IR 05000440/1986011
| ML20211D994 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 06/06/1986 |
| From: | Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20211D963 | List: |
| References | |
| 50-440-86-11, GL-81-01, GL-81-1, IEIN-82-25, IEIN-83-83, IEIN-85-035, IEIN-85-35, NUDOCS 8606130193 | |
| Download: ML20211D994 (23) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-440/86011(DRP)
Docket No. 50-440 License No. NPF-45 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: April 9 through May 12, 1986 Inspectors: J. A. Grobe K. A. Connaughton J. W. McCormick-Barger Approved By:
RF(Da&W R. C. Knop, Chief (/f/f'4 Reactor Projects Dht4 Section IB Inspection Summary Inspection on April 9 through May 12, 1986 (Report No. 50-440/86011(DRP))
Areas Inspected: Routine unannounced inspection by resident and region based inspectors of previous inspection items, I.E. Bulletins, I.E. Circulars, Allegations, Operational Safety, Nonroutine Events, Licensee Event Reports, Onsite Review Committee Activity, Maintenance Program Implementation, and other activitie Results: Of the nine areas inspected, three violations were identified in one area (failure to comply with technical specifications concerning Plant Vent
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Radiation Monitor operability, Paragraph 7a; two examples of failure to maintain containment integrity during core alterations, Paragraphs 7b and 7c; and failure to implement the fire protection program, Paragraph 7d. While the identified violations were of minimal safety significance, this lack of safety significance was due in large part to plant operating conditions at
+ the time of their occurrence. Initial fuel loading was completed on April 24, 198 Ma'MABMdA_
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DETAILS
- 1. Persons Contacted M. R. Edelman, Vice President, Nuclear Group
- - A. Kaplan, Vice President, Nuclear Operations Division
- C. M. Shuster, Manager, Nuclear Quality Assurance Department (NQAD)
M. D. Lyster, Manager, Perry Plant Operations Department (PPOD)
- R. A. Stratman, General. Supervising Engineer, Operations Section, PPOD
- J. J. Waldron, Manager, Perry Plant Technical Department (PPTD)
- S. F. Kensicki, Technical Superintendent, PPTD
- R.'P. Jadgchew, General Supervising Engineer, Instrumentation and Controls Sntion, PPTD B. D. Walrath, General Supervising Engineer, Operational Quality Section, 3 NQAD
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- F. Silakoski, Operations Section, Senior Engineer, PP0D
- J. Takacs, General Supervisor, Maintenance Section, PP0D
- A. Russ, Compliance Engineer, PPTD <
- L. Heatherly, Operations Engineer, PPTD
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- M. W. Gayrek,' Senior Operations Coordinator, PP0D T. E. Hicks, Operations Engineer, PPTD
- *F. R. Stead, Manager, Nuclear Engineering Department (NED)
- A..G. Migas, Senior Project Engineer, I&C, (NED)
- *D. R. Green, Senior Project Engineer, (NED)
- *T. K. Boyer, Shift Supervisor, (PP00)
- R. Kanda, Jr. , General Supervising Engineer, Technical (PPTD)
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- J. R. Novak, Lead Quality Engineer, Operational Quality Section, Mechanical Maintenance Unit (MU)
- R. H. Simmons, Unit Supervisor, Operational Quality Section
- E. M. Buzzelli, Senior Licensing Engineer, (NED)
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i * Denotes those persons. attending the exit meeting held on May 12, 1986.
{ The inspectors also contacted other individuals during the inspectio i Licensee Action on Previous Inspection Items (92701, 92702)
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i (Closed) Open Item (440/85018-01(DRP)): Completion of IEEE 384 raceway separation barrier At the close of this inspection, a
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single Design Change Package (DCP) remained to be issued and
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. implemente This DCP (No. 86-0452) was issued and implemented
- subsequent to the close of this inspection period but prior to the writing of this report on May 16, 198 .
