IR 05000354/1986020

Insp Rept 50-354/86-20 on 860317-0430.Violation Noted: Filtration,Recirculation & Ventilation Sys Inoperable During Core Alterations & Failure to Perform Surveillance Testing of Standby Liquid Control Sys

ML20198E225
Person / Time
Site:	Hope Creek
Issue date:	05/09/1986
From:	Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20198E192	List: IR 05000354/1986020 ML20198E187 ML20198E211
References
50-354-86-20, IEB-80-10, IEB-80-16, IEB-80-17, NUDOCS 8605270316
Download: ML20198E225 (18)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /86-20 Docket 50-354 License NPF-50 Licensee: Public Service Electric and Gas Company Facility: Hope Creek Generating Station Conducted: March 17 - April 30, 1986 Inspectors: R. W. Borchardt, Senior Resident Inspector D. K. Allsopp, Resident Inspector J. J. Lyash, Reactor Engineer r 'ect Engineer

E. L. Nim Conne t , Ra P, lation Specialist . 8,1 e l bse,RadiatonSecialist Approved: /_ ' A, -

[7-Dat6 L.' Norrpig', Chfef, Projects Section 2B Inspection Summary:

Inspection on March 17, 1986 - April 30, 1986 (Inspection Report Number

50-354/86-20)

Areas Inspected: Routine onsite resident inspection of the following areas: followup on outstanding inspection items, operational safety verification, preoperational test results review, surveillance testing, engineered safety feature system walkdown, licensee event report followup and NRC Commissioner's visit. This inspection involved 530 hours0.00613 days <br />0.147 hours <br />8.763227e-4 weeks <br />2.01665e-4 months <br /> by the inspectors.

i Results: This report documents two violations identified by NRC inspectors that indicate an increased focus on details may be warranted by the operating staff. The inoperability of the Filtration Recirculation and Ventilation system (FRVS) and the failure to perform required surveillance testing of the Standby Liquid Control System during core alterations could have been avoided had the control room operators paid more attention to the details of plant systems and Technical Specification requirements. As the plant approaches power operations the complexity of system operability requirements

increases considerably as does the need for strict attention to detail. This i

concept is equally applicable to all other departments at Hope Creek since their actions have the potential for increased consequences as power operation l is commence DR 860520 ADOCK 05000354 PDR t

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Details 1. Persons Contacted

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Within this report period, interviews and discussions were conducted with members of the licensee management and staff and various contractor per-sonnel as necessary to support inspection activit . Followup on Outstanding Inspection Items 2.1 Violations

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(Closed) Violation (85-51-02); Reversed Transmitter Wires. Inspection Report 85-51 identified a violation of Appendix B, Criterion V in which wires from flow transmitter E51-FT-N051 and pressure transmitter E51-PT-N051, inside cabinet H11-P618, were reversed from that shown on the appropriate circuit diagram. PSE&G responded to this violation by letters dated January 14 and March 11, 1986. Their corrective steps included the determination that the root cause was an error on the part of craft per-sonnel, the visual inspection of 870 field terminations supported by their

! QC group, the resolution of six discrepancies identified during the visual inspection, and the corrective actions to be accomplished prior to initial criticalit The inspector reviewed SDR 1047, GE Panel Terminations, and the related

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Hope Creek Deficiency Reports, HTE-86-066 through HTE-86-071. The visual inspection of 870 terminations was an approximate 10% sample of the EE-580 i computer generated listing of all Q listed GE terminations from the field (incoming wires to GE control cabinets). The sample was heavily weighted to terminations with two (black, white) and three (black, white, black-shield ground) wires since this is the type where the violation was foun The inspecter confirmed a random sample of 28 connections and talked to the QA people involved with the 10% sample. Although the apparent reason for the switching of termination wires is error on the part of craft i personnel, GE's failure to use the industry standard color coding system may be a contributing factor. Since the termination problem was identi-fied by I&C Startup Test Engineers during functional loop tests and all circuits have or will receive these tests, any other termination problems will be discovered. This issue is close (Closed) Violation (86-06-02); Failure to adequately test the core spray diesel power monitor. This Notice of Violation identified a failure to adequately test the described circuit. During the Preoperational Test Program three levels of testing were scheduled which could have detected the missing cables. In order of execution these tests were component

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level checks, system preoperational testing, and integrated ECCS initiation testin In discussions with the inspector the licensee stated their ' view that the logic need not be tested in the system preoperational test because it was tested in the integrated ECCS initiation test. The inspector stated that this did not relieve requirements to perform ade-quate component level tests. While the licensee's response did not

