IR 05000275/1992016

Intervenor Exhibit I-MFP-137,consisting of Insp Rept Re Dockets 50-275/92-16 & 50-323/92-16,dtd 920707

ML20059D059
Person / Time
Site:	Diablo Canyon
Issue date:	08/24/1993
From:	Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
References
OLA-2-I-MFP-137, NUDOCS 9401060375
Download: ML20059D059 (24)

Text

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g y 93 U.S. NUCLEAR REGULATORY COMMISSION

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REGION V

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Report Nos:

50-275/92-16 and 50-323/92-16

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Docket Nos:

50-275 and 50-323 License Nos:

DPR-80 and DPR-82 i

Licensee:

i Pacific Gas and Electric Company

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77 Beale Street, Room 1451 San francisco, California 94106 Facility Name:

Diablo Canyon Units 1 and 2 Inspection at:

Diablo Canyon Site, San Luis Obispo County,

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California

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Inspection Conducted:

April 21 through June 1, 1992 Inspectors:

P. Morrill, Senior Resident inspector M. Miller, Resident Inspector

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Approved by:

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P. H.[/ohnson, Chief

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React 4r Projects Section 1 Date Signed j

Summary:

inspection on April 21 through June 1. 1992 (Report Nos. 50-275/92-16 and l

50-323/92-161

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Areas Inspected:

Routine inspection of plant operations, maintenance and surveillance activities, followup of onsite events, open items, and licensee event reports (LERs), as well as selected independent inspection activities.

i Inspection Procedures 93702, and 94600 were used as guidance during this inspection.37

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Safety Issues Manaae nent System (SIMS1 Items:

None Pesults

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i General Conclusions on Strenoths and Weaknesses Strengths:

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A more conservative approach to the scheduling of Instrumentation and Control surveillances was noted.

This involved a licensee initiative to discontinue routine use of the 25% grate period allowed by the Technical

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Specifications for the performance of surveillances (paragraph 6.a).

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Weaknesses:

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f-The NRC identified several problems which had not been identified or

appropriately documented by the licensee. These included the installa-

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tion of plastic tarps in front of Appendix R lighting (paragraph 8.a) and j

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the violations identified below.

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Sionificant Safety Matteri: None i

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Sumary of Violations:

involving the loading of two one-ton chainfalls to approxim j

i pounds each wile installing a resin cask lid (paragranh 3.b(2)).

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l nnn-cited violations were also identified, involving inconsistent Two

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1. astener terques for two residual hest removal (RHR) syst j,

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em valves (paragraph 8.b) and failure to declare a Unit 2 main steam safety valve

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inoperable after test equipment used to establish its setpoint was found j

to have failed (paragraph 10.k).

i Open Items Sumary:

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9 were closed, I remains open. pened. Ten followup items were reviewed;

.nree new followup items were o J 7 closed, I remains open.

Twenty LERs were reviewed; 19 were il

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DETAll5 1.

Persons Contacted Pacific Cas and Electric'ComDany G. M. 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 Te;hnical Services D. B. Miklush, Manager, Operations Services

• M. J. Angus, Manager, Technical Services

• B. W. Giffin, Manager, Maintenance Services

• W. G. Crockett, Manager, Support Services

• J. E. Molden, Instrumentation and Controls Director

• W. D. Barkhuff, Quality Control Director.

• R. P. Powers, Mechanical Maintenance Director T. L. Crebel, Degulatory Compliance Supervisor H. J. Phillip Electrical Maintenance Director

• J. A. Shoulders, Onsite Project Engineer

• S. R. Fridley, Operations Director

• R. Gray, Radiation Protection Director J. J. Griffin, Senior Engineer, Regulatory Compliance J. V. Boots, Chemistry Director

• T. A. Moulla, Assistant to Vice President, Diablo Canyon Operations

• R. Kohout, Safety, Health and Emergency Services Director

• D. P. Sisk, Regulatory Compliance Engineer D. R. Stermer, Power Production Engineer M. R. Tresier, Project Engineer R. Clark, Assistant Project Engineer R. Gagne, Acting Radwaste foreman U. A. Farradj, Fire Protection Engineer i

R. A. Waltos, Senior Power Production Engineer S. F. Shrefler, Mechanical Maintenance Engineer B. D. Pogue, System Engineer R. Ortega, System Engineer j

R. Watson, Quality Assurance Engineer San Luis Obispo County Officials R. Hendrix, Administrator L. Williams, Deputy Administrator V. Morici, Emergency Response Coordinator

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• Denotes those attending the exit interview.

The inspectors interviewed several other licensee e maintenance personncl, plant technicians and engineers, and quality assurance personnel.

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2.

Doerational Status of Diablo Canven Unit s 1 anu

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During the inspection period, Unit 1 operated at 100% power, except for April 25-27, and May 2-3, 1992.

On April 25, Unit I reduced power to 50%

for routine condenser cleaning. While conducting maintenance on the 1-1 main feedwater pump, vacuum in the condenser was lost, causing the main turbine and the reactor to trip.

was at full power on April 27.

Unit I was restarted on April 26 and

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This event is described in paragraph 4.a l

below.

During the period May 2-3, 1992, Unit I was curtailed to 50%

power for routine condenser cleaning.

Unit 2 operated at 100% power during the reporting period.

3.

Doerat tor,a1 Safety 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 examin.tions 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 Logs, instrumentation, recorder traces, and other oper

records were examined to obtain information on plant conditions and to evaluate trends.

to determine whether regulatory requirements were utisfied.T turnovers were observed on a sample basis to verify that all perti-Shift

nent 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:

(li General plant and equipment cond.:ons (2) Fire hazards and fire fighting equipment (3)

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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 i

(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

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i plant conditions, procedures, security, training, and other aspects

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of the work activities, i

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The inspectors also accompanied auxiliary operators during their rounds in the plant and reviewed the associated log entries.

On two occasions the inspector verified that auxiliary operators were completing their rounds consistent with the log sheet entries.

