat The inspectors walked down portions of the protected and vital area boundaries to ensure that they were complete and that security personnel were properly posted where known deficiencies existe The inspectors also cbserved protected area access control, personnel screening, badge issuing and maintenance on access control equipmen '

Access control was observed. Personnel entering with packages were properly searched and access control was in accordance with licensee procedures. Personnel allowed access to the site are, by procedure, '

badged to indicate whether they have unescorted or escorted access authorizatio The inspectors observed no obstructions in the isolation zone which could conceal a person or interfere with the detection / assessment syste Protected area illumination appeared adequat :

On June 22, 1988 the licensee placed the B2 Transamerica DeLaval In (TDI) Emergency Diesel Generator (EDG) out of service for scheduled maintenanc Technical Specification (TS) 3.7.2 permits this generator to be out of service for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with no impact on plant operations. On

June 23, 1988, the licensee placed the 82 EDG Control Room Essential Heating, Ventilation and Air Conditioning (HVAC) out of servic ,

TS ?.28 permits this unit to be out of service for 3 days without l impacting plant operation !

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The inspector questioned licensee operators about whether the essential HVAC unit was a required auxiliary for the TDI EDG even though it has a separate technical specificatio The EDG essential HVAC is located in the diesel generator room and performs the function of maintaining the diesel control room temperature less than 105 F, which is the maximum allowable equipment temperature. Under normal plant conditions, room cooling is accomplished by the normal HVAC system which is not safety relate Assuming a loss of offsite power, concurrent with the hottest day air temperature, cooling for the local TDI control rooms is designed to be provided by the EDG essential HVA The inspectors observed that the licensee's equipment clearance personnel did not appear to recognize that the essential HVAC was a required auxiliary to the diesel generator and had to be returned to service with the ED Licensee engineering and management personnel concurred that the EDG essential HVAC was a required auxiliary system to the TDI diesel generators, and that the dit.sel could not be declared operable until the essential HVAC was returned to service. At the management exit meeting, a licensee representative stated that the operations department had issued a memorandum to clarify diesel system interrelationships and would also review the emergency safety features listing of equipment for enhancement and inclusion into the USAR to clarify further what subsystems were to be required auxiliary systems. The inspector concluded that this item had been adequately resolve No violations or deviations were identifie . ESF System Walkdown (71710)

During the inspection period the inspectors walked-down the Emergency Diesel Generators (EDG) portion of the vital electrical system and the accessible areas of the Auxiliary Feedwater (AFW) Syste l The inspectors concluded that:

U All observed hangers and supports were properly made up and aligne Housekeeping was adequat No excessive packing leakage was observed on valve No prohibited ignition sources or flammable materials were observe Major system components were properly labelled, lubricated and cooled. No excessive leakage was apparen The instrumentation appeared to be properly installe Out of calibration gauges were identified with an alternate means to determine the parameter needed in one cas Flow path components appeared to be in the correct positio .

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Required support systems were availabl Proper breaker and switch positions were verifie No violations or deviations were observe . Monthly Surveillance Observation / System Review and Test Program (SRTP)

(61726) (72700)

Technical Specification (TS) required surveillance tests were observed ,

and reviewed to ascertain that they were conducted in accordance with these requirement The following surveillance activities were observed:

STP.660, ICS Tuning SP.498, Monthly EFIC Logic SP.20, AFW Monthly Surveillance The following items were considered during this review: Testing was in accordance with adequate procedures; test instrumentation was calibrated; limiting conditions for operation were met; removal and restoration of the affected components were accomplished; test results confirmed with TS and procedure requirements and were reviewed by personnel other than the individual directing the test; the reactor operator, technician or engineer performing the test recorded the data and the data were in agreement with observations made by the inspector, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne No violations or deviations were identifie S. 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 Maintenance activities observed included troubleshooting of main feedwater pump speed oscillations, review of the main feedwater block valve closure incident, auxiliary feedwater (AFW) puv packing adjustments, cleaning of the fuel oil injectors nn the Bruce-GM diesel generators, and a review of electrical deficiencie No significant observations were made concerning the main feedwater pump speed

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I oscillation troubleshooting, the AFW pump packing adjustments, nor the i the diesel generator maintenance, a, Main Feedwater Block Valve Closure Incident

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In order to determine potential causes for the main feedwater block j valve closure (Paragraph 6), the licensee initiated Work Request - #01486100 to gather data on the main feedwater pump discharge pressure switch response as a function of pump speed. All equipment involved with this work was non-safety related. However, the work planning process at Rancho Seco for non-safety related and safety related work is simila , The work request used stated that maintenance was to provide general support to engineering and install test equipment as required. The

