IR 05000440/1985078
| ML20137B666 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 01/08/1986 |
| From: | Holtzman R, Knop R, Oestmann M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20137B661 | List: |
| References | |
| 50-440-85-78, TAC-59166, NUDOCS 8601160016 | |
| Download: ML20137B666 (27) | |
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O U. S. NUCLEAR REGULATORY COMMISSION
REGION III
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Report No. 50-440/85078(DRP)
' Docket.No. 50-440 License No. CPPR-148
. Licensee:
Cleveland Electric Illuminating Company Post Office Box 5000 Cleveland, OH 44101 Facility Name:
Perry Nuclear Power Plant, Unit 1 Inspection At:
Perry Site, Perry, OH Inspection Conducted: ' November 7, 1985 through December 30, 1985 Inspectors:
J. A. Grobe K.'A. Connaughton D. E. Miller C. F. Gill J. W. McCormick-Barger P. L. Hartmann A. Morrongfello M.J.Q./)jdim u w 7 11 0'es ann R.S %
i ww R. B. Ho man J. S. Wiebe M. E. Cashatt B. Thurmond ($ C b s1g lg/sh Approved By:
R. C. Knop, Chief Reactor Projects Section 1C Date koj Ihk jo
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. Inspection Summary Inspection on November 7 through December 30, 1985 (Report No. 50-440/85078(DRP))
- Areas Inspected:
Routine, unannounced inspection by resident and region based inspectors and the technical training staff of previous inspection findings, 10 CFR Part 21 reportable deficiencies, preoperational test results, an allegation, regional requests, TMI Action Plan items, emergency instructions, maintenance instructions, surveillance test instructions, control room readiness to control operating activities, operating staff training, system operating and valve lineup instructions, fire protection and prevention program, safety committee activities, management meetings, preoperational test-program implementation, and administrative procedures for operations.
The inspection involved a total of 576 inspector-hours onsite by 12 NRC inspectors
. including 72 inspector-hours during off-shifts.
Results:
No violations of regulatory requirements were identified in any area.
During the review of system operating and valve lineup instructions, the inspectors noted many deficiencies which brought into question the
- attention to detail and technical adequacy of the preparation and review
- process. While the instructions were deemed to be minimally adequate, the applicant has undertaken a 100% re evaluation of the those instructions to resolve the identified discrepancies and ensure technical adequacy of the
- instructions (Paragraph 13).
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-DETAILS 1.
Persons Contacted
- M.!D. Lyster, Manager, Perry Plant Operations Department
- J. J. Waldron, Manager, Perry Plant Technical Department (PPTD)
- F. R.-Stead, Manager, Nuclear Engineering Department
-*C. M. Shuster, Manager, Nuclear Quality Assurance Department
- R. A. Stratman, General Supervising Engineer, Nuclear Design and Analysis Section
- K. R. Pech, General Supervising Engineer, Nuclear _ Construction Engineering Section
- B. D. Walrath, General Supervising Engineer,' Operational Quality Section
- W. R. Kanda, Jr., General Supervising Engineer, Technical Section
- R.-'J. Tadych, General Supervisor, Operations Section
- S..F. Kensicki, Technical Superintendent, PPTD During this inspection period, the inspectors also contacted numerous-other applicant personnel and contractors and consultants acting on behalf of the applicant.
'* Denotes those persons in attendance at one or more of the exit interviews conducted throughout the inspection period and at the conclusion of the inspection period on December 27, 1985.
. 2.
Applicant Action on Previous Inspection Findings (92701)
a.
'(Closed) Open Inspection Item (440/84016-02(DRSS)):
Review of QC/QA results of spiked and split samples provided to chemistry technicians.
The applicant has implemented procedure RAP-0204, " Chemistry Unit-Analytical Quality Control Program," Revision 1, November 7, 1985.
Subsequent to the issuance of the last inspection report in this area, (Inspection Report No. 440/85070(DRSS)), the applicant submitted evidence.that the six technicians to be qualified to meet the ANSI /ANS-18.1-1971 and RAP 0204 requirements, had completed the tests on the analyses for low-level chloride using the ion chromatograph and the potentiometric boron method for samples from the standby' liquid control system.
This item is considered closed.
b.
(0 pen) Unresolved Item (440/85006-04(DRSS)):
Resolve with NRR the use of silicone sealant on ventilation ductwork.
The applicant and NRR met onsite to discuss this issue on November 25, 1985.
The technical concerns and regulatory positions were discussed and the meeting participants agreed to hold further meetings via conference calls after the applicant and NRR each held internal caucuses.
c.
(Closed) Open Inspection Item (440/85010-07(DRP)):
General employee training course deficiencies. An interim inspection on this item is documented in paragraph 2 of Inspection Report 440/85046(ORP).
The remaining open aspect of this item concerned completion of
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re-training those personnel already qualified under the General Employee Training Program with the revised training lesson plans.
-Approximately 1200 personnel required retraining. -As.of November 1,
'1985, approximately 100 personnel had not been retrained with the revised lesson plans.
Consequently, their access badges to the facility were pulled and to regain access to the facility the individual has to complete the retraining program.
The inspector has no further concerns in this area.-
d.
(Closed) Open Inspection Item (440/85022-07(DRP)):
TMI Items I.C.1-and I.C.9, Emergency procedures generation in accordance with Safety Evaluation Report (SER) Sections 4.4.7.2, 6.3.3 and 13.5.2.2.
The inspectors reviewed the Plant Emergency Instructions (PEIs) during this inspection period. The results of this inspection activity are contained in Paragraph 7 of this report.
A critical aspect of this review was verifying instruction adequacy in accordance with Operations Administrative Procedure (0AP)-0507, Revision 2,
" Preparation of Plant Emergency Instructions." 0AP-0507 is a portion of the NRR approved emergency Procedures Generation Package (PGP) containing methodology for organizing, preparing, reviewing, approving, revising and validating the emergency instructions.
The inspectors verified that the reviewed PEIs were prepared in accordance with 0AP-507.
The inspector also examined the Perry Specific Technical Guidelines in the PGP for reactor-pressure vessel and containment control when reviewing those emergency instructions.
The Senior Resident Inspector discussed the inspection scope in response to this issue with the NRR Licensing Project Manager, and NRR stated that this inspection was sufficient for closeout of the item.
The inspector has no further concerns in this area.
e.
(Closed) Open Inspection Item (440/85022-28(DRP)):
This item was opened to track Region III's review of TMI Action Plan Item No.
II.K.1.22.
This Action Plan Item concerns the applicant action required in item 3 of I.E. Bulletin No. 79-08 dated April 14, 1979, which requires the applicant to describe the actions, both automatic and manual, necessary for proper functioning of the auxiliary heat
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removal systems (e.g., RCIC) that are used when the main feedwater system is not operable.
For any manual action necessary, the bulletin required the applicant to describe in summary form the procedure by which this action is taken in a timely sense.
In response to item II.K.1.22, the applicant, in a letter from D. R. Davidson (CEI) to R. Tedesco (NRC), dated October 30, 1981, described actions necessary for proper functioning of the auxiliary heat removal system as required by the bulletin.
NRR reviewed the applicant response as documented in section 5.4.2 of the Perry SER (NUREG-0887).
