IR 05000354/1987016

Insp Rept 50-354/87-16 on 870609-0713.Violations Noted. Major Areas Inspected:Followup on Outstanding Insp Items, Operational Safety Verification,Surveillance Testing,Maint Activities,Esf Sys Walkdown & LER Followup

ML20236N384
Person / Time
Site:	Hope Creek
Issue date:	07/20/1987
From:	Swetland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20236N252	List: IR 05000354/1987016 ML20236N249 ML20236N267
References
50-354-87-16, NUDOCS 8708110481
Download: ML20236N384 (8)

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Public Service Electric

and Gas Company

U. S. NUCLEAR REGULATORY COMMISSION i

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REGION I

050354-870508.

l 050354-870509 j

Report No.

50-354/87-16 l

l Docket 50-354 License NPF-57

l Licensee:

Public Service Electric and Gas Company Facility:

Hope Creek Generating Station

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Conducted:

June 9, 1987 - July 13, 1987 Inspectors:

R. W. Borchardt, Senior Resident Inspector

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D. K. Allsopp, Resident Inspector

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R. R. Brady, Reactor Engineer Approved:

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P. Swe'tland, Chief, Projects Section 2B Dite

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Inspection Summary:

Inspection on June 9,1987 - July 13,1987 (Inspection Report Number 50-354/87-16)

Areas Inspected:

Routine onsite resident inspection of the following areas:

followup on outstanding inspection items, operational safety verification, surveillance testing, maintenance activities, engineered safety feature system walkdown, electro-hydraulic control system "A" pressure regulator failure, and licensee event report followup. This inspection involved 212 hours0.00245 days <br />0.0589 hours <br />3.505291e-4 weeks <br />8.0666e-5 months <br /> by the inspectors.

Results: One violation of tf.chnical specifications regarding thermal overload bypass configuration was cited in this report and is discussed in paragraph 3,-2.

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Public Service Electric

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Details 1.

Persons Contacted Within this report period, interviews and discussions were conducted with Mr. S. LaBruna and members of the licensee management and staff and various contractor personnel as necessary to support inspection j

activity.

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2.

Followup on Outstanding Inspection Items

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(Closed) Inspector Follow Item (86-40-01); " Scram Dump Valve" (F 182 A&B) Incorrectly Labeled " Backup Scram Pilot Valve"..The inspector verified the correct labeling was properly installed on the scram dump valve.

This item is closed.

3.

Operational Safety Verification 3.1 Inspection Activities On a daily basis throughout the report period, inspections were conducted to verify that the facility was operated safely and in conformance with regulatory requirements.

The licensee's management control system was evaluated by direct observation of

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activities, tours of the facility, interviews and discussions l

with licensee personnel, independent verification of safety i

system status and limiting conditions for operation, and review l

of facility records.

The licensee's adherence to the

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radiological protection and security programs was also verified on a periodic basis. These inspection activities were conducted in accordance with NRC inspection procedures 71707, 71709.and 71881 and included weekend and backshift inspections conducted on June 24 (0:30-6:00 a.m.), June 27 (1:10-5:10 p.m.), and July 9 (3:30-6:00 a.m.).

3.2 Inspection Findings and Significant Plant Events i

l The unit entered this report period at approximately 92% power I

as limited by the transmission network stability curves

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generated after the damage to the Keeney 500 KV transmission i

lines.

The unit remained at the maximum allowable power levels throughout this report period except for short power reductions during thunderstorms or in order to perform maintenance or

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surveillance activities.

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4 At 9:30 a.m. on June 6, while discharging sodium hypochlorite onsite from a delivery truck, the hose connection broke spilling 150-200 gallons on the ground.

The spill area was blocked off and flushed with water to maximize the dilution effect. The licensee utilizes sodium hypochlorite in the service water i

system to minimize marine bio-fouling.

The licensee's

calculations indicate that the sodium hypochlorite will break i

down in the soil before reaching the Delaware river, and 'ill cause no adverse environmental impact.

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l Public Service Electric

• and Gas Company

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l On June 16, during surveillance testing of the MOV_ overload bypass

J feature on the service water intake screen spray wash

isolation'.

