IR 05000354/1986015

Insp Rept 50-354/86-15 on 860210-0316.No Violation Noted. Major Areas Inspected:Followup on Outstanding Insp Items, Plant Tour,Preoperational Phase Activities & Insp Program Status

ML20202G377
Person / Time
Site:	Hope Creek
Issue date:	03/26/1986
From:	Strosnider J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20202G354	List: IR 05000354/1986015 ML20202G351
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.C.1, TASK-1.C.6, TASK-2.B.2, TASK-2.D.3, TASK-2.E.4.2, TASK-2.F.1, TASK-2.K.1, TASK-3.D.3.3, TASK-TM 50-354-86-15, IEB-73-06, IEB-73-6, IEB-80-14, IEB-80-17, IEB-83-05, IEB-83-5, IEC-77-12, NUDOCS 8604140336
Download: ML20202G377 (21)

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U. S. NUCLEAR REGULATORY COMMISSION

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REGION I

Report N /86-15 L

Docket 50-354 License CPPR-120 Licensee: Public Service Electric and Gas Company Facility: Hope Creek Generating Station Conducted: February 10 --March 16, 1986

' Inspectors: R. W. Borchardt, Senior Resident Inspector D. K. Allsopp, Resident Inspector J. J. Lyash, Reactor Engineer E. L. Conner, Project Engineer F. J. Crescenzo, Reactor Engineer R. L. Nimitz, Radiation Specialist S. K. Chaudhary, Lead Reactor Inspector M. T. Miller, Radiation Specialist .K tangRadiationSpecialist Approved: U [fsdg 6 6 J Strosnt'deF, Chief, Projects Section IB Date Inspection mary:

Inspection on February 10, 1986 - March 16, 1986 (Inspection Report Number 50-354/86-15)

Areas Inspected: Routine onsite resident inspection of the following areas: followup on outstanding inspection items, plant tour, preoperational phase activities, construction deficiency reports, training program, inspection program status, safety evaluation report followup and management meetings. This inspection involved 501 hours0.0058 days <br />0.139 hours <br />8.28373e-4 weeks <br />1.906305e-4 months <br /> by the inspector Results: Significant progress has been made toward resolving the backlog of NRC open inspection items required to be closed prior to fuel loa Although a significant amount of work remains on open items,  !

preoperational tests (including test exceptions) and surveillance testing, we do not see any obstacles to licensing once this work is satisfactorily complete ,

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I Details 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 activitie . Followup on Outstanding Inspection Items 2.1 Violations (Closed) Violation (85-61-01); Performance of Preoperational Testing (PTP). During performance of a retest on PTP-AB-1, Main Steam System, quality control mandatory witness . points were deleted without proper documentation and approval. The inspector notified the test engineer of this problem and immediate corrective actions were take The inspector reviewed the applicant's response to this violation dated March 11, 1986 including the long term corrective actions take In addition to specific training of all test engineers, the following corrective actions were initiated due to this and other similar violations identified in Inspection Report 86-10.

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Deletion of the on-the-spot procedure change program l

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Use of mandatory procedure review checklists for the Public Service Startup Group (PSSUG)

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PSSUG review of PTP comments after Preoperational Review

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Committee (PORC) comments This item is close (Closed) Violation (85-61-02); Performance of QC Inspection. This violation identified three instances where final QC inspections had accepted equipment that did not meet design requirements. The applicant responded to this violation in a letter to the NRC dated March 11, 1986. The applicant corrected the specific discrepancies

and performed a limited sampling program to determine that these QC

! errors were not indicative of a more widespread problem. The inspector will continue to pay particular attention to QC inspection l activities during future inspections. This item is closed.

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l l (Closed) Unresolved Item (85-19-01); Primary Containment Operability 1 Issues. This item was updated in Inspection Reports 85-42 and 85-45.

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Issues addressed by this item are as follows:

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1) Existence of non-Q instruments and instrument piping on systems identified as part of the extended primary containment boundar ) Deletion of the third feedwater, RCIC, HPCI and RWCU isolation valves for containment penetrations 2A and 2B from the FSAR/ Technical Specification ) Extension and deletion of containment isolation valve stroke times. Of specific interest is the deletion of the HPCI and RCIC steam supply line isolation valve stroke time requirement The applicant's analysis of extended primary containment integrity identified twelve non-Q instrument / instrument piping runs which required upgrade. The inspector reviewed Design Change Request (DCR)