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address the test control issue, the inspector's continuing review of the preoperational test program and the conduct of surveillance testing, as well as the fact that the preoperational test program is essentially complete, provide assurance that similar situations will not occu The-inspector reviewed the results of audits conducted by the licensee to identify any other similar problems. The audits conducted include review of design documents from elementary drawings through actual cable installation. These audits in combination with the inspector's own activ-ities indicate that the violation was an isolated incident. This item is close .2 Unresolved Items (0 pen) Unresolved Item (85-44-14); Proposed deferral of portions of the Radiation Monitoring System (RMS). This item was reviewed during Inspection Report 86-07 and the applicant's plans were discussed at a -

meeting on January 17, 1986. At the meeting, the applicant indicated that Detailed Test Procedures (DTP) would be completed for components of the RMS to ensure their operability. During this inspection, completed /in progress DTPs for the following monitors were reviewed relative to commitments in various applicant letters, the Hope Creek Generating Station Final Safety Analysis Report and the draft Technical Specifications:

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Cooling Tower Blowdown Monitor (RE-8817);

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Liquid Radwaste Monitor (RE-4861);

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South Plant Ventilation Monitors (RE-4875A/RE-4875B);

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North Plant Ventilation Monitors (RE-4873A/RE-48738);

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Off gas Post-treatment Monitors (RE-6281/RE-6282); and

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Filtration, Recirculation and Ventilation System Vent Monitor (RE-4811)

Within the scope of this review, the applicant appeared to be meeting previous commitments for the monitors reviewe (Closed) Unresolved Item (86-06-03); Control Room Labeling for Rosemount 1151 Transmitters. During inspection 85-56 and 86-06, the inspector reviewed PSE&G's response to IE Bulletin 80-16, Potential Misapplication of Rosemount Models 1151 and 115 Inspection Report 86-06 documents closure of the bulletin concerns except for the labeling of control room indications that should not be relied upon during transient / accident conditions because their level signal is provided by Rosemount 1151 pressure transmitters. DCP 7110 was written to do the labelin The inspector confirmed, by drawing review and personal observation that the appropriate level indications / recorders have been labeled in accord-ance with DCP 7110. The inspector has no further question .

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(Closed) Unresolved Item (86-15-02); Installation of safety relief valve acoustic monitors. During inspection report period 86-15 the inspector identified four uninstalled acoustic monitor accelerometers. These com-ponents-had previously been installed and tested under Nuclear Department Work Order (WO) 86-02-14-061-4. The licensee could produce no documenta-tion authorizing their removal or tracking reinstallation / retest. The licensee cited ongoing painting / cleanup activities in the drywell as a possible cause for the damaged monitors. Greatly reduced levels of activity in the drywell will preclude / minimize this type of damag Surveillance testing as well as area walkdowns should assure detection of any similar damage. The inspector, at a later date, verified proper installation of the 14 instrument Proper retest will be verified during future inspection of open item 86-10-0 .3 Inspector Follow Items (Closed) Inspector Follow Item (85-44-13); Gaseous Radioactive Waste System Valve (V)-246 fails open and the radiation monitor upstream of V-246 was not installed. An engineering review of the function of V-246 (0-HA-HV 10285) provided by the applicant was reviewed and discussed with applicant's representatives. Other valves in the system provide isolation by closure upon failure of the gaseous recombine The post treatment radiation monitors had been installed. This item is close (Closed) Inspector Follow Item (85-52-02); Actions regarding NRC Bulletin No. 80-10. The applicant revised Chemistry Technical Instruction (CH-TI.ZZ)-012, " Chemistry Sampling Frequencies, Specifications and Sur-veillances," to include sampling frequencies, gamma isotopic analyses and appropriate action levels for normally nonradioactive system Station Administrative Procedure (SA-AP.ZZ)-020, "Nonconformance Program," provided instructions for conducting safety reviews for systems with detectable radioactivity following analysis under CH-TI.ZZ-01 (Closed) Inspector Follow Item (85-52-12); Licensee to establish and implement a neutron monitoring program. The licensee established and implemented a neutron monitoring program. The program is described in Procedure RP-TI.ZZ-014Q, Revision 1, " Personnel Monitoring Program."

The licensee will provide and use neutron dosimetry when the presence of neutrons is detected. The licensee will also provide for estimation of personnel neutron dose received using neutron survey meter readings in conjunction with time and motion studie (Closed) Inspector Follow Item (85-52-16); Verify calculation of line loss for_ particulate sampling of the drywell. The applicant completed line loss tests which indicated that particulate losses for the drywell sampling line exceeded calculated values. Based upon these results the licensee does not plan to use this sampling line for particulate analysis for radiation protection purpose The sampling line could still be used for other sampling purposes. The inspector has no further question .