The inspector compared the applicant's response to item II.K.1.22 with Perry Plant Emergency Instruction (PEI)-B13, " Reactor Pressure Vessel Control", PEI-G42, " Suppression Pool Level Control", PEI-E12,
" Suppression Pool Temperature Control", and other plant operating
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cooling.
From this review the inspector determined that the actions described by the applicant's response to item II.K.1.22 are reflected in plant procedures.
This' item is considered closed.
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. f.1-(Closed)'Open Inspection Item (440/85022-43(DRP)):
Discrepancies on-Reactor Core Isolation Cooling (RCIC) system logic testing.
This item was identified by NRR as requiring further evaluation by the inspection staff for resolution.
Regional test program inspectors performed RCIC preoperational test procedure and results verifications demonstrating that the design was in accordance with the SER.
Following a telephone conversation between the Senior Resident Inspector and the NRR Licensing Project Manager on November 26, 1985, NRR concluded that no_further inspection activity was necessary regarding this item.
g.
(Closed) Open Inspection Item (440/85022-46(DRP)):
During the NRC's comparison of the completed plant with the applicant's as-built drawings and FSAR descriptions, many minor discrepancies were noted between the FSAR diagrams and the plant.
These changes were due to posted design changes against the current controlied Piping and Instrument Diagrams.(P& ids) which had not been incorporated into the FSAR. The applicant had previously committed to provide updated P& ids in the last FSAR amendment prior to fuel load.
The applicant submitted Amendment 22 to the Perry FSAR on November 20, 1985.
This amendment included the remaining updated FSAR P& ids.
During the inspector's review of the applicant's program for updating the FSAR P& ids, it was determined that the FSAR diagrams are matted duplicates of the controlled Plant P& ids at the time of the amendment submittal.
After the final prelicense P&ID FSAR submittals were issued.the applicant has been tracking the Plant's controlled P&ID drawings including all revisions, and change paper.
An ongoing review of all drawing changes is conducted by the applicant to assure that any significant change to the drawings, that alters the plant as described in the FSAR, results in a FSAR change submittal.
Minor changes to the drawings, which do not alter the description of the
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plant as described in the FSAR, will result in FSAR diagram updates no later than two years after the issuance of the operating license and annually thereafter in accordance with 10 CFR 50.71.
The drawing discrepancies identified between the FSAR diagrams and the plant, which resulted in this open item, were minor and did not require an immediate (prior to licensing) FSAR update.
This item is considered closed.
h.
(Closed) Open Inspection Item (440/85033-19(DRP)): Technical Specification instrument setpoint methodology and limiting conditions for operation.
By memorandum dated December 2, 1985, from R. M. Bernero (NRR) to C. E. Norelius (Region III), NRR acknowledged that the PNPP, Unit 1, technical specifications are consistent with the BWR 6 Standard Technical Specifications in that no readjustment of trip setpoints is required by the technical specifications if the
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setpoint is found between the required " trip setpoint" value and
" allowable value" as defined in the technical specification.
NRR also acknowledged that in discussions with the applicant, the applicant intends to readjust the trip setpoint if found nonconservative with respect to the applicant defined tolerance band, the " leave-ac-is-zone" (LAIZ).
NRR further accepted this practice provided that the applicant's setpoint methodology confirms that there is ample allowance between the required and allowable values to accommodate the LAIZ.
By letter dated October 17, 1985, the applicant c amitted to submit a setpoint methodology report prior to startup 'ollowing the first refueling outage for Unit 1.
In the interim, the applicant implemented Special Project Plan (SPP) 0301, " Coordination of Setpoints and Interrelated Documents", to clearly state the definition of and calculational techniques utilized to prescribe the LAIZ as well as other setpoint values and terms.
SPP 0301 defines the LAIZ as the setpoint plus or minus the square root of the sum of the squares of the vendor supplied accuracies for all loop components.
Also, the LAIZ shall not exceed the allowable value.
The technique is in accordance with Instrument Society of America (ISA) Standard S67.04-1982, "Setpoints for Nuclear Safety Related Instrumentation used in Nuclear Power Plants".
Plant Administrative Procedure (PAP)-1403, Revision 1, " Control of Setpoints", requires in Section 5.3 that " Devices for which the setpoint is found to be outside the leave-as-is-zone shall be recalibrated".
The inspector reviewed the channel functional LAIZ and loop calibration LAIZ for the reactor pressure vessel (B21) pressure and level instrumentation and found the calculated LAIZ's comprise less than 27% and 38%
respectively of the difference between the appropriate required setpoint and allowable value.
The eighteen month instrument drift allowance provided by the vendor is less than the calibration tolerance used to calculate the LAIZ.
Consequently, there appears to be ample margin between the required setpoint and allowable value to accommodate the LAIZ.
The inspector has no further concerns in l
this area.
i.
(Closed) Open Inspection Item (440/85033-21(DRP)):
Use of out of service switch caps in the control room.
This item was initiated when the inspector observed that the Main Steam Isolation Valves (MSIV) had been rendered inoperable by isolating the compressed air to the MSIV actuators.
The isolation air supply valves were tagged out of service in the field.
No indication on the control panel in the control room existed to alert th's operators that the MSIVs were out of service.
This occurred during a preoperational testing activity.
The purpose of the testing activity was to verify actuation logic following manipulation of the control room switches.
Because of this unique testing activity, switch caps were not placed on the MSIV switches in the control room.
The inspector reviewed Plant Administrative Procedure (PAP)-1401, Revision 1, " Equipment Tagging," and found that during the operational phase of the plant,
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whenever a system is taken out of service, the first tag placed would be the switch cap or tag in the control room and the last tag removed would be the switch cap or tag in the control room.
This is referenced in sections 6.2.6.5, 6.2.9.4, 6.2.11.6 and 6.2.13.6 of PAP 1401.
Because of this administrative control and the low probability that the unique preoperational testing that resulted in the concern would recur during operations, the inspector has no further concern in this area.
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(Closed) Open Inspection Item (440/85046-03(DRP)):
During a review of the proposed Perry Technical Specifications, the inspectors noted that there were no technical specifications proposed for the
Redundant Reactivity Control System / Alternate Rod Insertion System.
In a letter from Harold R. Denton, Director, Office of Nuclear Reactor Regulatian (NRR) to John M. Fulton, Chairman, BWR Owner's Group, dated August 19, 1985, Mr. Denton noted that technical specifications do not currently address the backup scram system and alternate rod insertion system and that technical specification requirements for Anticipated Transients Without Scram (ATWS)
mitigating systems are under review as part of the NRR Technical Specification Improvement Program.
Due to the generic nature of this concern (affects all BWRs) and the ongoing work being performed by the NRC, which will result in specific guidance to the applicant, this item need not be tracked as a plant specific open item.
Changes, if any, to the Perry Technical Specifications, resulting from NRR's Technical Specification Improvement Program, will be reviewed at the time NRC guidance is provided.
This item is considered closed.
k.
(Closed) Open Inspection Item (440/85056-03(DRP)):
This item concerns a discrepancy, discovered during testing, between the plant and Technical Specifications 4.8.1.1.2.e.14.A(2) and 3.
Those specifications indicate that having either the diesel generator
" local / remote" switch in the local position or the "inop/ normal" switch in the inop position would prevent the diesel generators from starting.