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valves, it was discovered that these valves did not have an overload-bypass circuit installed.

The subject valves (IEP-HV2225lA/B/C/0)

were being tested per technical specification (T.S.)'3.8.4.2.

-Im-i mediate corrective action consisted of installing a 'TMOD jumper to bypass the thermal overload circuit and return the system to an operable status..The licensee stated that the overload bypass of these valves was not tested prior to initial plant startup because they were added to T. S. 3.8.4.2 following the conduct of_the. initial T. S.

surveillance tests. Upon issuance of the final T.

S., the-licensee

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did not detect the omission of initial surveillance tests for these four valves. The' licensee conducted a 100% review of-all other valves listed in the technical specification to ensure the proper ~ thermal overload configuration. This-verification of.approximately' 280 valves required to have a bypass circuit,. identified 4 additional deficiencies. An improper design change implemented during the j

startup program and prior to implementation of the final T. -S., dis-

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abled the overload bypass circuit on the 4 turbine auxiliary cooling system (TACS) to safety auxiliary cooling system (SACS) return isolation valves (1EG-HV-2496 A/B/C/D).

These valves were immediately corrected by the licensee.

The inspector noted that upon failure of i

any of these 8 valves to operate on demand, an operator could be j

dispatched to manually operate each valve.

The QA organization

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conducted an audit of past and current versions of T.S. and verified.

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service water intake screen spray wash isolation valves were per -

manently hardwired with the proper thermal overload bypass circuit l

and the TMOD cancelled. Although the licensee identified this jl problem during routine surveillance testing, eight valves did not.-

meet TS requirements because appropriate surveillance tests were not'

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performed prior to reliance on these valves ^for plant ope ation, j

This occurrence is cited as a violation to assure adequate licensee

corrective action to prevent recurrence of similar events.

(354/87-16-01)

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At 1:00 p.m. on June 24, the "A" electro-hydraulic control (EHC)

system pressure regulator circuit was removed from service when it was determined the circuit would periodically malfunction.

This malfunction would cause the turbine control valves to momentarily open causing a small down power plant transient.

The EHC system modification is discussed in detail in Section 7 of this report.

i-At 3:48 a.m. on June 26, the licensee received a channel

"C"

high pressure coolant injection (HPCI) isolation signal which

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shut the HPCI inboard steam supply valve.

The HPCI. isolation was caused by a faulty room thermocouple.which erroneously indicated high area differential temperature and actuated the

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steam leak logic.

The licensee declared HPCI inoperable and left the system isolated until the thermocouple was replaced.

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The HPCI system was decla' ed operable later on June 26.

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Public Service Electric

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and Gas Company At 3:00 p m. on June 26, the licensee experienced a "B" reactor

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feed pump (RFP) trip and a "B" recirculation ~ runback when the

"B" vital bus sensed an undervoltage condition and slow-transferred to the alternate infeed source._ The undervoltage condition was sensed when the primary infeed breaker status light unit in the control room was removed to change a light bulb. Upon' replacement, a current surge caused an undervoltage condition to be sensed and initiated automatic transfer to the alternate power supply.

The l

momentary loss of power to the "B" RFP lube oil pump during the I

transfer, tripped the "B" RFP. The ensuing level oscillation reached a momentary minimum value at the level 4 (30 inches) setpoint and l

caused the "B" recirculating pump to runback. The li.censee has j

concluded that the "A" recirculation pump did not run-back due to the l

momentary nature of the level oscillation coupled with the difference j

in relay response times for the individual runback channels.

The i

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runback channel relays do not have a seal-in feature. The recir-culation pump runback reduced power from approximately 100% to 85%.

Power was returned to approximately 100% at 5:21 p.m.

The licensee

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is considering modifications to eliminate current surges in control /

indicator units to prevent recurrence of such events.

At 9:35 p.m. on June 29 and 4:11 p.m. on July 1 the licensee experienced reactor' water cleanup system (RWCU) isolations.on high delta flow due to flow oscillations induced during routine recycling of the "A" filter / demineralized (F/D) unit.