4-BCM-86-069 and Design Change Package (DCP) 7155 which detail the planned upgrade rework of these instruments. The inspector also reviewed Nuclear Department Work Order 86-02-28-082-3 which has been initiated and approved to implement the DCR and DCP referenced abov This work order specifies that work completion is required prior to fuel loa Subsequent to discussions between the inspector and the applicant, a change to the Technical Specifications and FSAR was processed; adding the five primary containment penetration 2A and 28 isolation valves to the primary containment isolation valve list. While reviewing the additions to the Technical Specifications the inspector noted that incorrect leak rate testing requirements had been specified for these isolation valves. Discussions with the applicant indicated that the discrepancies had been identified and that a change would be processed to correct the requirement Information supporting the extension and deletion of containment isolation valve stroke times has been submitted by the applicant for NRR review. The inspector will review and discuss the deletion of time requirements for the HPCI/RCIC steam supply isolation valves with appropriate NRR personnel during the next inspection period (86-15-01).

(Closed) Unresolved Item (85-24-08); Ruskin Fire Damper Closure Under Air Flow Conditions. Inspection Report 85-24 documents PSE&G's commitment to respond to the subject concern and implement hardware modifications and/or administrative actions, as necessary. This information was provided by letter dated December 3, 1985 and approved in SER Supplement No. Most of the Ruskin fire dampers were found acceptable as is or as modified based on the information provided. However, to compensate for four unqualified dampers located in the floor slab at elevation 178 feet of the Emergency Diesel Generator building, the applicant needed to protect the related HVAC ductwork with a fire rated materia l l

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The inspector physically observed the four sections of ductwork that have been coated with 10 layers of fire rated material to provide 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> protection. .This effectively removes the ductwork from its original fire area and places it in the fire area being supplied or exhausted. This work is completed, this item is close (Closed) Unresolved Item (85-24-09); Fire Proofing of Structural Steel. In Inspection Report 85-24, PSE&G committed to provide a list of areas where structural steel will not be fire proofed and justification for omitting this fire proofin This information was provided by letters dated August 27 and December 16, 1985 and found acceptable in SER Supplement No. The inspector reviewed DCP No. 4-ECM-86-036 package and physically observed a number of the coated structural steel members. This work has been completed. This issue is resolve .3 Inspector Follow Items-(Closed) Inspector Follow Item (85-44-08); Installation of various liquid sampling stations. The inspector toured the facility with a chemistry department representative and verified that the radwaste, reactor, and condensate system sampling stations were completely '

installed and operable. The inspector also verified that the preoperational test procedures listed below have been field completed, PORC approved and all open test exceptions are assigned a startup deviation repor RC-1 Auxiliary Building Sampling RC-2 Turbine Building Sampling RC-3 Reactor Building Sampling RC-4 Post Accident Sampling The inspector has no further question (0 pen) Inspector Follow Item (85-45-01); Surveillance and Calibration of ECCS Room Cooler The inspector toured the facility and examined temperature switch covers installed by the applicant. The inspector reviewed procedure OP-FT.ZZ-001(Q), Emergency Area Cooling System (EACS) Room Coolers Functional Test. The inspector noted that procedure OP-FT.ZZ-001(Q) did not include steps to functionally test the station service water ventilation temperature switches. The station service water system is an engineered safety system and proper functioning of the ventilation system is required to ensure its operabilit c.

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> u The applicant informed the inspector that calibration of the room cooler temperature and flow switches would be scheduled via the inspection order system and performed using procedure

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MD-DC.ZZ-001(Q), General Instrumee.t Calibration. The inspector questioned the applicant as to whether this rt.ethod of scheduling and calibration

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provided adequate assurance that calibration frequency would nol. be extended or calibration missed, and that shift personnel would be aware of out of calibration instrument The inspector-discussed the above concerns with the applicant and will review the applicant's resolution in a future inspectio (Closed) Follow Item (85-59-0!); Leboratory QA Program. The inspector reviewed procedure CH-AP'.'lZ-017(Q), Chemistry Quality Control Program. This procedure as issued included radiochemistry as well as instrument quality control programs. This procedure also includes anion exchange resin analysis progra (Closed) Follow Item (85-59-02); Reaision of Atomic Absorption and j Ion Chromatography procedures. The inspector reviewed procedure CH-CA.ZZ-039(Q), Anions and Cations b/ Ion Chromotography, and procedure CH-CA.ZZ-006(Z), Metals by Graphite Furnace Atomic Absorption. The inspector noted that these procedures were revised to include field validation comment l (Closed) Inspector Follow Item (85-64-01); Control Room Instrument Label. The inspector verified that control room instrument SS-PR-4960A3 on panel 10C650E now reads "Drywell Presture". This item is close (Closed) Inspector Follow Item (85-64-02); Reactor?8uilding to Torus Vacuum Relief Valve. The inspector verified that Reactor Building -

Torus Vacuum Relief Valve HV-5029 label is now properly engrave This item is close (Closed) Inspector Follow Item (85-64-03); Reactor Building Ventilation System (RBVS) Operatio The inspector reviewed the applicable sections of Technical Specifications and found that the l licensee must verify at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the RBV.S exhaust flow rate is greater than the supply flow rate and the reactor

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building is at a negative pressure during refuel operations and in {.