No instances of improper log entries were identified.

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Radiolocical Protection (1) The inspectors periodically observed radiological protection practices to determine whether the licensee's program was t,eing implemented in conformance with facility policies and proce-dures and in compliance with regulatorp equirements.

The inspectors verified that health physics supervisors and professionals 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 t,f each RWP (Radiation Work Permit).

(2)

Radioactive Waste Shipment on May 28-29. 1992

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The inspectors observed the licensee's preparation to ship a high integrity container (HIC) of radioactive resin to Ricnland, Washington for disposal.

The licensee planned to survey ar.d remove the HIC from storage in a permanent plant shield cask and place the HIC in a transportation cask carried on a sole-use flatbed truck.

A shielded boom crane was used for this movement.

The evolution was significant in that it occurs infrequently (1 to 2 times a year at most) and the HIC had a surface dose rate of over 30 R/Hr. The inspectors observed the licensee's pre-brief meeting which covered communications, surveys, exclusion areas, ALARA considerations, trane ;perations, and contingency plans.

The inspectors cbserved the HIC transfer, which proceeded without incident. P w ver when placing the lid on the shipping cask, licensee per,sonnel were initially unable to get it to seat properly. The lid was supported by three slings and restrained horizontally with respect to the cask by two guide pins.

With the lid resting on the top of the task, the licensee's riggers replaced two of the slings with chainfalls.

This allowed the lid to be leveled by an individual standing on top of the lid and adjusting the chainfalls.

The rigger leveling the lid was shielded from the HIC by the lid itself.

The lid was approximately 3 inches higher than the seated position during the evolution.

After several tries the lid seated and the rigging equipment was removed.

During this evolution, the inspector questioned licensee personnel as to whether the chainfalls being used had adequate capacity.

The licensee personnel present stated that they were confident that the riggers knew what they were doing.

The inspector stated that the chainfalls looked like one-ton units and the cask lid looked substantially heavier.

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The next day May 29. 1992, the inspector conducted a survey of

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the shipping cask on the truck, Radiation levels appeared to be within regulatory limits and the cask appeared to be proper-ly secured and labeled for shipment.

The inspector observed that the weight of the cask lid was approximately 7300 pounds.

The inspector discussed the chainfall capacity with the Mecha-nical Maintenance Director who determined that the chainfalls used had been one-ton units.

The licensee's quality evaluation found them to have a safe working load of 2000 pounds each.

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The load on each chainfall would have been approximately 2400

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pounds.

The inspector noted that this was a personnel safety issue and that the individual standing on the cask lid could

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have been seriously hurt if the chain had parted.

of Maintenance Services and Mechanical Maintenance DirectorThe M agreed that a mistake had been made.

Licensee personnel stated that when the next shipment was made the correct size chain-falls would be available and that thei'r use would be required by written instructions.

This is an apparent violation of Technical Specification 6.8.1 and procedures for control of rigging and load handling equipment.

-(50-275/92-16-01)

i The inspector observed that this event involved riggers and contained elements similar to the loss of offsite power event

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in March 1991 which was caused by a boom crane under the Unit 1 500XV lines.

The Manager of Maintenance Services stated that the boom crane event in 1991 had been caused by personnel who were not riggers, but who had been trained to use the equip-

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ment.

The inspector acknowledged that two different groups of

personnel were involved, but noted that both events appeared to i

I involve weakness in the preplanning and control vi lifting or rigging activities.

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c.

Physical Security l

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Security activities were observed for conformance with regulatory

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requirements, the site security plan, and administrative procedures, including vehicle and personnel access screening, personnel ba site security force manning, compensatory measures, and protected

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and vital area integrit backshift inspections. y.

Exterior lighting was checked during

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No violations or deviations were identified.

Onsite Event Followon (93702)

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Trio of Uni _t_1 Due to loss of Vacuum on Acril a.

25. 1992

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Between 9:00 p.m and 11:00 p.m. on April 24, 1992, plant operators decreased Unit I power to approximately 50% to allow routine

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cleaning of the east main condenser water boxes.

Main feedwater pump (MfP) 1-1 was shut down and cleared to make repairs to steam stop valves to the MfP turbine.

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At approximately 3:45 a.m., vacuum in the operating west condenser had degraded to an absolute pressure of approximately 4 The air in-leakage which degraded condenser vacuum was a.5 inches Hg.

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pparently.

due to leakage through the large 1-1 MFP exhaust butterfly valve.

The on-shift crew attempted to stop the air in-leakage by reseating the closed 1-1 MfP exhaust valve and by placing a second air ejector

in service.

At 4:00 a.m., plant power was being decreased at 10 MW per minute and gland seal steam was re-established to the HFP l-1 turbine.

The decrease in power appeared to be helping condenser At 4:07 a.m.

vacuum.

mechanical vacuum pump, plant operators attempted to place the Hash in service. When the vacuum pump suction valve from the west main condenser was opened, backpressure iricreased rapidly and at 4:08 a.m. a turbine trip and reactor trip occurred.

The operators subsequently stabilized the plant in accordance with Emergency Operating Procedures. All safety related equipment functioned as expected.

four hour non-emergency report to the NRC.At 5:10 a.m.,

the lice tsee made a The inspector evaluated the licensee's investigation, discussed the event with the en-shift crew, and observed the subsequent plant startup.

At approximately 1:00 p.m. on April.27, after evaluation of the event and correction of the cause of the trip, the licensee restarted Unit 1.

full power the morning of AprilThe startup was uneventful and Unit I re 28, 1992.

The inspector observed that the licensee's investigation into the trip identified the following information:

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After the operators had begun to reduce power at 10 MW per

minute, condenser vacuum was stabilizing.

corrected itself without further action.

The problem may have Placing the Nash mechanical vacuum pump in service was

consistent with the licensee's procedure AP-7, " Loss of Condenser Vacuum".

steps necessary to place the Nash pump in service,Howeve The operators in the turbine building at the Hash pump did not

have a local copy of the procedure to place the pump into i

service.