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(Closed) Inspector Follow Item (85-52-18); Licensee to establish proce-dures for issue, use and control of personnel lapel air samplers. The Itcensee established and implemented procedures for issue, use and contcol of lapel air sampler (Closed) Inspector Follow Item (85-52-25); Licensee to establish a mechanism to upgrade radiation work permit radiological controls in the event of changing conditions. The licensee established and implemented Department Directive RP-DD.ZZ-013. The directive provides guidance relative to action to be taken when significant changes to plant radio-logical controls are identifie (Closed) Inspector Follow Item (85-52-27); Review procedures for opera-tions of GeLi detectors. The applicant completed procedures for routine operation, calibration, operability and quality contro (Closed) Inspector Follow Item (85-52-33); Review radwaste operating procedure development. During Inspection Report 86-05, the applicant stated that certain operating and alarm response procedures would be completed by initial fuel load with the remaining procedures to be completed by initial criticalit The completion of radwaste procedures was reviewed and determined to be adequate to support initial fuel load and criticalit (Closed) Inspector Follow Item (85-58-03); Misapplication of scram dis-charge volume (SDV) valve tags. During the original inspection the licensee provided Startup Deficiency Report (SDR) BF-270 correcting the discrepancy identified. During this report period the inspector reviewed the licensee's overall valve tagging program and discussed with applicable personnel the implementation of this program. Site Engineering Instruction (SEI) 7.4, Revision 1, Preparation of P&ID Valve Drawings was reviewed to determine the method used to assure all valves are correctly tagged. This procedure assigns site engineering the responsibility of performing system walkdowns to ensure drawing / valve number accuracy. Any untagged or mis-

. tagged valves identified are to be identified and corrected. The above is a continuing program currently in its final stages. The inspector, during

routine inspections, has verified tagging accuracy of main flow path l components of safety system The inspector will continue these observa-

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tions in future inspection This item is close (Closed) Inspector Follow Item (85-64-04); Calibration of HPCI/RCIC High

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Steam Flow Isolation Time Delay. The inspector reviewed approved proce-dures IC-FT.FC-002(Q), IC-FT.FC-001(Q), IC-FT.FD-002(Q) and IC-FT.FD-001(Q);

HPCI/RCIC Steam / Instrument Line Break Auto-Isolation. These procedures include steps and acceptance criteria which implement the Technical Speci-fication requirements concerning the high steam flew time delay setpoin These surveillance tests are scheduled for completion during power ascen-

. sion, prior to declaring HPCI/RCIC operable. This item is closed.

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(Closed) Inspector Follow Item (85-64-09); Low condensate storage tank level HPCI suction transfer. The applicant has processed a Technical Specification change specifying the transfer setpoint and allowable value in gallons of water. The inspector reviewed calculation number SC-AP-0001 to ensure-that the new Technical Specification setpoint in gallons had been correctly translated into the required MA output trip signal. The inspector also reviewed the appropriate instrument calibration data cards and the functional test acceptance criteria to verify the incorporation of new values. All documents reviewed appear to be consistent with Technical Specifications. This item is close (0 pen) Inspector Follow Item (85-68-05); RCIC pump functional test and flow verification. This item is updated to include a similar issue con-cerning the HPCI pump functional test and flow verificatio Surveillance requirement 4.5.1.b.3 requires HPCI pump flow of 5600 gpm against a test line pressure of 1000 psig with steam supplied to the turbine at 1000 +20,

-80 psig. Procedure OP-IS.BJ-001 which implements this test requirement specifies necessary steam pressure as greater than 200 psig. This is not consistent with the 1000 +20, -80 psig requirement specified in the Tech-nical Specification Similar discrepancies in the RCIC system test had not been adequately corrected at the time of this report. The inspector identified the HPCI/RCIC concerns to the licensee and will evaluate corrections during a future inspectio (Closed) Inspector Follow Item (86-05-02); Temperature regulated con-ductivity measurements. The applicant had prepared and scheduled two design changes to provide temperature regulated conductivity measurements for the Condensate Demineralizer System. These changes will be completed prior to initial criticalit (Closed) Inspector Follow Item (86-05-06); Review results of preoperational tests of the Demineralized Water Syste Three preoperational test pro-cedures testing the Demineralized Water System had been completed and test results were reviewe (Closed) Inspector Follow Item (86-06-01); Safety relief valve accumulator leak rate testing. The inspector reviewed approved Nuclear Site Mainte-nance Inservice Inspection Group procedure M9-ILP-305, Revision 0 This procedure will be utilized to perform periodic leak testing of the fourteen safety relief valve accumulators. The inspector verified that all appro-priate accumulators are included for test and that applicable leak rate acceptance criteria have been specified. The inspector had no further question (Closed) Inspector Follow Item (86-07-09); Licensee to establish adminis-trative control for use of airborne radioactivity sampling equipment

consistent with calibration limitation The licensee established administrative controls to address this matter.