Modes 1, 2 and 3. There are no specified requirements that a minimum number of fans be in service during refueling operations. This item is close '

(Closed) Inspet' tor Follow Item (85-64-06); Incorrect Value of Diesel Generator Continuous Rating in Technical Specifications. The inspector reviewed the " Final Draf t" version of the proposed i Technical Specifications and verified that the proper diesel generator continuous rating value (4430 KW) is used. The inspector, has no further questions and this item is close .

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(Closed) Follow Item (85-52-37); Licensee to Update Final Safety Analysis Report (FSAR) to include correct isolation term for the reactor building isolation dampers. The licensee updated the FSAR (Amendment 12) to include the correct isolation terms (Change Notice 1064).

(0 pen) Follow Item (85-52-18); Licensee to establish and implement a program for selection and use of portable airborne radioactivity sampling equipment. The licensee has established and implemented a program for selection and use of portable airborne radioactivity sampling equipmen The following matters have yet to be addressed:

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selection of proper sampling equipment based on job duration

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guidance for use of labeling air samples

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operability checks for labeling air samples

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administrative control for use of equipment in the field to ensure use of the equipment consistent with the configuration of the instrument during calibratio These items will be reviewed in a future inspectio (0 pen) Follow Item (85-52-12); Licensee to establish and imp!ement procedures for determination of neutron dose. The licenses has not yet established and implemented procedures for determination of neutron dose. These should be established and implemented priur to

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initial criticalit l (Closed) Follow Item (85-52-13); Licensee to correct technical deficiencies ir, procedures for determination of dose to the ski The licensee corrected the identified technical deficiencies and I approved the procedure (Closed) Follow Item (85-52-40); Licensee to correct deficiencies in the Radiological Occurrence Report (ROR) Program procedures. The licensee corrected the deficiencies in the ROR procedures and approved the (Closed) Follow Item (86-05-01); Train and qualify chemistry technicians to operate the Condensate Demineralizer and Reactor Water Cleanup systems. The inspector reviewed system training records and discussed the training program for chemistry technicians. The inspector determined that an adequate number of shift technicians were qualified on the Reactor Water Cleanup system and three individuals were qualified to operate the condensate demineralizer.

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The licensee stated that additional staff would be qualified on the condensate demineralizer system by March 21, 1986, dependent on system testin (Closed) Follow Item (86-05-03); Determine resin capacity analysis frequency. The inspector reviewed procedures CH-SO.AK-001(C),

" Operation of Condensate Demineralizer System," Revision 10 and CH-AP.ZZ-017(Q), " Chemistry Quality Control Program," Revision 1, which described system operation, verification that resin had been tested prior to use, and frequency for testing resin after batch change out and total hours of operation of the condensate demineralizer. The licensee stated that the resin capacity for both l the condensate demineralizer and reactor water cleanup system were

! verifie Decision to discard resin from the RWCU system will rely on conductivity and differential pressure criteria rather than total hours of RWCU operatio (0 pen) Follow Item (86-07-02); Licensee to correct technical deficiencies in the airborne radioactivity sample analysis progra The licensee has not yet corrected the technical deficiencies in his program. The deficiencies should be corrected to provide for proper analysis of airborne radioactivity samples at initial criticalit (See report 86-07)

(0 pen) Follow Item (86-07-12); Licensee to evaluate the adequacy of his dosimeter for quantifying personnel exposure to beta radiatio The licensee has not yet evaluated the capabilities of his dosimeter to properly measure beta raatatio This matter should be resolved prior to initial criticalit (0 pen) Follow Item (86-07-01); Licensee to train and qualify a sufficient number of radiological controls technicians to support fuel load, initial criticality, and low power operation. Inspector discussions with personnel, review of documentation and review of training records indicated a sufficient number of radiological control technicians had been trained and qualified to support fuel load, initial criticality and low power operation. The licensee is currently training and qualifying additional staff to support full power operation. The licensee's efforts in the area will be reviewed during a subsequent inspectio (Closed) Follow Item (86-07-07); Licensee to ensure calibration and use of portable radiation survey instrumentation minimizes geometry and angular dependence effects. The li:ensee has reviewed this program with respect to these matters. Precautionary notes were placed in procedures were appropriat (0 pen) Follow Item (86-07-10); Licensee to address following:

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the wire screens holding in place the combination roughing / moisture removal elements in the recirculation filter system not secur l

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filter mounting frames for one of the recirculation filter '

system were ben completion of safety related ventilation system testin The licensee secured the wire screens and cold worked the filter frames. Subsequent filter bypass leakage test, performed on the systems were within Technical Specification, bypass leakage limit The licensee has not completed the review of all test results. See )

section 4 of this repor .4 IE Bulletins, Circulars and Information Notices (Closed) IE Bulletin (73-06); Inadvertent Criticality in a BWR. This bulletin provided information on an inadvertent criticality event at Vermont Yankee and was supplied to the applicant for-information only. The inspector reviewed the applicant's internal response and determined the bulletin's concerns were adequately addressed. The following procedures address the bulletin concerns:

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RE-ST.8F-00(Q) CRD Scram Time Determination

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SA-AP.ZZ-013(Q) Control of Temporary Modifications

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SA-AP.ZZ-046 Radiological Access Control

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OP-AP.ZZ-107 Shift Relief and Turnover This bulletin is close '

(Closed) IE Bulletin (80-14); This bulletin addresses degradation of BWR scram discharge volume (SDV) capability. The SDV modification at Hope Creek incorporates ATWS-3A modifications to improve the dependability of the system and satisfy the bulletin concerns. The SDV vent and drain valve operability and surveillance requirements have been incorporated in the " Draft" Technical Specifications, section 3/4.1.3, and station operating procedure OP-OL.ZZ-004. The station inspection program requires that SDV vent and drain valves be inspected at least once every twenty four hours, and cycled through at least one complete cycle once every thirty one days. The " Draft" Technical Specification, section 4.1.2.1.4, also requires verification of proper float switch actuation following each l " SCRAM' from a pressurized condition greater than 900 PSIG to detect crushed or damaged float level switches. This item is closed.

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(Closed) IE Bulletin (80-17); Failure of 76 of 185 Control Rods to Fully Insert During a Scram. This bulletin was provided to the )

applicant for information only and did not require a response. In j Inspection Report 85-5d the hardware aspects of this bulletin were '

inspected by a region based inspector. No deficiencies were identified. The inspector reviewed the applicant's internal response, SER section 4.6 and various startup test program procedures and is satisfied that the applicant has adequately addressed this bulletin. This item is close (Closed) IE Bulletin (83-05); This bulletin refers to the NRC concern regarding ASME nuclear code pumps and spare parts manufactured by the Hyward Tyler Pump Company (HTPC). The licensee has compiled a list of pumps supplied by HTPC and their service applications. The inservice testing of these pumps will be covered in the station inservice plan developed to satisfy ASME,Section XI, and the licensee has committed to performance tests in the FSAR for the station. General requirements and procedures for Hydrostatic testing are contained in specification 10855-P-590(q). The enhanced commissioning and endurance tests recommended by the manufacturer were incorporated in the preoperational test procedure and have been verified by the inspector (pre-op test #DTP-BJ-0001, Revision 1, paragraph 6.2.1.1.3). This item is close (Closed) IE Circular (77-12); This circular refers to the problem of dropped fuel assemblies at BWR facilities. To prevent such incidences, GE has made design improvement in the refueling platfor The station fuel grapple has also been redesigned such as to prevent opening under load, and the fuel grapple open/ closed switch has been provided with a cover to prevent accidental actuation. The licensee has incorporated controls in station maintenance and operating procedures to assure safe operation of this system. The inspection requirements have been incorporated in procedures: RE-FR.ZZ-001(Q);

OP-FT.KE-001(Q); MD-PM.KE-00Z(Q); MD-PM.KE-003(Q); and

, OP-ST.KE-001(Q). A requirement for daily visual inspection has been included in procedure OP-IO.ZZ-009. This item is close .5 TMI Action Plan Items (Closed) TAP Item I.C.1, Short-Term Accident and Procedures Revie The applicant's Emergency Operating Procedures (EOP) are based on the BWR Owners Group Emergency Procedure Guidelines (EPG), which are contained in General Electrical Topical Report NEDO-24934, Revision In SER draft supplement 5, section 13.5.2.3.1, the BWR owners group EPG was reviewed and accepted by NRR. A plant-specific system function and task analysis described in Detailed Control Room Design Review Summary Report and Supplementary Report have been reviewed and accepted by NRR in SER Draft supplement 5, section 13.5.2.3.1. The

, applicant's E0Ps were developed utilizing a Procedures Generation

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Package (PGP) which includes a plant specific technical guideline,

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writer's guide, E0P validation program, and a training implementation program. In SER draft supplement 5, section 13.5.2.4, the NRC staff con-cluded the PGP for Hope Creek meets the requirements of supplement 1 to