Due to the sense of urgency the operators in the control room did not review the procedure to place the Nash pump in service.

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Unknown to the operators, a seal' water l' solation valve for the

Nash pump was required to be opened before the pump could operate properly.

When the suction valve to the west condenser i

was opened, air entered the condenser through the Nash pump seals.

The condenser vacuum pump suction check valve appears to have

leaked sev: rely in the reverse direction, allowing air flow back to the cundenser.

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-6-s The inspector discussed the event with the Operations Director to

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determine the adequacy of the licenste's corrective actions.

licensee issued an incident Summary to alert other crews and added The of a r. > n feefwater pump. precautions to be taken in the procedure fo The licensee plans to inspect each unit's vacuum pump suction line check valve during outages IRS and 2R5 The licensee also plans to complete a review of all emergency an.

tors might be dispatched to perform equipment' op d

procedures.

postings are plar.ned to be added based on the review.By The inspector questioned the adequacy of the licensee's preparation for taking MfP l-1 out of service.

Licensee personnel stated that preparations had been adequate, but that the plant had not responded as expected.

This event was described in Licensee Event Report (LER) 1-92-004 issued May 22, 1992.

the licensee's, analysis of the event and corrective actions review of the LER.

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l No viola' ions or deviations were identified.

5.

Maintenance (62703)_

The inspectors observed portions of and reviewed records on, selected maintenance activities to assure com,pliance with Furthermore, the inspectors verified that maintenance activities were

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performed by qualified personnel, in accordance with fire protection

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housekeeping controls, and that replacement parts were appro certified.

These activities included:

Work Order No. C0096441, BATP1, Replace Mechanical Seal

Work Order No. C0100168,10-142, Cask, Provide Support to M

Concerns Raised Durino Observation of Boric Acid Transfer Pu a.

i Maintenance

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i During an observation of Work Order No C0096441, BATP1 Mechanical Seal," the inspector noted the following con,cern: lace

" Rep (1)

When questioned by the inspector, the me Control sign-off of the alignment would require removal of the optical alignment tool used to ali lation of a double dial indicator,gn the coupling, and instal-tool.

This was required because an older, slower alignment on optical alignment information. Quality Control would not rely Further review by the inspector and Mechanical Maintenance management found that Quality Control had accepted optical Maintenance foreman and mechanics had

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Although both alignment methods are valid, the inspector was concerned that the mechanics and foreman were not informed of processes which could reduce radiation exposure by reducing time in radiation fields.

The licensee is reviewing training i

and technical information provided to foremen to determine if I

timely and adequate information is provided to Hechanical Maintenance foremen.

(2) The work order required all workers to attend the tailboard l

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However,' the mechanic performing alignment of the pump and motor coupling had not attended the tailboard. 'The licensee stated that coupling alignment activities are highly specialized, and in this case did not require attendance of the taliboard, in that the mechanic had been briefed on the job and

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provided adequate information to perform the work safely.

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The inspector noted that the procedure requirements and expecta-tions for specialists regarding tailboard attendance were not

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specific, and may be subject to incorrect interpretation, par-titularly during times of high work activity such as an outage.

The licensee was planning to review expectations regarding attendance at tailboards before the next scheduled outage.

These issues will be followed during routine resident inspection activities.

No violations or deviations were identified.

6.

Surveillance (61726)

the inspectors assessed compliance with TS requirem procedures.

and that test results met acceptance criteria or were app dispositioned.

These tests included:

STP 1-18W2D, Revision 5. Isntopic Calibration of Plant Vent Hid

Range lodine Radiation Monitor RM-32 STP R-3A, Use of flux Happing Equipment (Unit 2)

STP H-11B, Heasurement of Station Battery Voltage and Specific

Gravity (Unit 1)

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Established (Unit 2)STP H-45B, Containment Inspection When Cont

STP H-51A, Routine Surveillance of Containment Fan Cooler Unit

(CFCU) for Reverse Rotation (CFCU 2-4)

STP 1-18W2D. Revision 5. Isotooic Calibration of Plant Vent Hid a.

Rance lodine Radiation Monitor RM-32 During review of this surveillance activity, the licensee noted that several In:trumentation and Control (I&C) calibrations had been

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performed during the TS grace period (25% beyond the surveillance interval specified in the TS.

The 'Tspector was informed by 1&C technicians that routine use)of the grace perivd was being sto by management, and that surveillance activities have more recently been and will in the future continue to be performed within the TS interval.

document QE Ho. 9766, dated MayThis effort was documented in p 27, 1992.

The inspector reviewed a sample of recent 11C surveillances, and found that more recent sur-veillance tests had been accomplished within the TS interval, rather than during the grace period, as was the case for some surveillance activities performed more than six months ago.

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"itiative was noted by the inspector to reflect a more conservative approach to the scheduling of surveillance activities.

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Procram to Unorade Radiation Monitor Calibr> tion Procedures Tb inspector observed implementation of a review of radiation ecnitor calibration procedures to identify vague or' conflicting guidance.

and several action requests.The licensee had identified this effort in an a appeared appropriate.

The review and associated effort part of routine resident inspection of surveillance activities. Fo i

No violations or deviations were identified, 7.

Enaineerina Safety Feature Verification (71710(

During the inspection period, selected portions of the safety injection system for Units 1 and 2 were inspected to verify that s tion, equipment condition, valve and e,ectrical lineups,ystem configura-breaker positions were in accordance with plant drawings and Technica and local Specifications.

No violations or deviations were identified.

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Iramoles_of inadeavate Problem Identification Plant Construction Obscures 10 CFR 50. Accendix R. Lichting a.

underway to strengthen masonry walls.in Unit 1, Buses G and was hung from floor to ceiling along the length of one wall to pr tect the switchgear from debris generated by the construction work On May 19, 1992, lighting for both rooms was covered by the tarps.the inspector

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operator actions in the event of a fire.ing is installed in the 4K Emergency light-and EP M-10, Fire Protection of Safe Shutdown E

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The licensee documented the problem in AR No.266808, and replaced of the lights with transparent plastic.an approximately 6 f t by 5 ft to evaluate operability of the lighting, and assigned systemThe licens m __

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engineering responsibility for Appendix R lighting to the fire protection and emergency services organization.