The RWCU system was

returned to service promptly following each event.

This has become a

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recurrent event because of the extreme sensitivity of system flow during the F/D backwash and renewal lineups.

The licensee is evalu-ating actions to prevent these system isolation actuations.

On July 11, the licensee discovered that the instrument root valve for the "C" RHR pump pressure transmitter was out of position (closed), thereby disabling an input into the ADS permissive logic.

Because the investigation into this event was not complete at the end of this inspection period, this item is

unresolved.

(354/87-16-02)

4.

Surveillance Testing 4.1 Inspection Activity During this inspection period the inspector performed detailed technical procedure reviews, witnessed in progress surveillance testing, and reviewed completed surveillance packages.

The inspector verified that the surveillance tests were performed in accordance with Technical Specifications, licensee approved procedures, and NRC regulations.

These inspection activities were conducted in accordance with NRC inspection procedure 61726.

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Public Service Electric

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The following surveillance tests were reviewed, with portions witnessed by the inspector:

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IC-CC.BC-007 RHf Pump Discharge Flow OP-ST.KJ-001

' Diesel Generator Monthly Operational

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IC-FT.SE-016 Functional Test of Power Range Neutron

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Monitor "C"

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OP-ST.ZZ-001 Power Distribution Verification No violations were identified.

5.

Maintenance Activities 5.1 Inspection Activity During this inspection period the inspector observed selected maintenance activities on safety related equipment to ascertain that these activities were conducted in accordance with' approved

.j procedures, technical specifications, and appropriate industrial

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codes and standards.

These inspections were conducted in accordance with NRC inspection procedure 62703.

5.2 Inspection Findings l

Portions of the following activities were observed by the inspector:

Work Order Procedure Description j

87-06-18-174-1 Special Instruction Troubleshooting and repair of HCU pilot solenoid valve

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87-05-29-040-6 MD-PM.ZZ-005 Inspection of electric motors.

Observed cleaning and inspecting on

"A" service water pump motor 87-04-13-030-8 MD-CM.EA-003 Service water strainer overhaul and repair 87-06-08-101-1 MD-GP.ZZ-003 Repacked

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1GSHV-5741A; "A" H202 Analyzer H2 supply.

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Public Service Electric

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I Work Order Procedure Description l

I 87-06-30-184-3 DCP 4HM-0126 Installation of i

overload bypass circuit for 1EG-HV-24968'

No violations were identified.

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6.

Engineered Safety Feature (ESF) System Walkdown i

6.1 Inspection Activity

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The inspectors independently verified the operability of selected ESF systems by performing a walkdown of accessible portions of the system to confirm that system lineup procedures l

match plant drawings and the as-built configuration.

This ESF j

system walkdown was also conducted to identify equipment conditions that might degrade performance, to determine that instrumentation is calibrated and functioning, and to verify that valves are properly positioned and locked as appropriate.

This inspection was conducted in accordance with NRC inspection procedure 71710.

6.2 Inspection Findings The emergency diesel generator plant consists of a Colt-Piestick-diesel engine and a directly connected Beloit power system alternator. The diesel engine is a super-charged 4 cycle engine with 12 cylinders in a "V" configuration and is rated for 4430 KW net. A walkdown of the system did not identify any conditions adversely _affecting system operability nor were any minor discrepancies identified that were not previously licensee identified.

The inspector verified that corrective actions were

l in progress for the minor discrepancies observed.

No violations were identified.

7.

Electro-Hydraulic Control System (EHC) "A" Pressure Regulator Failure During this inspection period the operations staff noticed spurious and unexplained oscillations of reactor water level.

The station's system engineering group initially diagnosed the problem to be related to a control valve that supplies steam to the "C" reactor feed pump turbine (RFPT) that was known to be sticking and operating sluggishly. However, on June 16, 1987 at 7:30 p.m., the on shift control room operators noticed a small power oscillation and ensuing-

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water level oscillation as a result of the power transient. The

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Public Service Electric

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shift performed a preliminary analysis of the available strip chart recordings and concluded that the transient was caused by a turbine control valve spuriously opening and not by the control valve to the

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l-RFPT sticking open.