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work instruction section of the work request provided direction to the maintenance personnel to connect and disconnect test equipment at the direction of the engineer and to record lifted lead No specific instructions were documented on the preprinted work request regarding the attachment of a recorder, position f auxiliary relays, removal of fuses or adjustment of main 'dedwater ,

pump speed. However, these steps were used to gather data for the  :

trouble shooting effort. The steps hsd been written within a note i book and discussed with maintenance, engineering and operations personnel prior to the conduct of the activit l The inspector observed that by not inserting these steps into the work request, no formal review of these occurred, and a potential existed that an improper or incorrect step may have been performe However, in this case, all steps given at the timo were correct and appropriate for the activity performe A licensee representative acknowledged the inspector's concern that the work request instructions were incomplete for this activit A licensee maintenance department representative stated that their expectation was that all work requests, except those needed in the event of a shift supervisor declared emergency, should be preplanned to the extent that all steps to be performed were documented and reviewed prior to conduct of the maintenance activit No explanation for this occurrence was offered other than this was belteved to have been an isolated breakdown of the work control The Quality Assurance department had also reviewed the work request

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prior to conduct of the maintenance and concurred with the documentation on the work request. Quality Assurance management stated that the level of preplanned documentation on troubleshooting work requests should contain individual steps describing attaching

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of instrumentation, lifting of fuses and maneuvering of plant equipment to gather data. No explanation for this occurrence was -

offered other than that this was believed to have been an isolated i breakdown, i

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t At the time of the exit interview, the licensee was reviewing the *

manner in which troubleshooting work requests were planned to determine whether additional written guidance on work request planning was necessary. The licensee had also =iniciated a quality assurance surveillance of troubleshooting work request plannin . The inspector stated at the exit interview that the maintenar. e planning activities observed appeared to indicate.a programastic weakness in troubleshooting work request planning, including qeality

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assurance surveillances of work request plannin '

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The inspector stated that future work request planning would be monitored to assess the effectiveness of the licensee's evaluatiuns and corrective actions'in this are No violations or deviations were identifie Electrical Wirino Deficiencies On April 11, 1988 licensee personnel initiated Potential Deviation from Quality (PDQ) 88-0102 concerning Turbine Bypass Valve Limit Switch problems. The wiring deficiency occurred due to a  :

discrepancy in the drawing. The licensee corrected the wiring l deficiency and drawing discrepancy and closed the PDQ on

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April 12, 198 ,

Out of a sample of ten PDQs reviewed, the inspector noticed seueral  !

other electrical wiring / wiring diagram problems of a minor natur Examples of the wiring deficiencies were:

Auxiliary building ventilation for running indicator light r l failed to light on a motor start,

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D. C. emergency lights in the auxiliary building did not .ava a f relay removed as required by Enginering Change Notice (ECN)

R-003 i i The number of wiring / wiring diagram problems seemed '

disproportionately high to the inspector and he asked the licensee quality organization if they had any significant findings in this are The licensee Quality Assurance Engineering had already identified that further &ctivities on the part of the licensee were required and an action plan was being developed and approved. The inspector's concern was that the performance level of electrical wiring tasks within the powar block may have degraded, and some i wiring errors were not detected during post maintenance /modificatien testing but returned to service with the error. The inspector reviewed the licensee's draft action plan and concluded that it

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appeared adequat flo violations or deviations were identified.

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. Onsite Followup of Events (93702) (92700)

On June 29, 1988, with the reactor at 45% power, the plant received a reduction in main feedwater (MFW) flow due to the spurious closure of both main feedwater b'ock valves. The operators received the "ICS Runback or Limit" annunciator and recognized that the problem was a reduction in the feedwater flow rate +o the steam generators. The ICS responded to rapidly reduce reactor power. Then, the operators took manual control of pressurizer heaters and spray to stabilize reactor pressure during the transien The operators determined that the block valves had closed and there was no apparent cause for the block valve closure. The operators took manual control of the MFW pumps and MFW valves and returned the feedwater system to operatio The cause of the block valve closure was determined to be a failed pressure switch'in the discharge of the "A" inain feedwater pump combined with the "B" MFW pump operating in minimum spee The "B" pump pressure switch was found to be set at 800 ps A review of the status of the "B" MFW pump at minimum speed indicated that the "B" discharge pressure switch could actuate at this spee A combination of the failed switch on "A" pump discharge and the actuation of the "B" switch would, by design, result in the Integrated Control (ICS) system shutting the block valve Licensee corrective actions included replacing both pressure switches with new components and decreasing the setpoint to 700 psi. Long term actions consist of removing this block valve closure from the ICS logic ,

due to the EFIC system installatio ~

The irspectors concluded that the operator response to this incident was time / and appropriat No violations or deviations were observed.