Appendix R,Section III.J requires emergency lighting in all areas needed for operation of safe shutdown equipment. The operability of the lighting, and root cause of the failure to identify and correct the problem until noted by the inspector, will be followed as Unresolved Item 50-275/92-16-02.

b.

Contradictory Toraue Documentation for Residual Heat Removal (RHR)

System Valves During review of the evaluation section of Action Request No.

230770, which identified recent indication of leakage fram Unit I valve RHR 8702, the inspector noted that a second problem was dis-cussed concerning evaluation of inconsistent torque ~ documentation.

Torques of 1500 and 2500 ft-lbs for bonnet fasteners were listed in separate reference documents.

.This inconsistency was not identified in a separate Action Request as required, and therefore may not have been re iewed by Quality Assurance.

Af ter the inspector identified the issue, the licensee acknowledged that the lack of a separate problem identification document was inappropriate, and initiated a second Action Request (267237) to identify the problem.

The inspector was concerned that, based on ' inconsistent information, fasteners with incorrect torque may have been installed in the pl ant.

Licensee investigation determined that the valve manufac -

turer initially recommended 1500 ft-lbs, but had later offered different valve trim designs.

The manufacturer then specified 2500 f t-lbs torque because it bounded all design options for this valve line, and documented this value in lata vendor manuals.

Recent vendor correspondence validated this assessment.

Installation records from 1973 for original design valves showed torques of 1800 ft-lbs, which the licensee considers an acceptable torque value.

Additionally, the licensee stated that these valves are specifically inspected for boric acid leakage at least every outage during sur-veillance test STP R-8C, Containment Walkdown for Evidence of Boric Acid Leakage, to comply with the licensee's response to Generic Letter 88-05 concerning boric acid corrosion of fasteners.

Because no leakage has been previously observed, the licensee concluded that this is further assurance that fastener. torque is adequate,

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The licensee stated that this issue had not been identified in

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past because no work had been performed on the two valves of concern, RHR 8701 and 8702, since original installation.

The failure to promptly identify a problem adverse to. quality is considered to be a violation of 10 CFR 50, Appendix B.Section XVI.

Because the torque inconsistencies did not appear to be safety significant, and because licensee corrective actions were prompt,

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this violation is not being cited because the criteria specified in Section V.A. of the Enforcement Policy were satisfied.

(NCY 92-16-03, Closed)

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Uncontrolled information in Electrical Panels c.

During routine inspection activities, the inspector observed uncontrolled, outdated, and in some cases inaccurate drawings and circuit lists attached to the inside of several electrical breaker panels.

The presence of uncontrolled information in electrical panels, particularly in Class 1 panels, was of concern.

After.

discussion with the inspector, the licensee determined that this information was inconsistent with the requirements of Administrative Procedure AP C-55, and indicated lack of a formal program for updating electrical labels as part of the design process.

Licensee work control and operations procedures reautre use of controlled drawings and information for work associated with these panels.

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l Inspector observations indicated that the existence of these

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uncontrolled drawings did not appear to have encouraged the licensee l

to circumvent the use of controlled drawings.

significance of uncontrolled drawings in panels appeared low.Therefore, the Removal of uncontrolled information from inside electrical panels will be followed by. routine inspection activities.

One non-cited violation was identified (paragraph 8 b).

9.

Potential tieakness in Understandina the Recirculation phase Desion Basis

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In recent months, the inspector has noted the occurrence of five issues which appear to have been partially caused by insufficient awareness of

design basis requirements for the ECCS recirculation phase.

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follows:

They are as Diaphragm Valve Leakage May Have Exceed Part 100 Limits (LER

92-09).

DC 0-92-tis-fiOO7). Charging Pump and Safety injection Pump Runo

Leakage Path Vulnerability Via volume Control Tank (VCT) Outlet

Check Valve (LER 50-275/92-01).

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Inaccurate Single Failure Analysis Allows the Possibility that Com-Bearing Temperature, Limits may be Exceeded (LERpon 50-275/91-18).

Inaccurate Single Failure Analysis Allows the possibility that Long

Term Containment Temperature Profile Assumptions may be Exceeded to Lack of Containment Spray (ifCR DCO-92-Eti-fiOO2),

The inspector identified to the licensee that these issues may indicate insufficient understanding and integration of ECCS recirculation phase design requirements, with the potential for additional safety issues which have not yet been identified.

The licensee agreed to consider and evaluate this possibility.

The licensee identified that'this may be a:

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generic Westinghouse design issue, and was planning to explore this issue with the Westinghouse Owners' Group.

evaluation will be tracked as followup ItemFollowup of the licensee's 50-275/92-16-04.

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10.

i Licensee Event Report followun (92700)

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Unit 1 Control Room Ventilation System Outside Desian Basis a.

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JLER 50-275/83-39. Revision 0). Closed

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On November 21, 1991, the licensee identified that the control room

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l ventilation system could be outside its design basis as a result of

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a single failure of a booster fan or its associated damper.

i The fans are positioned in parallel flow paths, and failure would allow i

unfiltered air into the control room during the pressurization mode.

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i As a temporary corrective measure, the licensee issued Operations i

Department night orders to the control room discussing-the vulnera-i

bility, and installed flow indicators on ducts in the control room.

The licensee also changed emergency operating procedures to require

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the ducts by observing the flow indicators. verification that

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The licensee has prepared and scheduled a design change to provide

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permanent resolution of the vulnerability.

Therefore, this item is closed.

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b.

Unit 1 Check Valve Inservice Testino Deficiencies (LER,_50-275/84-44.

Revision 1). Closed

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As a result of deficiencies in the check valve in

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i testing in the closed direction.