The GETAR computer was set up to capture the l

next level oscillation to provide a detailed trena of important l

l parameters needed to troubleshoot and verify the problem.

The GETAR

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trace of the next oscillation confirmed that the transient was caused by a turbine control valve opening.

Further troubleshooting narrowed

the problem to the pressure control unit (PCU) of the EHC system.

i The purpose of the EHC system is to provide normal reactor vessel pressure control by positioning the turbine control valves, passing an amount of steam to the turbines consistent with the amount being generated in the reactor vessel.

The PCU consists of two redundant pressure setpoint summing amplifiers which compare a pressure setpoint signal (pressure error), manually inserted by the operators, to the main steam equalizing header pressure (turbine inlet pressure).

The difference between these two inputs provides the driving signal to the turbine control valves.

During normal operation the "B" pressure regulator circuit is placed in backup mode by inputting a negative bias into the circuit.

The malfunction of.

the "A" pressure regulator circuit caused the EHC system to see a large pressure error, causing the EHC system to open the turbine control valve.

This caused steam demand to increase, main steam and reactor pressure to decrease, and reactur water level to increase due to swell caused by the pressure decrease.

A safety evaluation was performed to analyze the adverse effects of operating with the

"A" pressure regulator biased into the back up mode or removed from the circuit entirely.

The Final Safety Analysis Report addresses the loss of the pressure regulator, therefore an I

unreviewed safety question does not exist.

Due to the unpredictable nature of the problem with the

"A" pressure regulator, the component was removed from service.

There have been no water level oscillations since the regulator was removed from service.

During the next forced outage the following work will be performed to repair the EHC system:

The "A" pressure regulator circuit card will be replaced; the pressure setpoint motor driven potentiometer will be replaced; and all interconnecting wiring will be checked for abnormalities.

No violations were identified.

8.

Licensee Event Report Followup The licensee submitted the following event reports during the inspection period.

These event reports and periodic reports were reviewed for accuracy and timely submission.

The asterisked reports received additional followup by the inspector for corrective action implementation. The (#) items identify reports which involve

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Public Service Electric

• and Gas Company licensee identified Technical Specification violations which are not being cited based upon meeting the criteria of 10 CFR 2 Appendix C.

Monthly Operating Report for May, 1987 LER 87-021-00 Reactor Protection System (RPS) Bus "A"

Inadvertently De-energized during Preparation for Surveillance Resulting in Engineered Safety Feature (ESF) Actuation

1. LER 87-022-00 T.S. Violation - Cooling Tower Blowdown Radiation Monitoring System Sample Pum.o Out of Service V 12 Hours and Required Grab Samples Not Taken LER 87-021-00 details the events which resulted in RPS bus "A" being de-energized due to a voltage transient while shifting station vital bus "A" infeed breakers.

At the time of this occurrence, RPS bus

"A" was being powered from its alternate source which is a non-regulated 480 VAC/120 VAC transformer.

Loss of the RPS bus resulted in an ESF actuation (reactor water cleanup isolation, reactor recirculation sample isolation, main steam drain line isolation) and a half scram.

Licensee corrective actions consist of installing a regulating transformer during the next refueling outage and instructing operations personnel not to perform manual feeder breaker swaps when an RPS bus is being powered from its alternate source.

LER 87-022-00 describes the events that caused the radiation monitoring system (RMS) cooling tower blowdown sample pump to be off j

for a period of ten days without the required technical specification

I grab samples having been taken.

The sample pump had been l

l inadvertently shut down due to a personnel error when a radiation protection technician mistakenly changed the RMS low flow alarm setpoint for the sample pump.

Corrective actions included l

counselling / retraining the technician involved, utilizing a different niethod of changing RMS setpoints, and procedural revisions to

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i periodically verify pump status.

No violations were identified.

9.

Exit Interview The inspectors met with Mr. S. LaBruna and other licensee personnel periodically and at the end of the inspection report to summarize the

scope and findings of their inspection activities.

Based on Region I review and discussions with the licensee, it was determined that this report does not contain information subject to 10 CFR 2 restrictions.

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