Preoperational testing indicated that it was necessary to have both the " local / remote" switch in local and the "inop/ normal" switch in inop to prevent the diesel from starting.
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The applicant prepared Design Change Package (DCP) No. 850346 which modified applicable diesel generator starting circuitry in order to comply with the system design as reflected in the above referenced technical specifications.
The inspector reviewed the design change documentation and work order numbers 850010127 and 850010128 which performed the modifications specified in DCP No. 850346.
The inspector also reviewed the testing steps to verify that the modification was acceptable.
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During the review of this item, the inspector also reviewed Surveillance Instruction SVI-R43-T1327, " Division 1 Standby Diesel Generator 18 Month Functional Test" to verify that the instruction satisfied the requirements specified in Technical Specification 4.8.1.1.2.e.14.
This Technical Specification requires that the applicant verify, at least once per 18 months, that certain diesel generator lockout features prevent the diesel generators from starting on an automatic initiation signal.
The inspector determined that the modifications resolved the issue and in place surveillance instructions satisfy the requirements of the Plant Technical Specification in que tion.
This item is considered closed.
1.
(Closed) Open Inspection Item (440/85056-04 1RP)):
Conduct of safety evaluations required by 10 CFR 50.59.
As discussed in Inspection Report No. 440/85059(DRP), the applicant had initiated actions to assure that past and future safety evaluations were/are performed properly.
These actions included training of personnel to Revision 3 of Perry Administrative Procedure (PAP)-0305, " Safety Evaluations" and determining whether or not procedures and changes thereto previously evaluated needed to be re-evaluated.
The above actions, as well as re-evaluation of identified procedures, were completed on December 18, 1985.
Procedures identified as requiring re-evaluation were Off Normal Instructions (0NIs) and Perry Emergency Instructions (PEIs).
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(Closed) Open Inspection Item (440/85059-03(DRP)):
Revision of Surveillance Instruction (SVI) 821-T0061, " Reactor Vessel Water Level Low Level 1 and 2 Channel Functional for 1821-N681B" to correct inspector identified discrepancies.
The inspector reviewed Revision 1 of the subject SVI dated November 8, 1985, and Temporary Change Notice (TCN) No. 001 to Revision 1 of the subject SVI dated November 20, 1985.
These two procedure changes satisfactorily resolved the inspector identified discrepancies.
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(Closed) Unresolved Item (440/85059-04(DRP)):
Revision of Surveillance Instruction (SVI) B21-T0252, " Reactor Vessel High and Low Water Level Channel A Response", to correct inspector identified discrepancies in Section 1.1, " Scope" and the sequences for valving level transmitter 1821-N080A in and out of service.
The inspector reviewed Temporary Change Notice (TCN) No. 002 to the subject SVI which was issued on November 24, 1985.
This TCN corrected the noted discrepancies.
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(Closed) Open Inspection Item (440/85059-06(DRP)):
Improper valve position specified in Valve Lineup Instruction (VLI) P54 for valve IP544726.
The inspector reviewed Revision 1 to the subject VLI issued on November 7, 1985.
The VLI was revised to correctly specify valve IP54-F726 as " locked closed".
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(Closed) Open Inspection Item (440/85082-01(DRP)):
Technical specification setpoint discrepancies.
In response to this open
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item, the applicant submitted a letter dated December 12, 1985, from M. d. Edelman (CEI) to C. E. Norelius (NRC), describing actions they had taken to ensure the adequacy of technical specification setpoints.
The applicant further described that the apparent discrepancy identified by the inspection team was not inconsistent with the accident analysis analytical limit.
The value quoted in the referenced FSAR section was a nominal value.
In addition, the applicant described a sampling review of values in FSAR Chapters 6, 7 and 15 against the appropriate technical specification sections.
The inspector reviewed the results of this sample analysis in detail and found the sample to be representative and the technical specification values to be conservative in every case.
The inspector has no further concerns in this area.
3.
Followup on 10 CFR Part 21 Reportable Deficiencies (92700)
The following 10 CFR Part 21 reports were received by the applicant from vendors of Perry equipment / components.
The applicant evaluated the reports using its Deviation Analysis Report (DAR) program and concluded that the discrepancies identified on the 10 CFR Part 21 reports were not reportable to the NRC as potential construction deficiency reports under 10 CFR 50.55(e).
The inspectors reviewed the DARs to determine if the action taken, if any, was appropriate to correct the cause of the defect, and that the actions were complete.
A review for reportability under 10 CFR 55.55(e) was also conducted during the inspection.
a.
(Closed) 10 CFR Part 21 Reportable Item (440/81002-PP (DAR-065):
Potentially defective intake and exhaust valve springs manufactured by Melrose Spring Company used in Transamerica Delaval diesel engines.
On August 14, 1981, the applicant received notification from Transamerica Delaval Inc. (TDI) regarding the potential use of defective intake and exhaust valve springs in the TDI diesel engines supplied to Perry Nuclear Power Plant.
An inventory search was conducted and it was determined that the diesel engines, spare cylinder heads and spare parts supplied to Perry Nuclear Power Plant did not contain any of the potentially defective springs.
This item is considered closed, b.
(Closed) 10 CFR Part 21 Reportable Item (440/81003-PP (DAR-087):
Potential termination deficiency in Potter-Brumfield relays supplied by General Electric Company.
On December 7, 1981, the Nuclear Regulatory Commission was notified by General Electric Company that loose terminations were detected on Potter-Brumfield relays used in the main steam line isolation valve leakage control system panels at Grand Gulf Nuclear Plant, Unit 1.
PNPP had also been supplied with panels utilizing Potter-Brumfield relays.
As a result of the notification to Perry, a 100% inspection of Potter-Brumfield relays was implemented.
All of the relays were found to have terminations made with either saddle clamps or ring lugs precluding the identified deficiency.
The corrective action documented on Deviation Analysis Report 087 reviewed by the.1spector appear to be adequate.
This item is considered closed.
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c.
(Closed) 10 CFR Part 21 Reportable Item (440/82005-PP) (DAR 108):
A letter from Pacific Scientific concerning a potential reportable 10 CFR 21 incident received on October 1, 1982, reported that damage was found in some Pacific Scientific snubbers at another nuclear facility after the snubbers were tested by an outside testing facility and reported as " good units".
Pacific Scientific inferred that the damage was caused by improper testing and recommended that any Pacific Scientific shock arrestors tested and approved by testing sources other than Pacific Scientific Company be returned to Pacific Scientific for retesting to assure functional integrity had not been compromised as a result of improper testing.
The applicant initiated DAR 108 which identified the Pacific Scientific concern and reported that none of their snubbers were currently known to be damaged. The inspector was informed that no testing other than performed by Pacific Scientific had been conducted on their snubbers.
In addition, the applicant described their program to perform functional testing of snubbers which, with input from Pacific Scientific, assures that their snubbers are not damaged as described in the Part 21 report.
This item is considered closed.
d.
(Closed) 10 CFR Part 21 Reportable item (440/83003-PP) (DAR 132):
Comsip, 10 CFR Part 21, dated April 13, 1983, reported that the catalyst used in the Comsip models K-III and K-IV containment gas monitoring systems would remain useful for only 10 days following use with iodine concentrations meeting or exceeding what could be expected in a large BWR post accident.