Identification of these valve.;

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occurred after review of a Westinghouse Infon..ition Letter dated August 4, 1989.

The valves, MS-5166 and -5177. are in the steam

Testing was in place to verify the valves opened

j when required, but not that the valves closed in order to prevent an

intact steam generator from blowing down through the AFW pump s supply line to containment after a main steam line break.

i approved a relief r..aest to allow the licensee to disassemble and The NRC i

inspect the valves on a rotating refueling outage frequency.

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licensee has added this testing and inspection requirement to the The i

IST program.

This item is therefore closed.

e Unit 1 Emeroency Diesel Generator (EDG1 1-1 Failure to Start c.

150ecial Reoort 50-275/09-01.

l Closed Revisions 1 and 2. Open item 89-01-X0)

On February 1, 1989, during a routine surveillance procedure, EDG

1-1 failed to start.

Two of the four air start motors failed as a j

result of overload of the pinion retainers, pinion retainer bolting i

and rotor shafts.

The licensee entered TS action statements i

corresponding to an inoperable diesel generator, and commenced

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investigation and repair.

failure to inadequate fit-up of the pinion gear and rotor sha during plant maintenance, resulting in improper torque transmittal.

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The licensee identified several possible contributory causes, including excessive Train "A" starting air pressure, improper vendor

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tolerances for slot fabrication of the pinion retainer, inadequate vendor assembly instructions for the motor, and an inadequate pinion retainer bolt locking mechanism.

provided inspection results for the remaining 18 air start m which identified 11 additional air motors with incorrect fit-up and 10 with rotor cracks.

The licensee documented corrective action for each of the a the diesel air start configuration. listed causes, and perform

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air start system was ade The licensee concluded that the actions, was acceptable.quate and, in conjunction with corrective

i The licensee discussed their findings with the vendor.

indicated that they were not aware of any similar failures.The vendor licensee informed other licensees of their findings through an The industry problem communications system.

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subject of NRC Information Notice 89-84. This issue was also the l

l The above actions appeared appropriate This item is closed, d.

Unit l_FJel Handlino Buildina Ver.tilation Inocerable Durina Movement _ (LER 50-275/89-19. Revision 1). Closeg

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On Januvy 18, 1991, the Unit I fuel handling building failed to The licensee determined that the fuel handling bu

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system had been inoperable since October 15, 1989, when fuel

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movement occurred during the Unit I third refueling outage.

leakage to be gaps in sheet metal siding, unse penetrations missina or degraded door seals, gaps between the movable crane, wall and the fixed walls, through-wall oxidation o sheet metal siding, and reduced flow through the ventilation ducts

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due to accumulation of debris.

Because seali boundary provided adequate negative pressure,ng of the pressure the root cause was considered to have been degradation of the building seals.

Based on observation of the implementation of the design change the fuel handling building seals and licensee action which has maintained continuing awareness o,f fuel handling building ventilation operability, this item is closed.

Jnocerable Unit 1 Valve Due to limitoroue Sprino Pack Relaxation e.

(LER 50-275/90-09. Revision 0). Closed In March,1990, during overhaul of a safety injection system valve i

a spring pack for a motor operator was found to have a less than

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expected preload.

valves, and was verified to be the result of spring pac dation.

Nine valves in each unit were identified as potentially affected.

Update Bulletin which indicated that valve models o

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i may also have spring pack relaxation.

After review of maintenance j

records and valve diagnostic information, the licensee determined that no additional valves were affected at Diablo Canyon.

The

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licensee replaced each of the potentially affected springpacks, and evaluated the safety significance of the spring packs which had been i

identified as degraded.

Based on the as found evaluation, the i

licensee concluded that the affected valves would have performed j

their safety function.

corrective action, this item is closed. Based on this assessment and th

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j Failure to Perform Hourly Unit 1 Fire Watch -(LER 50-775/91-15.

Revision 0). Closed

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j On September 17, 1991, a roving hourly fire watch was not performed

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in some of the safety related equipment rooms. The fire watch was delayed leaving the radiologically controlled area, and was not able

to contact his supervisor in time to prevent violation of the Technical Specification requiren.ent.

To corrective actions, an incident summary was prepared and

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reviewed with fire watch personnel, and included in fire. watch

training.

Written instructions have been provided to personnel for O

actions to take if delays occur on rounds.

Therefore this item is closed.

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Inocerable Unit 1 EDGs While Fuel Pool Crane Was Operatino Over Fu g.

j Pool (LER 50-275/91-14 Revision 0). Closed

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On February 13, 1991, with the core offloaded, all three EDGs beca i

inoperable.

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During this time, irradiated fuel was moved in the

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spent fuel pool.

Although a specific operational mode was not in force, the licensee determined that the intent of TS 3.8.5.2.b.

(Electrical Power Requirements) was not met, since loss of offsite

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power would have resulted in loss of fuel handling buil."ag venti-j lation.

The licensee returned EDG 1-2 to service on February 13.

For corrective action, the licensee issued procedural guidance to i

interpret electrical power requirements to include operational i

conditions when the core is offloaded.

i This corrective action appeared appropriate.

This item is closed.

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Missed Rod Position Indication Surveillance. Unit 1 (LER

9]-17. Revision 0). Closed.

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j from November 14 to llovember 16, 1991, rod position surveillance tests were missed as a result of inadequacies in the plant process

computer (PPC) software.

i After the PPC was shut down and restarted breakers' status as open.on !!ovember 14, the PPC incorrectl i

the alarm function of the rod position deviation monitor.Th a

corrective actions included assessment that rod position had been Licensee j

PPC software, operator training to identify rod p j

monitor inoperability, and revision of procedure AP C-27, Comput i

Software Quality Assurance, to ensure significant computer functio

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are verified on an appropriate schedule.

appropriate, and this item is closed, These actions appeared i.

Unit I low Vacuum Turbine Trin and Subsecuent Reactor Trio D Procrammatic Deficiency (LER 50-275/92-04 Revision 0). Open.