, Review of Emergency C?assification of Rec cor Coolant Leak l Incidents (71707).

l l On March 22, 1988 the licensee had an occurrence .1 wh:-h the "A" letdown i cooler relief valve lifted and stuck ope The leak was .' excess of 50 gpm, and was promptly isolated within approximately 5 minu'es. The licensee Emergency Plan Implementing Procedure (EPIP) 5001, "Lo of .

Primary Ccolant", TAB 11, defined conditions which require emergency action level classification.

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In NRC inspection report 50-312/88-05, paragraph 6.e, this event had been discusse An issue had been subsequently raised on the procedure l criteria used in TAB 11 to classify the event.

l l At the time of the event, TAB 11 specified that an Unusual Event should l be declared with an unidentified Reactor Coolant System (RCS) leak greater than 1 gp Also, under TAB 11, an Alert was to be declared when the difference between makeup plus seal-injection flow and letdown flow l exceeded 50 gpm. These two criteria had caused some confusion the day of

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the event since the leak was determined to be approximately 100 gp The leak was isolated from the RCS within several minutes of occurrenc To rcsolve and clarify the criteria, the licensee issued procedure interim change notice (PICN) Number 04 which provided additional guidanc Specifically, the change to TAB 11 stated that RCS leaks greater than 50 gpm which could not be isolated or reduced to less than 50 gpm through the performance of procedure C.3, would be declared to be an Aler Af ter further discussions, the licensee further clarified procedure C.3 to require that if the leak could not be isolated within 15 minutes of its discovery, declaration of an Alert shoult be considere The inspectors and Region V concluded that the licensee's actions appear to have adequately resolved this issu No violations or deviations were identifie . Review of Power Distribution Limits (61702)

The inspector reviewed the power distribution limits for the Cycle 7 reload. The inspector verified that surveillance procedure SP.1, "Shift Surveillance and Instrument Checks" contained the correct core irabalance versus power level chart for fue! cycle Core power imbalance calculations using incore detectors were used and appeared to be correc Quadrant power tilt surveillances sere performed correctly per Technical Specification 3.5. The inspector verified that:

the incore detector calibration was being followed,

axial power imbalance was being maintained,

• the hot channel factors were within licensed limits, in the observed sample, there were no apparent anomalies, the quadrant power tilt limits were being observed and there were no anomolies, the plant computer retrieves this data as the normal means of acquisition by plant personnel, and the plant Nuclear Engineering staff had acequate knowledge of the I

computer code and its limitations.

l No violations or deviations were identifia . Calibration of Nuclear Instrumentation Systems (61705)

The inspector reviewed completed licensee nuclear instrumentation calibration procedures and verified that Surveillance Procedure SP.406 A/B/C/D "Monthly Reactor Protection System Channel Tests" were performed and appeared to meet Technical Specification 4.1. Testing methods and acceptance criteria were specified and met at the completion of the

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surveillanc T's inspector also verified that the following requirements were met:

All-precautions and prerequisites were met,

power supply voltages were within tolerance,

calibration frequency was within Technical Specification guidance,

the respective instruments had their voltage properly set,

calibration instruments were recorded and traceable, licensee procedures ensured that setpoints were within TS requirement *

the results were reviewed, approved, and documented in accordance with the proper administrative procedure No violations or deviations were identifie . Determination of Reactor Shutdown Margin (61707)

The inspector reviewed the licensee's shutdown margin procedure. The procedure appeared to be technically adequate. The inspector verified that procedure B.6, "Reactivity Balance Procedure" met the requirements

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of Technical Specifications 3. Critical boron concentration'had been determined and reviewed by an on shift senior reactor operato Provisions in the procedure provided for calculation of shutdown margin with a known inoperable control rod. Shutdown margin calculations appeared to be correct and were determined at the reactor coolant system temperature of 532 F with a 1% Ak/k shutdown boron wort The inspector also verified that:

The proper conditions prio" to shutdown were recorded,

.the calculations cf shut down margin were correctly performed and .

