IR 05000354/1987014

Insp Rept 50-354/87-14 on 870512-0608.Deviation Noted.Major Areas Inspected:Outstanding Insp Items,Operational Safety Verification,Surveillance Testing,Maint Activities,Esf Sys Walkdown & LER Followup

ML20235N958
Person / Time
Site:	Hope Creek
Issue date:	07/06/1987
From:	Gallo R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20235N855	List: IR 05000354/1987014 ML20235N851 ML20235N905
References
50-354-87-14, NUDOCS 8707200283
Download: ML20235N958 (12)

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U. S. NUCLEAR REGULATORY COMMISSION l

REGION I

050354-870413 050354-870421 Report No.

50-354/87-14 Docket 50-354 License NPF-57 Licensee:

Public Service Electric and Gas Company Facility:

Hope Creek Generating Station Conducted:

May 12, 1987 - June 8, 1987 Inspectors:

R. W. Borchardt, Ser.ior Resident Inspector D. K. Allsopp, Resident Inspector R. J. Summers, Project Engineer Approved:

M 2$ J R. M. Gallo, Chief, Projects Branch 2

' Date Inspection Summary:

Inspection on May 12, 1987 - June 8, 1987 (Inspection Repcrt Number 50-354/87-14)

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, and licensee event report followup. This inspection involved 182 hours0.00211 days <br />0.0506 hours <br />3.009259e-4 weeks <br />6.9251e-5 months <br /> by the inspectors.

Results: One apparent deviation of the final safety analysis report (FSAR) was identified during this inspection period (paragraph 6.2)

pertaining to the failure to he.ve installed a control room annunciator to detect a malfunction of the hydrogen / oxygen analyzer heat trace system.

Increased management attention is needed to ensure the hydrogen / oxygen analyzer heat trace system is working adequately. The number of control room overhead annunciators in alarm during power operation continues to be of concern to the NRC.

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I Details 1.

Persons Contacted Within this report period, interviews and discussions were conducted with i

Mr. S. LaBruna and members of the licensee management and staff and various contractor personnel as necessary to support inspection activity.

2.

Followup on Outstanding Inspection Items a.(Closed)

Violation (87-08-01); Incorrect main steam line high radiation trip setpoints.

During the previous inspection period, the licensee took immediate action to correct the trip setpoints on all four main steam line (MSL) radiation monitors to within technical specification limits. The licensee responded to this violation in a May 22, 1987 letter and stated that MSL radiation levels will now be trended on a routine basis to verify that proper setpoints are in effect. The inspector has no further questions concerning this violation and will continue to verify proper instrumen-tation setpoints during routine inspection activities.

b.

(Closed)

Unresolved Item (87-08-04); Several deficiencies in the measuring and test equipment program existed in the radiation protection department and project installation department.

These deficiencies were reviewed and audited by the licensee QA organization and corrective action verified by the resident inspector.

This item is closed.

c.

(Closed)

Inspecte Follow Item (87-08-07); Root cause of main steam

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isolation valve (MSIV) failure. While the unit was shut-down during February 1987, the "A" inboard MSIV failed to

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close from the control room (ref: inspection report 87-05).

i An inspecticn of the MSIV found that one of the three

"A" MSIV solenoids was burned out and was the apparent cause of the MSIV failure. As immediate corrective action the f

licensee replaced all MSIV solenoids (24).

The solenoids from the "A" inboard MSIV were then sent to General Electric (GE) for a failure analysis. The following is an excerpt from a May 7,1987 GE letter.

"The cause of failure appears I

to have been the accidental inclusion of a foreign material, probably a loose seal, in the solenoid cavity between the plunger and the upper orifice.

Consequently the plunger for the solenoid was prevented from completing its magnetic metal to metal flux path.

The plunger for the solenoid probably oscillated at full power at 60 to 120 cycles per second, drawing in-rush current and bringing the temperature j

of the solenoid up to 200 degrees F thereby carbonizing the I

foreign material lodged between the plunger and its stop.