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I (Closed) Open Item (440/85057-01(DRS)): Procedure No. PAP-1107 did not reference Procedure No. PAP-0507.for preparation of test I instructions. -The inspector reviewed Temporary Change (TC)-001 to Revision 0 of Procedure-No. PAP-1107, "Special Test Control."
This procedure change required special test instructions be
'. prepared in accordance with Procedure No. PAP-0507, " Preparation,
. Review, Approval, Revision, and Cancellation of Instructions."
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The format for special test instructions will be that utilized for other test instructions (e.g. surveillance test instructions).
Procedural content will be that specified in the Perry Nuclear Power Plant Quality Assurance Pla c. (Closed) Open Item (440/85059-05(DRP)): Performance of slow bleeddown of instrument air pressure to verify check valve integrity. NRC Information Notices (IENs) 82-25 and 85-35 discussed certain safety related air system check valves which failed to seat upon a slow loss of instrument ai The inability of the check valves to maintain system pressure under these conditions was not identified during preoperational testing at other sites because such testing was not performed as required by NRC Regulatory Guide 1.80, "Preoperational Testing of Instrument Air Systems," Regulatory Position The inspector verified that the licensee was committed to Regulatory Guide 1.80 and, as such, had incorporated slow loss of air pressure testing as part of preoperational test procedures for the safety related and non-safety related instrument air systems. Further, the inspector determined that the check valves utilized in the safety related instrument air system were Borg-Warner, 1" swing-check valves. This make and model of valve was not involved in the failures reported in IENs 82-25 and 85-3 Tne licensee did not specifically intend to periodically perform slow bleeddown testing of the safety related instrument air system in response to the subject IENs. Surveillance test procedure P57-T2002, " Safety-Related Instrument Air Check Valve Operability Test," did, however, require that when upstream instrument air pressure is vented to verify check valve integrity, it is to be bled off in a controlled manner by throttling the vent isolation valv This is necessary for personnel safety since nominal system pressure is approximately 2500 psig. Based upon the preoperational testing to be performed, the type of check valves utilized, and the method to be employed for periodic leak checking the valves, the inspector found that the licensee had adequately addressed the issues raised by IENs 82-25 and 85-35 for the Perry, Unit 1 safety-related instrument air system. Surveillance test procedure P57-T2002 is,
! therefore, considered adequate.
f d. (Closed) Unresolved Item (440/85084-01(DRS)): Coatings test to be performed by Oak Ridge National Laboratory (ORNL) to resolve
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questionable application. The licensee's evaluation of test results received from ORNL in late December '.985 concluded that the environmental qualification of Carboline 191 HB coating when i
applied to bare steel, was not supported. Based upon this conclusion,
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the licensee performed field inspections to ascertain the surface area covered by Carboline 191 HB coating applied in this manne Once obtained, this surface area was added to previously identified
, unqualified coatings in containment. The total was found to exceed the analyzed quantity of unqualified coating reported in Section
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6.1.2 of the Perry FSAR as 5,500 sq. ft. The licensee performed an evaluation of the increased surface area of unqualified coatings identified by the ORNL test results and field inspections and concluded that the quantity of unqualified coatings utilized in containment remained acceptabl A proposed FSAR change reporting the increased quantity of unqualified coatings was transmitted by the licensee to the NRC Office of Nuclear Reactor Regulation by l letter dated March 13, 1986. A preliminary review by the NRC staff j found the proposed change acceptable for low power licensing. The NRC staff documented these findings in a letter from W. R. Butler to M. R. Edelman, dated March 18, 1986. The NRC staff review will be completed and documented in a future supplement to the Perry SER (NUREG-0887) as necessary to support full power licensing of Perry, Unit e. (0 pen) Violation (440/86006-01(DRP)): System Operating Instruction deficiencies. Prior to fuel load, detailed NRC reviews of Perry Plant System Operating Instructions (S0Is) identified .iignificant technical errors as documented in Inspection Report N /86006(DRP). As a result of the inspectors' findings, the licensee committed to perform a detailed technical review of all S0Is for systems defined as safety-related in PAP-0205, Revision 3,
" Operability of Plant Systems," prior to use under the operating license. This commitment was documented in a letter from Mr. Murray R. Edelman, Vice President, Nuclear Group, CEI, to Mr. James G. Keppler, Regional Administrator, Region III, NRC, dated March 4, 1986. To assure that the S0Is required to have a detailed technical review prior to initial criticality were adequate, the inspectors performed the following detailed review of a sample of the S0Is included in this group. This review was performed using the licensee's Operations Administrative Procedure (OAP)-0502, Revision 0, " Preparation of System Operating Instructions," and the applicable Piping and Instrumentation Diagrams (P& ids).