used the boron concentration, rod worth, temperature, xenon and poison buildup and fuel burnup, the determination met TS requirements,

boron changes were verified by chemical analysis, and

the licensee has reviewed all data supplied by the fuel vendor that is used in the shutdown margin determinatio No violations or deviations were identified 11. Follow-up of NRC Open Items IE Bulletins _(92701)

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1) IEB 85-03 (CLOSED), "Motor-0perated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings" As requested by action item e. of Bulletin 85-03,

"Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings", the licensee identified the selected safety-related valves, the valves'

maximum differential pressures, and the licensee provided its program to assure valve operability in their letters-dated May 16, September 5, and November 5, 198 Review of these responses by NRR indicated the need for additional information which was requested in Region V letter dated September 22, 198 Following NRR review of the licensee's November 2 and 12, 1987, responses to this request for additional information, NRR determined that the licensee's selection of the applicable safety-related valves to be addressed, and the valves' maximum differential pressures met the requirements of the bulletin and that the program to assure valve operability requested by action item e. of the bulletin was acceptabl Tnis documents the completion of the NRR review of the licensee's initial response to IEB 85-03, Action Ites.; e, in accordance with Section 04.01 of IE Manual Temporary instruction 2515/73. Inspection Report 87-44 previously closed TI 2515-73 and Bulletin 85-0 B. Follow-up Jf Enforcement Items (92702)

1) 88 06-01 (CLOSED), "Expired Shelf Li fe Items" This violation resulted from the issuance by the licensee of items with an expired shelf lif To prevent a recurrence the licensee issued procedure K4DI-WO36, "Shelf Life", dated February 11, 1988. Tne inspector reviewed this procedure and concluded that it should be adequate to prevent a recurrence of tais violatio This item is close ) 88-12-04 (CLOSED), "Failure to Report ESF Actuation" The licensee replied to the Notice of Violation for failure to make a required 4-hour report by letter dated June 21, 198 The licensee committed to revise AP 23.08,

"Reporting / Notification", by September 9, 1988. In addition the licensee stated that they would conduct an assessment of plant systems to specify which systems are Engineered Safety Features (ESF) system These actions appeared adequate to prevent a recurrence of this reporting proble This item is close C. Follow-up Items (92701)

1) 87-20-01 (CLOSED), "TDI Diesel Vibration Root Cause Evaluation"

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This item concerned the lack of a formal root cause investigation by the licensee-into the. vibration related problems that were being discovered during the.1987 testing of the Transamerica DeLaval (TDI) Diesel Generators. Subsequent to the discovery of these problems, the licensee conducted extensive investigation and corrective actions for each of the discovered vibration. problems. The licensee's activities are documented in a letter, dated June 22, 1988, titled, "TDI Diesel Generator Vibration Final Report". Most of the vibration problems and resolutions documented in the report were inspected by the inspectors during the applicable inspection periods. The licensee is continuing to monitor the TDI diesels for vibration problems during each period of operation. The licensee's investigations appeared adequat This item is close ) 87-24-02 (CLOSED), "Remote Shutdown Test" This item concerned the lack of an integrated test of the licensee's remote shutdown capability. As documente'J in inspection report 50-312/88-15, the licensee subsecaently satisfactorily completed testing of the plant's remote. shutdown capabilit The test procedure and goals had been reviewed by the NRR prior to performance of the tes This item is close ) 88-12-01 (CLOSED), "Use of Commercial Grade UV Devices" This item concerned the procurement and subsequent use of commercial grade undervoltage devices for reactor trip breakers. The licensee's prccurement process was inspected by the NRC in March, 1988. This inspection resulted in significant changes and enhancements to thu licensee's cocmercial grade dedication process. The licensee decided to procure only safety related undervoltage devices (UVs) in the future. Tnis item is close ) 88-12-03 (CLOSED), "Letdown System Quality Class" This open item concerned the issue of why the licensee's Letdown Syston was not a Quality Class 1 system although by the current 10 CFR 50 definitions and Appendix A requirements it appeared appropriate that it 3hould be one to the first isolation valve outside of f.he reactor buildin The licensee noted that the casign of Rancho Seco was based on the requirements of the 1963 issue of 10 CFR 50 that were in effect at the time of issue of Construction Permit CPPR-56 on October 11, 196 It was designed and constructed to meet the intent of the General Design Criteria (GDC) as originally proposed in July 1967. The plant was subsequently assessed against the revised 1971 GDCs (Appendix A to 10 CFR 50) and was found to conform to the intent of the revised GDCs as indicated in the Safety Evaluation Report (SER).