Additional associated damage was caused by this condition within the solenoid.

It appears that this was a unique t

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l occurrence. The vendor records since 1972 only show one other occurrence of a similar nature in the early 1970's."

l This item is closed.

3.

Operational Safety Verification 3.1 Inspection Activitles On a daily basis throughout the report period, inspections were conducted to verify that the facility was operated safely and in

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conformance with regulatory requirements. The licensee's management i

control system was evaluated by direct observation of activities, tours of the facility, interviews and discussions with licensee personnel, independent verification of safety system status and limiting conditions for operation, and review of facility records.

These evaluations included backshift inspections conducted on May 28 (1:45 - 6:00 a.m.) and Saturday, May 30 (10:30 - 2:30 p.m.).

The licensee's adherence to the radiological prc+ection 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.

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j 3.2 Inspection Findings and Significant Plant Events The unit entered this report period at approximately 92% power as limited by the transmission network stability curves generated after

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the d6 mage to the Keeney 500 KV transmission lines. The unit l

remained at the maximum allowable power levels throughout this report l

period except for short power reductions in order to perform maintenance or surveillance activities.

On May '18, Stanley LaBruna relieved Roger Salvesen as Hope Creek's General Manager.

Roger Salvesen has been reassigned to work at INPO for the next two years.

On May 21 at 1:20 p.m., while replacing a Tobar pressure transmitter on the channel "0" instrument rack, an I&C technician inadvertently bumped an isolation valve with a wrench causing a spurious channel

"D" LOCA signal. The channel "D" ESF systems actuated as designed including "D" diesel,

"D" RHR, "D" core spray, filtration re-circulation and ventilation fans and control room emergency ventilation.

RCIC also initiated but was secured before it injected any water into the reactor vessel. All plant systems were returned to normal by 2:30 p.m.

1 On May 29 at 12:15 p.m., the licensee reduced power from j

approximately 90% to 70% as directed by the load dispatcher due to a damaged transformer on the W223 transmission line out of New Freedom.

At 3:09 p.m.,

the "A" recirculation pump tripped during maintenance on its MG set and caused a power reduction from /0% to approximately

55%. While replacing brushes on the "A" recirculation MG set, a

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brush pig tail contacted the outer ring causing a short which tripped l

both the "A" recirculation pump and its MG set. All systems functioned

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E as designed. The "A" recirculation pump was returned to service at

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.3:32 p.m.' and power returned to 70%. At approximately 4:00 p.m. the W223 transformer was restr ed to service and power was progressively increased by the load dispatcher until power was returned to (92%) at 12:37 p.m. on'May 31.

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On May 30.at 12:10 p.m., while conducting a plant tour the nuclear l

shift supervisor fsund the reactor roving patrol, a contractor fire l.'

watch, sleeping in the. reactor building. The fire watch was-p

' immediately relieved from fire watch duties and escorted offsite.

It-p var later deterrained that the fire watch had not missed any of his

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.howrly rounds. The fire watch was terminated and returned to his k

employer, Protection Technology Incorporated.

On June 5 at 0:12 a.m., the licensee discovered' low diesel air. start pressure (300 psig) on the "C" diesel due to the air start compressor'.

'having been. turned off. An investigation of the other. diesels revealed that the "B" diesel air start pressure was low (300.psig)

and ii.s associated compressor had also been turned off. The "A" and

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"D" diesal air compressor alignment was found to be proper. All four dieself were test run satisfactorily following the event and detailed inspections of all four diesels identified no further discrepancies.

The licensee utilized both on-site and off-site safety review members to assist in the investigation of this event. The licensee determined the cause of the "B" and'"C" diesel air start compressors being turned off was personnel error on the part of.the June 4, 4 to 12 p.m. shift eccipmect operator (E0).

The licensee empirically calculated that the air start compressors were turned of f on that shift based on the receiver bleed down rate.