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NUREG-0737 and provided acceptable methods for accomplishing the objective of NUREG-0899. A Special As-Built Team Inspection conducted December 2-13, 1985, concluded that systems inspected were constructed in confor-mance with the FSAR description. The training program consists of a combination of classroom instruction, simulator exercises, and training on revisions to the E0Ps. The training program effectiveness has been reflected in the emergency preparedness appraisal during a full scale exercise and throughout operator licensing exams. This item is close (Closed) TAP Item I.C.6, Verifying Correct Performance of Operating Activities; II.K.1.5 and II.K.1.10, Proper Safety Features Functioning. The applicant's approach to resolution of these issues

was reviewed and accepted by NRC/NRR in the SER. The inspector reviewed implementation of the program in Inspection Report 85-5 The single issue remaining open concerned the applicant's independent verification program. Station Administrative Procedure SA-AP-ZZ-002(Q), Station Organization and Operating Practices, has been revised to clarify the meaning of independent verification and to provide station level guidance on those systems requiring independent verification. 0ther station and department procedures are being revised to ensure consistency with the governing procedure, SA-AP-ZZ-002(Q). The inspector reviewed the method prescribed for

conducting independent verification, and the list of systems for

- which it is required. The applicant's program appears to be conservatively structured and consistently applied to those systems affecting safet Based on the above these items are close (Closed) II.B.2, Plant Shielding. The concern of this requirement is to ensure that personnel can perform necessary post-accident operations in plant vital areas and not be overexposed. This issue was evaluated by NRR as reported in the Hope Creek SER, Section 12.3.2, which concluded that Hope Creek meets the criteria of Item II.B.2 in NUREG-0737. The NRR review was based on information provided in the FSAR. At the time of review, the final location of the Operations Support Center (OSC), the Filtration, Recirculation and Ventilation System (FRVSV) RMS Post-Accident Skid, and the Emergency Assembly Points had not been determined. In addition, the FSAR analysis assumed leakage rates of 0.5 percent from the containment and 10 percent from the reactor building. The licensee, through Bechtel Construction, Inc. calculated new post-accident exposures and dose rates for all vital areas using leakage rates of 1.0 percent and 100 percent from the containment and reactor building, respectively. This new data

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was supplied to the NRC in Amendment 14 to the FSAR. The recently completed ILRT showed a containment leak rate of 0.193 volume percent per day.

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l The inspector compared the original and revised post-accident exposures and dose rates and found whole body gamma and skin beta dose rates and total dose rates, as revised, higher but within the acceptance criteria. The total doses for the guard house, HP access control areas and the remote shutdown panel are lower in the revised analysi The licensee explained this was due to revised X/Qs for these areas. The inspector suggested this change should have been identified in FSAR Amendment 14. In addition, a couple of the thyroid dose rates and total doses were calculated to be slightly lower in the revised than the original. The licensee explained that this was due to minor corrections and assured the inspector that the Amendment 14 values are correct. The inspector had no further

! questions. Therefore, NUREG-0737, Item II.B.2 is close (Closed) TAP Item II.D.3, Direct Indication of Safety Relief Valve In the Hope Creek Safety Evaluation Report the NRC/NRR staff reviewed and accepted the applicant's overall relief and safety valve position indication system design. By memorandum from G. C. Lainas to Wagner dated December 31, 1985, the results'of the Instrumentation and Control Systems Branch (ICSB) site visit were documented. During this ICSB inspection the staff reviewed the installation of the system and the calibration procedures available. Based on the visit the staff found that the installation meets the requirements of NUREG-073 The inspector reviewed the approved acoustic monitor functional test and channel calibration procedures to ensure that proper vendor recommendations had been included. The inspector also reviewed power I

ascension tests TE-SU.BB-262(W) and TE-SU.BB-261(Q) and verified that indication response testing during power operation is planne Completed Nuclear Department Work Order (WO) 86-02-14-061-4 tracked installation of the acoustic monitor accelerometers. This WO required QC to witness installation and initial system response tests. The WO indicates that all accelerometers, with the exception of IAB-XE-4507J, were installed. On March 10, the inspector examined system components in the field and identified that four accelerometers, 1AB-XE-4507J, G, D, and E, were not installed. When questioned, the applicant could not produce documents authorizing the removal of the additional accelerometers. The inspector will review the applicant's actions in response to this concern during a future inspection (86-15-02).

