IR 05000275/1993032
| ML16342C037 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 01/07/1994 |
| From: | Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML16342A385 | List: |
| References | |
| 50-275-93-32, 50-323-93-32, NUDOCS 9402020204 | |
| Download: ML16342C037 (32) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION REGION V
Report Nos:
Docket Nos:
License Nos:
Licensee:
Facility Name:
Inspection at:
50-275/93-32 and 50-323/93-32 50-275 and 50-323 DPR-80 and DPR-82 Pacific Gas and Electric Company Nuclear Power Generation, B14A 77 Beale Street, Room 1451 P. 0.
Box 770000 San Francisco, California 94177 Diablo Canyon Units
and
Diablo Canyon Site, San Luis Obispo County, California Inspection Conducted:
November 4 through December 8,
1993 Inspectors:
M. Miller, Senior Resident Inspector M. Tschiltz, Resident Inspector S.
Peterson, Project Manager, NRR Approved by:
~Summar:
nson, ie React Projects Section
/7e~
ate cygne Ins ection on November 4 throu h December
1993 Re ort Nos.
50-275 93-32 and 50-323 93-32 Areas Ins ected:
Routine, announced, resident inspection of plant operations; maintenance and surveillance activities; followup of onsite events, open items, and licensee event reports (LERs);
and selected independent inspection activities.
Inspection Procedures 37700, 51332, 60705, 61726, 62703, 71707, 92701 and 92712 were used as guidance during this inspection.
Safet Issues Mana ement S stem SINS Items:
None Results:
General Conclusions on Stren ths and Weaknesses:
Strengths:
~
The licensee continued plant operations during this inspection period with no events, and no significant equipment failures.
940202020 05000275 cy40}07 pgR
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A review of the licensee's design change program found the licensee to have produced timely and technically sound
CFR 50.59 reviews.
Meaknesses:
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In four instances identified by NRC inspectors, licensee personnel failed to comply with plant procedures related to fire protection, control of lubricants, documentation of gauge oscillations during surveillance testing, and control of equipment configuration (Paragraph 4).
Si nificant Safet Matters:
None Summar of Violations:
A level IV violation was cited for four instances of failure to follow plant procedures (Paragraph 4).
DETAILS Persons Contacted Pacific Gas and Electric Com an G.
H. Rueger, Senior Vice President and General Manager, Nuclear Power Generation Business Unit J.
D. Townsend, Vice President and Plant Manager, Diablo Canyon Operations W. H. Fujimoto, Vice President, Nuclear Technical Services
- R. P.
Powers, Hanager, Nuclear guality Services J.
H. Arhar, Nuclear Regulatory Engineer, Operator Licensing
- J. S.
Bard, Director, Mechanical Maintenance A. G. Barriga, Engineer, Nuclear Engineering and Construction Services
- K. L. Bych, Senior Engineer, Independent Safety Evaluation Group R. S.
Cahn, Nuclear Safety Evaluation Group Leader, Nuclear Engineering and Construction Services W.
G. Crockett, Manager, Technical and Support Services W. D. Ellis, Mechanical Engineer, Nuclear Engineering and Construction Services
- S.
R. Fridley, Director, Operations N. S.
Gowrish, Nuclear Generation Engineer, Nuclear Engineering and Construction Services J.
D. Grammer, System Engineer, Systems Engineering T. L. Grebel, Supervisor, Regulatory Compliance
- B. W. Giffin, Manager, Maintenance Services
- C. R. Groff, Director, Plant Engineering
- J.
R. Hinds, Director, Nuclear Safety Engineering
- K. A. Hubbard, Engineer, Regulatory Compliance T.
P.
Lee, Nuclear Engineering Group Supervisor, Nuclear Engineering and Construction Services
- T. L. HcKnight, Senior Engineer, guality Control H. H. Heleis, Nuclear Engineer, Nuclear Engineering and Construction Services
- D. B. Hiklush, Director, Operations Services R.
H. Nakao, Engineer, Nuclear Engineering and Construction Services
- S.
R. Ortore, Director, Electrical Maintenance
- J. L. Portney, Senior Engineer, System Engineering J.