501-P86, Revision 4 " Nitrogen Supply System (Unit 1)"
f S01-P50, Revision 3 " Containment Vessel Chilled Water
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System (Unit 1)"
In addition, a non-detailed review of the following 50I1 was performed. This inspection included a step by step review to assure a logical progression, but did not include comparing the instruction with controlled drawings or other reference material.
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l 50I-C22, Revision 2 " Redundant Reactivity Control System l (Unit 1)"
l l 50I-R44/E228, Revision 3 " Division 3 Diesel Generator Starting Air System (Unit 1)"
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The inspectors' review of the above S0Is identified no significant technical errors. At the close of this inspection period, all S0Is for initial criticality had not been reviewed and revise However, prior to issuance of this report, NRC inspectors verified that all 32 S0Is required for the initial criticality milestone had been completed. Additional inspector reviews of licensee actions relative to 50I quality will be tracked by this ite (Closed) Violation (440/86006-02(DRP)): Failure to adequately specify and implement housekeeping / cleanliness controls. The inspector verified by review of Plant Administrative Procedure (PAP)-0204, Revision 2, and numerous plant tours conducted subsequent to February 10, 1986, that the licensee had taken corrective actions to remove the inspector-identified debris from the suppression pool and had established and implemented interim and permanent administrative controls as described in the licensee's April 23, 1986 response letter. Specifically, the inspector observed that the licensee had established the containment and drywell as ANSI N45.2.3 cleanliness zone 3 areas. Protective coverings were visually observed to be in place to prevent incursion of debris into the suppression pool. Personnel and material accountability were established as require With regard to equipment tagging, the inspector verified during plant tours that equipment was properly tagged and that the expiration dates for tagged equipment in the plant had not expire The inspector verified by review of training session attendance lists that appropriate licensee personnel had received the training specified in the licensee's response lette Inspector observations of plant housekeeping / cleanliness during this inspection indicated that the foregoing licensee actions have, thus far, been effective.
l I g. (Closed) Unresolved Item (440/86006-04(DRP)): Review and evaluations of NRC Generic Letters (GLs) not requiring licensee action. As a result of inspector concerns identified with documentation of licensee reviews of selected GLs, the licensee performed a re-review
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of actions taken with respect to all GLs beginning with GL 81-01.
