IR 05000354/1986056

Resident Insp Rept 50-354/86-56 on 861118-1231.No Violations Noted.Major Areas Inspected:Followup on Outstanding Insp Items,Operational Safety Verification,Surveillance Testing, Maint Activities,Esf Sys Walkdown & LER Followup

ML20207P972
Person / Time
Site:	Hope Creek
Issue date:	01/12/1987
From:	Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20207P968	List: IR 05000354/1986056 ML20207P967
References
50-354-86-56, NUDOCS 8701200498
Download: ML20207P972 (8)

Text

.

7_ ,

.

.

U. S. NUCLEAR REGULATORY COMMISSION

REGION I

050354-861010 050354-861018 Report N /86-56 050354-861019 050354-861022 Docket 50-354 050354-861027 050354-861028 License NPF-57 050354-861111 050354-861114 Licensee: Public Service Electric and Gas Company 050354-861114 050354-861117 Facility: Hope Creek Generating Station 050354-861118-050354-861119 Conducted: November 18 - December 31, 1986 Inspectors: R. W. Borchardt, Senior Resident Inspector :

opp, Resident Inspector Approved: / / L/ 7 L.LNorrholm, Chief, Projects Section 2B / Dat'e Inspection Summary:

Inspection on November 18, 1986 - December 31, 1986 (Inspection Report Number 50-354/86-56)

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 GE HGA Relay applications. This inspection incolved 151 hours0.00175 days <br />0.0419 hours <br />2.496693e-4 weeks <br />5.74555e-5 months <br /> by the inspector Resul_ts: No violations were identifie PDR ADOCK 05000354 0 PDR ._ .

.

Details 1. Persons Contacted Within this report period, interviews and discussions were conducted with members of the licensee management and staff and various contractor personnel as necessary to support inspection activit . Followup on Outstanding Inspection Items 2.1 TMI Action Plan Items (Closed) TMI Action Plan Item III.0.1.1; Primary Coolant Leakage Reduction Outside Containment The licensee has committed in FSAR section 1.10 to implement a leakage reduction program for those systems containing reactor coolant or radioactive gases outside containment that must function during an accident. The following systems are included in the Hope Creek leakage reduction program:

Reactor Core Isolation Cooling (RCIC) System Residual Heat Removal (RHR) System Core Spray (CS) System High Pressure Coolant Injection (HPCI) System Hydrogen /0xygen Analyzer System Post Accident Sampling System Hydrogen Recombiner System Control Rod Drive System (Scram discharge portion)

Design features that are included in the above systems to minimize leakage include: (1) pumps have mechanical seals; (2) welded pipe construction; and (3) systems are isolabl The inspector reviewed station administrative procedure SA-AP.ZZ-051(Q) " Leakage Reduction Program" and also reviewed the baseline leakage data. The station leakage reduction coordinator is a member of the plant technical staff who is responsible for overall coordination of the program and for maintaining app'icable record The leakage inspection for all systems is performed during periodic surveillance tests which require operation of the system. The maximum interval between inspections is 18 months, with many of the fluid systems inspected quarterly. This item is considered close . Operational Safety Verification 3.1 Documents Reviewed

-

Selected Operator's Logs

-

Senior Shift Supervisor's Log

-

Jumper Log

-

Radioactive Waste Release Permits (liquid & gaseous) _ _ . _ _ _ _ .

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - - _ _ _ _ ,

-.

, .

-

-Selected Radiation Work Permits (RWP)

-

Selected Chemistry Logs

-

Selected Tagouts

-

. Health Physics Watch Log 3.2 The inspectors periodically toured the plant during regular and backshift periods. These tours included the control room, Reactor, Auxiliary, Turbine and Service Water buildings, and the drywell (when access is possible). During the inspection activities, discussions were held with operators, technicians (HP & I&C), mechanics, super-visors, and plant management. The purpose of the inspection was to affirm the licensee's commitments and compliance with 10 CFR, Technical Specifications, and Station Procedure . On a daily basis, particular attention was directed to the following areas:

-

Instrumentation and recorder traces for abnormalities;

-

Adherence to LCO's directly observable from the control room;

-

Proper control room shift manning and access control;

-

Verification of the status of control room annunciators that are in alarm;

-

Proper use of procedures;

-

Review of Logs to obtain plant conditions; and,

-

Verification of surveillance testing for timely completio . On a weekly basis, the inspectors confirmed the operability of selected ESF trains by:

-

Verifying that accessible valves in the flow path were in the correct positions;

-

Verifying that power supplies and breakers were in the correct positions;

-

Visually inspecting major components for leakage, lubrication, vibration, cooling water supply, and general operating conditions; and,

-

Visually inspecting instrumentation, where possible, for proper operabilit . On a biweekly basis, the inspectors:

,_ . _ _ _ _

.-

I

.

