ML20138D156

Insp Rept 50-440/97-02 on 970204-0321.Violations Noted.Major Areas Inspected:Operations,Maint & Engineering

ML20138D156
Person / Time
Site:	Perry
Issue date:	04/23/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20138D089	List: ML20138D084 ML20138D123 ML20138D156
References
50-440-97-02, 50-440-97-2, NUDOCS 9705010049
Download: ML20138D156 (22)

See also: IR 05000440/1997002

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. U. S. NUCLEAR REGULATORY COMMISSION

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REGION lli

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Docket No: 50-440

License No: NPF-58 -

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Report No: 50-440/97-02 -

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Licensee: Centerior Service Company

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Facility: Perry Nuclear Power Plant

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Location: P. O. Box 97, A20O

Perry, OH 44081

Dates: February 4, through March 21,1997  !

Inspectors: D. Kosloff, Senior Resident inspector

R. Twigg, Resident inspector

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Approved by: J. M. Jacobson, Chief, Projects Branch 4

Division of Reactor Projects

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9705010049 970423 TL

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EXECUTIVE SUMMARY

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Perry Nuclear Power Plant, Unit 1

NRC inspection Report 50-440/97-02

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This inspection included aspects of licensee operations, engineering, and maintenance.

The report covers a 6-week period of resident inspection.

j Ooerations

e There was good control of minor plant transients caused by equipment failures:

loss of LF-1-C resulted in power reduction due to bus duct cooling fault

(Section 02.1), rod control transformer failure (Section 01.2), and minor FCV

fluctuations (Section 04.1). There were no significant personnel errors.

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1 e Three-legged communications and control room annunciator call-outs remained

i excellent (Section 01.1). Projob briefings improved (Section 01.1) and there was

excellent control of circuit breaker polarity testing (Section M1.1).

e Management involvement in operations improved with active participation at on-

shift briefings and training sessions (Section 01.1). There was good reinforcement

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of management expectations related to work being performed in other EDG rooms

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with Div 2 EDG out of service (Section 01.1). However, the inspectors identified a

non-conservative equipment restoration decision (Section M4).

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Maintenance

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e There was good field performance of testing activities, including some improvement

in procedures and briefings. However, some weaknesses remained in follow

through for identified problems (Section M1),

e Although there have been significant improvements in material condition, two

observations indicated equipment problems needed continued vigilance. A

! Division 2 outage was well planned and executed (Section M2).

e There was good maintenance support for restoration of equipment after transients

(Sections 01.2, 02.1, and 04.1).

e During planning for maintenance of an EDG fuel oil transfer pump, non-conservative

_ schedule was not identified. Also, a related deviation from an industry standard

was not initially evaluated (Section M4).

Enoineering

! e Inspector observations of testing led to questions concerning Unit 2 abandonment

and its impact on Unit 1 (Sections E1.1 and E2.2.2).

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e Surveellence failures of the EDG Testable Rupture Discs indicated possible design

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problems (Section E2.1).

e Safety related transformer evaluation was good, identified possible shortened

service life of 4160 to 480 VAC transformers (Section 08.3).

e The inspectors identified a violation of 10 CFR 50.59 after the emergency closed

, cooling system was modified to improve plant safety (Section E.3).

, o Several discrepancies were identified during an inspector walk down of emergency

food, water, and medical supplies located in the control room envelope as described

in the USAR (Section E.2.2.1).

e There was good follow through on breaker L1108 (LF-1-C) ground relay fault

(Section 02.1) and FCV control signal fluctuations (Section 04.1).

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Resort Details

Summary of Plant Status  !

, The plant operated at full power until February 11,1997, when power was reduced due to

an electrical bus duct cooler problem. Power was increased again on February 12. For the 3

remainder of the inspection period the plant operated at full power except for short power l

reductions for testing and control rod realignments. l

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l. Operatiorw

01 Conduct of Operations

01.1 General Comments (71707)

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Using inspection Procedure 71707, the inspectors conducted frequent reviews of l

ongoing plant operations. Conduct of operations continued to be safety-conscious. )

Verbal communications in the plant continued to be excellent, as shown by j

consistent use of three-legged communications and verbal acknowledgment of all )

control room annunciators. Control of reactivity changes improved with direct  ;

senior reactor operator oversight of all planned reactivity changes. The frequency  ;

of projob and protest briefings increased. The content of the briefings was  !

improved, while personnel participation continued to be good. Examples included

briefings for breaker polarity testing, residual heat removal system testing, and

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electrical switching activities. Frequently, operations management actively

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participated in shift turnover briefings and shift training sessions. During

maintenance on the Division (Div) 2 Emergency Diesel Generator (EDG), operations

j management clarified and reinforced management expectations by writing a

potential issue form (PIF) on failure to prevent concurrent maintenance on

equipment in the Div 2 and 3 EDG rooms. The inspectors confirmed that the Div 2

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and 3 maintenance had no impact on EDG operability. The inspectors observed

that operations management provided immediate follow through on the PlF by

encouraging open and positive discussion of the issue at the next managers'

meeting. A failure to identify nonconservative control of maintenance is discussed

in Section M4.

01.2 Failed Power Sunolv Caused Loss of Monitorina Parameters for 32 Control Rods

a. Insnection Scone (71707,92901)

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The inspectors observed the licensee's response to a failed power supply for the

rod control and information system (RC&lS).