l This documented re-review determined that the applicable issues I
raised by the GLs had been ultimately addressed either in response to the GLs or subsequent NRC licensing reviews against NRC regulatory requirements. Additionally, by procedural revision to Project Administration (PA)-1602, " Evaluation of NRC/INP0 Documents,"
the licensee established a more formal and clearly defined program for the receipt, distribution and evaluation of GLs, including requirements for documenting evaluations of GLs not requiring action by the licensee. The inspector reviewed the licensee's evaluation of previous GLs and the procedural change to PA 1602 and found them acceptabl "
l h. (0 pen) Open Item (440/86008-01(DRP)): Control rod position indication anomalies. Following a reactor scram on March 31, 1986, which resulted from spurious upscale trips of the Intermediate Range Monitoring (IRM) system it was observed that the full core display did not update position indication for a rod selected at the time I of the scra Further, full in-2 (full in-overtravel) indication was not received for any rod. When the scram was reset, full in indication was restored for all rod Licensee investigation into this matter disclosed that two printed circuit cards in the Rod Control and Information System (RC&IS)
referred to as Probe Data Receive Cards were of an earlier revision than that required by the Perry BWR 6 design. The cards receive combined and multiplexed control rod identity and position information for all rods. The cards serve to partially demultiplex the incoming signals and separately output multiplexed rod identity and multiplexed rod position signals. These signals are then sent to demultiplexers for further processing and use by the RC&I With the earlier revision card installed, in the BWR 6 design, a full in-2 position signal from the rod position indicating probes caused the cards to inhibit any output signals. In earlier designs such a signal would have been acknowledged by the system as invalid (earlier designs did not have full in-2 rod position indication) and
" Response Fail" annunciation would have been generated. With the proper revision card installed in the BWR 6 design, the full in-2 position signal is recognized and processed to provide, among other things, full in-2 rod position indication on the full core displa Further, the Probe Data Receive card will continue to process and update rod position data with a full in-2 rod position signal presen Following the licensee's investigation, the correct revision cards were procured and installed. The inspector will further review licensee actions to confirm RC&IS performance and will attempt to determine how the incorrect revision of the cards were supplie i. (0 pen) Open Inspection Item (440/86008-02(ORP): Preventive maintenance program deficiencies. Inspector's followup of licensee acticns on this item are discussed in Section 10 of this repor j. (0 pen) Violation (440/86008-04(DRP)): Failure to provide controls necessary to establish the operating status of safety related instruments. The inspector reviewed the status of licensee actions taken in accordance with Special Project Plan 1401, " Instrument Valve Lineup Verification," and the licensee's letter from M. R. Edelman to J. G. Keppler, dated March 14, 1986, both of which were incorporated into Attachment 1 to the Perry, Unit 1 operating license. As of the close of this inspection, the licensee reported that more than 90% of the instrument valve verification walkdowns and interim as-built drawings required prior to initial criticality (0perational Condition 2) were
complet The inspector will verify during a future inspection prior to initial criticality, that the actions required for that milestone have been complete . Inspection and Enforcement Bulletin (IEB) Followup (92703)
(0 pen) IE Bulletin (440/79018-BB): " Audibility Problems Encountered on Evacuation of Personnel From High-Noise Areas." By letter dated May 9, 1986 from M. R. Edelman to J. G. Keppler, the licensee submitted a description of the program to address the issues identified by the subject IEB. The submittal included an implementation schedule for interim and permanent corrective actions identified as necessary by noise level surveys to be conducted during plant startup and power ascension phases. Review of this submittal by NRC resident and cognizant regional office-based inspection personnel has determined that it satisfied the requirement contained in Item B.2 of Attachment 1 to the Perry, Unit 1 operating license relative to the subject IEB. The subject IEB will remain open pending review of licensee implementation of the program described in the May 9, 1986 lette . Inspection and Enforcement Circular (IEC) Followup (92703)
(Closed) IE Circular (440/80009-CC): " Problems With Plant Internal Communications Systems." This item was previously reviewed during an inspection documented in NRC inspection report (440/85080). The inspector found that the licensee had not taken action with regard to additional information and recommendations on this same issue contained in NRC Information Notice (IEN) 83-83, "Use of Portable Radio Transmitters Inside Nuclear Power Plants."
Following that inspection, the licensee has conducted an additional review of Unit 1 and common areas and equipment to identify those areas in which the use of portable radios should be prohibited. This and previous efforts had identified a total of 11 areas and approximately 32 equipment panels / racks containing equipment susceptible to radio frequency interference (RFI). Subsequently, the identified areas and/or equipment were suitably posted with signs which clearly convey the prohibitions on portable radio use. The inspector independently verified by visual observations during plant tours that affected areas and equipment were, in fact, posted as described in the licensee's response file for the subject IEC. These additional actions were based, in part, upon specific consideration of the examples cited in IEN 83-3 . Followup on Allegations (99014) (Closed) Allegation (AMS-RIII-86-0037): The NRC received two
, concerns on March 6, 198 These items were labeled concerns by the individual who called because the individual did not know if the rework relating to the alleged discrepancies had been completed.