Based on the review documented in the SER, ICSB site inspection report and the inspector's observations, this item is close (Closed) TAP Item II.E.4.2, Containment Isolation Dependabilit This item was updated in Inspection Report 86-06. The inspector identified that six inch nitrogen purge inlet valve, HV-4978, had not been qualified. The inspector further noted that " Final Draft" Hope Creek Technical Specifications did not include adequate leak rate I

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test requirements for the resilient seat butterfly valve. The applicant has initiated qualification testing of the subject valve, with final test results anticipated prior to fuel load. The applicant has committed to revise the Technical Specifications to

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include proper leak rate testing. The inspector has discussed with NRC/NRR the need for correction of this specification, and will verify institution of this change in a future inspection (86-15-03).

Based on the above this item is considered close (Closed) II.F.1, Accident Monitoring Instruments. The requirement was to have the following additional accident monitoring instruments:

1) noble gas effluent radiological monitor; 2) iodines and particulates continuous sampling; 3) containment high-range monitor; 4) containment pressure monitor; 5) containment water level monitor; and 6) containment hydrogen concentration monitor. Numbers 4, 5 and 6 above were inspected and found acceptable as documented in Inspection Report 85-56 and 86-06 (Nos. 4 and 5 in 85-56, No. 6 in 86-06). In SER Sections 11.5.1 and 12.3.4.1, NRR documented its finding that the appropriate noble gas, iodines, particulates, and high-range containment monitors had been provided and located to meet NRC requirement The inspector observed the north and south plant vent and the FRVS radiation monitoring systems. These skid mounted units, described in FSAR Sections 11.5.2, were being tested during the inspection. The inspector also confirmed that the two containment high range monitors were operational. Next, the inspector reviewed the Emergency Event Classification Guide, the Emergency Plan and its implementing procedures. The subject monitors are integrated into the Emergency Plan documents. This concluded the open issues under NUREG-0737, Item II. Therefore this item is close (Closed) III.D.3.3, Improved Inplant Iodine Instrumentation Under Accident Conditions. This requirement was for equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident. The NRR review, documented in SER Supplement No. 3, found the two types of portable instruments, their location and use acceptable. In addition, the portable instrumentation contained in the emergency supply cabinets was surveyed as documented in Inspection Report 86-0 The inspector observed the evacuated Marinelli beakers with silver zeolite cartridges for the sample medium and the ac powered and dc powered air samplers contained in the TSC and control room access emergency supply cabinets. The instruments were functional with the Marinelli beakers at about 25 inches of vacuum and the dc batteries charged. The inspector also reviewed EP IV-112, Emergency Grab Air Sampling; EP IV-113, Radioiodine and Noble Gas Activity Sample Analysis During an Emergency; and EP IV-104, Emergency Exposure

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Authorization and Emergency Task ALARA Review. These procedures, the dose evaluation-sheet and the other attachments are judged adequate to perform the needed sampling during an emergency. This NUREG-0737 item is close (Closed) 85-SC-03; TMI Action Plan Items requiring closure by fuel load. This report closes the last TMI Action Plan Items required to be addressed by the applicant prior to fuel loa .6 Construction Deficiency Reports (Closed) Construction Deficiency Report 86-00-01; Circuit Design Error on Tobar Instrument Loops. Through an administrative error this CDR was previously identified as CDR 85-00-11. The subject design error and the applicant's corrective action was reviewed, and the CDR-closed in Inspection Report 86-01. This item is close . Plant Tour The inspector periodically toured the plant and performed walk-through inspections during this period. Special emphasis was placed in the areas of the drywell, reactor building, torus /wetwell, and the diesel generator buildin These inspections were carried out to assess the level of general workmanship in the areas of piping and pipe support; effectiveness of the cleanliness and housekeeping program; and general conformance to project procedures for work in progress and completed work. The inspector also toured the control room on regular and backshifts. He interviewed operations personnel regarding testing scheduled or in progress, reviewed logs and night orders, and observed alignment and indications of systems undergoing tests. He checked on tests and operations in progress, observed equipment conditions, and interviewed personnel involved in ongoing activitie On February 15, 1986, while attempting to run the "D" diesel generator for surveillance testing, the generator experienced an electrical fault. The fault was caused by a metal cover plate which was mistakenly left inside of the generator after maintenance activities. When the generator was started the loose metal cover impacted and damaged the stator and field windings. The applicant's evaluation determined that extensive repairs were necessary which would require shipping the generator off site. The inspector periodically reviewed the work activities associated with the generator removal and the inspection of diesel generator bearing The re-assembly and post-assembly testing will be reviewed during a future inspectio No violations were identifie . -. - . - -.. .. _ _ . .__

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e 14-4. .Preoperational Phase Activities '

i 4.1- Preoperational Test Witnessing The inspector witnessed ' testing in progress on regular and backshifts and verified that: 1) testing was conducted using approved procedures . tar qualified individuals, 2) controlled, calibrated measuring and test equipment was available_for required data gathering, 3) adequate quality control coverage was provided, 4) proper coordination between test engineers and operations existed and 5) test exceptions and changes were

