IR 05000354/1987005

Insp Rept 50-354/87-05 on 870210-0309.Violation Noted: Failure to Obtain & Analyze Grab Samples of Cooling Tower Blowdown Line.Unresolved Item Noted Re Environ Qualification of MSIV Terminal Boxes

ML20205E528
Person / Time
Site:	Hope Creek
Issue date:	03/20/1987
From:	Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20205E495	List: IR 05000354/1987005 ML20205E491
References
50-354-87-05, 50-354-87-5, NUDOCS 8703300632
Download: ML20205E528 (10)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

050354-860821 050354-870123 Report No.

'50-354/87-05 050354-870123 050354-870123 Docket 50-354 050354-870126 050354-870127 License NPF-57 050354-870130 050354-870203 Licensee:

public Service Electric and Gas Company 050354-870207 Facility:

Hope Creek Generating Station Conducted:

February 10, 1987 - March 9, 1987 Inspectors:

R. W. Borchardt, Senior Resident Inspector D. K. Allsopp, Resident Inspector Approved:

k. -. b-- -- M 3/2./67 L. Norrholm, Chief, Projects Section 2B

/ [ Tate Inspection Summary:

Inspection on February 10, 1987 - March 9, 1987 (Inspection Report Number 50-354/87-05)

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, licensee event report followup, and security inspection. This inspection involved 146 hours0.00169 days <br />0.0406 hours <br />2.414021e-4 weeks <br />5.5553e-5 months <br /> by the inspectors.

Results: The failure to obtain and analyze grab samples of the cooling tower blowdown line is a licensee identified violation of a Technical Specification action statement (Paragraph 7). Although this event was not cited, it rep-resents the tenth violation of Technical Specifications due to personnel error in the radiation protection and chemistry areas since the plant was granted a low power license.

Increased management attention is needed to prevent ad-ditional Technical Specification action statement violations, especially in the radiation protection and chemistry areas. One unresolved item pertaining to environmental qualification of MSIV terminal boxes was also identified (Para-graph 3.2).

0703300632 070323 PDR ADOCK 05000354 O

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Details 1.

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

2.

Followup on Outstanding Inspection Items 2.1 Unresolved Items (Closed) Unresolved Item (86-17-01); During a previous inspection, a number of gauges and meters in the control room and relay room were found to be lacking permanent labels. The inspector verified that the inst-ruments in question have been properly identified and labeled. This item is closed.

(Closed) Unresolved Item (86-41-01); The inoperability of the reactor building to suppression chamber pressure relief system has been cited as a violation.

The licensee has corrected the sensing line inputs to the pressure differential transmitter (PDT) and has successfully tested the PDT for proper operation. The licensee's corrective action included a review of similar vacuum application of safety related PDTs with no problems identified.

The inspector had no further questions and this item is closed.

2.2 Violation (Closed) Violation (86-47-01); The inspector reviewed the licensee's response to this violation dated February 27, 1987. The safety relief valve acoustic monitor panel is now supplied from a IE, uninterruptible power supply which has been successfully tested during several subsequent loss of offsite power tests. The inspector had no further questions and this item is closed.

2.3 Inspector Follow Item (Closed) Inspector Follow Item (86-36-02); The inspector opened this item to followup on the licensee's long term corrective action to a through wall leak on a service water pipe elbow.

The licensee replaced all four service water elbows with piping which had been treated with a protective lining of Belzona epoxy.

Flow balancing valves, located just upstream of the through wall leak, will continue to be positioned in accordance with the startup test program which adjusts these valves to ensure adequate cooling flow during worst case flow conditions. A design change request has been submitted by the station which incorporates recommendations by i

Stone and Webster to minimize the erosion effect by reducing the differ-ential pressure drop across the flow balancing valves. The Itcensee's response to quality assurance corrective action report HS-86-020-0, was l

to revise Station Administrative Procedure AP.ZZ-020 to ensure all

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deficiency reports are reviewed by both the line supervisor and the senior nuclear shift supervisor for operability consideration. The inspector had no further questions and 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 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 inspection activities were conducted in accordance with NRC inspection procedure 71707.

3.2 Inspection Findings and Significant plant Events The unit entered this report period at 100% power and with unidentified drywell leakage slowly increasing.