The report identified a new catalyst bed configuration which was shown to be useful after 5 months of use in the prescribed iodine concentration.
Comsip recommended to the applicant, in a letter dated June 6, 1983, to replace the standard catalyst beds with the new configuration which showed no degradation after 5 months of continuous testing.
The applicant initiated DAR No. 132 which identified Comsip's deficiency, and the affected component as analyzer cells M51-N714A and B (Post-Loca Containment Gas Monitoring System).
The applicant concluded that the 10 CFR Part 21 was not a significant deficiency since it only monitors hydrogen concentration and does not provide any automatic control functions.
The applicant went on to state that should the hydrogen monitors be rendered inoperable, the other components of the combustible gas control system (hydrogen recombiners and back-up purge systems), could be started, allowing sufficient time for repairs of the hydrogen analyzers or to manually obtain and analyze samples for hydrogen.
The applicant also stated that they intended to replace the catalyst with the Comsip recommended new catalyst beds.
The inspectors reviewed the DAR and verified that the catalyst was replaced with the new catalyst by review of Work Authorization (WA)
NTS-85-2552 and associated Nonconformance Report (NR) NDS-20. The inspector considers this item closed.
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e.
~ (Closed) 10 CFR Part 21 Reportable Item (440/84002-PP) (DAR 192):
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. On July 13, 1984, Transamerica Delaval Inc. (TDI) reported a potential defect, in accordance with the requirements of 10 CFR Part o
L 21, concerning the Standby Diesel Generators (R43) Engine Intake and
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Exhaust Valve Springs which could result in engine non-availability.
TDI concluded in the report that the failure identified on a non nuclear marine engine application was isolated and "no corrective action is required".
The applicant initiated DAR 192 which identified the potential
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defect and concluded that the failure to the non-nuclear marine engine was isolated.
The DAR also stated that an inspection of the
diesel valve springs for surface imperfections is included in the
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site inspection plan to be performed during the Diesel Inspection Program as required by the Diesel Owners Group.
The inspector reviewed the above documentation, including NCR 0QC-0970 which documented the potential defect and dispositioned use-as-is based on the conclusion that the reported failure was
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isolated.
This item is considered closed.
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Subsequent to the applicant's review and disposition of this 10 CFR
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Part 21, an additional 10 CFR Part 21 report was issued by TDI on November 6, 1985, concerning the same subject.
Apparently other l
failures of diesel intake and exhaust valve springs were identified l
in October of 1995.
Review of the applicant's actions concerning L
- this new 10 CFR Part 21 is addressed in item 3.h of this report.
(440/85002-PP).
f.
(Closed) 10 CFR Part 21 Reportable Item (440/84003-PP) (DAR 193):
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On-July 13, 1984, Transamerica Delaval Inc. (TDI) reported a l
potential defect, in accordance with the requirements of 10 CFR Part 21, concerning the Standby Diesel Generators (R43) engine fuel
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injection pumps.
TDI stated that the cause of a fuel injection pump i
L failure at the Catawba Nuclear site, was due to a material defect in the delivery valve holder associated with the fuel injection pump.
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TDI informed the applicant that they believed the failure to be an isolated case and that they have'used the Bendix fuel injection pump
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on all DRS and DSRV engines manufactured for at least the last 15 years with no other known failures of this type.
The applicant initiated DAR 193 which identified the potential defect and concluded that the failure of the fuel injection pump was
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isolated and the pumps are to be inspected and re-calibrated as part
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l of the standard, pre-engine start-up and test program.
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i This inspector reviewed the above documentation, including NCR l
0QC-0971 which documented the potential defect and dispositioned
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f use-as-is based on the conclusion that the reported failure was isolated. This item is condidered closed.
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(Clored) 10 CFR Part 21 Reportable Item (440/85001-PP) (DAR-257):
Diesel engine intake silencer deficiency.
On November 1, 1985, the applicant was notified by Transamerica Delaval Inc. of a September 3, 1985, 10 CFR Part 21 report filed by American Air Filter Company regarding a potential manufacturing deficiency on the PNPP diesel engine intake silencers.
The deficiency involved an internal part, silencer centerline bullet end cap, not welded in place.
The applicant inspected the eight American Air Filter intake silencers used at PNPP and verified proper welding of the end caps (Reference:
Nonconformance Reports 0QCS-0176 and 0QC-3245).
The applicant concluded that no further action is required in response to that report.
This item is considered closed.
h.
(Closed) 10 CFR Part 21 Reportable Item (440/85002-PP) (DAR-258):
Diesel engine valve springs.
On November 6, 1985, the applicant received a copy of a 10 CFR Part 21 report filed by Transamerica Delaval Inc. (TDI) regarding failures of diesel engine intake and exhaust valve springs manufactured by Betts Spring Company.
The applicant initiated Deviation Analysis Report (DAR) 258 in response to that report.
The applicant had visually inspected all valve springs as part of the engine Design Review / Quality Revalidation program and found no surface discontinuities which could result in stress risers.
This was the action recommended by TDI.
The applicant concluded in DAR 258 that no further action was required.
This item is considered closed.
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(Closed) 10 CFR Part 21 Reportable Item (440/85003-PP) (DAR 249):
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General Electric (GE) notified the NRC of a reportable defect per 10 CFR Part 21 on June 17, 1985, concerning the misapplication of switch CR2940 on the Standby Liquid Control System (SLCS) (installed in a potential harsh environment but qua:ified for only mild environments) which may cause failure of Class 1E equipment.
' he applicant' initiated DAR 249 which identified the G.E. concern i
and stated that the switches in question had already been relocated (in Unit 1) to a mila environment prior to notification by G.E. due to other design considerations.
The inspector reviewud the above documentation and Field Deviation Disposition Request (FDDR) No. KLI-964 which directed the movement of the CR2940 switches to a mild environment prior to notification by GE of their reportable condition.
The inspector verified that the location of switches C41-C001A, (B) and C41-F001A (B) (the CR2940 switches) were in a mild environment location.
This item is considered closed.
4.
Preoperational Tests Results Reviews (84522, 84523, 84524)
Preoperational testing has been completed on the following test packages.
There were no outstanding exceptions at the end of testing.
No problems were noted during the inspector's review.
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1021-P-001 Area Radiation Monitoring System
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1017-P-001 PRMS Non-GE Channels
OG50-P-001 Liquid Radwaste System (Sumps)
1M98-P-001 ESF System Inplace Filter Testing Preoperational testing has been completed on the following test package.
A deficiency report has been generated to correct a high background noted on one channel.
The deficiency report is being tracked by the applicant's tracking system.
The needed correction does not preclude adequate operation of the system.
No additional problems were identified during the inspector's review.
No further inspection review is planned.
- 1017E-P-001 PRMS Main Steam Line Subsystem In a letter to NRR, dated November 20, 1985, the applicant requested deferral of the following preoperational test until initial criticality.
- 0G51-P-001 Solid Radwaste Disposal System No violations of regulatory requirements or deviations from commitments were identified.
5.
Followup on an Allegation (99014)
(Closed) Allegation (AMS-RIII-85-A-0125):
This allegation was received from a third party and concerns alleged bad welds found on the polar cranes of both units.
It was alleged that some welds were radiographed and some welds on the Unit 1 crane were repaired.
However, other welds were ignored after only a few repairs were made.