E. Skaggs, Senior Operations Engineer, Operations T.
W. Soohoo, Contract Engineer, General Office Training Group
- Denotes those attending the exit interview.
The inspectors interviewed other licensee employees, including shift supervisors, shift foremen, reactor and auxiliary operators, maintenance personnel, plant technicians and engineers, and quality assurance personnel.
0 erational Status of Diablo Can on Units I and
During this inspection period, Unit 1 operated at 100% power except for a reduction to 50 percent power for approximately 3 /~ days from December
through December 6,
1993.
The initial power reduction was necessitated by a failure of two of the three traveling screens associated with circulating water pump 1-2 while the third screen was out of service for maintenance.
Following restoration of the traveling screens, a scheduled curtailment was commenced to clean the condenser and circulating water pump 1-1 conduit and forebay.
Unit 2 operated at 100 percent power except for a curtailment to 50 percent power for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on November
and 14, 1993, to conduct condenser cleaning.
3.
0 erational Safet Verification 71707 a
~
General During the inspection period, the inspectors observed and examined activities to verify the operational safety of the licensee's facility.
The observations and examinations of those activities were conducted on a daily, weekly or monthly basis.
On a daily basis, the inspectors observed control room activities to verify compliance with selected limiting conditions for operation (LCOs)
as prescribed in the facility Technical Specifications (TS).
Logs, instrumentation, recorder traces, and other operational records were examined to obtain information on plant conditions and to evaluate trends.
This operational information was then evaluated to determine whether regulatory requirements were satisfied.
Shift turnovers were observed on a sampling basis to verify that all pertinent information on plant status was relayed to the oncoming crew.
During each week, the inspectors toured accessible areas of the facility to observe the following:
(1)
General plant and equipment conditions (2)
Fire hazards and fire fighting equipment (3)
Conduct of selected activities for compliance with the licensee's administrative controls and approved procedures (4)
Interiors of electrical and control panels (5)
Plant housekeeping and cleanliness (6)
Engineered safety features equipment alignment and conditions (7)
Storage of pressurized
.gas bottles The inspectors talked with control room operators and other plant personnel.
The discussions centered on pertinent topics of general plant conditions, procedures, security, training, and other aspects of the work activitie b.
Unit 1 Circuit Breaker in Incorrect Position C.
On November 16, 1993, during a tour of Unit 1, the NRC inspector found a normally closed circuit breaker in the "tripped off" posi-tion.
This issue is discussed further in Paragraph 4.d of this inspection report.
Probabilistic Risk Based Auxiliar Feedwater Walkdown Ins ection The inspector performed a walkdown of the Unit
AFW system using the guidelines of NUREG/CR-5616, "Auxiliary Feedwater System Risk-Based Inspection Guide for the Diablo Canyon Unit 1 Nuclear Power Plant," Risk Important Walkdown Table 3. 1.
No discrepancies were noted.
d.
Radiolo ical Protection e.
The inspectors periodically observed radiological protection practices "to determine whether the licensee's program was being implemented in conformance with facility policies and procedures and in compliance with regulatory requirements.
The inspectors verified that health physics supervisors and technicians conducted frequent plant tours to observe activities in progress and were aware of significant plant activities, particularly those related to radiological conditions and/or challenges.
ALARA considerations were found to be an integral part of each RWP (Radiation Work Permit).
Ph sical Securit Security activities were observed for conformance with regulatory requirements, the site security plan, and administrative procedures, including vehicle and personnel access screening, personnel badging, site security force manning, compensatory measures, and protected and vital area integrity.
Exterior lighting was checked during backshift inspections.
No violations or deviations were identified.
4.
Failures to Com
With Procedure Re uirements 71707 During this inspection report period, NRC inspectors observed several instances which involved the licensee's failure to comply with administrative procedures.
a ~
Failure to Follow Administrative Re uirements For Dealin With Gau e
Oscillations During the performance of surveillance test procedure (STP P-3B) for residual heat removal (RHR)
pump 1-1 on November 10, 1993, an NRC inspector noted differential pressure gauge oscillations up to 50 psid during pump operation.