. b. Observations and Findinas

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At 12:17 a.m. on February 18, the control room received an alarm indicating that l

control rod motion was inhibited. The operators entered Off Normal Instruction ONI i

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C11-1,' " inability to Move Control Rods." investigation revealed that one of

i 16 power supplies associated with the RCailS had failed. Control room remote

indication of the scram valve position and hydraulic control unit accumulator

4 pressure and water level were lost for 32 of 177 control rods. Technical

j Specification (TS) 6.1.2 was entered and all 32 rods were declared inoperable.

Also, in accordance with TS 3.1.4, the operators entered a 12-hour to hot

i shutdown action statement at 1:17 a.m. The power supply was replaced and the

action statement was exited at 11:03 a.m. Other options included starting a

1. reactor shutdown by reducing flow at 11:30 a.m. followed by a scram, or

i completing a 10 CFR 50.59 change to the TS 6.1.2, located in the Operations *

Requirement Manual of the TS.

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During the removal and replacement of the power supply, 88 control rods could not

be monitored from the control room due to the tagout for the work. Root cause

investigation revealed a failed capacitor in the transformer. At the end of the

inspection period, engineering was conducting an extent of condition review.

j There was a delay of several hours in replacing the transformer because the spare

i transformer was located at a warehouse in Pennsylvania. As a result of the delay,

the licensee began a review of past decisions on which spare parts would be stored

in Pennsylvania and which would be stored on site.

c. Conclusions

Operations, engineering, maintenance, and management responded appropriately to

the event.

! 02 Operational Status of Facilities and Equipment

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02.1 Loss of Las F-1-C Caused ECCS Actuation and Power Reduction

f a. Insnaction Scone (71707. 92901)

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Plant operators responded to the loss of bus F-1-C at 3:50 a.m. on February 11,

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1997. The inspectors evaluated th, Jant staff's response to the event, including

troubleshooting.

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b. Observations and Findinas

The operators identified the breaker (L1108) that had tripped and verified the status

of plant equipment. There had been no personnel near the breaker when it tripped.

The operators determined that main generator isophase bus duct cooling had been

! lost and promptly reduced power. By 4:44 a.m. plant power had been reduced to

58% so that forced cooling was no longer required for the isophase busses.

Maintenance subsequently identified that a sticking flow switch had caused the

standby isophase bus duct cooling fan to fail to provide cooling. Cooling was

restored and the plant returned to full power at 5:36 p.m.

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No problems were identified during extensive troubleshooting of the L1108 breaker

t and the F-1-C bus. The L1108 ground relay, which had actuated with no evidence

of a ground, was considered the cause of the breaker trip. The relay was replaced

and sent to the manufacturer for further testing. The manufacturer determined that

the relay had a fault in its silicone control rectifier (SCR) that could not be detected

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with test equipment that had been available on sts. The SCR was well within its

, expected service life. At the end of the inspection period, engineering was

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continuing its evaluation of the expected service life of the SCRs. Bus F-14C was

energized on February 12 from its alternate source.

i c. Conclusions

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Operator response to the loss of F-1-C was appropriate. Engineering and

maintenance provided prompt support for troubleshooting and restoration of F-1-C.

) The event was appropriately reported to the NRC.

l 04 Operator Knowledge and Performance

! 04.1 Flow Control Valve Oscillations

a. Inanaction Scone (71707 and 37551)

The inspectors observed the operations response to an observed fluctuation in a

i flow control valve (FCV) control signal,

b. Observations and Findinas:

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On February 5, the operators saw an unexpected fluctuation in the indicated 'B'

FCV control signal and requested engineering and maintenance assistance in

! evaluating the problem. An electronic data acquisition system was used to provide

more detailed information on the inputs and responses of the control system and

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position indications. After reviewing the data for several days, the engineers

concluded that small differences in two position monitoring inputs was causing

i- actual periodic movement of the FCV that was too small to be detected by

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observed changes in reactor flow, FCV position, or power indication. Engineering

i established long term monitoring to insure that the movement did not increase

enough to cause premature wear of the FCV.

c. Conclusions

Plant staff responded promptly and appropriately to an unexpected change in a

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plant indication and developed an appropriate long term action plan.

08 Miscellaneous Operations losues (92700, 92901, and 92903)

, 08.1 (Closed) Licensee Event Renort (LER) 50-440/94-022-00: " Failure to Perform

Diesel Generator Testing Utilizing Staggered Test Basis." On November 11,1994,

the licensee determined that testing of the emergency diesel generators (EDG) did

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not meet the definition of Staggere<1 Test Basis as required by Technical

Specification (TS) Surve!!Ionce Requirement 4.5.1.1.2.a. The root cause of this ,

event was an error in developing the eurveillance test program to properly schedule

EDG testing. 'The sched6*T h:J Men based only on required individual EDG

surveillance test frequencies sad had not spaced the EDG testing for the three

divisions equally (Staggered Teit Basis). The schedule for EDG testing was

modified to correct the problem. Tne inspectors verified that testing was correctly i

staggered until the requirement was eliminated when the improved TS were

implemented in July 1996. The requirement was eliminated because it provided

minimal safety benefit. This licensee-identified and corrected violation is being

treated as a Non-Cited Violation (NCV 50-440/97002-01(DRP)), consistent with

Section Vll.B.1 of the NRC Enforcement Policy, NUREG-1600.

08.2 (Clomad) LER 50-440/96-004-00: " Maintenance Under inadequate Tagout Resulu ,

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in Partial ESF Actuation." Personnel errors in judgement and poor communicatio*:

during implementation of the safety tagging and work order processes resulted ir. A

blown fuse, causing an unexpected closure of two containment isolation valves.