The 'hift E0 admitted to turning off the s

air compressors when he checked the air compressor oil level. When directly questioned, the E0 could not recall positively that he had returned the."B" and "C" diesel air compressors to the "on" position.

As corrective action, the licensee has issued new direction to the E0 for cheching air compressor oil level which precludes operation of the air c6mpressor on-off switch. The licensee has also modified E0 logs to include recording air receiver pressure and air compressor on-off switch position daily.

On June 8 at 9:51 a.m.,

the licensee received an "A" channel primary containment isolation syttem (PCIS) isolation signal. This PCIS signal isolated the following systems:

torus water cleanup, "Ar, hydrogen-oxygen analyzer, primary containment instrument gas, and portio ~ns of the filtration recirculation and ventilation system. The

"A" channel PCIS isohtion was generated by concurrent surveillance s testing on NSSS$ and FCIS high drywell pressure. All systems functioned as designed and were returned to operation within six minutes of the isolati.on.

The impcrtance of coordinating surveillance tests was re-emphasized to both the operation <, and I&C Groups.

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No violations were identified.

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4.

Surveillance Testing 4.1 Inspection Activity During this inspection period the inspector perforsned 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.

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

- IC-FT.BB-008 Functional Test of Reactor Pressure Vessel Level Trips

- IC-CC.BF-011 Channel Calibration of Scram Discharge Volume Level I: truments

_ IC-FT.SE-019 Functional Test of Power Range Neutren Monitor

"C" No violations were identified.

5.

Maintenance Activities 5.1 Inspection Activity During this inspection period the inspector observed selected main-tenance activities on safety related equipment to ascertain that these activities were conducted in accordance with approved procedures, Technical Specifications, and appropriate industrial codes and standards. These inspections were conducted in accordance with NRC inspection procedure 62703.

5.2 Inspection Findings j

Portions of the following activities were observed by the inspector:

J Work Order Description

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l 87-05-14-132-0 Preventative

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Maintenance on "A" H2 Recombiner MOV Breakers

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87-05-27-046-4 Replace "A" Hydrogen-Oxygen Sample Pump No violations were identified.

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6.

Engineered Safety Feature (ESF) System Walkdown

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6.1 Inspection Activity

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The inspectors independently verified the operability of selected ESF l

systems by performing a walkdown of accessible portions of the system

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to confirm that system lineup procedures match plant drawings and the j

as-built configuration. This ESF system walkdown was.also conducted

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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.

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6.2 Inspection Findings i

The hydrogen-oxygen (H202) analyzer / monitor system was inspected.

The H202 analyzer utilized at Hope Creek is a model number K-IV built by Comsip, Incorporated. This system is designed to meet seismic category I criteria and'is powered from a class 1E power source. The H202 system continuously draws an atmospheric sample from three locations in the primary containment for analysis of hydrogen and oxygen content. All sample lines are heat traced to maintain sample line temperatures between 250-270 degrees F to prevent condensation formation. The inspector reviewed the H202 analyzer technical manual, surveillance procedures, FSAR commitments, and conducted an in plant system evaluation. Amendment 14 to the FSAR, Section 7.3.1.1.6.3 hydrogen-oxygen analyzer system, states that sample line heat trace failures "are individually indicated at each heat trace panel and

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provided as a common trouble alarm for both panels in the main control

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room." Contrary to this FSAR commitment, there is no control room

alarm for a sample line heat trace failure. This discrepancy is being j

sited as a deviation. (87-14-01) Several "B" H202 local heat trace

circuits (CN3, CN4, CN5) appear to be malfunctioning in that they are in almost constant alarm and cannot be reset. Corrective action on these circuits will be tracked as an inspector follow item.

(87-14-02)

7.

Licensee Event Report Followup The licensee submitted the following reports during the inspection period. These event reports and periodical reports were reviewed for accuracy and timely submission.

The asterisked reports received additional followup by the inspector for corrective action implementation.