At 4:05 a.m. on February 11, 1987, the licensee commenced a reactor shutdown and, at 6:35 a.m., declared an unusual event when the drywell unidentified leak rate exceeded the 5.0 gpm Technical Specification limit. The actual leak rate was calculated to be approximately 6 gpm. The unidentified leak rate gradually increased since February 9, and all attempts to identify and stop the source were unsuccess-ful. At approximately 1:00 p.m.,

reactor power was maintained at less than 20% power and the primary containment was deinerted to allow personnel entry into the drywell to conduct an inspection.

The inspection found that the leak was coming from a 300 degree crack in a 3/4 inch drain line from the "A" recirculation pump discharge valve (HV-F031A). The crack was located in the heat affected zone of the drain line to valve body weld and had been vibration induced. At 4:30 p.m., the reactor was manually scrammed to complete the reactor shutdown and the plant was taken to a cold shutdown condition.

The recirculation pump isolation valves at Hope Creek are Lunkenheimer flexible disc gate valves with full penetration welded, double valved drains. As a result of this failure the licensee took the following actions:

Inspected the 3 other recirculation pump isolation valve

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drain lines for related damage.

penetrant tests provided no indication of cracking.

The cracked pipe was sent to a laboratory for analysis and

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determination of failure mode.

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l The vendor and other BWRs were contacted to determine if this

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was a generic problem. The licensee is not aware of any previous similar failures.

A repair was made to the damaged line which included reducing

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the distance from the valve body to the first drain valve from approximately 6 to 3 inches. The licensee believes this will minimize any vibration induced problems.

At the completion of repair efforts a successful hydrostatic

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test was conducted on the valve body drain line.

The licensee completed repairs to the "A" recirculation pump dis-charge isolation valve drain line and the reactor was taken critical at 2:12 a.m. on February 14. A failure of the

"B" secondary conden-sate pump restricted power to approximately 88% until repairs were completed on day shift and power returned to 100%.

On February 16, the licensee officially declared commercial opera-tions.

At approximately 4:00 p.m. on February 18, 1987, the "E" filtration, recirculation, and ventilation system (FRVS) recirculation fan was observed to be running. The fan was stopped and all ventilation systems returned to normal. The licensee has been unable to deter-mine the cause for the fan start and is continuing the investigation.

At 10:22 a.m. on February 24, 1987, the plant inadvertently scrammed from 100% power while performing an I&C surveillance test of the main turbine and feed pump turbine trip systems for high reactor vessel water level. The licensee's initial determination of the cause of the scram was that an I&C technician apparently connected his test apparatus to the wrong set of contacts which actually caused a main turbine trip on indicated high vessel water level.

The ensuing transient caused actual vessel water level to decrease to level two, resulting in actuations of high pressure coolant injection, reactor core isolation cooling, and level two primary containment isolation system (PCIS). All systems performed as expected except that the "C" filtration recirculation and ventilation recirculation fan failed to auto start and 2 of 3 sample line valves associated with the "A" hydrogen / oxygen analyzer failed to shut in response to the PCIS signal. The licensee has determined the failure of two isolation valves to shut in response to a level two PCIS signal and the failure of the "C" FRVS fan to start were due to the extremely short duration of the initiating signal (Agastat relays are used for these components). The initiating signal was a reactor water level two, sensed immediately after the scram when all level instruments showed indications of ringing. The duration of the signal was suf-ficient to pick-up the Potter Brumfield relays but not the Agastat

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relays.

Licensee testing has verified that a signal duration of 56 milliseconds is needed for the Potter Brumfield relays to actuate; however, signal duration of 65 milliseconds is necessary for the Agastat relays to actuate. A review of the transient data indicates that the initiating signal lasted approximately 60 milliseconds and therefore-the equipment did respond as designed. The inspector will review the licensee's final root cause determination and corrective measures after the Licensee Event Report is submitted.

While the reactor was shut down, the licensee determined that the HPCI steam emission valve (F001) DC Limitorque motor operator may contain the degraded motor leads discussed in IE Information Notice 87-08. On February 28, the licensee replaced the entire Limitorque motor with a spare which does not contain the degraded motor leads.