The safety consequence of the valve closure was minimal in that the valves'

required safety position was closed. The inspectors reviewed the corrective actions  ;

for this event. Personnel were retrained on the tagout process and requirements.

The weaknesses in decision-making and communications were also addressed.

08.3 (Closed) LER 50-440/96-005-00: " Automatic Reactor Scram Following Auxiliary

Transformer Fsilure." A " sudden pressure" signal from the auxiliary transformer

caused a main generator trip and rsector scram when a balance-of-plant (BOP)

13.8 Kilovolt (Kv) breaker was closed after post maintenance testing. When the

13.8 Kv breaker was closed the BOP 13.8 Kv to 480 volt alternating current (VAC)

transformer (LF-1-C) it supplied failed. The auxiliary transformer should have

survived the resulting electrical transient. However, because of an uninsulated

splice, it also failed. Corrective actions included replacement of the auxiliary

transformer and LF-1-C with Unit 2 transformers. Plant engineers could not identify

j any testing that could detect an uninsulated splice in the replacement auxiliary

, transformer. Since there was no proof that the manufacturer had improved its

manufacturing technique before the Unit 2 transformer had been manufactured,

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eng5eering began procurement of a new auxiliary transformer.

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LF-1-C failed because of corona-related insulation damage. Engineering concluded

that the replacement LF-1-C and several similar load center transformers were

susceptible to corona effects. All of the Unit 1 BOP load center transformers were

, later replaced with new transformers from a different manufacturer. Another plant

l had reported similar failures of two BOP load center transformers and one safsty-

i related transformer. Initial evaluations had indicated that transformers with

! voltages less than 10 Kv would not be affected by corona. Safety-related 4160

VAC to 480 VAC transformers of similar construction were inspected and

! corona-related damage was observed. The licensee also removed a safety 1olated

transformer to determine if such transformers were susceptible to similar insulation

i failure. A destructive examination indicated that there had been no corona-induced

msulation damage. However, there were indications that the secondary insulation

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had degraded more than was expected due to the transformer having been operated

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still determining the expected service life of the safety-related transformers.

08.4 (Closed) Violation 50-440/96011-02: An error in an emergency diesel generator

operating instruction had not been identified or corrected during multiple uses.

Personnel errors and ineffective communication of management expectations were

j the cause of the violation. During simulated use of the same instruction, only two l

! of eleven nonlicensed operators identified the instruction errcr. Corrective actions

j have been incorporated in the corrective action program and include an operations

l self assessment and a follow-up to evaluate its effectiveness. The inspectors

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observed that the heightened awareness of operators to problems in procedures

was reflected by increased identification of errors for correction.  ;

ll. Maintenance l

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M1 Conduct of Maintenance  ;

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M1.1 General Comments l

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a. Inanaction Scone (62707. 61726. 92902) ,

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The inspectors observed all or portions of the following work and surveillance

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testing activities with no concerns identified. Additional items are discussed under

l Observations,

e Surveillance instruction (SVI) D51-T304A, Triaxial Response Spectrum

Recorder Channel Calibration for D51-R160 (Reactor Building Foundation)

e SVI B21-TOO97A, Main Steam Line Condenser Vacuum Channel 'A'

Calibration

e SVI E12-T2003, RHR C Pump and Valve Operability Test

e SVI E22-T2101, Operability Test and Maintenance of Diesel generator

Testable Rupture Disc

e SVi R42 T5202, Weekly 125 V Battery Voltage and Category A Limits

Check (Unit 1)

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e SVI R42-T5220,125 V Battery Category B Limits, Termine! Corrosion and

Electrolyte Check (Unit 1, Division 2)

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b. Observations and Findinas:

During the test activities the inspectors observed that personnel correctly used )

written instructions, that measuring and test equipment used was within its l

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calibration limits, and that test results were appropriately evaluated. Specific

observations were as follows. i

e PTI-P45-P0002 ESW System Loop B Flow and differential Pressure Test.  !

Step 5.2.a and 5.2.b directed the opening of valves 2P42-F05298/5308,

"ESW FLOW TO FPCC HT EXCH INSTRUMENT ROOT VALVE." The actual

field labels were "ECC Loop B Flow" (see Section E1.1). This PTl weakness

was similar to a procedure violation cited in the previous inspection report '

(50-440/96011-02(DRP)) for which corrective actions had not yet been

completed.

e SVI E22-T2101, Operability Test and Maintenance of Diesel Generator

Testable Rupture Disc (TRD). The Division 3 Emergency Diesel Generator  ;

(EDG) TRD lifted at 750,620, and 590 pounds (lbs) consecutively and was l

required to lift at between 357 and 380 lbs. The set point of the TRD was '

adjusted so that it would lift within the required band. During the post

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maintenance test with the EDG running, the rupture disc lifted unexpectedly

(see Section E2.1). Other minor discrepancies were identified with the

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surveillance instruction.

e A high voltage test of the LF-1-C transformer was performed after an

unexpected breaker trip occurred. The activity was handled cautiously and

appropriate independent verification was used for wiring lineups. '

e SVI E22-T2001, HPCS Pump and Valve Operability Test. When this SVI l

' was observed during the previous inspectic,r, period (IR 50-440/96018)

i several problems were identified. The procedure was then completely

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revised. As a result, performance of the surveillance this inspection period

was considerably improved.