-

Verified the correct application of a tagout to a safety-related system;

-

Observed a shift turnover;

-

Reviewed the sampling program including the liquid and

. gaseous effluents;

-

Verified that radiation protection and controls were properly established;

-

Verified that the physical security plan was being implemented by monitoring:

1) Fixed and mobile guard posts; 2) Personnel, vehicle, and package control into the protected area; 3) Protected area and vital area detection aids; 4) Compensatory action taken when a security barrier is degraded; and, 5) Use of photo identification badges within the protected are Reviewed licensee-identified problem areas; and,

-

Verified selected portions of containment isolation lineu .3 Inspector Comments / Findings:

The unit entered this report period in operational condition 4 '

(cold shutdown) while conducting short term outage activities, including replacement of the "B" RHR pum On November 19, 1986, the reactor water cleanup (RWCU) system isolated when an I&C technician inadvertently bumped and disconnected a loose wire in the 10C609 cabinet while performing surveillance test IC-FT.SK-039, "RWCU Steam Leak Detection".

The RWCU system performed as expected for plant conditions (operational condition 4). The RWCU isolation was reset and the system returned to normal operatio ,

At 3:55 p.m. on December 4, the reactor was taken critical for continuation of the power ascension test program. At 9:25 on December 5, 1986, the "P" Safety Relief Valve (SRV) lifted on

.. .

-

..

e 1 A - r 5 e

,

^3 ;

$, : k. d low low set point due to personnel error while performing a surveil-' /

lance test. The reactor was at 70% power when the event occurre The SRV was shut from the control room and the plant continued to:make preparations for the load reject scram which was conducted on December During the load reject scram, High Pressure Coolant Injection, Reactor-

- Core Isolation Cooling, and the "D" Primary Containment Isolation System actuated when ringing in the reactor vessel level instruments produced spurious -38 inch, and +54 inch _ signals. Actual reactor-vessel level never went below -30 inches. Although instrument ringing

- has been observed at other BWR facilities, it appears to be more sig-nificant at Hope Creek. The licensee and General Electric conducted reviews which show that ringing is most noticeable after a turbine trip - load reject, HPCI actuation, and a safety relief valve lif The "A" and "0" instrument racks were backfilled in order to r'emove any air that could have possibly been trapped in the sensing line The "C" rack was not backfilled since it~has not experienced ringing and the "B" rack was left "as is" in order,to use as a reference to evaluate the corrective actions on "A" and "D". The licensee is also evaluating the use of instrument transmitters with a slower response time than the model 1153 Rosemounts currently in use. A longer res-ponse time would prevent the actuation of systems due to a momentary spike. The inspector will continue to follow the licensee's action in this are At 5:30 a.m. on December 9, 1986, an RHR shutdown cooling isolation occurred due to an I&C technician personnel error. The technician was backfilling an instrument line using a 120 PSI water source and exceeded the high pressure isolation setpoint of 82 PSI. The isol tion was reset and shutdown cooling returned to service by 6:00 h..a- The reactor was taken critical on December 10, and the turbine generator was synchronized to the grid on December 12. The unit remained in operational condition 1 for the remainder of the report perio ~

During this report period, the plant was inspected for cold weather preparation. As a result of this inspection, the licensee determined that the hydrogen / oxygen analyzer heat trace alarm was not electrically connected to the control room common heat trace alarm circuit. The licensee initiated a work order to include the hydrogen / oxygen analyzer alarm in the control room common heat trace alarm. A problem also exists with '

equipment operators not fully understanding operation of the heat trace control panels in that they did not initiate corrective action for power failure lights or alarm lights on these heat trace control panels. In addition, numerous heat trace controller setpoints were found not to be consistent with the Instrument Calibration Data (ICD) card settings. These items will be followed up during the next inspectio (86-56-01)

-

..