The licensee evaluated this condition and determined that the oscillations were acceptable for the conditions which existed during the performance of the tes However, the inspector noted that during the surveillance, the operator did not follow all of the requirements of Administrative Procedure AP C-3S3, Revision 1, "Dealing With Gauge Oscillations During the Performance of ASHE Section XI Required Pump Tests."
AP C-3S3 requires the use of averaging techniques for gauge oscillations which can not be reduced to less than or equal to two percent of the midpoint of the reading without isolating the gauge.
AP C-3S3 also requires that the use of averaging techniques be annotated in the remarks of the procedure, and that an Action Request (AR) be initiated to ensure that a review of the instrumentation is performed to determine if the gauge oscillations can be reduced or eliminated by a modification.
Although the operator followed the procedure by throttling the gauge isolation valves and using the averaging techniqu'e, the operator did not make the required annotations in the remarks and an AR was not initiated, as required by procedure.
Failure to Follow Administrative Re uirements For the Control of Plant Lubricants On November 18, 1993, during a tour inside the radiologically con-trolled area (RCA) an NRC inspector noted that lubricant lockers were not being maintained in accordance with licensee procedures, as follows:
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Flammable liquid lockers used for the storage of lubricants were observed to have improper materials stored with the lubricants.
Administrative Procedure AP C-251, Revision 7, Paragraph 4.2.2, requires that flammable liquid cabinets be separated from other hazardous materials.
The inspector observed that corrosive material was stored in a lubricant locker, and lubricants that were required to be stored inside the lockers were stacked on top of the lockers.
~
The weekly review of the RCA lubricant storage log book, required by Paragraph 5.2.2 of Plant Procedure AP D-753, Revi-sion 20, was not performed for the period between November
and November 18, 1993, a period of 13 days.
These conditions were corrected promptly after they were brought to the licensee's attention.
Unit
480 Volt Bus 1F Breaker 52-1F-46R in Incorrect Position On November 16, 1993, during a tour of the Unit 1 Bus 1F 480 volt load center, an NRC inspector found Circuit Breaker 52-1F-46R in the
"tripped off" position.
This breaker provides power, in series with Circuit Breaker 52-1F-46, to the Accumulator 1-1 outlet valve to the Loop 1-1 cold leg (SI-)-8808A).
The breaker is normally closed, with the valve normally open.
The function of circuit breaker 52-1F-46R is to provide backup overcurrent protection for the containment electrical penetration.
Technical Specification (TS) 3.5. 1 requires that SI-1-8808A be open, with power removed, when in modes 1, 2, and 3 (above 1000 psig).
In order to comply with the TS
requirement, Circuit Breaker 52-1F-46, the primary breaker, is sealed in the "off" position when in these modes, and the series circuit breaker, 52-1F-46R, is normally closed.
The operations staff quickly returned Circuit Breaker 52-1F-46R to its proper closed position.
Emergency procedures which require operation of the accumulator outlet valve were reviewed by the licensee and the NRC to ascertain the adequacy of actions for restoring power to the valve during an event.
Based on the review it was determined that emergency procedures requiring operation of SI-1-8808A included operator actions to restore power to the accumulator isolation valve by verifying, and closing if required, both Circuit Breakers 52-1F-46 and 52-1F-46R.
The inspector noted that scaffolding had been erected in Unit 1 near the Bus 1F 480V load center, within several feet of Circuit Breaker 52-1F-46R, for pre-outage work involving the replacement of inverters.
The inspector concluded that the installation of scaffolding and the ongoing pre-outage work significantly increased the amount of activity in that area and possibly resulted in the inadvertent opening of the circuit breaker.
In response to this concern, the licensee immediately halted all construction work and reviewed the importance of preventing inadvertent operation of equipment and the need for immediate notification of the control room whenever plant equipment is disturbed.
The accumulator is required to be operable in Nodes 1, 2, and 3 when pressurizer pressure is above 1000 psig.
Valve SI-1-8808-A is required to be operated during depressurization after certain accidents.