This event is described in paragraph 4.a of this report.

will be icf t open for followup of licensee corrective actions The LER related to posting local procedures.

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Inonerable Unit 2 Fuel Handlino Buildino Ventilation System Durina (

fuel Movement (LER 50-323/90-02. Revision 2). Closed As a result of inadequate planning and understanding of fuel handling building ventilation system operability requirements, the licensee moved fuel in the spent fuel pool while the fuel handling ventilation boundary doors were blocked open with temporary hoses used for outage work.

l was not adequately maintained.As a result, the required negative pressure

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The licensee reviewed this event

with Operations and Maintenance personnel, and posted signs on all affected ciocrs announcing the need to ensure that doors are closed during fuel 4.ovement.

therefore, this item is closed.This corrective action appeared adequate; be addressed in Followup ItemCorrective action for the generic l

50-275/91-27-02, documented later in t

this report.

This item is therefore closed.

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Two Unit 2 Main Steam Safety Valves Set With Inoperable Test l

fouipment (LER 50-3?3/91-02. Revision 0). Closed On August 26, 1991, main steam safety valve setpoints.the licensee performed routine verificat it became apparent that the test equipment had failedAfter two valves we had been adjusted +3 flats, rougP, a 33 pounds per square inch First, FV-60

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(PSI) change in setpoint.

roughly corresponding to 143 psi setpoint change.Then, RV-225

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Because of the unexpectedly high magnitude of adjustment, the licensee suspected and promptly confirmed improper performance of the test equipment.

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Yalve RV-225 was declared inoperable.

l declare valve RV-60 inoperable, based on the small, expecte tude of adjustment.

Subsequent testing 4 psi above its +/- 1% tolerance range, found that RY-60 lifted while setting RV-225, and had been operable wh

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l The licensee also concluded that there was no firm evid since setpoint drift is common and expected up to +

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RV-60 was not declared inoperable after test equ Because confirmed, the NRC concluded that this was a nonipment failure was (50-323/91-24-02),

cited violation based on the low safety significance of the as found setpoint of RV-60, the licensee's prompt corrective action to

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points, and continuing improvement in documentation of the rationale for operability decisions.

Therefore, this item is closed.

1.

Jnadvertent Removal of Power From Unit 2 RHR Puen_Sumn Suttign Valve Op.tra_ tors Before Enterina Mode 5 (LER 50-323/91-03. Revision 0).

Closed On September 1, 1991, control room operators determined that both RHR sump suction valves and both containment spray (CS) pumps were inoperable.

The RHR valve breakers hi.d been opened locally, and control power to the CS pump breakers had been removed.

Both events occurred after entry into hot shutdown earlier in the day, and were corrected within 15 minutes after identification by operators.

For the RHR valves, power must be removed from the valves at the control room contactor toggle switches (rather than the local breakers) to satisfy safety analysis requirements that power be able to be restored from the control room.

Procedure OP L-5 required power be removed from the valves at the local breaker, thus violating safety evaluation requirements.

L-5 and reviewed all other L-series procedures.The licensee revised OP For the Containment Spray pumps, power was removed before entry into Mode 5 as a result of inadequate control and review of clearances by a Senior Control Operator.

Specific licensee corrective actions included issuing an Operations incident Summary and memoranda stressing the significance of this event and the need for ensuring TS commitments are met.

l action for the generic concern of personnel errors will be addressed Corrective in followup Item 50-275/91-27-02, This item is therefore closed.

documented later in this report.

Shift of Unit 2 Control Room Ventilation Mode (LER m.

Revision 0). Closed 50-323/91-06._

On October 3,1991, an inouvertent engineered safety features (EST)

actuation occurred.

two operators who lifted an incorrect lead while ittempting

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remove an inverter from service.

Power was interrupted to inverter lY-23, resulting in a shift of the control room ventilation system to the pressurization mode.

expected.

All safety systems functioned as The operators followed generic procedure guidance instead of detailed information provided in an attachment.

As corrective and the procedure was revised to delete the generic emphasize detailed information.

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l Corrective action for the generic concern of personnel errors will be addressed in followup Item 50-275/91-27-02, documented later in this report.

This item is therefore closed.

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_ Inadvertent Unit 2 Safety In_iection While in Mode 5 (LER 50-3?3/91-

,07. Revision 0). Closed On October 6, 1991, during a refueling outage (Mode 5), technicians inadvertently caused a safety injection signal by performin prctection switch positioning out of the correct sequence. gThe cause of the event was inattention to detail. All expected ESF actuation signals occurred. No water was injected into the core, since emergency core cooling pumps were secured for maintenance.

The technicians were counselled according to the licensee's positive discipline program, and a memorandum was issued by the Plant Manager emphasizing the need for procedure ccmpliance and verification.

Lorrective action for the generic concern of personnel errors will be addressed in Followup Item 50-275/91-27-02, documented later in this report. This item is therefore closed, linit 2 Valve leakace Hav Have Resulted in Exceedino 10 CFR 100 Do o.

RER 50-3?3/91-09. Revisions 0 and 1). Closed 09 October 4, 1991, 1.3 gallons per minate leakage from diaphragm vsives CVCS-2-8471 and -548 was evaluated by the licensee.

The evaluation concluded that the control room and exclusion area boundary 10 CFR 100 limits could have been exceeded during the recirculation phase of a loss of coolant accident (LOCA).

The licensee determined that the diaphram service life for CVCS-2-8471 had been exceeded and that diaphram degradation had resulted in leakage. The LOCA analysis had not identific CVCS-2-8471 as a valve in the recirculation phase flow path, as a result, the valve was not included in the licensee's preventive maintenance program for diaphragm replacement.

had not been retorqued at normal operating pressure a following maintenance, as was required for that type valve.perature

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The licensee ccamitted to revise the procedure to include this step.

While this specific issue is closed, the inspector was concerned that other issues concerning the design requirements of the recirculation phase may not have been adequately-addressed, and documented the n' ecd for NRC followup of this concern earlier in this report.