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(Closed) Inspector Follow Item (86-15-03); Containment nitrogen purge valve position / leak rate test requirements. The inspector reviewed applicable sections of Hope Creek Technical Specifications as issued with the station operating license. Adequate revisions have been made requiring this unqualified valve to be locked closed and requiring necessary leak rate testing. The inspector also reviewed approved procedure OP-ST.GS-002, Drywell and Suppression Chamber Purge System Valve Verification - Monthly, Revision 2. Steps had been added to seal closed, and periodically verify closed, the subject valve. The inspector had no further question (Closed) Inspector Follow Item (86-15-04); RPS wiring discrepancies. The licensee informed the inspector that the concerned circuit had been rewired in accordance with the approved drawings by a Design Change Pack-age processed on the same trip units. The inspector was provided, and reviewed, Nuclear Department Work Order (WO) 86-04-21-015- The WO documents inspections conducted to ensure that the subject circuit is installed in accordance with the design documents. The inspector had no further question .5 Construction Deficiency Reports (Closed) Construction Deficiency Report 86-00-04; Anchor Darling Testable Swing Check Valves. The licensee identified a problem with certain Anchor Darling testable swing check valves that would stick open and not close when fluid flow was stopped. The analysis of this deficiency indicated that when the packing gland was tightened to eliminate packing leakage, the weight of the check valve disc could not overcome the friction between the packing and the actuator shaf Design Change Package 7189 was initiated to modify the actuator shaft to allow the valves to function as a simple swing check valve with assist open levers. This modification allows the valves to open and close due to flow conditions without causing the actuator shaft to move. The inspector reviewed the licensee's safety evaluation report, deficiency report HMD-86-099, DCP-7189, the applicable work orders used to accomplish the modifications and the applicable retest  :

procedures. No problems were identified and the inspector has no further question (Closed) Construction Deficiency Report (86-00-05); Nutherm Supplied GE THC325 Disconnect Switches. On March 20, 1986, as a result of vendor supplied information, PSE&G reported the defective switches as a possible significant construction deficiency. On April 14, 1986 the licensee, based on engineering evaluation, determined that the defect was not reportable and the CDR was withdrawn. The inspector reviewed the refer-enced engineering evaluation to verify that had the defect gone undetected ,

there would be no impact on system operation. The subject switches are used as power disconnect switches for standby liquid control (SLC) room

duct heater control panels. The SLC storage tank is provided with a tank

heater and the associated piping is equipped with electrical heat trac The heater / heat tracing alone are sufficient to maintain solution temper-atures at the required leve Loss of the duct heaters would not result i

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in impairment of SLC. In review of the deficiency reports (DR) generated to document the defective switches, the inspector noted that these DRs did not specify rework / repair requirements for the defective switches. The inspector expressed concern that although the defect is not reportable, appropriate measures should be taken to correct the problem. In response to the inspector's concerns the DR dispositions were revised to specify that the defective switches will be scrapped and replaced. The inspector had no further question . Operational Safety Verification 3.1 Documents Reviewed

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Selected Operators' Logs

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Senior Shif t Supervisor's (SSS) Log

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Jumper Log

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Radioactive Waste Release Permits (liquid & gaseous)

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Selected Radiation Work Permits (RWP)

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Selected Chemistry Logs

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Selected Tagouts

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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, Tech-nical Specifications, and Station Procedure (1) On a daily basis, particular attention was directed to the following areas:

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Instrumentation and recorder traces for abnormalities;

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Adherence to LC0's directly observable from the control l

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Proper control room shift manning and access control;

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

Proper use of procedures;

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Review of Logs to obtain plant conditions; and,

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Verification of surveillance testing for timely completion.