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The inspector observed that neither the 1968 10 CFR 50 nor the GDCs.(1967 and 1971 versions) defined reactor coolant pressure boundary. This was' defined in 1971. 10 CFR 50.2v and 50.55 Also, USAS B31.7, "Nuclear Power Piping,'". June 1968, which was used for the reactor coolant system did not' define the reactor coolant pressure. boundar Until 1971, the reactor coolant pressure boundary was defined by the vendors' interpretation of 10 CFR 50 and applicable industry codes and standards. For Rancho Seco, the reactor coolant pressure boundary (or Class 1 boundary) was defined as follows:

"Class 1 (USAS B31.7) will be applied to the reactor coolant system pumps and piping, and the pumps, valves, and piping connected to the reactor coolant system out to and including the first isolation valve if the valve is normally closed, otherwise out to and including the second isolation valve."

This arrangement was the original design agreed upon by the NSSS Supplier (B&W), the AE (Bechtel), SMUD, and the NR The resultant designation for the piping systems is 11N1 - Quality Class 1, Seismic Category 1, Nuclear Class 1 (USAS B31.7) for the reactor coolant pressure boundar Based on the above definition, the reactor coolant pressure boundary was considered to on'ly extend to the letdown cooler inlet isolation valves (SFV-22005 and SFV-22006), inside the reactor building. The piping downstream of the valve insid? the reactor building, was not considered subject to the requirements of 10 CFR 50, Appendix A. Therefore, the licensee concluded that classification of the letdown system in containment met or exceeded the guidance available at-the time of system design and, complied with the requirements of 10 CFR 5 The inspector agreed with the licensee's explanatio This item is close . Allegation RV-88-A-0016 closed Characterization a) The concerned individual stated that he was a pipe welder and was attempting to qualify as a welder completing a weld test specime During the performance of the first test specimen, the individual etated he had been interrupted and could not complete the test specimen. When he returned, the weld test specimen was completed plus an additional specimen had been made. The individual stated that he had not completed the first nor started on a second specime .-. .-,

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b) The individual stated that he had been hired as a welder and had been assigned to weld on U-bolt supports on the Transamerica DeLaval (TDI) diesel radiators. His concern was that another individual had ground out his welds, and then rewelded without proper work authorizatio c) The individual was concerned that his welding was being damaged after he had completed his veldin Implied Significance to Design, Construction or Operation The welding in question by the concerned individual was on onsite emergency diesel generator cooling system structural supports. A failure of the welds might affect both trains of onsite emergency power supplies and could potentially lead to a loss of AC power onsite if offsite AC power was coincidentally los Assessment of Safety Significance a) This concern addressed the possible falsification of a welders' test specimen during welder performance qualification tasting. The inspector reviewed licensee procedure M.305, "Welder and Brazer Qualifications", which specified the steps which were to be taken to qualify a welde The procedure did permit the welding inspector to terminate the testing at any stage of testing when it had become apparent to the welding inspector that the welder had not produced an acceptable weld. Documentation of the welding inspector's findings were required to be made on an inspection form. A review of the records file by the licensee did not produce the inspection for Licensee personnel stated that such forms are not kept since the individual was not accepted as a welde However, the licensee retrirved a Quality Control department inspection memo which documented the welder performance testing of <

the concerned individual and an inspection by a welding inspector who had written the inspection memo and documented the welding performance failur The cause for failure, as documented by the welding inspector, was stated as unacceptable weld undercut and profil Discussion with Quality inspection personnel and welding engineering personnel indicated that when welding performance qualification testing is in process, it is required by procedure for inspection personnel to observe welding in process until test coupons are complete Had a welder needed to leave the welding test booth for a moment, Quality inspection personnel stated they would not have permitted any other welder in the boot Welding engineering personnel stated that, to the best of their knowicdge, all welders tested had completed their test coupons within one work shift with the coupon inspected by the weld inspector during the shif The weld inspector who had inspected ,

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the concerned individual's weld coupons was no longer employed by the licensee or available-for discussion with the inspector, b) One month later the concerned individual was retested for welder