Monthly Operating Report for April, 1987 LER 87-019-00 Technical Specification Violation - Main Steam Line

Radiation Monitors Set Above Allowable Limits Due to Administrative Errors i

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LER 87-020-00 Isolation of RWCU While Placing Filter / Demineralized in " HOLD" Mode-Probable Design Deficiency LER 87-019 describes the events which resulted in the main steam line

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radiation monitors "B" and "C" trip setpoints being established in excess

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of allowable technical specification limits. This was sited as a violation anc is discussed in detail in paragraph 3.2 of NRC inspection report 87-08.

8.

Protection From Flooding of Equipment Important to Safety (TI 2515/88)

An inspection was conducted to verify that equipment important to safety would not be damaged by flooding caused by the rupture of non-Class I system components or piping to the extent that engineered safety features could not perform their design function.

This inspection consisted of plant walkdowns, and reviews of drawings, design specifications and separation data sheets.

Particular attention was given toward evaluating the licensee's compliance with the following guidelines:

1.

Separation for redundancy - single failure of non-Class I system components or pipes shall not result in loss of a system important to safety.

Redundant safety equipment shall be separated and protected to assure operability in the event of a non-Class I system or component fails.

2.

Access doors and alarms - watertight barriers for protection from flooding of equipment important to safety shall have all access doors or hatches fitted with reliable switches and circuits that provide an alarm in the control room when the access is open.

3.

Sealed water passages passages or piping and other penetrations through walls of a room containing equipment important to safety shall be sealed against water leakage from an postulated failure of non-Class I water systems. The seals shall be designed for the SSE, including seismically induced wave action of water inside the affected compartments during the SSE.

4.

Class 1 watertight structures - walls, doors, panels, or other compartment closures designed to protect equipment important to safety from damage due to flooding from a non-Class I system rupture shall be designed for the SSE, including seismically induced wave action of water inside the affected compartment during the SSE.

5.

Water level alarms and trips - rooms containing non-Class I system components and pipes whose rupture could result in flood damage to equipment important to safety shall have level alarms and

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pump trips (where necessary) that alarm in the control room and limit flooding wi. thin the design flood volume.

Redundancy of switches is required.

Critical pump (i.e., high volume flow, such as condenser

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circulating water pumps) trip circuits should meet IEEE 279 criteria.

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Class 1 equipment should be located or protected such that rupture of a non-Class I system connected to a tower containing water or body of

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water (river, lake, etc.) will not result in failure of the equipment l

from flooding.

i The inspection of selected plant areas, design specification 10855-D7.3 (Plant Separation Criteria), and various separation review data sheets indicates that the licensee satisfies the above guidelines with one exception. Watertight doors throughout the plant are kept closed (except for passage) by administrative guidelines, but they do not have switches or alarm circuits as recommended in item 2 above.

Because the use of alarm switches is not a requirement by regulation, nor is it a commitment in any correspondence known to the inspector, no followup action is anticipated.

9.

Scram Discharge Volume Capability (TI 2515/90)

The inspector reviewed the licensee's actions to ensure that the scram discharge volume (SDV) capability is in accordance with the BWR Owners Subgroup long term program and the plant Final Safety Analysis Report (FSAR) and Safety Evaluation Report.

The Hope Creek design of the contrci rod drive hydraulic system ensures good hydraulic coupling between the SDV and the instrument volume (IV).

i The SDV consists of a 12 inch pipe of approximate 20 linear feet for each of the two banks of hydraulic control units (HOUs). The IV is a vertical run of 12 inch pipe that is integral i.o the SDV piping. The SDV is sloped to ensure proper draining to the IV. The SDV vent and drain valves I

consist of series isolation valves that are normally open/ fail closed, l

which are on common legs of the vent and drain piping associated with the j

two SDV/IVs.

Following are the inspection criteria and findings:

1.

Criterion:

The scram discharge headers shall be sized in accordance with GE OER-54 and shall be hydraulically coupled to the instrument volumes in a manner to permit operability of the scram level instrumentation before loss of system function.