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(2) On a weekly basis, the inspectors confirmed the operability of selected ESF trains by:

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

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

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Verifying that de-energized portions of these systems were de-energized as identified by Technical Specifications;

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

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Visually inspecting instrumentation, where possible, for proper operabilit (3) On a biweekly basis, the inspectors:

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

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Observed a shift turnover;

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Reviewed the sampling program including the liquid and gaseous effluents;

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Verified that radiation protection and controls were properly established;

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Verified that the physical security plan was being implemented;

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Reviewed licensee-identified problem areas; and,

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

During the first half of this report period the resident inspectors and the region I staff performed an in-depth review of the Hope Creek station's readiness for issuance of an operating license and com-mencement of fuel load activities. This review included inspections in the areas of preoperational test results, test exception resolution, surveillance testing, mode 5 required system operability determinations, safety evaluation reviews and individual department readiness. The inspector attended a number of Station Operations Review Committee (SORC) meetings to verify that SORC fulfilled its

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oversight role as defined by Technical Specifications (TS). As further discussed in paragraph 5 of this report, the inspector wit-nessed and conducted technical reviews of numerous surveillance tests to independently verify that TS surveillance testing requirements would be satisfied and to provide an added degree of assurance that required systems would be operabl On April 11, 1986, Facility Operating License NPF-50 with Technical Specifications and the Environmental Protection Plan was issued to PSE&G for the Hupe Creek Generating Statio License NPF-50 authorizes operation of the reactor at power levels not in excess of 3293 megawatts thermal (100*;) with power restricted to ST4 pending completion of low power testing and Commission approva Continuing problems with the "B" Source Range Monitor (SRM) and the licensee's inability to quickly resolve those problems required the fuel load procedures to be modified to allow the use of a fuel loading chamber (FLC) in place of an SRM. The inspector verified that the implementation of these modified fuel loading procedures satisfied the requirements of TS. After resolving problems with the control room emergency filtration system and the "C" source range nuclear instrument, fuel loading activities commenced on April 15. The inspectors provided 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day coverage of fuel load activities during the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, at which time normal inspection activities were resumed. During the first two days of fuel load activities the licensee received two inadvertent RPS actuations. An I&C technician working on the "B" SRM preamp bumped the "F" IRM cable, causing an IRM upscale RPS trip. At a later time, an individual LPRM failed upscale. Both the LPRM and APRM gains are set high, resulting in the single LPRM failure causing an "F" APRM trip. The shorting links were removed, as required, so that each of the above described incidents resulted in a non-coincident reactor scram. No rods were withdrawn at the time of either scra On April 16, 1986, during a routine control room tour, the inspector noted that all six Filtration, Recirculation and Ventilation System (FRVS) recirculation units and both ventilation units had been placed in lock-out/ manual. Technical Specification 3.6.5.3 requires five FRVS recirculation and two ventilation units to be operable while performing core alteration Initial fuel load activities had begun April 15, 1986, and had been in progress for approximately twenty four hours. The inspector questioned the control room staff regarding the status of FRVS. The control room staff informed the inspector that the equipment had been placed in lock-out/ manual during secondary containment painting activities to prevent system auto starts and protect the charcoal beds from paint fumes. The inspector pointed out that by disabling the FRVS auto start capability the system is effectively made inoperable, and that core alterations were progress-ing in violation of Technical Specification 3.6.5.3. After discussion with Operations Department management the FRVS units were returned to

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auto, and an ENS call was placed reporting the inciden In discus-

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sion with the Operations Department Manager the inspector expressed three concerns:

1) The secondary containment painting effort will be a continuing effor If FRVS charcoal damage is a significant concern some alternate means of protection must be devised.

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2) A non-conservative Technical Specification interpretation had

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been made at the operating shif t level. This interpretation had apparently not been relayed through department managamen ) In conversation with the operating shift it was not apparent that the comprehensive meaning of system operability had been understood. The shift did not appear to be conscious of the fact that for a system to be operable all auto start functions must also be operabl In response to the inspector's concerns the applicant halted painting activities in secondary containment. Operations Department management stated that conservative interpretation of the Technical Specifica-tions, as well as the implications of system operability, would be stressed with shift personnel. The inspector informed the licensee that the failure to maintain FRVS operability during core alterations constitutes a violation of Technical Specifications (86-20-01).

However, since only non-irradiated fuel was moved, no significant

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radiological effects would have resulted if a fuel handling accident had occurred at this time. Therefore, the important aspect of this violation is the apparent lack of awareness on the operating shift's part for equipment operability requirement Fuel loading activities continued until April 20 when a faulty refueling bridge power supply and data link cable failed causing the bridge to become inoperabl An unusual event was declared at 11:27 p.m. on April 21, 1986 when the "A" Control Room Integrated Display System (CRIDS) experienced

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a failure. The failure was caused by an I&C technician running a diagnostic test on the NSSS computer which has inputs to CRIDS and