Nevertheless, existing licensee emergency procedures require operator actions which would have resulted in the discovery that breaker 52-1F-46R was open and would have restored power to SI-1-8808A without an adverse effect on plant operation.
The existence of licensee emergency procedures which would have identi-fied and corrected the improper positioning of breaker 52-1F-46R lessened the safety significance of this condition.
Licensee Operations Procedure OP B-3B: II, Revision 8, "Accumulators Alignment Verification Checklist for Plant Startup," requires that Circuit Breaker 52-1F-46R be maintained in the closed position to provide redundant overcurrent protection for the containment penetration.
This failure to maintain circuit breaker configuration as required by OP B-3B:II was in violation of TS 6.8.1, which requires that activities recommended in Appendix A of Regulatory Guide 1.33, Revision 2, be impl,emented by approved procedures, including procedures associated with startup, operation and shutdown of safety related systems.
In response to the inspectors'bservations, and in addition to the actions documented in Inspection Report 50-275/323/93-29, licensee management significantly increased its involvement with Haintenance and Operations personnel regarding expectations and requirements associated
with procedure compliance.
For example, a
human factors review and a
focused evaluation of the barriers to procedure compliance were included in the resolution of the nonconformance (NCR DCO-93-PG-N048)
concerning lack of procedural compliance.
The inspector noted that management involvement and direct interaction have been observed within all levels of the plant staff regarding this issue.
Each of these four instances was of low safety significance.
However, these examples of failure to follow plant procedure requirements are a
violation of TS 6.8. 1, which requires that procedures be implemented governing the fire protection program and the activities recommended in Appendix A of Regulatory Guide 1.33, Revision 2, including procedures for surveillance testing; preventive maintenance; and startup, operation, and shutdown of safety-related systems (Violation 50-275/93-32-01).
One violation with four examples was identified.
Maintenance 62703 During the inspection period, the inspectors observed portions of, and reviewed records on, selected maintenance activities to assure compliance with approved procedures, Technical Specifications, and appropriate industry codes and standards.
Furthermore, the inspectors verified that maintenance activities were performed by qualified personnel, in accordance with fire protection and housekeeping controls, and that replacement parts were appropriately certified.
The inspectors observed portions of the following maintenance activities:
Descri tion Dates Performed CARDOX Post-Modification Test for Diesel Generator Room 2-1 (PMT 18.08 Revision 0)
Diesel Generator Starting Air Compressor 1-2A Replacement (W/0 R0102749)
November 4, 1993 November 8-9, 1993 Diesel Generator Fuel Oil Line Replacement November 10, 1993 (W/0 C0116143)
Br idge Crane Cal ibr ati on (Unit 1)
(HP 1-4.4-2.A, Revision 4, W/0 R0103138)
Battery Charger 1-1 Routine Maintenance (HP E-67.3A Revision 19, W/0 R0102585)
Ventilation System VAC-1-HOD-2A HOV Overhaul (W/0 C0114834)
No violations or deviations were identified.
November 19, 1993 November 17, 1993 November 19, 1993
Surveillance 61726 The inspectors reviewed a sampling of Technical Specifications (TS)
surveillance tests and verified that:
(1)
a technically adequate procedure existed for performance of the surveillance tests; (2) the surveillance tests had been performed at the frequency specified in the TS and in accordance with the TS surveillance requirements; and (3) test results satisfied acceptance criteria or were properly dispositioned.
The inspectors observed portions of the following surveillance tests on the dates shown:
Procedure STP P-3B STP H-12B STP H-21C Descri tion Routine surveillance of RHR pumps (Unit 1)
Battery Charger Performance Test (Battery Charger 1-1)
Hain Turbine Valve Testing (Unit 2, W/0 R0122044)
Dates Performed
November 10, 1993 November 18, 1993 November 14, 1993 During observation of the RHR pump surveillance, the inspector identified a procedure violation involving the disposition of pressure gauge oscillations.
This issue is discussed in Paragraph 4.a of this report.