, documented and dispositioned properl During the report ~ period the inspector sitness=d sections of the 7,

following preoperational test:

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SB-2 Reactor Protection System Response Time Testing No violations were identifie .2 Preoperational Test Results Review

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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

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technical and administrative adequac For the following tests the inspector verified the licensee's evaluation of test results by review of test changes, test exceptions, test deficiencies, "As-Run" copy of test procedure, 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, identification of personnel conducting and evaluating test data, and verification that the test results have been approved:

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SM-1, Nuclear Steam Supply Shutoff System and Primary Containment Isolation System

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SB-1 Reactor Protection System Logic

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GU-1 Filtration Recirculation Ventilation System i

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4. Reactor Protection System Logic Testing During review of SB-1 the inspector noted that Test Exception 17 had been written to document missing wiring between trip units B21-N6808 and C11-N601 In order to facilitate test completion, a temporary modification was processed to install temporary connections between the trip units. A Startup Deviation Report was initiated and approved making these temporary connections permanent. The inspector reviewed General Electric elementary diagrams and identified that the as-designed circuit should include a diode, identified as CR28, between the two trip units. The temporary modifications and the approved, installed wiring effectively bypassed this diod Discussions with the applicant indicate that the circuit will be rewired to return it to its as-designed configuration. The inspector will continue review of this matter during the next inspection period (86-15-04).

4. Safety Related Ventilation System Testing The inspector reviewed testing of the control room emergency ventilation system; the filtration, recirculation, ventilation system, and the technical support center emergency ventilation system. The review was with respect to criteria contained in the following:

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Technical Specifications

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Applicable ANSI and ASTM standards

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. Hope Creek Final Safety Analysis Report (FSAR) Chapter 14 The evaluation of the licensee's performance in the area was based on discussions with personnel, review of test data, and visual inspection of the syste The following matters were reviewed:

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Auto isolation capabilities

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Technical Specification surveillance

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heater testing was applicable

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filter testing l

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charcoal testing

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condition of systems

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status of preoperational testing

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The evaluation of the licensee's performance was based on review of test data, discussions with personnel, review of procedures,

-and inspector walkdown of syste Findings The licensee has completed prescribed preoperational testing of the system. The following matters needing licensee's attention were identified in Inspection Report 86-07 (Inspector Follow Item 86-07-10):

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Complete applicable system Technical Specification surveillances of the systems to support fuel load. All applicable surveillances have not yet been performe Review and approve the remaining tests of the system Some tests (e.g. retest of "B" control room emergency ventilation system) have not yet been reviewed and approve Initial corrective actions for housekeeping and other concerns associated with the Technical Support Center Emergency Ventilatio: , Syste Actions associated with the above items are not complete and will be reviewed in a future inspectio . Radiation Monitoring System Testing The inspector reviewed the status and adequacy of testing of the radiation monitoring system (RMS) to support fuel load, initial criticality and low power operation. The status of the following systems were reviewed:

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effluent monitoring system

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process monitoring system

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area radiation monitoring system The evaluation of the licensee's performance in the area was based on discussions with cognizant licensee and contractor personnel, review of test results, and observation by the inspecto ,

Findings l

The licensee has completed testing of the RMS consistent with commitments to the NRC to support fuel load with the exception

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of the filtration, recirculation and ventilation system (FRVS)

exhaust. The testing of the monitoring system is about 60%

complet Review of calibration data-and discussions with personnel indicate acceptable calibrations have been performe The following matters needing licensee attention were previously identified in Inspection Report 84-44 (Inspector Follow Item-84-44-14):

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complete review of test results for those monitoring .

systems with completed tests

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complete testing of the FRVS exhaust monitors

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complete applicable Technical Specification surveillance for the monitor

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provide all primary calibration data for NRC review These matters will be reviewed during a future inspectio No violations were identifie . Construction Deficiency Reports ,

t At the end of this inspection period three construction deficiency reports (CDR) remained open. A brief description of each CDR is given belo Tobar Pressure Transmitter Term'inal~ Connections One hundred and eighteen Tobar pressure transmitters were incorrectly installed such that the transmitters were not environmentally

' qualified. The correct design called for the use of butt splices at the connection of the external cable conductors to the pigtails of the transmitters. Contrary to the design, the transmitters were

, installed utilizing the internal terminal blocks. This CDR will be l reviewed by the NRC after the applicant completes the planned j corrective actio Diesel Air Start System Colt Industries, manufacturer of the applicant's Emergency Diesel l Generators has determined that the Diesel Air Start System, which  !