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- 1 The first item concerned mechanical penetrations located in either the containment or drywell where insulation material between piping and penetration sleeves was faulty. Reportably, the mechanical penetration sleeves were approximately 3" in diameter and the piping 2-7/8" in diameter. The individual stated that the insulation had to be dug out and this was very difficult because of the small annular ring that was filled with the insulation materia The individual indicated that approximately 2000 of these penetrations had to be reworked and the individual doesn't believe that they were all fixe The second item concerns small piping approximately 2" in diameter which is in clusters of 50 to 60 pipes at a penetration through a wall. The wrong type of caulking material was allegedly used to seal around these piping clusters. Again, the material had to be dug out and "it was like digging concrete with a knife." The individual believed that this was Control Rod Drive System pipin The individual indicated that 10% of this work had been done prior to the alleger leaving the sit The alleger reportably left the site in April 198 This allegation was transmitted to the licensee for their review in a letter to Mr. Murray R. Edelman, Vice President, CEI, from Mr. Charles E. Norelius, Director, Division of Reactor Projects, NRC, dated March 26, 1986. The licensee responded to this allegation in a letter from Mr. Edelman to Mr. Norelius, dated April 30, 1986. The following is the result of the licensee's review of the two concerns raised by the individua Licensee Investigative Results for Allegation No. 1
"Due to the lack of specific details provided in this concern, it was difficult to pinpoint the exact nature of the allegatio However, the magnitude of the alleged problem is questionable since only 319 mechanical penetrations exist in the entire Reactor i Building. Of this number, 13 are 3" in diameter and contain insulated pipes. With the information provided, a review was made by the Nuclear Construction Engineering Section of the Reactor Building penetrations. The results indicated that the only condition that may be related was the standard practice of removing l insulation prior to installing a penetration seal. This was done l
primarily because the silicone sealants could not effectively seal against the fiberglass insulation. In some cases, such as those penetrations requiring radiation seals, it was necessary to remove all the existing insulation within the opening. Because some of the annular gaps between the pipe and sleeve were small (min.1/2"),
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insulation removal was somewhat tedious. This condition affected i less than 50 penetrations of various diameters ranging from 3.00"
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to 48.00" and all work has been completed and inspected by Quality
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Licensee Investigative Results for Allegation No. 2
"A review of the Penetration Seal Schedule for the Reactor Building was performe This review revealed that there are four penetrations which each contain from 75 to 105 runs of 1" and 1-1/4" diameter piping of the Control Rod Drive System, (Cll). The penetrations are PRB-3045, PRB-3047, PRB-3051, and PRB-305 A review of documentation for these penetrations revealed that they were all inspected and accepted by both the contractor's QC and the CEI quality inspector. None had any nonconformances written against i Caulking material was used around these pipes to temporarily seal the interface between the pipes and a steel plate at the outside face of these penetrations. It was used to prevent leakage of sealant while still in its liquid stat The temporary caulking was verified by final inspection as having been removed after the seal was installed. Some permanent caulking (DC-790) was also used, documented and left in plac An inspection was performed by an engineer from the Nuclear Construction Engineering Section and the Responsible Quality Engineer on April 4, 1986. The result of this inspection supported the previous inspection which revealed that no caulking remained around these pipes except for some DC-790 caulking. The permanent DC-790 caulking was documented as such on the original installation documentation."
The licensee concluded that they were unable to produce any indications of deficient penetrations or caulking materials and, therefore, they were unable to substantiate the concern NRC Followup The inspector reviewed the licensee's response and met with licensee personnel that were responsible for overseeing penetration sealing construction activities and reviewing the individual's concerns.