Findings:

The inspector reviewed design drawings (FSAR Figure 4.6-10 and P&ID M-47-1, Control Rod Drive Hydraulic - Part B) and determined that there is good hydraulic coupling in the Hope Creek design of the

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scram discharge volume and instrument volumes since the two instrument volumes are vertical risers integral to the SDV with

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proper sloping designed in the associated piping to assure adequate

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drainage of water in the SDV to the IV.

Section 4.6 of the FSAR states that the SDV has been designed with sufficient capacity to receive 3.34 gallons of water for each control rod during a scram.

The inspector independently calculated the capacity of the SDV and

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verified that a sufficient volume exists to fulfill the scram I

function if the IV high level reactor scram instrumentation l

functions.

Therefore the GE OER-54 criteria have been met.

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2.

Criterion:

Level instrumentation shall be provided for automatic scram initiation while sufficient volume exists in the SDV.

Findings: The inspector reviewed the FSAR, plant technical specifications and plant drawings (reactor protection systems (RPS)-

logic diegrams 10855-N-C71-1010-1(3)-5 and PN 1-C11-1010-0154(3)-04)

and confirmed that the automatic scram function exists for high instrument volume water level. As' discussed above, sufficient SDV

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j capacity exists if this high level scram functions properly.

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Criterion:

Instrumentation taps shall be provided on the vertical IV and not on the connected piping.

t Findings:

The inspector reviewed P&ID M-47-1, Control Rod Drive Hydraulic - Part B and in addition, field verified that the associated instrument volume level instrumentation taps are on the J

vertical IVs and not on connected piping.

4.

Criterione. The scram instrumentation shall be capable of detecting water accumulation in the IVs assuming a single active failure in the instrumentation systems or the plugging of an instrument line, i

Findings: The inspector field verified for each of the two instrument volumes that the system configuration precludes a single i

line plugging or other single failure that would cause a failure of I

the instruments to detect water in the IVs.

Each instrument volume f

has two water level float switches and also two level transmitter

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instruments,-which provides a diverse method of monitoring level and providing RPS input.

The two float switches share a common instru-

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ment tap and the two level transmitters share a common instrument

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tap. Therefore a single line plugging would only potentially fail

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two of the four instruments. The RPS logic arrangement (drawing No.

C71-1010) is such that each instrument inputs to a separate logic subchannel and is therefore electrically independent.

The RPS logic arregement is designed such that one float switch and one pressure switch input to the "A" protection train and the redundant instru-

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ments input to the "B" protection train.

Since RPS logic is "one-

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out-of-two-taken-twice", this arrangement precludes common mode failures due to manufacture defects of the instruments and also those introduced during operational testing, calibration, or repair of the instruments by human error.

5.

Criterion:

Vent and drain functions sha:1 not be adversely affected by other system interfaces, such as the capability of water backup into the scram IVs which could cause a spurious scram, i

i Findings:

The inspectors reviewed system P& ids (M-47-1) and field reviewed accessible portions of the vent and drain piping associated with the scram IVs.

This review could not conclude that backflow into tne IVs is precluded by design.

This item will be reviewed during a future inspection.

(87-14-03)

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6.

Criterion:

The pawer-operated vent and drain valves shall close

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under loss'of air and/or electric power.

Valve position indication i'

shall be provided in the control room.

Findings: The inspector reviewed Figure 4.6-6 of the FSAR and P&ID M-47-1 and verified that the vent and drain valves are designed to fail closed on a loss of air.

The valve position is indicated in the control room and is provided by a proximity switch on the valve stem.

7.

Criterion:

Instrumentation shall be provided to aid the operator in the detection of water accumulation in the IVs before scram initiation.

Findings: The inspector verified that alarms exist in the control room that would indicate the presence of water in the IVs.

Two additional float-type : level switches are provided for each of the IVs.

These instruments provide iriput to the process computer for high level alarm, to the rod block monitor for rod withdrawal block and to the control room overhead annunciators Alarm response procedures are available for operator actiun in the event that water is detected in the IVs as indicated by these level instruments.

8.