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l e The pre test briefing for SVI E12-T2003, RHR 'C' Pump and Valve

l Operability Test, was thorough with good participation by those involved in

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• Operations conducted detailed projob briefings for the 480 VAC circuit

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breaker trip circuit polarity testing. A comprehensive written work plan was

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used during the briefings and the work. The inspectors verified that spare

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breakers had been brought to the plant to be used as replacements for any

breakers with identified problems. Operations provided direct management

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control of the breaker work at the job site. The inspectors observed that the

work was well-controlled with excellent communications. This minimized

the time the plant was in a TS LCO while the breakers were out of service.

i Maintenance personnel were familiar with the testing and promptly

completed their tasks. An engineer was present to provide support if

! needed; however, all breakers tested satisfactorily.

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c. Conclusions ,

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There was go xi field performance of testing activities, including some improvement -

in procedures and briefings. However, some weaknesses remained in follow

through for identified problems. l

M2 Maintenance and Matedal Condilon of Facilities and Equipment l

a. Irsaaaetion kana (82707. 71500. 71707, and 92720)

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The inspectors observed plant conditions during plant walk downs. Maintenance

activities were also observed.

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b. Observations and Findings

The inspectors observed that the noise level emanating from transformer R14SO18

for Vital Panel V-1-B (supplies power to the plant data computer in the control

room) was considerably worse than in August 1996, when Deficiency Tag #39649

had been written because of the noise. The shift supervisor discussed the

following with maintenance and engineering staff the next morning: 1) the

potential iot the transformer to fail and cause a fire (the transformer was located in

a safety related cable spread room), 2) the potential loss of the plant data computer  !

in the control room,3) no action plan was in place for a loss of the transformer and >

no parts were available for repair. Also, there had been considerable discussion

recently of a motor operated valve " hot short" issue which should have made plant

personnel more sensitive to a potential for fire in a cable spread room. The work

order was upgraded to a priority 3, work within 3 weeks.

The inspectors observed replacement of E12-F0581, vent valve for the shutdown i

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cooling (SDC) suction header for both RHR divisions. Replacement of the valve i

l was required due to concerns that the valve would begin to leak because of 1

repetitive use. Leakage and thermal conduction past the 20" diameter SDC suction

j header isolation valve for the last several months forced operators to vent the SDC

suction header about three times a day because of pressure buildup. Similar

1 problems have occurred with the RHR low pressure coolant injection valves as *

discussed in NRC inspection report 50-440/96-18.

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! A Div 2 equipment outage was well planned and executed. All work was

l completed successfully and equipment out of service time was minimized.

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! Replacement of Reactor Feedwater (FW) Booster Pump 'C' which began during the

last inspection period was completed and all four FW booster pumps have remained

i available.

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c. Conclusions

Although there have been significant improvements in material condition, these ',

[ observations indicated continuing equipment problems needing additional vigilance j

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on the part of Perry personnel. Of particular concern were the work arounds

associated with the RHR system leakage and the potential for development of

addetional equipment problems as a result. A successful Div 2 equipment outage

was an indication of improvements in work planning and coordination among plant

organizations.

M4 Maintenance Staff K.w '9 and Performance (62707 and 92902)

a. Scone

The inspectors observed maintenance on a Div. 3 EDG Fuel Oil Transfer Pump.

b. Observations and Findinas

Maintenance on Pump R45 COO 2C (one of two) was scheduled and worked on

Thursday and Friday (February 27 and 28) with tags cleared on Friday. Post

maintenance testing (PMT) was not scheduled until the following Monday. The

ability of the EDG to function after a fuel oil pump failure was degraded during this

time. However, there was no requirement in the Improved Technical Specifications

or the original Technical Specifications to have more than one pump operable. The

original plant safety evaluation report stated that the fuel oil transfer system

conformed to ANSI N195, " Fuel Oil Systems for Standby Diesel Generators,"

except that the fuel oil day tank contained less than the amount of fuel oil required

for 60 minutes of operation. Operations management believed the day tank had

enough fuel oil for 34 minutes of operation. The NRC, in the original safety

evaluation report accepted the design because there were two fuel oil transfer

pumps. However, no requirement to maintain the second fuel oil transfer pump

was included in the Final Safety Analysis Report, only the statement that Perry took

exception to the ANSI day tank size requirement. After the inspectors discussed

this issue with the plant manager, engineering performed an operability

determination. Engineering calculations revealed that the day tank, in fact, had

approximately 64 minutes of fuel supply available.

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c. Conclusions

! Although the fuel oil transfer pump was not required for diesel operability, and no

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regulatory compliance issues were identified, the 2-day delay in performing the PMT

was non-conservative. The inaccurate understanding by operations management of

l the fuel oil transfer systora day tank capacity for the Div 3 EDG (also the Station

Blackout Diesel) indicted potential weaknesses in the overall understanding of the

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plant's design basis.

M8 Miscellaneous Maintenance issues (62707, 92700, and 92902)

l M8.1 (Closed) LER 50-440/94021-00: " Inoperable Div. 2 Diesel Generator Results in a

{ Technical Specification violation and Operating License Violation." On November 8,

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1994, a maintenance mechanic identified that a bridge clamp was installed on the

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Div 2 EDG testable rupture disc (TRD). Installation of the clamp could have

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prevented the TRD from performing its safety function, making the diesel

generator inoperable. The circumstances surrounding this event were investigated ,

in Inspection Report (IR) 94-14 and a violation was issued. The violation, j

94014-01, was closed in IR 96-18. )

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111. Enainsedng

E1 Conduct of Enginsedng

E1.1 Unclear Unit 1 and Unit 2 (abandoned) Boundary Conditions

a. Insoection Scone (37551. 40500)

. The inspectors observed performance of Periodic Test instruction (PTI) P45-POOO2, l

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Emergency Service Water (ESW) System Loop B Flow end Differential Pressure  ;

i. Test.  !