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Jnprerable Unit 1 Vide Rance Containnient Sumo level Indication p.

JLfR 50-3?3/91-10. Revisions 0 and 1). Closed On October 22, 1991, with Unit 2 in Mode 2 at Ot power, TS Action Statement 3.3.3.6 was exceeded due to the reactor cavity sump wide

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channel was returned to service about 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> later. range The

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failure to met the TS action statement was caused by inadequate corrective action for a previous similar occurrence.

During investigation of a similar failure on March 15, 1992, the licensee determined that only the zero range was affected by the failure, and all the ranges required for the indicator to fulfill its safety function were operable. The licensee considers previous failures to be similarly of low safety significance.

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In Inspection Report 50-323/92-01, the NRC issued a violation concerning the inoperable sump level indication.

This item is closed.

Failure to Test a Unit ? ECCS Valve After Maintenance (LER 50-323/

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91-1). Revision 0). Closed

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On October 17, 1991, while in Mode 5, the packing on ECCS valve

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SI 2-8802B was adjusted and the packing gland follower nuts were torqued to a value greater than the torque used for the last successful stroke time test.

Both these operations result in requiremc..ts to perform a post-maintenance test before returning the valve to operable status.

However, neither requirement was immediately recognized. During mode change review on October 22, the deficiency was recognized, and the test was performed.

As corrective action, training on mode transition was provided for the Engineering Test Group, an Operations event. report was issued, and work planning issued enhanced work policies.

Corrective action for the generic concern of personnel errors will be addressed by Followup Item 50-275/91-27-02, This item is therefore closed, documented later in~this report.

11. Doen item followup (927031 Public Address (PA) System Powered by Non-Vital Circuitry a.

(Followuo Jtem 50-275/91-09-03). Closed During the loss of offsite power event on March 7,1991, operators were unable to quickly inform workers in the' containment of plant status because the PA system was powered from non-vital power.

Containment was evacuated in an orderly manner without use of the PA system.

The licensee plans to implement control of PA system power sources during outages.

item is closed.

Based on planned licensee actions, this b.

EDG Slow to load (Followuo item 50-275/91-09-04). Closed This item discussed the failure of EDG 1-1 to load onto the vital bus within 10 seconds during the loss of offsite power event on March 7, 1991.

After NRC review of the circumstances, the licensee received a violttion associated with determinstion of the validity of the start failure (50-275/91-07-10).

i No other issues were identified. Therefore, this item is closed.

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Refuelina Procedures Chanon_(Followuo item c.

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50-275/91-09-05). Closed Juring the loss of offsite power on March 7,1991, a concern was

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a loss of offsite power. identified regarding the safety of fuel asse The licensee identified a comitment based on INP0 SOER 85-01 to change procedures to include actions to place fuel assemblics in a safe location durin cooling concurrent with a loss of power.g loss of refueling cavity.

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The inspector reviewed Revision 11 of procedure OP S-8DS2, Core Loading Sequence, which was revised to include a step which does not allow a fuel assembly to be li'ted from the spent fuel pool until the previous assembly has beer unlatched in the core.

The inspector verified that a similar requirement existed in the core unloading procedure. Based on the licensee's corrective action, this item is closed.

d.

Inonerable Wide Rance Reactor Cavity Sumn (Followup 01-03). Closed Item _50-323/92-The NRC issued a Notice of Violation due to inoperable wide range reactor cavity sump level indication.

has successfully identified and corrected failure of the sump leve indication on two occasions, as a result of more attention to Technical Specification operability requirements.

As permanent corrective action, continued troubleshooting and repair work on the level indication has occurred during unscheduled outages.

change has been reviewed for installatien during the next scheduledA

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outage.

planned corrective action, this item is closed. Based on th

Increase in Personnel Errors (Followun item e.

50-275/91-27-021. Ooen number of personnel errors, foiiow of correctiv

iten will include the generic corrective actions for licen reports documented above which were caused by human error.see event tion of successful actions will include reduction of the n Indica-i LERs caused by human error. This item remi. ins open.

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Spurious Actuation of Carbon Dioxide Fire Suporession System in J-3 Room (Unresolved Item 50-275/92-05-01). Closed

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dioxide (cardox) suppression system actuated.During The ventilation for generator coolitig was then cut off as a result of the cardox system closing the west rolldown doors.

Operators shut down the diesel to l

prevent overheating.

failure of the diesel. The NRC questioned whether this was a valid Because the root cause is still under failure evaluations in response to an NRC notic issue is closed.

Followup of the issue of validit

, this be part of routine resident inspector activities. y of failure will

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Comnariton of ticuid Radwaste Analysis Results (followuo item 10-275/88-33-01). Closed This item concerned an inter-comparison of iron-55 (Fe-55) activity in a liquid radwaste sample that was split between the licensee and the NRC contract laboratory. The initial inter-comparison of analytical results for Fe-55 did not agree, and a followup inspection was conducted at the licensee's off-site laboratory.

The inspector found that the licensee's laboratory sampling and measure-ment techniques were fundamentally sound.

To resolve the issue, the inspector, in coordination with the licensee, devised an experiment to independently verify both NRC's and the licensee's analytical methods for determining Fe-55 in liquid radwaste.

A liquid radwaste sample, spiked with FE-55, was split between the HRC contract laboratory and the licensee. The amount of FE-55 activity was known only by the inspector.

Based on the results of the inter-comparisons, the inspector concluded that the licensee's method for analyzing liquid radwaste for FE-55 is satisfactory. This item is closed.

12. Followup on Unresolved items (927011 Auxiliary Salt Water (ASW) Floodino Due to Failure to Follow Proce-a.

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dures (Unresolved Item 50-275/91-24-01). Closed On August 2,1991, flooding of the component cooling water heat exchanger area occurred as a result of maintenance workers' failure to follow clearance instructions in the order specified.

Review of the safety significance of this issue determined that the other trains of CCW and ASW were available and would have performed their safety functions.