Criterion:

Vent and drain line valves shall be provided to contain the scram discharge water with a single active failure and to minimize operational exposure.

Findings: The inspector verified through review of P&ID (M-47-1) and i

RPS logic diagram (C71-1010) that the vent and drain line valves are provided in series (redundant valves for both vent and drain) such that they will contain the scram discharge water with a single active failure.

9.

Criterion: Vent and drain valves shall be periodically tested.

Findings:

The inspector verified that procedures exist that test operability of the vent and drain valves. The valves are routinely cycled (closed / opened) once per 31 days.

In addition, the valves are response time tested once per 18 months.

The closure response time acceptance criteria is consistent with GE specification (less than 30 seconds).

The inspector also reviewed the most recently completed response time surveillance procedure (OP-ST.BF-006(Q)) dated March 10, j

1986, whirh was satisfactory.

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10.

Criterion:

Level detection instrumentation shall be periodically tested in place.

Findings: The inspector reviewed the licensee's level detection

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surveillance procedures ard verified that both the protection

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instruments and the control and alarm instruments are periodically j

tested in place.

These surveillance activities are conducted in I

accordance with the plant technical specifications.

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state the appropriate test acceptance criteria and also include

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necessary precautions, limitations and restoration to service for the i

affected instrumentation.

11.

Criterion: The operability of the entire system as an integrated

'I whole shall be demonstrated periodically.

j Findings:

The inspector reviewed the licensee's technical specifications and the following surveillance procedures:

a.

RE-ST.BF-001(Q), Coni.rol Rod Drive Scram Time Determination b.

IC-CC.BF-001(Q), Channel Calibration - Scram Discharge Volume Water Level, Channel C11-N601A, RPS Channel Al c.

IC-CC.BF-002(Q)', Channel Calibration - Scram Discharge Volume Water Level, Channel C11-N601B, RPS Channel S1 d.

IC-CC.BF-003(Q), Channel Calibration - Scram Discharge l

Volume Water Level, Channel C11-N601C, RPS Channel A2

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IC-CC.BF-004(Q), Channel Calibration - Scram Discharge Volume Water Level, Channel C11-N6010, RPS Channel B2 f.

IC+CC.BF-005(Q), Channel Calibration - Scram 01scharge Volume Water Level, Channel C11-NO.13A, RPS Channel Al g.

IC-CC.BF-006(Q), Chanr.el Calibration - Scram Discharge Volume Water Level, Channel C11-N013B, RPS Channel B1 h.

IC-CC.BF-007(Q), Channel Calibration - Scram. Discharge Volume Water Level, Channel C11-N013C, RPS Channel A2 1.

IC-CC.BF-008(Q), Chann'el Calibration - Scrar Discharge'

Volume Water Level, Channel C11-N013D, RPS Channel 82 j.

IC-CC.8F-010(Q), Channel Calibration - Scram Discharge Volume High Level Alarm and Control Rod Withdrawal Block, Channel C11-N013E and C11-N013F, Division 1 k.

IC-CC.BF-011(Q), Channel Calibration - Scram Discharge Volume High Level Alarm and Control Rod Withdrawal Block, Channel C11-N013G and C11-N013H, Division 2 1.

OP-DL.ZZ.026(Q), Operations Daily Log m.

OP-ST.BF-003(Q), Scram Discharge Volume Vent and Drain Valve Operability Test - Monthly

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OP-ST.BF-006(Q), Scram Discharge Volume Vent and Drain Valve Functional Test - 18 Months i

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The inspector verified that the procedures implement the surveillance requirements of the technical specifications, which have incorporated the changes transmitted by Generic Letter to all BWR licensees from NRR on July 7,1980. Although normally-rod scram testing is accomplished one rod at a time, the-licensee has the capability to conduct unscheduled full core scram timing using the Sentinel Option of the'GETARS computer system for any actual reactor scram. This method is incorporated into the control rod drive scram time determination surveillance procedures as one of three available options.

No violations were identified.

10. 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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