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performance qualification and was able to successfully meet welding requirements. The individual was concerned that grinding and rewelding was performed on a number of pipe supports in the TDI radiator yard on welds which he had already complete .The inspector reviewed three nonconforming reports (NCRs), numbers 7289, 7290, and 7309. The NCRs document several welding problems on the pipe supports in questio The issue involved the amount of weld undercut which had been permitted. The licensee's procedure did not permit undercut, however the applicable welding code, American Welding Society (AWS) D1.1, had permitted an acceptable level of undercu From the NCRs, it appeared that the welding for the pipe supports in question had contained undercut, as documented on the NC Procedure revisions were undertaken by the licensee to permit some undercut as allowed by AWS D1.1, and two work requests (01360230 and 01360240) were used to document welding performed on the support Licensee welding engineering personnel stated some grindins had been done to correct some weld deficiencies and documented on the above stated NCR All involved welds on the U-bolt supports for the radiator were documented as inspected, and were found acceptabl The NRC inspector, accompanied by licensee quality and welding engineering personnel, reinspected all U-bolt support weld The welds in question had been painted for some time. However, no weld deficiencies were identified by visual examination through the paint laye Some of the welds on the TDI radiator U-bolt support had been reworked. However, it appeared that required documentation had been completed with all quality control final inspections completed and found acceptabl c) This concern involved welds in the same area as in concern b) abov The inspector reexamined the welding in this area, as noted above, through a paint layer, and found it to be acceptable with no damage identified to the final wel This allegation was not substantiate Conclusion This allegation is considered closed.

I a) The inspector concluded that there was no apparent indication of weld coupon falsification in this occurrenc This allegation was not substantiated, b) This allegation was not substantiate , .

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c) This allegation was not substantiate Actions'Needed for Resolution None 1 Part 21 (92701) P (CLOSED), "GE Relays on TDI Diesels" The 10 CFR Part 21 notification indicated the General Electric (GE)

type NGA 15AG3 relays, used in the licensee's TDI diesel generator control circuitry may be damaged by a wiring error. The-licensee corrected the wiring erro Relays with burnt resistors were replaced with new ones, others were repaired, retested and reinstalled in the control panel This item is close . Licensee Event Reports (92700) LO (CLOSED), "NSEB HVAC Design Error" This LER identified a potential failure of the Nuclear Service Electrical Building (NSEB) HVAC isolation dampers to close if required due to the fact they were a "normal open, fail-open" design. If the dampers failed to isolate, it could prevent the NSEB essential HVAC system from performing its safety functio To correct this potential problem, the licensee changed the damper logic to a normal-open, fail-closed logic. This was done per ECN No. R-2401 and was completed in January 1988. This item is close L0 (CLOSED), "Missed Surveillance" This item concerned a missed weekly battery surveillance which was performed one day late. The reason for missing the weekly surveillance was due to its being combined with the monthly surveillance when the monthly was required to be performed. In this case, the monthly was delayed and thus the weekly was performed one day lat The licensee has revised their method of scheduling these surveillances and has taken the weekly surveillance from the monthly. This item is close L0 (CLOSED), "Missed Surveillance" On April 28, 1988 the licensee discovered that the requirements of Technical Specification (TS) 4.6.4.F were not being performed. This specification required weekly checking of vital electrical bus breaker lineups and voltages and referred to the list given in TS 3.7.1.G & H. TS 3.7.1.G and H were applicable only in the heatup mod A Potential Deviation from Quality (PDQ), #88-0434, was writte The PDQ initiator, reviewer and Operational Technical Advisor reviewer did not recognize that TS 3.7.2.H and 3.7.2.I, which also

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referred to the 3.7.1.G and 3.7.1.H lists, were also applicable during startup through power operations. The OTA did recogniz'e that the event was reportable since the technical specification had not been performed during the startup phase of operations. During the reportability review by the licensing staff, the technical specification requirements were recognized. In the interim, the j plant procedures group had already entered the' procedure change incorporating this technical specification into SP-3, "Weekly System Verifications." SP-3 was performed with this required surveillance on May 1, 1988, before the licensing reportability review was complete The licensee attributed the failure to include the weekly inspection in SP-3 in a timely manner to confusion that occurred when the technical specification was issue The licensee had undertaken a review of the PDQ system which included identification of a weakness in the review cycle for reportability under 10 CFR 50.72 and 10 CFR 50.7 Program enhancements were being evaluated for procedure modifications by the license The inspector concluded that the licensee's corrective action should preclude missing this technical specification agai This LER is close . Exit Meeting (30703)

The inspector met with licensee representatives (noted in Paragraph 1) at various times during the report period and formally on July 8, 198 The scope and findings of the inspection activities described in this report were summarized at the meetin Licensee representatives acknowledged the inspector's findings at that time.

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