Pre aration for Refuelin 60705 On December 7-8, 1993, an NRC inspector observed portions of licensee activities involved with the inspection of new fuel per Operations Procedure OP B-BC, Revision 6, "Inspection of New Fuel,"
and OP B-BA, Revision',
"Handling and Storage of New Fuel Assemblies."
New fuel was offloaded from shipping containers and prepared for inspection.
The inspector observed appropriate rigging practices for lifting heavy, hazardous loads and effecting storage of new fuel.
As previously discussed in Paragraph 4.c, the inspector observed a violation of administrative procedure requirements for housekeeping and cleanliness practices and postings.
These were corrected promptly after the inspector identified the concerns to the refueling senior reactor operator.
One violation was identified and discussed in Paragraph 4.c.
NRR Review of 10 CFR 50.59 Evaluations 37700 The NRR Project Hanager conducted a review of 10 CFR 50.59 evaluation packages prepared by the licensee for Units
and 2.
The effort included discussions with the Pacific Gas and Electric (PGLE) staff familiar with the licensee's training, procedures and preparation of 10 CFR 50.59 packages, as well as review of the licensee's
CFR 50.59 design change review program procedures and completed package The licensee considered the program to conform to the Nuclear Safety Analysis Center (NSAC)-125, "Guidelines for 10 CFR 50.59 Safety Evaluations,"
which was referenced and partially included in procedure TS 3.ID2, Revision 0, dated September 17, 1993,
"Licensing Basis Impact Evaluations."
The inspector also reviewed the licensee's Nuclear Engineering Services administrative procedure CF3.NE9, Revision 0,
"Diablo Canyon Power Plant Design Changes,"
dated July 3, 1993, and CF3. ID9, Revision 0,
"Design Change Package Development,"
dated September 11, 1993.
Although the licensee's guidance in this area indicated that persons preparing or reviewing the evaluation packages should have completed the licensee's
CFR 50.59 training, the inspector noted that two of the evaluation packages reviewed were prepared by licensee staff who had not received the appropriate training.
However, the inspector noted that the packages were thorough and technically sound.
The licensee stated that the engineers would receive required formal training within the year.
In general, a good program had been established with good engineering support of activities.
Review of 10 CFR 50.59 Evaluation Pro ram Packa es In general, the licensee produced timely and technically sound
CFR 50.59 reviews.
Based on conversations with members of the Nuclear Engineering Services organization, the appropriate level of guidance and feedback appeared to have been provided to the
CFR 50.59 reviewers to ensure that complete safety evaluations were prepared.
The following is a summary of the
CFR 50.59 packages reviewed.
Desi n Chan e Packa e
DCP E-45581 Revision 0 This DCP addressed the removal of the emergency diesel generator (EDG) 1-3 tie-in to Unit 2 to ensure that no interconnection existed between the EDGs of Unit
and those of Unit 2.
The tie-in removal was a result of the licensee's installing a sixth EDG in Unit 2.
The
CFR 50.59 determination correctly identified the need for a safety evaluation because the tie-in removal caused a change to the facility as described in the Safety Analysis Report (SAR).
The SAR, Technical Specification (TS),
and licensing basis document references were complete.
In addition, the associated design change notices (DCNs)
to revise the simulator (DCN-DC1-EJ-45648),
drawings (DCN-DCI-EH-45581),
and control board (DCN-DC1-EJ-45581)
were appropriately addressed.
The
CFR 50.59 items (a)(2)(i) through (a)(2)(iii)
were adequately addressed according to procedures and the regulations.
OCP-E-47690 Revision 0 - This package addressed a Unit
modification to the power supply for the steam generator wide range level (SGWRL) instruments, loops 1-3 and 1-4, from vital instrument AC channel IV to III. This DCP included the installation of new cable in dedicated conduit between Hagan control racks.
The modification provided a minimum of two power sources for single failure protection of Regulatory Guide 1.97, Variable 46 (i.e.,
steam generator level).
Two design change notices (DC1-EE-47690 and
DC1-EJ-47690) for the associated SGWRL recorders in the turbine steam supply and post-accident monitoring systems were also reviewed.