. apparently overloaded CRIDS. CRIDS did not auto swap to the "B" computer because the auto fail function was defeated to perform software modifications to the "B" system. The "A" CRIDS was imme-diately reloaded and the unusual event was terminated at 11:27 At 5:27 a.m. on April 22, an unusual event was again declared due to failure of the "A" CRIDS. The cause of this failure is unknown and the licensee is investigating. CRIDS is not required by Technical Specifications and it is not a safety related system but because it supplies control room indication the emergency classification guide requires declaration of an unusual event when CRIDS is los The "B" CRIDS system was started and the unusual event terminated at 9:25 , _ _ _ _ _ _ _ . _ . __ - . _ _ _ _ , _

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Repairs to the refueling bridge were completed and fuel loading recommenced on April 22. During a review of the April 22 and April 23 control room log (OP-DL.ZZ-026) the inspector noticed that the steps used to satisfy the daily surveillance requirements for the Standby Liquid Control (SBLC) system had been marked as N/A (not applicable). Technical Specification 3.1.5 requires the SBLC system to be operable in Mode 5 when any control rod is withdrawn. Sur-veillance requirement 4.1.5 states in part that the SBLC system shall be demonstrated operable at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that the available volume of sodium pentaborate solution is within the limits of Figure 3.1.5-1 of T During fuel load activities control rods had been withdrawn on both April 22 and April 23 without documenting the verification of SBLC system operability. The inspector brought this situation to the attention of Operations Department personnel and verified that the SBLC system had been properly determined to be operable on April 2 Normal operations department routine verifies SBLC system operability on the 11:00 p.m. to 7:00 a.m. shift daily. On April 22 and April 23, fuel load activities were not in progress during this shift so the operators marked the system operability verification steps N/ However, later during each day fuel loading recommenced and the operators failed to verify SBLC system operability prior to with-drawing a control rod as require The inspector informed the licensee that the failure to perform the surveillance test as required by Technical Specification 4.1.5 was a violation. (86-20-02)

On April 23, 1986, at 10:35 p.m., the licensee declared a third unusual event due to a loss of the CRIDS which was caused by a failure of the CRIOS printers. The system was returned to service and the unusual event terminated at 10:40 Fuel load was completed on April 27, and preparations were commenced for reactor vessel assembly and the operational hydrostatic test. At 12:55 a.m. on April 26, a spurious low reactor water lev 21 signal caused an inadvertent ESF system actuation. The "A" Diesel Generator started, NSSSS and PCIS isolation signals were received, ECCS pumps received start signals, and the alternate rod insertion (ARI) system was activated. All rods were inserted at the time of the ARI initiation and no water was injected from the ECCS due to the system tagout statu . Preoperational Test Results Review The inspector reviewed test results during this inspection to verify that adequate testing had been conducted to satisfy regulatory guidance, licensee commitments and FSAR requirements, and to verify that uniform criteria were being applied for evaluation of completed test results in order to assure technical and administrative adequac _ _ _

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For the following tests the inspector verified the licensee's evaluation of test results by review of test changes, test exceptions, test defi-ciencies, acceptance criteria, performance verification, recording conduct of test, QC inspection records, restoration of system to normal after test, independent verification of critical steps or parameters, identifi-cation of personnel conducting and evaluating test data, and verification that the test results have been approved:

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PTP-GR-1, Reactor Building Ventilation System The procedure appeared to be technically adequate and had received the required level of revie PTP-KJ-2, Emergency Diesel Generator (EDG) - B Preop Test Review The inspector reviewed the preop test results package for EDG-B. The package included 84 multiple item Test Exceptions (TEs) and 22 Change Notices (CNs) making approximately 100 individual changes to the Preoper-ational Test Procedure (PTP). Fifty-four (54) Startup Deviation Reports (SDRs) were generated to track the TEs. Of that number, 14 remain open at the time of NRC revie The inspector's review identified a number of minor concerns that were subsequently resolved during discussions with a test engineer and the QA staff. Although the package was difficult to review, because of the large number of TEs,- CNs and SDRs, no concerns remained at the end of the revie No violations were identifie . Surveillance Testing During this inspection period the inspector placed special emphasis on the licensee's surveillance test program. This in-depth and extensive review was performed to provide an additional degree of assurance that Mode 5 required systems would be operable prior to fuel load activities. The inspector performed detailed technical procedure reviews, reviewed in-progress surveillance testing as well as completed surveillance package The inspector verified that the surveillances were performed in accordance with licensee approved procedures and NRC regulations. The inspector also verified that the instruments used were within calibration tolerances and that qualified technicians performed the surveillance The following surveillances were reviewed, with portions witnessed by the inspector:

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IC-CC.SE-006(Q), Channel Calibration Nuclear Instrumentation System -

Channel Intermediate Range Monitor CSI-K601B

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OP-ST.KJ-007(Q), EDG CE400 Integrated EDG Test - 18 Month

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OP-ST.GK-002(Q), CREF ISLN/ Actuation Functional - 18 Month

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IC-TR.SM-009(Q), NSSS System A, Logic C B21-N681C (B21-N684C) Main Steam Line Isolation Reactor Vessel Low Level

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IC-FT.SP-035(Q), Functional Test Process Radiation Monitoring -

Non Divisional Monitor RY-4861 Liquid Radwaste Discharge Monitor

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OP-ST.KJ-002(Q), "B" EDG Operability Test

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IC-FT.BF-002(Q), C11-N601B Control Rod Drive Hydraulic - Division 2,

Channel 81/B, Scram

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IC-CC.FB-002(Q), C11-N601B Channel Calibration CRD Hydraulic, Division 2, Channel 81/B, Scram

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IC-SC.BF-002(Q), C11-N0128 CR0 Hydraulic, Division 2, Channel B1/B, Scram -

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IC-CC.BE-001(Q), E21-N654A, Division 1, System A, Core Spray Discharge Line Flow

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IC-CC.SE-011(Q), Nuclear Instrumentation System, Channel G-IRM

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OP-ST.BJ-001(Q), HPCI System Piping and Flow Path Verification -

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OP-ST.BJ-002(Q), HPCI System Functional Test - 18 Month

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OP-IS.BJ-001(Q), HPCI Main and Booster Pump Inservice Test

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OP-ST.SN-001(Q), ADS Manual Operability Test - 18 Month

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OP-ST.SN-002(Q), ADS Manual Initiation Functional Test - 18 Months

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OP-DL.ZZ-026(Q), Surveillance Log

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IC-FT.BB-007(Q), B21-N691D Reactor Vessel Level Trip 1 Functional Test

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IC-SC.BB-007(Q), 821-N091D Reactor Vessel Level Transmitter

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IC-CC.BB-007(Q), 821-N091A Reactor Vessel Level Transmitter Sensor Calibration

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IC-FT.ZZ-003(Q), Logic System Functional Test: Emergency Core Cooling System

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IC-GP.ZZ-035(Q), General Procedure: Channel Setpoint Evaluation .

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MD-ST.KL-001(Q), Containment Instrument Gas System Preventive Maintenance

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FC-FT.AB-033(Q), ABXISH 450E Acoustic Monitor With the exceptions noted below the inspector found the quality of pro-cedures and test conduct to be adequat The inspector reviewed several Operations Department functional tests for the High Pressure Coolant Injection System (HPCI). Procedure OP-ST.BJ-001, HPCI System Piping and Flow Path Verification - Monthly, implements the requirements of Technical Specification surveillance 4.5.1.a. This specification states that at least once per 31 days the system shall be vented at the high point vents to verify that the system discharge piping is filled with wate The HPCI discharge at Hope Creek is split; with a portion of system flow routed to feedwater and a portion to core spra '

Procedure OP-ST.BJ-001 specifies venting only on that piping leg pene-trating feedwater and does not address the leg penetrating core spra This method does not satisfy the Technical Specification requirement in that the HPCI discharge piping to core spray is not verified fille The inspector reviewed procedure OP-SO.BJ-001, High Pressure Coolant Injection System Operations to determine if adequate system fill / vent procedures for placing the system in service had been developed. The inspector noted that while steps for start of the HPCI Jockey Pump had been included, venting of the system high points had not. The combination of apparent inadequate fill / vent procedures and the lack of surveillance for the HPCI injection to core spray creates a potential for a water hammer even Although the subject procedures have not been implemented, and HPCI is not required to be operable in modes 4 and 5, the inspector expressed concern that these oversights, gone undetected, could have caused system degradatio During review of procedure OP-ST.BJ-001, HPCI System Functional Test -

18 Months, the inspector noted that air operated valve F025 was verified closed prior to system manual initiation, and verified closed after initiation. This valve receives an automatic closure signal on system initiatio In order to verify the operability of this auto closure function the initial position of the valve must be open. The inspector will review the licensee actions in response to the above two concern (86-20-03)

The inspector performed a detailed review of a sample of Automatic Depressurization System (ADS) surveillance procedures. The review was conducted to verify that ADS Technical Specification surveillance requirements had been adequately addressed and to ensure that procedures were technically and administrative 1y correct. During review of several Operations Department tests the inspector noted that while the test is conducted by operations personnel, I&C personnel are directed to simulate various signals / conditions. In such cases the procedure outlines the required signal / condition but does not specify steps to achieve nor restore from the situatio Procedure OP-ST.SN-002, ADS Logic Functional