. The inspector also performed a sample field check of containment
! penetrations including the Control Rod Drive piping penetrations and found no apparent deficient sealing configurations associated with penetration Based on the licensee's detailed review and the inspector's followup, these concerns were not substantiated and are considered closed.
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b. (Closed) Allegation (AMS-RIII-A-86-0045): Electrician at Perry site worked while under the influence of narcotic The NRC was informed that an individual, formerly employed as an l electrician for L. K. Comstock and Co. , Inc. at the Perry site, had l been indicted by law enforcement authorities for illegal possession and use of narcotics. The indicted individual apparently admitted to having been under the influence of narcotics while working at Perry from January 3, 1983 through January 198 _
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I At the NRC's request, the licensee conducted a review to establish the subject individual's dates of employment, and the nature of work activities in which the individual participate This effort established by document review that the individual was employed frcm August 23, 1978 through June 20, 1981 and again from August 17, 1981 through January 24, 1986. During the latter term of enployment, the individual took three extended leaves of absence; fron May 10, 1982 through August 28, 1982, February 27, 1984 through October 1, 1984, and April 25, 1985 through January 24, 1986, his termination dat Through interviews with Comstock supervisory personnel present at the Perry feite for the duration of the subject individual's employment, it was determined that the individual had been, for most of his employment period, assigned te a group which installed, moved, and removed temporary power and lighting equipment in support of ongoing construction activities. The individual was, however, believed to have been assigned to a cable pulling crew for approximately 1 to 1-1/2 years in the 1979-1981 time frame. The individual's training file indicated that he had not received formal procedure training for safety related work. Training to Comstock safety related cable pulling procedures was required for individuals performing this activity at the time the subject individual was employe In order to substantiate the foregoing information, the inspector contacted the subject individual by telephone and inquired directly as to what types of work activities the individual had participated in. The individual corroborated the recollections of Comstock personnel but added that he had also been assigned during the 1978-1979 time frame to a pipe gang which place PVC conduit for cable duct banks embedded in concrete. He did not believe that he had been involved in safety related cable pulling activities and did not recall ever attending any formal training session Based upon the foregoing information covering activities performed by the subject individual and the time frames in which they were performed the inspector has determined the following:
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- Temporary power and lighting work and equipment is not l safety-related and as such would not be expected to adversely l impact public health and safety. Further NRC action regarding j this work activity is thus not warrante * Cable pulling activities during the 1979-1981 time frame were predominantly non-safety related. In 1979, no safety related
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cable was pulled. In 1980 less than 1% of the cable pulled
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was safety related. In 1981, only about 3% of the cable
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pulled was safety related. If the individual had participated in safety related cable pulling, the activity was subject to;
in process quality control inspection by Comstock QC, and over-inspection by the licensee's quality assurance organiza-tion. Further assurance of the adequacy of such activities l
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were provided by post installation testing (cable meggaring)
and system precperational testin The inspector's review concluded that the individual was in all likelihood not involved in safety related cable pullin Further, if some isolated activities during the time frame in questicn had involved safety related cable pulls, these activities would have been subjected to routine quality control inspections and overinspection scrutiny. Construction installation /precperational testing confirm the suitability of the work performed. Based on the above, no further action by the NRC is warrante _
PVC pipe was used as part of the form for embedded conduit duct bank installations. After placement of the PVC pipe by production personnel, Comstock quality control inspection personnel verified by direct inspection prior to concrete placement that the correct number of PVC pipes were installed, aligned laterally and longitudinally, and that appropriate spacing between the ducts and reinforcing steel was established. Licensee quality control personnel performed overview inspections for each placement covering the foregoing quality attributes on a sampling basis. The craft activity of PVC duct form placement was a simple task with considerable independent overview prior to each concrete placement. Were the subject individual to have been involved in safety related cable duct form placement, the possibility of a condition adverse to safety having been created and then having gone undetected is considered to be so remote that no further action by the NRC is warrante (Closed) Allegation (AMS-RIII-A-86-048): Allegation regarding cold water piping insulation inside containmen Allegations were received from two individuals on March 13 and 14, 1986, at the PNPP resident inspectors office and the Region III