Several attempts were made to contact the alleger.
Since the third party who provided us details of the allegation would not reveal to us the allegers name and was unable to persuade the alleger to contact us directly, no specific information concerning the location of the welds in question were obtained.
A review of the history of the installation of structural steel associated with the Perry plant polar cranes was conducted by the applicant to determine if a connection between the allegation and identified discrepancies could be made.
The results of the review identified that prefabricated Unit 1 polar crane box girder shop welds were rejected during the receipt of the girders due to various visual defects (porosity, underfill, undercuts, convexity and concavity).
These defects were identified and repaired on Nonconformance Report (NCR)17-136, Revisions 0, 1, 2, and 3.
Subsequent to the repairs of Unit 1 box girders, Unit 2 box girder shop welds were found to have major amounts of linear indications (lack of fusion), documented in NCR 17-136, Revisions 6 and 7, which resulted in
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the applicant notifying the NRC of a potential 10 CFR 50.55(e) reportable defect (applicant Deviation Analysis Report (DAR) No. 71).
Repairs of defective shop welds on the Unit 2 box girders were performed at Perry and reviewed by the NRC resulting in the closure of the 10 CFR 50.55(c)
(NRC item 441/81016-EE) in Inspection Report No. 440/83009.
Due to the defects found on the Unit 2 box girders, a limited informational f
reinspection-(NDE) of the Unit 1 box girders was performed per Revision 8 i
of NCR 17-136.
The ultrasonic examinations (UT) performed on the Unit 1 box girders revealed that three minor indications existed but were not similar to the Unit 2 defects in that no linear indications or lack of fusion was detected.
The minor defects discovered during the reinspection were evaluated by the applicant and determined to be within the allowable code and therefore not repaired.
The primary reason Unit 1 box girders were not found to have similar defects found on the Unit 2 box girders was that the assemblies were fabricated by different vendors using different welding techniques.
The NRC review of the Unit 1 l
reinspection was documented in Inspection Report 440/82006.
Without specific information as to the location of the welds questioned by the alleger, it has been assumed that the problem identified above and documented by the applicant in NCR 17-136 is the concern to which the alleger was referring.
It is therefore the inspector's conclusion that, although some defects were identified and repaired while others were not repaired due to the nature of the defects and the evaluation by the applicant's engineering staff that no rework was necessary, there is no safety significance to this allegation.
6.
Followup on Regional Requests (92705)
a.
Regional management requested the resident inspection staff determine the applicability to PNPP Unit 1 of a River Bend Nuclear Station previously unanalyzed containment negative pressure scenario involving a Reactor Water Clean Up system line break event.
(Reference:
Letter from J. E. Booker (GSU) to R. D. Martin (NRC)
dated April 19, 1985).
The inspector reviewed this scenario with the applicant and concluded that the scenario is not applicable to PNPP due to a different containment design.
PNPP containment design as described in FSAR Section 6.2.1.1.4.2.1 includes a vacuum relief system designed to prevent containment negative pressure from exceeding the design value of 0.8 psi.
The inspector has no further concerns in this area, b.
Regional management informed the inspectors that a fuel assembly lifting rig failed at the Hatch Nuclear Plant, Unit 2 on November 8, 1985. The failure was due to an improperly labeled cable terminal assembly.
The assembly was indicated to contain a 7/16 inch threaded female connector intended to mate with a 7/16 inch threaded eye bolt.
In fact, the terminal assembly female connector was 1/2 inch and resulted in failure due to improper thread engagement.
The applicant examined all of their fuel handling equipment and found that they had only equipment with indicated 1/2 inch and 3/4 inch lifting connectors.
The applicant further tested each assembly and verified that all equipment was properly sized.
The inspector has no further concerns in this area.
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7.
Followup of TMI Action Plan (NUREG-0737) Items (25401)
a.
(0 pen) TMI Item I.A.1.3.2.A, " Shift Manning".
As discussed in
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paragraph 18 of this report, the inspector reviewed Perry Administrative Procedure (PAP)-0110.
This procedure contained the shift manning requirements specified by this item as well as the Perry Technical Specifications.
The inspector was informed by the applicant that the procedure would be fully implemented some time before initial fuel load.
Implementation will be verified in a future inspection prior to fuel load, b.
(Closed)TMI Item I.D.2.2, " Plant Safety Parameter Display (SPDS)
Installation".
The inspector determined by discussions with applicant personnel, review of the Perry SER and direct examination of the SPDS display console that the applicant had installed the SPDS hardware and software described in applicant licensing submittals.
The Perry SPDS is a subsystem of the Emergency Response Information System (ERIS) computer.
The SPDS inputs to ERIS and associated alarm and display functions were preoperationally tested in preoperational acceptance test C95-A001, " Emergency Response Information System".
c.
(0 pen) TMI Item I.D.2.3, " Full Implementation of Safety Parameter Display Console".
In Supplement 7 to the Perry SER the NRC staff reported that it had reviewed the proposed Perry SPDS and concluded that no serious safety questions were posed and therefore implementation could continue.
The staff reserved judgment as to whether or not the SPDS fully met the requirements of this item pending a planned post implementation audit to be conducted in the future.
This item therefore remains open pending the subject audit and resolution of any identified deficiencies.
d.
(Closed) TMI Item II.K.3.22.A, " Reactor Core Isolation Cooling Suction Switchover - Verify Procedures".
This item does not apply to Perry since the Perry design incorporated the automatic suction switchover (see the following item),
e.
(Closed) TMI Item II.K.3.22.B. " Reactor Core Isolation Cooling Suction Switchover - Modifications".
In Section 7.4.2.1 of the Perry SER the staff reported that the automatic RCIC suction switchover included in the Perry design met the requirements of this item and was therefore acceptable.
The inspector verified that this design feature was demonstrated to function per design in preoperational test E51-P001.
f.
(Closed) TMI Item II.K.3.25.A. " Consequences of a Loss of Cooling Water to the Reactor Recirculation Pump Seal Coolers - Propose Modifications".
In section 15.1 of the Perry SER the NRC staff reported its review and acceptance of the applicant's adoption of a BWR Owners Group evlauation to address this item.
The evaluation concluded that a total loss of pump seal cooling water to the recirculation pump seals would not result in a serious safety problem.
Design modifications were therefore not proposed, s
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(Closed) TMI Item II.K.3.25., " Consequences of a loss of Cooling Water to the Recirculation Pump Seal Coolers - Modifications".
As discussed in the preceding item, modifications to the Perry design c,
, i were not required.
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(Closed) TMI Item II.K.3.27, " Common Reference Level".
The
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inspector reviewed Supplement 7 to the Perry SER, Section 18.5 which reported the NRC staff's acceptance of the applicant's commitment to reference all the vessel level indications to the top of active fuel
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prior to fuel load.
The inspector reviewed completed Field-
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Deviation Disposition Requests (FDDRs) No. KL1-3368, Revision 0,
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Revision 1, Revision 2, Revision 3 and Revision 4 which implemented the applicant's commitments by specifying range changes to level instrument scales.
The inspector visually examined a sample of reactor vessel level indicators in the control room to verify that physical work associated with the subject FDDRs was completed.
Level indicators observed were: IN691B, IN691F, R606A, R606C and R608.
The instruments observed were in accordance with the applicant's commitment.
i.