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b. Observations and Findings l

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Steps 5.2.a and 5.2.b of the PTl directed the opening of volves 2P42-

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F05298/5308, ESW FLOW TO FPCC HT EXCH INSTRUMr.NT ROOT VALVES. The  ;

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actual field labels were "ECC Loop B Flow." Further investigation noted that a

1990 design change package (DCP 90-000868) allowed the use of ESW as the

alternate cooling source for the fuel pool cooling and cleanup (FPCC) system heat l

exchangers via the Unit 2 emer9sncy closed cooling (ECC) heat exchangers inlet l

and outlet piping. The USAR referred to the system lineup resulting from the

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change as "not the normal supply to FPCC" and as " awaiting U2 completion" when

the Unit 2 ECC would provide the alternate FPCC heat exchanger cooling. Unit 2 >

was abandoned in 1995. l

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l The inspectors were concerned that: 1) the DCP was either not completed or was

j not reviewed in light of Unit 2 abandonment, 2) long term use of ESW as the  ;

alternate FPCC heat exchanger cooling may present additional concerns not yet i

reviewed, and 3) ESW Div 1 has similar concerns as Div 2 and has degraded flow

as documented in IR 96-04 (Section E4.3). These concerns will be addressed

during the resolution of Unresolved item (50-440/96004-04).

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E2 Enginsedng Support of Facilities and Equipment

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! E2.1 Potential Weakness of Emeroency Diesel Generator Exhaust Desion i

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a. Manection Scone (37551) l

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The inspectors observed performance of SVI E22-T2101, Operability >ct fmd

Maintenance of Diesel Generator Testable Rupture Disc (TRD) for the Dission 3

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Emergency Diesel Generator (EDG). The TRD provided over pressure protection of

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the exhaust system upstream of the unprotected (against seismic events and

tornado missiles) exhaust piping.

b. Observations and Findinas

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[ The Div 3 EDG TRD lifted at 750,620, and 590 pounds (lbs) consecutively and ,

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was required to lift at between 357 and 380 lbs. The set point of the TRD was

, adjusted so that it would lift within the required band. During the post maintenance

test with the EDG running, the rupture disc lifted unexpectedly. This indicated that

it lifted with less force than had been indicated by the testing. There had been

i several similar surveillance failures in the past. ,

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c. Conclusions ,

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Engineering efforts to resolve past surveillance test failures and problems had not

been fully effective. An NRC engineering inspection team, on site during the

testing, was already evaluating the design and performance of the Div 3 EDG TRD,

which is similar for all three EDGs. The inspectors reviewed the licensee's initial ,

i operability determination and concluded that it was adequate based on the

information that was available at the end of the inspection period. However, the

engineering inspection team identified additional information that was needed to

i complete the evaluation of the long term operability of the EDGs. This issue will be

addressed in the engineering team's inspection report.

E2.2 Review of Undated Safety Analysis Reoort (USAR) Commitments ,

. The inspectors reviewed applicable portions of the USAR that related to the areas

inspected; inconsistencies were identified, two are discussed below. The

inspectors also reviewed items that the licensee had identified during its reviews of

i the USAR. The licensee included the inconsistencies in its corrective action

l program. These may be reviewed in a future inspection based on the NRC's policy

(61 FR 54461, October 18,1996) for the review of licensee-identified USAR

l mconsistencies.

I

l The inspectors also reviewed current safety evaluations for some of the identified

i USAR inconsistencies. The safety evaluations were timely and appropriate for the

, identified issues. It appeared that the licensee had addressed the inconsistencies

) appropriately in accordance with safety significance. See Section E3 for a related

l item.

,

E2.2.1 USAR Section 6.4.1

The inspectors conducted a walk down of emergency food, water, and medical  ;

supplies located in the control room envelope. USAR Section 6.4.1 stated that the  !

!

normal occupancy level of the control room was six people following an accident. ,

- USAR Section 6.4.1 d., stated that a 7-day supply for seven people was . stored.

l The observed supplies appeared adequate for 7 days. However, USAR Table 6.4-1

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noted the location of the food to be the " kitchen" within the control room envelope.

~*

The actual location is the store room, also within the control room envelope. The

inspectors also noted that there are normally more than seven people in the control

room and that during an event requiring the use of the supplies there would be

more than seven people in the control room; e.g., normal control room crew (six

people), plus a security officer, communicator, several nordcensed operators and

support personnel, management oversight personnel, and an NRC inspector.

Additional operators also normally report to the centrol room to provide assistance

if needed. Engineers also often directly support operations in the control room.

E2.2.2 USAR Section 9.2.

A design change to allow use of ESW for cooling of FPCC via the Unit 2 ECC heat

exchangers inlet and outlet piping (sw Section E1.1) was considered permanent

even though the USAR referred to the system lineup resulting from the change as I

"not the normal supply to FPCC" and as " awaiting U2 completion." l

E3 Engineering Procedures and Documentation

a. Inanection Scoon (37001 and 37551)

The inspectors observed an operability determination related to temperature control l

valve (TCV) P42-F0665B for the emergency closed cooling (ECC) water system.