,- actuates the governor rack boost is not vented from the rack boost 1 promptly after start signal is terminated. This causes the boost to fight the governor and creates a potential overspeed condition and resultant overspeed trip. An overspeed trip had been reported to

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Colt Industries by Public Service of New Hampshire, Seabrook Nuclear Plant, Unit 18. Colt Industries has recommended that all Colt-Pielstick PC-2 engines in nuclear service be modified to positively vent the air from the rack boost cylinder, and its associated shuttle valves. The applicant plans to implement this change when the parts arrive. This CDR will be reviewed by the NRC after the applicant completes the planned corrective actio Anchor Darling Testable Swing Check Valves Testable swing check valves supplied by Anchor Darling company have experienced sticking open problems. Preliminary evaluations indicate that the valves are sticking open due to the extremely tight packing at the manual operator shaft. The applicant has proposed a fix that has been approved by Anchor Darling which allows the swing check valve to open and close due to flow without causing the manual operator to change position. This CDR will be reviewed upon completion of corrective actio . Nuclear Operator Training Program An inspection of the Hope Creek licensed operator training program was conducted in conjunction with the operator licensing exams given during the week of February 24, 1986. This inspection included a review of the following:

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Individual training records to ensure documentation of compliance with the FSAR cold license training program

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Current training materials to ensure changes to the plant design and procedures have been properly incorporated

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Interviews with currently licensed operators The results of this inspection are discussed below and were provided to the applicant during the license examination exit interview on Februa ry 28, 198 The records maintained by the training staff are easily retrievable and well organized. All training required by the FSAR was documented as complete for those individuals whose records were inspecte Many of the procedures /PIDs used by the training staff for cold license training are not current. This was discussed with the training staff and it was explained that the material was

" frozen" prior to the NRC exam for question response consistency. It was further explained that significant design / procedure changes were trained on despite the " freeze" and that all changes would be trained on during requalification

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trainin It was noted that the material used for i

requalification training was current. This appears to be an acceptable approach and the inspector had no further questions f

in this are The operators who were interviewed, in general, had positive

attitudes towards the training program. All felt they had been adequately trained on plant systems and system operations. They also felt the lecture and simulator programs were excellen Although varying opinions were observed as to the technical adequacy of the written training material, it was agreed that the readability of these materials could be greatly improve It was generally felt that the training program was weighted more to successful completion of a NRC exam rather than what was actually required of a licensed operator on a daily basis. This-attitude was particularly evident with the SRO's who felt that they could have used more training on supervisory / administrative skill No violations were identifie . Inspection Program Status Preoperational Test Program Inspection completion status is approximately as follows

Area  % Inspection Complete Overall Program 80 Procedure Programs Mandatory 100 Primal 100 Test Witness Mandatory 90 Primal 100 Results Review Mandatory 46 Primal 100 Inspection status is consistent with applicant test program progres Operational readiness inspection status is approximately as follows:

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Area  % Inspection Complete OPS Staffing & Procedure 100 Tech Spec Review 100 QA 100 Maintenance 100 Fire Protection 100 Surveillance 100 Rad. Controls 85 Rad. Waste 70 Security 100 Emerg. Planning 90 8. Safety Evaluation Report Followup During the course of the preoperational inspection program (2513) and as a result of specific requests from NRC/NRR the inspector verified that the applicant has satisfactorily implemented selected commitments as found in the Safety Evaluation Report (SER) and supplements. Although this activity will continue into the startup and operational phase inspection programs the inspector considers this module closed for the 2513 progra . Management Meetings On February 27, 1986, the Systematic Assessment of Licensee Performance (SALP) management meeting was held between the NRC:

Region I and PSE&G management. This meeting was held on the Hope Creek site and was open to the public. Prior to the SALP meeting the Region I management toured the facility with the resident inspector On March 10, 1986, the NRC Region I Regional Administrator and members of his staff met with PSE&G management to discuss the readiness of Hope Creek for a low power operating license. On March 11, 1986, a similar meeting was held between NRC/NRR management and PSE&G followed by a site tour. The topics discussed during these meetings included the following:

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Construction readiness

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Organization, staffing, training

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Operational experience / personnel qualifications

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Operational readiness / procedures

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Lessons learned from other plants

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Preoperational and Startup Testing

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Technical Specification Certification Status

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Quality Assurance Program

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Third Party Audits On March 10, 1986, both the Region I and NRC headquarter's staffs witnessed a simulator exercise at the applicant's 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 their inspection activities. Written material was not provided to the applicant during the exi During the course of this inspection, the applicant 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 determined that this report does not contain information subject to 10 CFR 2 restrictions.