(Closed) TMI Item II.K.3.28, "Autodepressurization System Accumulator Capacity".
In Supplement 2 to the Perry SER the NRC staff reported its review and acceptance of the Perry design to meet the requirements of this item.
No further inspection of this item is planned.
8.
Emergency Procedures (42452)
The inspectors examined the current revisions of the following instructions (procedures) developed for responding to abnormal and emergency conditions to verify that the instructions were developed in accordance with the applicant's administrative controls the facility technical specifications, ANSI N18.7-1976, and Regulatory Guide 1.33, Revision 2, and to assess whether or not the instructions were technically adequate to accomplish their stated purposes.
7]
Off Normal Instructions (ONIs)
B21-1
"SRV Inadvertant Opening / Stuck Open (Unit 1)"
B33-2
" Loss of One or Both Recirculation Pumps (B33-2)"
C11-1
" Inability to Move Control Rods (Unit 1)"
C11-2
" Uncoupled Control Rod (Unit 1)"
C11-3
" Control Rod Drop (Unit 1)"
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's C51
" Unexplained Change in Reactor Power or Reactivity (Unit 1)"
C61
" Evacuation of Control Room (Unit 1)"
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C71
" Reactor Scram (C71)"
C85-1
" Pressure Regulator Failure - Closed (Unit 1)"
C85-2
" Pressure Regulator Failure - Open (Unit 1)"
~ D51.
" Earthquake (Unit 1)"
E12-1
"Inadvertant Initiation of ECCS (E12-1)"
'J11-2
" Fuel Bundle Rupture During Fuel Handling (Unit 1)"
-N27
"Feedwater Pump Trip (Unit 1)"
N32
" Turbine and/or Generator Trip" N36
" Loss of Feedwater Heating (Unit 1)"
N62-
" Loss of Main Condensor Vacuum (Unit 1)"
P41
" Loss of Service Water (P41)"
PS2
" Loss of Service and/or Instrument Air (PS2)"
P54
" Fire (Unit 1)"
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R10
" Station Blackout (Unit 1)"
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R23-1
" Loss of Essential 480 V Bus" ZZZ-1
" Tornado or High Winds (Unit 1)"
ZZZ-2
" Natural Gas Line. Leak" Plant Emergency Instructions (PEls)
B13
" Reactor Pressure Vessel Control"
- G42
" Suppression Pool Level Control"
- E12
" Suppression Pool Temperature Control"
- D23-1
" Containment Temperature Control"
D23-2
"Drywell and Containment Pressure Control",
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- D23-3
"Drywell Temperature Control"
- These PEIs were only reviewed to verify proper preparation, review and approval in accordance with the applicant's administrative controls.
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In general, the ONIs and PEIs were found.to be adequate.
Two minor discrepancies were noted in the ONIs which resulted in Temporary Change Notices (TCN) being issued on ONI-N36 and ONI-E12-1.
One discrepancy was noted on PEI-B13 which resulted in the issuance of a TCN.
The ONIs and PEls have been and will continue to be used on the plant specific simulator which provides further assurance of their adequacy.
In assessing.the technical adequacy of PEI-B13 and PEI-D23-2, the inspector considered the Perry Specific Technical Guidelines "RPV Control Guideline" and " Primary Containment Control Guideline" contained in Attachment 6 to the applicant's Procedures Generation Package submitted to NRR by letter dated September 11, 1985, from M. R. Edelman to B. J. Youngblood, as supplemented, including the applicable deviation sheets.
No violations of regulatory requirements or deviations from commitments were identified in this area.
9.
Maintenance Procedures (42451)
The inspectors examined current revisions of the following instructions which contain specific direction for accomplishing selected electrical, mechanical and instrumentation maintenance activities.
GMI-0013 " Overhaul of ChD Scram Pilot Valves" GEI-0019 " Emergency Core Cooling Pump Motors" GMI-0002 " Turbo-charger Removal and Installation" GMI-0015 " Repair of Safety Relief Valves" GMI-0058 " Generic Maintenance and Repair Procedures for Gate Valves" GMI-0063 " Instructions for the Installation and Removal of the Reactor Pressure Vessel Head" GMI-0067 " Instructions for the Installation and Removal of Control Rod Drives and Thermal Sleeves" IMI-E2-1 " Instrument Valve Lineups" IMI-E2-4 "IRM/SRM Detector Drives" PMI-0028 " Sling Examinations and Tests" PMI-0030 " Maintenance of Limitorque Valve Operators" The review was conducted to ascertain whether or not instructions were developed, reviewed and approved in accordance with applicable administrative controls, ANSI N18.7-1976 and Regulatory Guide 1.33, Revision 2, and to assess the technical adequacy of the instructions to
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control and accomplish the activities described.
Documentation utilized in the review were the Maintenance Administrative Procedures and appropriate Vendor Manuals.
The instructions appeared to be technically adequate.
No violations of regulatory requirements or deviations from commitments were identified in this area.
10.
Surveillance Test Instruction Review (42450)
The inspectors reviewed current revisions of the following instructions developed for the conduct of surveillance testing activities to verify that the instructions were developed in accordance with the applicant's administrative controls and to assess whether or not the instructions were technically adequate to accomplish their stated purposes.
Surveillance Instructions (SVIs)
821-T0067
"DW & Cntmt Isol Viv Man Isol Logic System Func. Test" B33-T1168
" Idle Recirculation Loop Temperature and Flow" C11-T0044A
" Scram Discharge Volume H 0 Level High Channel A
Functional for IC11-N601A" C11-T0045A
" Scram Discharge Volume Water Level High Channel A Calibration for 1C11-N012A" C11-T5376A
" Scram Discharge Volume Water Float Switch Level High Channel A Functional / Calibration for 1C11-N013A" C41-T1420A
"SLCS-RWCU Isolation Initiation Channel Functional Test" C61-T0333
" Residual Heat Removal System Flow (REM Shtdn Mon) Channel Calibration for IC61-N001" C71-T0038A
" Main Steam Line Isolation Valve Closure Channel A Calibration for 1C71-N701A" C71-T0046
" Turbine Stop Valve Closure for Channel Functional for IC71-N006, A, B, C, D, E, F, G, and H" C71-T0048
" Turbine Control Valve Fast Closure Channel Functional for 1C71-N005A, B, C, and D" 017-T8031
" Unit 1 Vent Radiation Noble Gas Activity Monitor Calibration for 1017-K786" D17-T8037
" Unit 2 Vent Radiation Noble Gas Activity Monitor Calibration for 2017-K786" D17-T8038
" Unit 2 Plant Ventilation Radiation Monitor Functional for 2017-K786"
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D23-T1214A
" Suppression Pool Water Temperature Channel Functional for 1023-K052A, ID23-K062A, 1023-K072A and 1023-K082A" D23-T1215A
" Suppression Pool Water Channel Calibration for 1D23-N051A, 1023-N061A, ID23-N071A, and 1D23-N081A" E12-T1182
"LPCI Valve Lineup Verification and System Venting" E51-T2001
"RCIC Pump and Valve Operability Test" M15-T1240
" Annulus Exhaust Gas Treatment System Flow and Filter Operability Test" These surveillance instructions were reviewed to ensure conformance with the applicable administrative controls, the facility technical specifications, ANSI N18.7-1976 and Regulatory Guide 1.33, Revision 2.