Licensee corrective actions included issuing a maintenance event sumary, counseling the individuals involved using positive counseling, and emnhasizing the personal and plant safety aspects of properly following clearances.

The failure to follow clearance instructions appears to be a viola-tion of Technical Specification 6.8.1,~ which requires that work be implemented by procedures. Based on the licensee's corrective action and the low safety significance of the concern,-the violation is not being cited because the criteria specified in Section V.A. of the Enforcement Policy were satisfied (92-16-05, closed),

b.

Senaration of Safety Related Electrical Circuits (Unresolved item 50-275/91-11-01). Closed The licensee had previously identified that two independent Class IE i

direct current (dt) supplies were electrically connected in a non-l Class IE fire horn relay panel.

within the panel.

There was no electrical separation

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The licensee concluded that connection of the two Class IE de supplies in the relay panel was within the separation criteria established by their license.

An inspector also previously noted that both Class IE cables and non-Class IE cables were being routed through a new Conax penetra-tion.

The licensee stated that they were using Regulatory Guide (RG) 1.75, " Physical Independence of Electrical Systems," when feasible during design modifications.

were using coax cables, for which there were no spare penetrations.

Therefore, the licensee concluded that separation per RG 1.75 was not feasible.

The inspector determined that the routing of both Class IE cables and non-Class IE cables through the new Conax penetration would be further reviewed for acceptable electrical separation.

The review would be performed in conjunction with the review of the accepta-bility of electrical separation within the fire hora relay panel.

The inspector evaluated both of the previously identified items discussed above.

predated RG 1.75. The inspector noted that the Diablo Canyon design 308-1971, " Criteria for Class IE Electrical Systems for Nucle Power Stations." Paragraph 5.3.5 of IEEE 308-1971 stated that protective devices shall be provided to isolate failed equipment automatically.

The inspector reviewed the circuit design and concluded that the Diablo Canyon design for protective devices was Class IE de supplies. adequate to prevent faults in non-lE Class pane The inspector al.o concluded that the installation of bth Class IE criteria 'or Diablo Canyon.and r.on-Class IE cables in a new Cona The inspector concluded that lack of a spare penetration was a reasonable evaluation for concluding that

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i installing the new cables per RG 1.75 criteria was not feasible.

This item is closed.

4 Kilovolt Switchaear Fault Current Ratino (Unresolved Item c.

and 50-323/91-07-01). Closed The Electrical Distribution System functional Inspection identified ratings during certain plant operations.that calculated fault cu exceeded switchgear ratings when one or more emergency dieselCal generators (EDGs) were operated in parallel with the main generator.

The licensee took action to minimize the tests which required para-llel operation of the main generator and one or more EDGs.

licensee also took action to minimize the fault capability of the The main generator during parallel operation with an EDG.

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The licensee perfonned a calculation which showed that a maximum (bolted) fault was a low probability event during the limited time the main generator was operated in parallel with an EDG.

The NRC staff calculated the core damage frequency caused by the

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occurrence of a bolted fault during a time when the main generator was operated in parallel with an EDG.

The calculation showed that the core damage frequency caused by this bolted fault was in the order of 1.0E-9 per reactor year or_ less.

The inspector reviewed the licensee's actions and concluded that these actions to minimize the risk from a bolted fault were adequate.

The inspector reviewed the licensee's calculation and concluded that this calculation was adequate to demonstrate that a bolted fault during parallel operation of the main generator and an

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EDG was a low probability event. The inspector also concluded that the core damage frequency calculation demonstrated that the bolted fault condition discussed above had little safety significance.

This item is closed.

13.

Verification of As-Built Drawinas (37051 and 500731 On April 24-25, 1992, the inspector entered containment to examine containment fan cooler damper 2-4 to verify the licensee had restored The inspector used the following documents as the basis

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inspection.

American Warming & Ventilating, Inc., Drawings SHW-D-9098, PG&E

Revision 6, record 663079, Sheet 37 81010-001-000 Revision A, PG&E recor;dSHW-D-9099, Revision B; and 663079, Sheet 75 Design Change Notice DC2-EH-44664, CFCU Backdraft Damper Helpe

Springs Non-Conformance Report DCO-92-MM-N007, Containment Fan Coolers

The inspector observed that the backdraft damper appeared to conform to the design documents with two exceptions.

First, several of the Allen

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vane linkage arms appeared loose in the holes in the

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I did not contact the rubber block shock absorbers whe closed. The gap was 1 specify any clearances /16 inch or less.

related to these observations.The design documents did not When questioned by the inspector, the Mechanical Hair.tenance Engineer

showed the inspector an Action Request which had been written regard the loose fit of the shoulder bolts in the linkage arms.

concluded that this deficiency would not have preventeo the damper from The licensee working and planned to completely overhaul these dampers during the next refueling outage.

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The inspector re-entered containment to observe the surveillance testing (M-51A) of damper 2-4 to determine if the lack of contact with the shock absorber block would affect the operability of the damper.

observed that : tarting and stopping the associated containment fan coolerThe insp unit caused the damper to swing open and shut over a period of several seconds.

would not be significantly affected by the counterweights the shock absorber block.

The inspector requested that the licensee evaluate the effect of the observed gap on the post-LOCA performance of the dampers.

The licensee evaluated the observation and concluded that during a LOCA with the gaps measured by the licensee, the flexibility of the linkage,

mechanism and the dampers was sufficient so that the shock absorbers would atsorb any significant shock to the linkage arms.

No violations or deviations were identified.

14.

Infomation Meetino with local Officials (94600)

i On May 4 County Administrator, Deputy Administrator, and Emergen Coordinator.

The purpose of the meeting.was to introduce the NRC

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inspectors, to offer the NRC's openness to discuss inspection report findings with the County, and to discuss general issues of emergency planning.

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15.

Exit Meetina An exit meeting was conducted on June 12, 1992, with the licensee repre-sentatives identified in Paragraph 1.

The inspectors summarized the

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scope and findings of the inspection as described in this report.

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