They also reviewed an associated safety evaluation.

b. Observations and Findinas

On March 14,1997, the inspectors observed that a system engineer had

questioned the operability of the 'B' Emergency Closed Cooling (ECC) Water

System temperature control valve (TCV) with the remoto (control room) manual

function unavailable. The operators had determined that the valve would not

respond to manual operation from the control room on March 12,1997,and

verified that the TCV controlled temperature in automatic, its normal mode of

operation. On March 14, a maintenance technician had written PlF 97-0522 on the

condition of too controller and operations had requested an operability

determination. The inspectors reviewed the operability determination and Safety

i Evaluation 97-001 and concluded that the TCV was operable. Safety Evaluation 1

97-001 hed been written on December 6,1996, to meet the requirements of

i 10 CFR 50.59 after the TCV had been installed for both trains of ECC. It

superseded two earlier safety evaluations.

l The design of the plant, as licensed, assumed that the ECC system would operate  ;

i at a temperature that would support operation of all cooled components with no

TCVs in the ECC piping and no bypass available around the heat exchangers. ECC

! was cooled by emergency service water (ESW). The licensee later determined that

the TCVs and bypass lines were needed to maintain the' safety related control 4

complex chillers operable when ESW was less than about 55* Fahrenheit (F). The

"

TCVs were not needed to maintain operability of the other ECC heat loads.

!

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The NRC had licensed the plant with no TCVs or bypass lines, therefore a safety

evaluation was required to evaluate a change in the plant, as oeiginally described in

the USAR. This change involved the addition of active components (TCVs) that

could cause the heat exchangers to be completely bypassed during certain active

failures. For example, loss of power to the TCV would cause it to fail "ac-is."

Typically, during most of each winter, ESW temperature has been between 32' and

45' F. With ESW that cold the TCVs would be in the full bypass position and

would have to reposition (an active function) during an accident as increasing heat

loads caused ECC temperature to increase above 70* F (TCV control design set

point). If this active function were to fait early in an accident, ECC heat loads on

the affected train would be without cooling. The original licensed design did not ,

include an active component that could malfunction and bypass the heat i

exchanger. The safety evaluation did not support the conclusion that the change I

did not increase the probability of occurrence of a malfunction of equipment or ,

create the possibility of a different type of malfunction (bypass of heat exchanger). l

l

c. Conclusions I

The failure to provide a safety evaluation with sufficient bases to determine that the ,

'

change to the facility did not increase equipment malfunction probability or create

the potential for a different type of malfunction was considered a violation (50-

440/97002-02(DRP)) of the requirements of 10 CFR 50.59, changes, tests, and l

experiments.

]

E8 Miscellaneous Enginsedng issues (92720 and 92903)

E8.1 ICIDSAdLLif,AQs  : " Slow Control Rod Scram

insertion Times Result in Technical Specification Violation." On December 12,

1994, during lechnical Specification (TS) surveillance testing of control rod scram i

insertion times, TS Limiting Condition for Operation (LCO) 3.1.3.2 ACTION c.1 was

entered due to t% number of " slow" control rods exceeding 20% of a 10%

sample. The cause of the event was attributed to the degradation, over time, of

the seating materiallocated in the disk holder sub-assembly used in some scram

solenoid pilot valves (SSPVs). Enforcement discretion was granted by the NRC.

l The suspect SSPVs were replaced and the licensee has continued to monitor the

! SSPVs for any evidence of a similar problem.

!

.

E8.2 (Closed) LER 50-440/96007-00: " Design Modification Program Weakness Results

in Missed Surveillance Requirement and Technical Specification Violations." The

. licensee identified the omission of isolation logic verification within two TS

! surveillance instructions that resulted in a condition prohibited by plant TS. The

i omission occurred during a design change of the leak detection system. Corrective

actions included development of a formal method of verifying surveillance test

methodology against electrical drawings. The inspectors verified that the

i procedure for the verification of test methodology was implemented. Violation 50-

!

440/96006-08 was issued.

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- E8.3 (Closed) LER 50-440/97002-00: " Breaker Trip Causes Loss of Electrical Power to

Non-Class 1E Bus and Results in Engineered Safety Feature At ion." A non-

class 1E feeder brosker to a 13.8 Kv/480 VAC transformer triL _ unexpectedly on

February 11,1997. The trip caused a loss of power to Reactor Protection System

Motor-Generator 'B' and a resultant half scram. This event is discussed in

Section 02.1 and the LER is closed based on that discussion.

E8.4 (Closed) Violation 50-440/96006-08: Surveillance testing inadequate after leak

detection system modification. This item is closed based on the discussion in

Section E8.2.

E8.5 (Clomad) Innoaction Follow-un item 50-440/94011-06: This item was opened when

delays in the opening of the scram solenoid pilot valves (SSPVs) for the reactor

control rods resulted in slow scram insertion times. This problem was first

identified at Grand Gulf. The licensee has demonstrated diligence over the last ,

2 years in addressing this problem. The root cause was sticking between the valve

seating surface and the seating material. Recent industry development of a new

seating surface combined with increased testing and replacement of the SSPVs by

the licensee have resolved this problem. See Inspection Reports 94-13/15 and

95-02/04.

V. Manaaement Meetinas

X1 Exit Meeting Summary

l

The inspectors presented the inspection results to members of licensee management at the i

,

conclusion of the inspection on March 21,1997. The licensee acknowledged the findings '

presented.