Also, appropriate vendor manuals were utilized in review of these instructions.
These instructions were found to have been prepared in accordance with the relevant administrative controls and with one exception were technically adequate.
Concurrent with the inspector's review, the applicant noted that SVI B21-T0067 was faulty in that the procedure established a test condition having the main steam isolation valves being open in modes 4 and 5.
In modes 4 and 5 the main steam isolation valves would be closed.
Consequently, the procedure could not be performed as written.
The applicant has rewritten this procedure and the inspector has concluded a. review of the revised procedure with no further concerns.
No violations of regulatory requirements or deviations from commitments were noted in this area.
11.
Control Room Readiness to Control Operating Activities (71302)
On December 26 and 27,1985, the inspector monitored control room operations to ensure that activities were being conducted in accordance with the facility administrative procedures and to evaluate control room readiness for an operating license.
a.
The inspector noted that applicant personnel are not aware of the requirements of administrative procedures which had recently become effective.
In addition, personnel are not aware of which administrative procedures are presently in effect or when not yet effective administrative procedures will become effective.
Examples are noted below.
1.
The inspector noted that following the "B Residual Heat Removal (RHR) Pump and Valve Operability" surveillance the second (independent) verification of the valve lineup was performed using the same control panel electrical position indication used for the original verification.
This does not appear to be consistent with Perry Administrative Procedure (PAP)-0205, Revision 2, paragraph 6.5.3, dated 7-3-85.
After further discussion with the operations organization, the inspector determined that this practice was not isolated to one
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operator.
This item is considered open pending licensee review to determine the effect of this practice on valve lineups used to declare systems operable and subsequent NRC review of licensee findings and resolution.
Open Item (440/85078-01).
-2.
An administrativa prccedure that became effective about December 20, 1933, requires a Radiation Work Permit (RWP) to enter the containment.
On December 27, 1985, the operating shift noted that personnel were entering the containment and no RWP was in effect.
3.
An administrative procedure that became effective about December 16, 1985, requires stamping charts with a specific stamp and filling in the required information.
The required stamp could not be located and the individual changing the charts indicated that he had not seen the procedure.
He also indicated that he was being routed procedures for training purposes that had been effective up to one month ago.
Because of the short inspection time, the inspector did not obtain details concerning these administrative procedures.
During discussions with the operating crews, however, the inspector noted that they were quite uncomfortable because they were not aware of the status of the many administrative procedures.
The applicant agreed to identify the administrative procedures which were now in effect, determine when the other administrative procedures would be effective, and determine if the effective administrative procedures were being complied with.
The status of the administrative procedures will be discussed during a meeting between the NRC and applicant on or about January 15, 1986.
b.
The inspector observed shift turnovers at the Shift Supervisor, Unit Supervisor, and Supervising Operator levels.
The turnovers were complete and were in accordance with the applicant's requirements.
c.
The inspector observed operator response to alarms in the control room.
The inspector noted that in several instances alarms were not investigated for excessive periods of time.
One instance occurred where the audible alarm was silenced and the visual alarm was not noted nor acknowledged for six to seven minutes.
The alarm was finally noted when other duties (marking of chart paper) required the operator's attention at the panel.
The inspector was also informed of a recent event where a low level alarm was not noted until cooling and seal water were apparently lost to equipment which then had to be secured. Although the short inspection time did not allow the inspector to obtain details and verify the cause of this event, it is consistent with the inspector's observations.
The apparent inattentiveness to alarms appeared to coincide with nuisance alarms which would periodically plague the operators.
The Unit Supervisors became aware of the nuisance alarms very quickly and initiated corrective action in an expeditious manner.
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.The inspector believes that continued emphasis in eliminating the nuisance alarms will allow the operator to increase his attentiveness to the alarms.
The applicant agreed to re-evaluate this area and.to discuss it during a meeting with the NRC on or about January 15, 1986.
The inspector's:overall impression of control room operations was that s1though it appears that good progress is being made, control room operations at the time of this inspection were not yet at the required level for receipt of an operating license.
During the inspection, the inspector determined that personnel knowledge of administrative procedure content and status may not be adequate to support an operating license and the nuisance alarms being received in the control room is sufficiently distracting that the operators may not recognize and take action on important alarms.
No violations of regulatory requirements or deviations from commitments were identified in this area.
12.
Operating Staff' Training (41301)
=As part of a continuing review of FSAR Chapter 13 and applicant staff
_ qualifications and training,_the inspector noted that the only PNPP, Unit 1, systems training provided to the Plant Manager was a one week simulator class in 1982.
Credit was taken for the individual's previous training at the Vermont Yankee Nuclear Plant necessary te his Senior Reactor Operators license and the individual's educational background.
After discussing this situation with the applicant, the applicant committed to provide the Plant Manager with detailed PNPP, Unit 1, systems training before the end of the first refueling. outage.
The inspector has no further concerns in the interim period based on the individual's on the. job experience and the training of the Operations General Supervisor. The' applicant is tracking this commitment through its computerized training scheduling system.
Inspection closeout of this issue will be tracked as an open inspection item with a completion
. milestone at the end of the first Unit 1 refueling outage (440/85078-02(DRP)).
No violations of regulatory requirements or deviations from commitments were identified.
13.
System Operating and Valve Lineup Instruction Review (42450)
The inspector reviewed the following system operating instructions (50Is)
to verify that they were prepared, reviewed and approved in accordance with the applicant's Operations Administrative Procedure (0AP)-0502, Revision 0, " Preparation of System Operating Instructions".
a.
SOI-C11(RCIS), Revision l'
" Rod Control and Information System"
-501-C51(APRM), Revision 1
" Average Power Range Monitoring System (APRM) (Unit 1)"
,
o 50I-E21, Revision 2
" Low Pressure Core Spray System (Unit 1)"
S0I-E32, Revision 1
" Main Steam Isolation Valve Leakage Control System (E32)"
S0I-G33, Revision 1
" Reactor Water Cleanup System (Unit 1)"
S0I-M25/26, Revision 2
" Control Room HVAC and Emergency Recirculation System" S0I-M39, Revision 2
"ECCS Pump Room Cooling System (M39)"
S0I-M51/56, Revision 2
" Combustible Gas Control System and Hydrogen Ignitors (Unit 1)"
S0I-P42, Revision 2
" Emergency Closed Cooling System (Unit 1)"
S01-P57, Revision 2
" Safety Related Instrument Air System (Unit 1)"
S01-R23, Revision 1
"480 Volt Load Centers (R23)"
50I-R45, Revision 0
" Division 1 and 2 Diesel Generator Fuel
,
Oil System (R45)"
The inspector performed detailed reviews of the following S0Is and Valve Lineup Instructions (VLIs) to evaluate their technical adequacy.
Employed during this review were OAP 0502, OAP 0503, Revision 1,
" Preparation of Valve Lineup Instructions", and the applicable piping and instrumentation diagrams (P& ids) and elementary wiring diagrams.