The licensee's
CFR 50.59 evaluation addressed all parts of 10 CFR 50.59, items (a)(2)(i) through (a)(2)(iii).
The package included a thorough evaluation of electrical design considerations such as the power supply, cable system and raceway design, load study, circuit design, and interfaces.
The licensee's conclusions, evaluation, and considerations appeared thorough.
DCP-P-48672 Revision
This package addressed the addition of a 1.5-inch isolation valve to the positive displacement pump discharge relief valve line for Unit 2.
The valve was needed to provide positive isolation between the volume control tank and the relief valve to support installation of pulsation dampers during power operation.
Although the package specified appropriate administrative controls for ensuring the gate valve was locked open to aid in preserving the overpressure protection function of the relief line, the package did not properly identify the need for relief from the requirements of ASNE Section III, Subsection NC-7142.
The licensee subsequently identified this need, and requested relief in a letter dated April 13, 1993.
The NRC Staff is reviewing the licensee's relief request, and will communicate with the licensee on this matter in separate correspondence.
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Packa e DC1-EN-48572 Revision 0 This package addressed changes to the design drawing for the connection of the lower internals stand seismic restraint plates to the lower internals stand guide stud stabilizing struts.
The connection was changed from a bolted to a
pinned connection.
The change also added a lifting lug to the plates to allow for easier installation and removal with a defueled reactor.
This package was screened by the licensee and a
CFR 50.59 evaluation was deemed unnecessary.
The licensee's screening appeared acceptable.
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Packa e DC1-EP-47794 Revision 0 - This package addressed a
modification to Diablo Canyon Unit 1 steam generator feedwater nozzle transition piping from the nozzle to the first elbow.
The licensee had detected erosion on the outer diameter of the thermal sleeves and decided on the replacement due to the potential for fatigue cracks developing at the weld junction.
The
CFR 50.59 evaluation appropriately addressed the material differences and feedwater flow changes due to the new tuning-fork design of the feedwater nozzle transition piping.
The inspector found that the licensee's evaluation addressed all parts of 10 CFR 50.59, items (a)(2)(i) through (a)(2)(iii), and was acceptable.
No violations or deviations were identified.
9.
U date of Instrument Power Su
Isolation Devices Anal sis 37700 The licensee is conducting a review, of instrument power supply isolation devices.
This review was previously discussed in Inspection Report 50-275/323/92-25, Paragraph 11, which documented a licensee meeting with NRC Region V management on this subjec During followup of this review the licensee stated, to update and clarify the Inspection Report 50-275,323/92-25 documentation, that the review program which is in progress stemmed, from the licensee's determination during the Regulatory Guide (RG) 1.97 project reviews that some safety and non-safety related (NSR) instruments shared common power supplies.
This review included all loads fed from the following 120 VAC breakers:
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All 120 VAC breakers that feed instrument Class-1A loads.
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All 120 VAC Class-1E breakers that supply devices that are within the current scope of the Regulatory Guide (RG) 1.97 Review Pro-ject, which includes all variables listed in the licensee's TS Table 3.3-10,
"Accident Monitoring Instrumentation".
The review encom-passes all TS variables with identified breaker installation issues.
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A random sample of 60 of the total population of 800 RG 1.97 instrumentation loops from both Diablo Canyon Units.
Since the discussions documented in inspection report-50-275,323/93-25, the licensee decided to increase the review to 100K of the population RG 1.97 instrumentation fed by vital power.
These analyses, to date, have not identified a general concern.
In addition to earlier information and preliminary conclusions, the licensee has reached the preliminary conclusion that reactor protection and safe shutdown circuits are not affected (i.e., there are no associated NSR components or they have been provided with appropriate isolation devices).
The licensee submitted a progress review of the RG 1.97 review project to NRR, and has also discussed the results with NRR in meetings and telephone conversations.
No violations or deviations were identified.
10.
Licensee Event Re ort LER Followu 90712 The inspector performed an in-office review of the following LER associated with an operating event.
Based on the information provided in the report, the inspector concluded that the licensee had met the reporting requirements, had identified root causes, and had taken appropriate corrective actions.