The inspectors asked the licensta whether any matericts examined during the inspection )

'

should be considered propriate.ty. No proprietary infoimation was identified. l

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• PARTIAL LIST OF PERSONS CONTACTED

.

Licensee

J. P. Stetz, Senior Vice President, Nuclear

L. W. Myers. Vice President, Nuclear

R. D. Brandt, General Manager, Nuclear Power Plant Department

W. R. Kanda, Director, Quality and Personnel Development Department

N. L. Bonner, Director, Nuclear Maintenance Department

J. J. Powers, Director, Nuclear Engineering Department

L. W. Worley, Director, Nuclear Services Department

- J. Messina, Operations Manager

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INSPECTION PROCEDURES USED I

e  ;

IP 37001: 10 CFR 50.59 Safety Evaluation Program i

IP 37551: Oneite Engineering

,

IP 61726: Surveellence Observations

IP 62707:

'

Maintenance Observation

IP 71500: Balance of Plant inspection

IP 71707

Plant Operations ,

1 IP 71750: Plant Support Activities l

l lP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power  !

'

Reactor Facilities

I

IP 92720: Corrective Action

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IP 92901:. Followup - Operations 1

IP 92902: Followup - Maintenance  !

! IP 92903: Followup - Engineering I

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ITEMS OPENED, CLOSED, AND DISCUSSED  :

l Onened l

i  !

1

50-440/97002-01 NCV Failure to Test EDGs on a Staggered Test Basis ,

50-440/97002-02 VIO Insdequate 10 CFR 50.59 Safety Evaluation

.

Closed

!

50-440/94011-06 IFI Scram Solenoid Pilot Valves

50-440/94021-00 LER ' Inoperable Div. 2 EDG in a TS Violation

'

l 50-440/94022-00 LER Failure to Test EDGs on a Staggered Test Basis

50-440/94023-00 LER Slow Control Rod Scram insertion Times Result in TS Violation

50-440/96004-00 LER Maintenance With Poor Tagout Causes Partial ESF Actuation

j 50-440/96005-00 LER Automatic Reactor Scram Following Auxiliary Transformer

Failure

l '

50-440/96006-08 VIO GE NUMAC Surveillance Testing inadequate

50-440/96007-00 LER Design Modification Program Weakness Results in Missed l

Surveillance Requirement and TS Violations l

50-440/96011-02 VIO Error introduced into EDG sol l

50 440/97002-00 LER Breaker Trip Causes ESF Actuation i

50-440/97002-01 NCV Failure to Test EDGs on a Staggered Test Basis l

i

Discussed

50-440/96004-04 URI Unit 2 Battery Room not Completed (E1.1)

)

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- . - -

.

.

-

LIST OF ACRONYMS USED

.-

ALARA AS LOW AS REASONABLY ACHIEVABLE

BOP BALANCE OF PLANT

CFR CODE OF FEDERAL REGULATIONS

DCN DRAWING CHANGE NOTICE

DCP DESIGN CHANGE PACKAGE

DIV DIVISION

ECC EMERGENCY CLOSED COOLING

ECCS EMERGENCY CORE COOLING SYSTEM

EDG EMERGENCY DIESEL GENERATOR

ESF ENGINEERED SAFETY FEATURE

ESW EMERGENCY SERVICE WATER

F FAHRENHEIT

FCV FLOW CONTROL VALVE

FPCC FUEL POOL COOLING AND CLEANUP

FR FEDERAL REGISTER

FW FEEDWATER

GE GENERAL ELECTRIC

HPCS HIGH PRESSURE CORE SPRAY

IFl INSPECTION FOLLOW-UP ITEM

IR INSPECTION REPORT

KV KILOVOLT

LBS POUNDS

LER LICENSEE EVENT REPORT

LCO LIMITING CONDITION FOR OPERATION

NCV NON-CITED VIOLATION

NPF NUCLEAR POWER FACILITY

NRC NUCLEAR REGULATORY COMMISSION

NRR NUCLEAR REACTOR REGULATION

ONI OFF NORMAL INSTRUCTION

OOS OUT OF SERVICE

PAP PERRY ADMINISTRATIVE PROCEDURE

PDB PLANT DATA BOOK

PDR PUBLIC DOCUMENT ROOM

PlF POTENTIAL ISSUE FORM

PMT POST MAINTENANCE TESTING

PORC PLANT OPERATIONS REVIEW COMMITTEE

PTl PERIODIC TEST INSTRUCTION

RC&lS ROD CONTROL AND INFORMATION SYSTEM

RFO REFUELING OUTAGE

RHR RESIDUAL HEAT REMOVAL

SCR SILICONE CONTROL RECTIFIER

SDC SHUTDOWN COOLING

SE SAFETY EVALUATION

SOI SYSTEM OPERATING INSTRUCTION

SSPV SCRAM SOLENOID PILOT VALVE

19

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.