SOI-C41, Revision 2
" Standby Liquid Control System (Unit 1)"
VLI-C41, Revision 3
" Standby Liquid Control System (Unit 1)"
501-P45, Revision 1
" Emergency Service Water System (Unit 1)"
VLI-P45, Revision 1
" Emergency Service Water System (Unit 1)"
S0I-M15, Revision 2
" Annulus Exhaust Gas Treatment (AEGT)
System" VLI-M15, Revision 2
" Annulus Exhaust Gas Treatment (AEGT)
System"
S01-E22A, Revision 1
"High Pressure Core Spray System (Unit 1)"
VLI-E22A, Revision 1
"High Pressure Core Spray System (Unit 1)"
The inspector found discrepancies in these instructions regarding the l
accuracy and completeness of the information provided bringing into
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question the attention to detail during the preparation, review and i
approval process.
The following list describes the types of discrepancies identified.
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Valve and component numt'ers incorrectly identified or missing.
- Action restoration steps missing.
- Instrument and switch numbers missing.
- Throttle valves without specific position identification.
- Inadequate control of electrical fused disconnects when going from
system secured status to standby readiness.
System secured status configuration different in the VLI and SOI.
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Improper or inadequate notes and cautions.
- Instruction steps requiring too many actions.
- In addition, in 50I-C41, the inspector identified a valving configuration which could potentially damage the standby liquid control system positive displacement pump.
The valves on the drain line on the back side of the pump piston were closed in one step and the pump was subsequently started in a later step. Without reopening the drain valves, potential leakage past the piston could result in a hydraulic lock and damage the pump when started.
After discussion of these findings with the applicant, the applicant performed detailed reviews of a more extensive sampling of S0Is and VLIs with similar conclusions.
Consequently, the applicant committed to review and revise as necessary prior to use under the operating license all S0Is and VLIs for systems defined as safety related in PAP-0205.
Followup inspection on this activity will be tracked as an open inspection item (440/85078-03(DRP)).
No violations of regulatory requirements or deviations from commitments were identified in this area.
14.
Fire Protection and Prevention Program (64703)
As a result of the licensing initiative directed at transferring the fire protection program from the technical specifications into the FSAR, the previously evaluated technical specification requirements will need to be re-evaluated to verify adequate incorporation of those requirements into Plant Administrative Procedure (PAP)-1923, " Actions on Inoperable Fire Protection Systems".
In conjunction with that review, the inspector will evaluate the proposed license condition, FSAR Amendment and administrative requirements to ensure that appropriate and enforceable requirements still exist in the fire protection area.
This review will be tracked as an open inspection item (440/85078-04(DRS)).
No violations of regulatory requirements or deviations from commitments were identified in this area.
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c 15.
Safety Committee Activity (40301)-
The inspector reviewed the~ minutes of the Plant Operations Review Committee (PORC) meetinos No.85-100 through 85-132 conducted during the inspection period to verify conformance with PNPP procedures and regulatory requirements.
These observations and examinations included PORC membership, quorum at PORC meetings, and PORC activities.
No violations of regulatory requirements or deviations from commitments were identified.
16. Management Meetings (30702)
Meetings were held at the NRC Region III offices in Glen Ellyn, Illinois, and licensing offices in Bethesda, Maryland, on December 3 and 17, 1985, between Messrs. J. G. Keppler and H. R. Denton and other members of the NRC Region III and licensing staffs and Mr. M. R. Edelman and other members of the applicant's staff to discuss the completion status of PNPP, Unit 1.
Major topics of discussion at those meetings included the preoperational test program, Master Deficiency List, Operations Manual, open inspection issues and system operational readiness.
A followup meeting is currently scheduled for January 21, 1985.
No violations of regulatory requirements or deviations from commitments were identified.
17.
Preoperational Test Program Implementation Verification (71302)
The inspector observed control room operation, maintenance activities and test coordination; reviewed applicable log books; and conducted discussions with control room operators, maintenance, and test personnel during the inspection period, to ensure activities were being conducted in accordance with regulatory requirements and facility procelures.
Tours of the Unit 1 reactor building, intermediate building, Unit 1 auxiliary building, control complex, and diesel generator building were conducted to observe test and maintenance work in progress, area housekeeping, equipment condition, and system cleanliness.
No violations of regulatory requirements or deviations from commitments were identified.
18.
Administrative Procedures for Operations (42450)
{
The inspector reviewed the following Plant Administrative Procedures (PAPS), Operating Administrative Procedures (OAPs) and Technical Administrative Procedures (TAPS) which contained administrative control activities directly relating to plant operation:
PAP 0110
" Shift Staffing", Revision 1 PAP 0201
" Conduct of Operations", Revision 1
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PAP 0205
" Operability of Plant System", Revision 2 PAP 0307
" Operation, Maintenance & Tasting of Fused Circuits",
Revision 0 PAP 0504
" Electrical Operating Rules & Practices", Revision 1 PAP 0508
"PNPP Operating Rules & Practices", Revision 2 PAP 0520
"FSAR & Tech Spec Changes", Revision 0 PAP-0603
" Licensee Event Reports" PAP 0604
"NRC-IE/INP0 Reports", Revision 0 PAP 0607
" Perry Plant Department Drawing Control", Revision 0 PAP 1103
" Master Deficiency Lists", Revis' ion 0 PAP 1106
" Inservice System Pressure Testing Program", Revision 2 PAP 1107
"Special Test Control", Revision 0 PAP 1113
" Surveillance Requirement Tracking", Revision 0 PAP 1403
" Control of Setpoints", Revision 1 PAP 1602
" Post Reactor Scram Evaluation", Revision 0 PAP 1703
"S'hift Reports, Logs and Records", Revision 1 OAP 0103
" Shift Relief & Turnover", Revision 1 OAP 1701
" Tracking' of Limiting Conditions for Operations",
Revision 0 OAP 1702
" Operations Sections Logs and Records", Revision 2 TAP 0101
" Duties of the Shift Technical Advisor", Revision 0 The review was performed to verify that activities covered by the procedures were prescribed in accordance with applicable facility technical specifications, Chapter 13 of the FSAR, ANSI 18.7-1976 as endorsed and supplemented by NRC Regulatory Guide 1.33 and other applicant commitments.
Operating Administrative Procedure (0AP)-1701, Revision 0, described the tracking systems to be employed to assure compliance with technical specification Limiting Conditions for Operation (LCOs).
The procedure defined " potential LCOs", in part, as plant conditions which would require entry into the applicable Action Statements upon a change in Operational Condition.
The inspector was concerned that the subject procedure did not acknowledge that for certain potential LCOs, a change
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in 0perational Condition may result _in a technical specification violation and that reliance on the provisions of the associated Action Statements may not be allowed (i.e., Technical Specification 3.0.4 may be applicable).
Applicant personnel informed the inspector that Revision 1 to the subject 0AP expanded the definition of " potential LCOs" to include those conditions which would prevent entry into another Operational Condition.
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No violations of regulatory requirements or deviations from commitments were identified.
19. Open Inspection Items Open inspection items are matters which have been discussed with the-applicant, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or applicant or both.
Open inspection items disclosed during the inspection are discussed in Paragraphs 11, 12, 13, and 14.
20.
Exit Interviews (30703)
The inspectors met with the applicant representatives denoted in Paragraph 1 throughout the inspection period and on December 27, 1985.
The inspector summarized the scope and results of the inspection and discussed the likely content of the inspection report.
The applicant did not indicate that any of the information disclosed during the inspection
,
could be considered proprietary in nature.
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