The following LER is closed:
LER NUMBER DESCRIPTION Unit 1,93-009, Revision
Actuation of Wrong Undervoltage Relay Causes Unplanned Diesel Generator Start Due to Personnel Error No violations or deviations were identifie.
Followu of 0 en Items 92701 a ~
Closed Information Notice 92-06 Reliabilit of Antici ated Tran-sient Without Scram ATWS Hiti ation S stem 50-275 92-13-INN b.
This Information Notice addressed situations at operating plants in which the ATWS system was not maintained in an operable status for significant lengths of time.
The licensee response to this Information Notice concluded that testing of the AHSAC system (the Westinghouse design ATWS mitigation system)
included appropriate overlap of circuitry, that operability of the ATWS system had been high, and that administrative controls in the Equipment Control Guidelines were appropriate to ensure continuing high operability.
Based on routine resident inspector walkdowns, review of operating logs, and inspections in accordance with NRC Temporary Instruction (TI) 2500/20,
" Inspection to Determine Compliance with the ATWS Rule," which was recently documented in Inspection Report 50-275/323/92-31, the operabil,ity of the ATWS system appears to have been high (greater than 95%).
Therefore, this item is closed.
Closed Information Notice 89-32 Surveillance Testin of Low Tem-erature Over ressure Protection LTOP S stems 50-275 91-10-INN c ~
This Information Notice addressed a concern for the lack of surveillance testing of LTOP systems in reactor plants.
The inspector reviewed the licensee's actions in response to this Information Notice.
Licensee staff personnel concluded that surveillance testing verified that power operated relief valves operated within design basis requirements of 3.5, seconds.
Additionally, the review of design criteria memoranda will continue to validate this an other design basis as part of an ongoing program This response appeared appropriate.
Therefore, this item is closed.
Closed Information Notice 91-13 Testin of Emer enc Diesel Generators EDGs 50-275 91-11-INN Information Notice 91-13 addressed the adequacy of testing of EDGs at operating plants, particularly with respect to maximum load testing.
The licensee's review in response to this information notice concluded that the load testing performed on EDGs appropriately addressed all required loads.
The licensee also concluded that continuing periodic reviews (on an 18-month refueling interval) of the configuration control of loads, and review of EDG capacity requirements, as required by licensee procedures, ensured that the EDGs would continue. to meet requirements.
The inspector noted that the NRC Electrical Distribution System Inspection 50-275,323/91-07 also reviewed several of the EDG surveillance tests for adequacy, and concluded that the tests appropriately addressed the EDG design basis requirements.
The licensee response appeared appropriate, and independent NRC assessment did not identify significant concerns with EDG testing.
Therefore, this item is close I/
-12-Closed Enforcement Item 50-275 92-20-3 Inade uate 0 eratin Instructions for Positive Dis lacement Char in Pum s
PDPs This violation addressed the licensee's failure to provide appropriate operating guidance for operating the PDPs in the event an Appendix R safe shutdown was required.
The inspector noted during followup inspection that revised proce-dures which had been issued or were pending appeared appropriate.
The licensee had revised the operability evaluation procedure to require that Engineering provide specific written operating limitations to Operations, and committed to revise operating and abnormal procedures for use by operators at the completion of the PDP design change installation.
Additionally, several PDP modifications have been implemented, improving the reliability of the PDP.
Based on the licensee's corrective actions, this item is closed.
Closed Enforcement Item 50-275 92-31-02 Inade uate Containment Cleanliness This violation addressed a failure to take appropriate corrective actions to ensure containment cleanliness near the end of the Unit
outage in 1992, after containment integrity was set.
Inspector walkdowns, personnel interviews, and review of procedures during the 1993 Unit 2 outage determined that containment cleanliness controls had improved to an appropriate level, particularly at the end of the outage when containment integrity was set.
Based on the licensee's corrective action, this item is closed.
Closed Enforcement Item 50-323 93-16-03 Untimel ualit Evaluations This violation addressed a programmatic lack of timeliness in issuing guality Evaluations
{gEs) within the 30-day limit required by procedures.