.

No . violations were identifie j Surveillance Testing During this inspection period the inspector performed detailed technical procedure reviews, reviewed in progress surveillance testing as well as completed surveillance packages. The inspector also verified that the surveillances were performed in accordance with licensee approved proce-dures and NRC regulations. The inspector also verified that the instru-ments used were within calibration tolerances and that qualified techni-cians performed the surveillance The following surveillance was reviewed, with portions witnessed by the inspector:

-

IC-FT.BB-33 SRV Low Low Setpoint functional test No violations were identifie . Maintenance Activities 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 Specifi-cations, and appropriate industrial codes and standard Portions of the following activity was observed by the inspector:

Work Order Procedure Description 86-12-04-033-7 MD-ST.PK-001 Preventative maintenance on station batteries No violations were identifie . Engineered Safety Feature (ESF) System Walkdown The inspectors verified the operability of the selected ESF system by performing a walkdown of accessible portions of the system to i

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 appropriat The Filtration, Recirculation, and Ventilation system was inspecte No violations were identifie '

n

- _ _

~

.

,

,

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 implementatio * LER 86-70 "C" Core Spray Pump Discharge Pressure Transmitter Isolated

• LER 86-77 Inadvertent Isolation of Reactor Water Cleanup System

• LER 86-78 Reactor Water Cleanup System Isolation

• LER 86-79 Isolation of Reactor Cleanup LER 86-80 Full Reactor Scram on Low Reactor Water Level 3 LER 86-82 High Pressure Coolant Injection System Inoperative LER 86-85 Full Reactor Scram Due to Reactor High Pressure Signal

• LER 86-86 Reactor Building Ventilation Isolation and Actuation of FRVS LER 86-87 Inadvertent Actuation of the "A" Channel NSSSS Isolation Logic

• LER 86-88 Loss of Residual Heat Removal Room Cooling LER 86-89 Inadvertent Isolation of Reactor Water Cleanup System Special Report 86-005; Failure to Submit Control Room Design Review, Supplement II, in Accordance with License Condition 2.C.1 LER 86-70 describes the discovery of the "C" core spray pump discharge pressure transmitter being isolate This transmitter being isolated prevented the Automatic Depressurization System from sensing eperation of one loop of the core spray system. Details of this event are discussed in paragraph 6 of NRC Inspection Report 86-4 LER 86-77, LER 86-78, LER 86-79 are all related in tSat they involve isolations of the Reactor Water Cleanup (RWCU) system due to high differential flow. LER 86-77 and LER 86-79 RWCU isolations resulted from false flow indication due to the blowdown line isolation valve indicating full shut when it was actually partially open. LER 86-78 RWCU isolation occurred while placing "B" filter demineralizer in

.

.

.

.

service with a significant RWCU leak which placed the system near its isolation setpoint. Details of these events are discussed in paragraph 3.3 of NRC Inspection Report 86-4 LER 86-86 described the isolation of the Reactor Building Ventilation

. System and automatic actuation of the Filtration, Recirculation, and Ventilation System (FRVS) from a valid high radiation signal. The high radiation signal was due to a steam leak from the Reactor Water Cleanup system in conjunction with high reactor coolant activity from recent testing and a scram induced crud burst. Details of this event are discussed in paragraph 3.3. of NRC Inspection Report 86-4 LER 86-88 describes the "A" and "D" residual heat removal (RHR) room coolers being declared inoperable. The low flow switch settings on these room coolers were calibrated to incorrect instrument calibration data (ICD) which had not been updated to correct calibration setpoint The corrective action included updating the ICD cards and correctly calibrating the low flow switches on the A" and D" RHR room cooler . GE HGA Relay Applications at Hope Creek Testing by Wyle Engineering company has shown the GE HGA relays are not suitable for use in safety related applications due to contact chatter during seismic testing. The inspector has verified through plant walkdowns, document reviews and discussions with system engineers that these relays are not used in any safety related applications at Hope Cree . Exit Interview The inspectors met with licensee and contractor personnel periodically and at the end of the inspection report to summarize the scope and findings of their inspectiE 7tivitie 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 restriction ,