While performing routine maintenance activities in the steam tunnel, the "A" inboard MSIV failed to close from the control room. A dry-well entry was made and it was found that one of the three "A" MSIV solenoids appeared to be burnt and the solenoid terminal box was not securely fastened to its mounting.

In addition, five MSIV terminal boxes were discovered which are larger than the standard MSIV terminal box and may not be in accordance with the vendor manual.

Two of these larger boxes had a hole drilled through one wall which violated the environmental qualification boundary. This matter is unresolved pending further inspection (87-05-02).

To ensure environmental qualification in the five larger MSIV terminal boxes, the licensee installed a Raychem splice in each box. All MSIV solenoids (24) were replaced with new solenoids and the loose terminal box was securely attached. The preliminary analysis indicates the cause of the "A" inboard MSIV burnt solenoid was vendor installation of a non-approved-solenoid poppet. The licensee has returned the burnt solenoid to the manufacturer for a detailed determination of the failure mode.

After replacing the MSIV solenoids, the licensee made the reactor critical on February 28, and synchronized the main generator to the grid at 4:16 p.m. on March 1.

Two minutes later, an empty oil tanker transiting the Delaware River collided with and damaged a tower supporting the 500 KV Keeney transmission line to Hope Creek. The Keeny line snapped and two isolation breakers in the Hope Creek switchyard opened isolating the line.

The accident did not result in any injuries. However, to maintain the grid within the transmission line stability curves, Hope Creek had to reduce power to 67% (700 MWE).

Even with the loss of the Keeney line, Hope Creek still has two

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independent offsite power supplies. The initial time estimate for repairs indicate that at least eight months will be required to restore the Keeney transmission line.

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On March 6, the control room received indication of a channel

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remote shutdown panel (RSP) takeover when a maintenance electrician was replacing a faulty cooling fan in 18D481, Class.1E channel "B" inverter. Plant c.ontrol never shifted to the RSP, however the control room received indication of a RSP takeover. During the inverter fan maintenance, the inverter transferred to its alternate source and returned back to its normal source. This electrical transient resulted in a blown fuse (FB044) in the RSP which caused the RSP takeover indication in the control room. The licensee is still investigating how the inverter fan maintenance initiated this event. The residents will review the licensee's investigation results in a future inspection.

On March 9, Salem Unit I shut down for a scheduled outage which allowed Hope Creek to increase power to approximately 92% and stay within the transmission line stability curves.

No violations were identified.

4.

Survefilance 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 perfo'rmed in accordance with Technical Specifications, licensee approved proce-dures, 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:

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IC-FT.BB-010.

Functional Test Main Steam Line Flow -

Div. 2 B21-N6888

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IC-SC.GS-008 Channel Calibration on Torus to Reactor Building Vacuum Breaker Transmitter 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 procedures, Technical Specifications, and appropriate industrial codes and standards. These inspections were conducted in accordance with NRC inspection procedure 62703.

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5.2 Inspection Findings Portions of the following activity was observed by the inspector:

Work Order Procedure Description 87-03-03-200-1 MD-GP.ZZ-031 Troubleshoot and Repair "D" Main Steam Line Outboard Sealing Valve No violations were identified.

6.

Engineered Safety Feature (ESF) System Walkdown 6.1 Inspection Activity 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 match plant drawings and the as-built configuration. This ESF 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 During the inspection of the control room emergency filtration (CREF) system, the inspector determined that the identical inline smoke detectors (XSH 9588A2, XSH 9588A1) on the "A" CREF train were mislabeled. The correct labeling of these "A" train smoke detectors will be tracked as open item 87-05-01.

No violations were identified.

7.

Licensee Event Report Followup The licensee submitted the following event reports during the inspection period. All of the reports were reviewed for accuracy and timely sub-mission. The asterisked reports received additional followup by the inspector for corrective action implementation.

The (#) items identify reports which involve licensee identified Technical Specification viola-tions which are not being cited based upon meeting the criteria of 10 CFR 2 Appendix C for licensee identified violations.