The inspectors verified that gEs are now being issued within the 30-day limit, and that the licensee's corrective action, increased administrative controls and management attention, appeared to appropriately manage gE timeliness.
No additional examples of untimely gEs have been observed.
This item is closed.
Closed Enforcement Item 50-323 92-22-01 Lack of Acce tance Criteri a for an EOG Test This violation described the lack of acceptance criteria for an EDG day tank fuel oil setpoint test following EDG 2-3 installation.
The licensee acknowledged the violation, corrected the test procedure, and performed a review of all related EOG tests to determine if appropriate acceptance criteria were included.
Licensee review determined that all other EDG tests included appropriate acceptance criteria.
Independent NRC inspection activities since that time have validated several instances of appropriate acceptance criteria in tests, and
0'
- 13-no instances of missing acceptance criteria have been observed.
Based on the licensee's actions, and on inspector observations, this item is closed.
0 en Followu Item 50-275 323 92-31-03 No Draina e for elief Valve Tail i es An earlier inspection report noted that the ASHE Piping Code required drains on relief valve tailpipes.
Inspectors noted that licensee commitments to piping code requirements included installation of drains on relief valve tailpipes.
The licensee is following this issue in Action Request A0283981.
The licensee stated that the Piping Code interpretation committee has discussed the clarity and intent of the code regarding the requirement for tailpipe drains on relief valves.
The committee is determining the requirements for relief valves in various services and in various installations.
Finalization of the code committee position is expected after April 1994.
The inspector will followup the licensee's resolution of this issue at that time.
Closed Followu Item 50-323 92-22-02 Lack of Desi n Basis for Diesel Fuel Oil Tank Low Level An inspector questioned whether the unusable volume of diesel fuel oil in the bottom of the EDG day tank had been inappropriately credited during the establishment of the minimum volume reference level for the EDG day tank.
The day tank minimum level was designed to hold a minimum of one hour of fuel for the EOG.
The test to establish minimum reference level introduced 200 gallons into the empty tank, without considering that about 30 gallons was unusable, leaving 170 usable gallons.
Under rated load, the EDG uses about 190 gallons per hour.'his issue is being tracked by Action Request A0299069, which docu-ments that the licensee has committed to submit a license amendment request to allow the minimum reference level to hold slightly less than an hour of fuel.
This license amendment was submitted to NRR on December 8,
1993.
The licensee concluded that a minimum day tank level setpoint of slightly less than one hour of fuel is not significant, since the safety-related diesel fuel oil system provides automatic makeup to the day tanks.
Based on the licensee actions to date, and the license amendment request submitted by the licensee, this item is closed.
Closed Enforcement Item 50-275 92-17-01 0 eration With Three Ino erable Containment Fan Cooler Units CFCUs This violation involved the inoperability of three CFCUs in Unit I for about II months.
An enforcement conference was held, as documented in Meeting Report 50-275,323/92-19.
At this conference the licensee noted that, although poor maintenance and engineering work had allowed the CFCUs to become degraded, detailed Westinghouse
-14-analysis had determined that operator actions within 55 minutes would have resulted in no safety significance or degraded condition during a design basis LOCA.
Although the safety significance of the degraded CFCUs was later determined to have been low, the root cause, poor maintenance and engineering work, was significant.
To prevent the, recurrence of poor maintenance and engineering, the licensee committed to strengthen the maintenance and engineering programs to preclude further instances of poor work.
Integrated system teams of maintenance, system, and design engineers were formed.
Several improvements and monitoring programs were also implemented specifically for the CFCUs.
These appeared to have directly improved CFCU performance and staff awareness of CFCU design requirements.
System teams were implemented for all plant systems.
Licensee performance in this area appears appropriate.
Therefore, this item is closed.
No violations or deviations were identified.
An exit meeting was conducted on December 14, 1993, with the licensee representatives identified in Paragraph 1.
The scope of the inspection and the inspectors'indings, as noted in this report, were discussed with and acknowledged by the licensee representatives.
The licensee did not identify as proprietary any of the materials reviewed by or discussed with the inspectors during this inspectio.4 0