LIST OF ACRONYMS USED (continued)

e

SVI SURVElLLANCE INSTRUCTION

TCV TEMPERATURE CONTROL VALVE

TS TECHNICAL SPECIFICATION

TRD TESTABLE RUPTURE DISC

TXI SPECIAL TEST INSTRUCTION

USAR UPDATED SAFETY ANALYSIS REPORT

i

URI UNRESOLVED ITEM

V VOLTS

VAC VOLTS ALTERNATING CURRENT

VIO VIOLATION

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PARTIAL LIST OF DOCUMENTS REVIEWED DURING THIS INSPECTION (9742)

Administrative Program Change, EMARP-0005, Rev.1, Monie,.ing the Shoreline Recession

and Bluff Erosion

Administrative Program Change, EMARP-0006, Rev.1, Monitodng the Settlement or l

Rebound of Safety Class Structures ,

1996 ALARA Report, Final,2/11/97 '

ALARA Report - February 1997,3/4/97

Audt Report PA 9744, Special Nucisar Material 3/14/97

Control Rod Mispositioning Quiz - 3/21/97 i

Control Room standng orders, various dates l

Control Room computer printouts, various parameters, various dates

Control Room daily instructions, various dates

Control room daily instructions, supplemental reading, various dates j

Control room safety tag log, various dates i

Control room strip charts, various parameters, various dates

Control room annunciator status books, revisable format various dates ,

Control room LCO log, various dates l

Cooper-Enterprise - R4/RV4 Preventive Maintenance Program (PMP) for Nuclear Standby )

Applications, Rev. 1,01/31/96, File 929

Daily Chemistry Report,02/10/97 ,

Deficiency tags, various locations, various dates

EMD-PS Owners Group Recommended Maintenance Program, Nuclear Standby Emergency

Diesel Generator, Rev. O, Feb.1993, Filo 285G

Fire extinguisher inspection tags, various locations, various dates

Forced Outages Meeting Agenda,03/14/97

GEK-63100, Operation and Maintenance instructions, Hydraulic Control Unit 4/80

Manager's Communication & Teamwork Meeting,2/5/97,2/7/97,2/10/97,2/12/97,

2/14/97,2/19/97, 2/21/97, 2/24/97, 2/26/97, 2/29/96, 3/3/97, 3/7/97, 3/12/97,

3/14/97,3/21/97

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Monthly Access Level Use Review For February, Dated 02/03/97 l

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Monthly Operations Report - January 1997

NRC Inspection 97002 Debrief Summary - 03/20/97

NSSS Heat Balance Computer Printouts 2/28/97,3/1/97,3/6/97,3/7/97,3/19/97 &

<

3/20/97

i 3D MONICORE periodc log, various dates

! Operations Section Operating Directives and Policies Handbook, Peer Observation Policy, .

Item 2-5, February 24,1997 l

Operations Administrative Control Tags, various locations, various dates

.

Operations infonnation Tags, various locations, various dates

l PAP-0204, Rev. 9, Housekeeping / Cleanliness Control Program

PAP-0230, Rev.1, independent Safety Engineering Group

PAP-0507, Rev.11, Preparation, Review and Approval of instructions'

,

PAP-1401, Rev. 9, Safety Tagging

PAP-1406, Rev. O, Equipment Control Tagging 1/16/95

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Perry Lines Weekly,2/13/97 & 3/20/97

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! PARTIAL LIST OF DOCUMENTS REVIEWED DURING THIS INSPECTION (9742), continued

'

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Perry Daily Report - Tuesdays and Thursdays l

Plant Log, Vol. 32,(02/03/97) Page No.92-138(03/21/97)

, Plant strip charts, various parameters, vadeus dates

Operational Surveillance Report No.96-021, Enginsedng Evaluation of Field Cladfication

j Requests (FCRs) (4/2/96)

, Operational Surveillance Report No. 97402, Evaluation of Maintenance Rule PlFs

! (2/27/97)  !

'

Operational Surveillance Report No. 97 004, R13 Isolate Phase Bus Duct Cooling Work

i Preparation, 3/12/97

Plan of the Day,2/03 thru 2/07/97,2/10 thru 2/14/97,2/17 thru 2/21/97,2/24 thru

,

2/28/97,3/3 thru 3/7/97,3/10 thru 3/14/97, and 3/17 thru 3/21/97.

,

PORC REVIEW ITEM Safety Evaluation #95-0123

PORC REVIEW ITEM Change Request #96-170 and Safety Evaluation 96-0155

'

i PORC REVIEW ITEM Safety Evaluation #97-0017

l PORC REVIEW ITEM Safety Evaluation #97-0019

PORC REVIEW ITEM (Safety Related) DCN 5620

,

PORC REVIEW ITEM Emergency Plan

! PORC REVIEW ITEM DCN 5628 for S.E.97-020 and Change Request j

i

PORC REVIEW ITEM DCN 5035 '

PORC REVIEW ITEM DCP 95-00228

j PORC REVIEW ITEM DCN 5600

j PORC REVIEW ITEM DCN 5626

. PORC REVIEW ITEM DCP 96-0049

PORC REVIEW ITEM PAP-0511

PORC REVIEW ITEM DCN 5425

PORC REVIEW ITEM TXI-0254

PORC REVIEW ITEM TXI 0255 l

PORC REVIEW ITEM PDB-R0001 _

PORC REVIEW ITEM Safety Evaluation 97-0018 )

PORC REVIEW ITEM Safety Evaluation 974030

Perry Operations Section Performance indcators, January - February 1997

Potential lasuo Forms No.97-001 through 974545 l

Radiation Work Permit 97006

Radiologically Restricted Area Radiation Surveys, various dates

RFO6 News - March 17,1997

Safety Tags, vadous locations, various dates

Shift Supervisor's Meeting,3/17/97

System Description Manuals - Various

Temporary Modfication Tracking Report,03/01/97

Unit Log, Unit 1, Vol. 91, (02/03/97) Page No. 79 -150 (02/27/97)

Unit Log, Unit 1, Vol. 92, (02/27/97) Page No. 1 - 54 (03/21/97)

Updated Final Safety Analysis Report

22