Special Report 87-001-00 North Plant Vent Radiation Monitor Inoperable For More Than 72 Hours

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LER 87-005-00 Primary Containment Isolation System Actuation

Due to Momentary Loss of Power to Radiation Monitoring Equipment LER 87-006-00 Eng'ineered Safety Feature (ESF) Actuation Due to Personnel Error During Surveillance

• LER 87-007-00 Unanticipated Actuation of HPCI Outboard Steam Supply Valve Caused by High Temperature Differential - Design Error LER 87-008-00 Inadvertent Start of Core Spray Pump "B" Due to Improper Performance of Surveillance Procedure

• 1. LER 87-009-00 Undetected Inoperability of the Cooling Tower Blowdown RMS Sample Pump Resulting in Violation of Technical Specifications - Personnel Error LER 87-010-00 Nuclear Steam Supply Shutoff System Channel A

Isolation Caused by Grounding Test Equipment in Steam Leak Detection Cabinets

1. LER 87-011-00 Delayed. Test of Four Licensed Radioactive Sources

- Technical Specification Violation Due to Personnel Error LER 87-012-00 Reactor Water Cleanup System Isolation Due to Spurious Signal Induced by Temperature Monitoring

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Modules LER 87-005 describes a primary containment isolation system (PCIS)

actuation which occurred when the reactor building exhaust radiation monitoring (RBERM) channel "A" received a momentary trip at the same time that channel "C" was in a tripped condition for a surveillance test. All systems responded to the PCIS actuation as expected. No root cause for the momentary and spurious channel "A" RBERM trip could be determined. Details of this event are discussed in section 3.2 of NRC Inspection Report 87-01.

LER 87-007 details the events which resulted in the high pressure coolant injection (HPCI) outboard steam supply valve isolating in response to a high temperature differential between the HPCI area ventilation intake and exhaust temperature.

This temperature differential is interpreted as a steam leak and occurred due to cold weather coupled with a low setting in the room supply thermostat.

Corrective action included increasing the intake air temperature from 40 degrees to 60 degrees F by adjusting the supply air heater controller.

LER 87-009 describes a personnel error which resulted in a Technical Specification violation after the cooling tower blowdown radiation monitoring sample pump froze. An annunciator alarmed in the control

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room when the sample pump froze but was incorrectly interpreted by the operators. Corrective action included counselling of the operational staff as to the appropriate investigative action to take in response to all alarms.

LER 87-010 explains a nuclear steam supply shutoff system (NSSSS)

isolation which occurred during a surveillance test. An I&C technician inadvertently grounded a test lead which resulted in a blown fuse which caused the NSSSS isolation. The root cause of the actuation is the insufficient work space inside the steam leak detection panels. Corrective action consisted of expediting a design change to improve access to monitoring points inside the panel.

8.

Security - Protected Area Physical Barriers The inspector condu'cted an inspection to verify that the licensee has installed and maintains physical barriers surrounding all protected areas in conformance with his physical security plan and regulatory requirements. The inspector also verified that the physical barriers

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were adequate and appropriate to perform their intended function.

No violations were identified.

9.

Allegation Closeout On August 27, 1985, an individual brought a number of concerns to the attention of the NRC relating to construction and testing activities at Hope Creek. As part of the normal NRC inspection activity at Hope Creek and as followup to the alleger's concerns, a number of independent NRC inspections have been conducted. The results of these inspections indicate that the end product of the Hope Creek construction program meets all regulatory requirements and if problems existed in the I&C calibration area they were corrected prior to licensing. These independent inspec-tions were conducted by the residents, the As-Built Team inspectors, and the Technical Specification inspection team.

In addition, the licensee's performance of all Technical Specification surveillance tests prior to entering the applicable operational condition provides added assurance of important system operability, including proper setpoints and calibrations.

In parallel with these inspection activities, the licensee was requested by the NRC to respond to a number of questions relating to the alleger's concerns.

The licensee's response and the inspector's observations indicate that the valid safety concerns raised were already being resolved or have been subsequently resolved in a manner consistent with the site quality and preoperational test programs. Although a number of the alleger's concerns were apparently valid, they have been adequately addressed by the licensee and no outstanding safety issues remain open.

The inspector has no further concerns pertaining to this issue at this time.

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10.

Exit Interview-The inspectors met with M.*. R. Salvesen and other: licensee personnel periodically and at thri.end of the inspection period 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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