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Test - 18 Months,_ step 5.1.10, states that I&C personnel will simulate RHR pump B discharge pressure at E11-N0558. It does not prescribe the method used to simulate this pressure and does not include steps for restoration and independent verification. The inspector discussed with the licensee ;

the need for more detailed guidance. This matter will be reviewed and discussed further with the licensee. (86-20-04)

Technical Specification surveillance table 4.3.3.1-1, Function 4.h, requires functional testing of the ADS manual inhibit switch every 18 month The licensee's Technical Specification / Surveillance Procedure System Cross Reference Matrix Report lists procedures IC-CC.SN-003 and IC-CC.SN-004 as implementing this test requirement. The inspector reviewed these procedures and determined that the ADS manual inhibit switch had not been addresse During review of the series of tests written to satisfy the requirement for 18 month ADS Logic System Functional Testing, the inspector noted that at no point were the ADS solenoids energized. Each solenoid receives input from two trip channels. Energization of both trip channels is needed to actually energize the solenoid. Testing reviewed by the inspector adequately tested each independent trip channel; but did not trip both channels simultaneously to actuate the solenoi Logic System Functional Testing, as defined by Technical Specifications, includes testing through and including the actuated device. The inspector pointed out that this testing, and the testing of the ADS manual inhibit switch, are Technical Specification requirements. The inspector will evaluate the licensee's response. (86-20-05)

Associated with each ADS logic train is a logic power monitor circui ADS is an " energize to trip" system. Loss of power to the actuation circuitry will disable the ADS function. In order to preclude an unde-tected loss of ADS logic power, and thereby loss of the ADS function, a logic power monitor circuit initiates main control room alarms if power is lost. The inspector noted that this circuit had not been included for test in any I&C or operations procedures. Discussions with the licensee indicate that there are approximately fourteen similar monitoring circuits on several logic systems. The licensee committed to develop a single operations procedure to periodically tests these circuits. This procedure will be approved prior to full power license issuance. (86-20-06)

No violations were identifie . Engineered Safety Feature (ESF) System Walkdown The inspectors verified the operability of the 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. The control room HVAC . .. _ _

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(including control room emergency filtration), Standby Liquid Control (SBLC), and Fire Protection systems were inspecte During the walkdown of the SBLC system the inspector observed that the pin used to secure the valve packing hold down bolt on storage tank drain valve-(1BHV-061) was missing and was replaced by an unauthorized bolt. The inspector informed the system engineer of this condition and work order 86-04-28-061 was written. The inspector reviewed this work order and has no further questions concerning this valve. It appeared to the inspector that the bolts used to secure the storage tank heater (10E277) were non quality although the SBLC is a safety related quality system. The inspec-tor's review of this system will continue into the next report perio No violations were identifie . Licensee Event Report Followup The inspector reviewed the following LER to determine whether report-ability requirements were satisfied, proper immediate corrective action was taken, and corrective actions to prevent recurrence _ appeared adequat LER 86-001 Damage of '!D" Diesel Generator On February 15, 1986, the "0" Diesel Generator was damaged when the brush rigging cover plate impacted the generator windings and rotor. The brush rigging cover had apparently been left inside the generator after the completion of maintenance activities which were being conducted under the architect / engineer's ( A/E) program. The generator was shipped off site for repairs and was returned to the site during the week of March 24, 198 A diesel engine inspection indicated that damage was limited to the generator. Although changes were implemented to the A/E work control program, as of March 17, 1986, all work is being conducted under Public Service work control procedures. As part of the no'rmal inspection program the reassembly and testing of the "D" Diesel Generator was inspected and the inspector has no further question No violations were identifie . NRC Commissioner's Visit Commissioner Bernthal met with licensee management and toured the Hope Creek facility on April 18, 1986. The tour incitided the control room, turbine deck, refueling floor, Drywell and the PSE&G Nuclear Training Cente . Exit Interview The inspectors met with applicant and contractor personnel periodically and at the end of the inspection report to summarize the scope and findings of the inspection activities. Written material was not provided to the applicant during the exit.

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l During the course of this inspection, the licensee was provided a written listing of NRC.open items from previously issued inspection reports. All of the information provided in the open item list was obtained from publicly available issued inspection reports and was provided to the licensee in order to more effectively address outstanding NRC concern Based on Region I review and discussions with the licensee, it was deter-mined that this report does not 'contain information subject to 10 CFR 2 restriction