, office, regarding the procedures and adhesive used to install antisweat insulation on chilled water piping inside containmen One individual also contacted the " Call for Quality" program. The individuals alleged that:
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Armstrong 520 adhesive was used to install the insulation instead of the procedurally prescribed adhesive,
- Insulation material was cut too small and stretched to fit prior to adhesive set up,
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Adhesive was applied to the piping instead of the insulation butt joint as procedurally prescribed, and
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Materials were authorized by persons signing other individuals name , _
The inspector reviewed these allegations and reviewed the licensee's
" Call for Quality" response to the allegaticos. The first consideration in reviewing this information is that the insulation was installed on the containment vessel chilled water system, a non-safety related system, and the insulation does not directly serve a safety related functio The only potential safety impact that the insulation could have would be potentially affecting the operability of the emergency sore cooling system (ECCS) by clogging the suction strainers in the suppression pool and starving the pump suction if the insulation material became dislodged. This potential impact is discussed in a draft revision to Sections 6.1.2 and 6.2.2.2 of the PNPP FSAR submitted to the NRC by letter dated March 13, 1986, from M. R. Edelman (CEI) to W. R. Butler (NRC).
The preliminary review and acceptance for lcw power licensing of PNPP Unit 1 is documented in a letter dated March 18, 1986, from W. R. Butler (NRC) to M. R. Edelman (CEI). The revised FSAR Sections described in the licensee's March 13, 1986 letter documents the acceptability and continued operability of the ECCS during a design basis accident with the dislodgement of all of the antisweat insulation material inside containmen Regarding the individual alleged activities, the licensee substantiated that Armstrong 520 adhesive was used instead of the specified adhesive, Thermo-Cell 950. Subsequent to the investigation, a nonsafety Noncomformance Report (NR) No. CQCN 287 was issued to document the deviation. Engineering evaluation and closecut of the NR documented that the only substantive difference between the adhesives is the evaporative carrier which does not affect the composition and strength of the final bond. A "Use as is" conclusion was reached. In addition, NR closeout activities included retraining the responsible individuals in the governing practices and procedures for material acquisition and contro The allegation regarding the improper insulation fit up could not be substantiated. The licensee performed visual inspections of accessible insulation, Inspection Report No. R86-1542, and did not
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alleged activity. Based on the safety analysis of the consequences of lack of insulation adhesion, no further inspection activities are warranted regarding this allegatio The allegation regarding the application of adhesive onto the pipe l was not substantiated. While it could not be determined if adhesive
! was applied to the pipe, the installation instruction, MIS-850436-1, l Revision 0, " System IP50 Antisweat Insulation Inside Containment,"
does not prohibit the practice, and requires adhesive be applied to the pipe at irregular insulation sections. The piping is carbon steel and the adhesive does not contain deleterious materials for l
carbon steel.
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substantiated by the licensee. Review of stores requisitions,
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l chemical permits, chemical requisition cards, and field return forms by the licensee disclosed no discrepancies by the individuals identified as controlling the materials. The inspector discussed the situation with the Site Protection Supervisor and was informed that guards at containment access had been and would again be instructed to ensure that materials brought into containment are .
signed in by the responsible individua While certain aspects of these allegations were substantiated and certain aspects remain unresolved, no further concerns exist and no additional actions will be pursued based on the lack of safety significance of the allegations and the adequacy of the licensee's review and corrective action . Operational Safety Verification (71707)
] The inspectors observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during this inspection period. The inspectors verified the operability of selected emergency systems, reviewed tag-out records and verified tracking of Limiting Conditions for Operation associated with affected components.
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Tours of the intermediate, auxiliary, reactor, and turbine buildings were conducted to observe plant equipment conditions including potential ;
fire hazards, fluid leaks, and excessive vibrations, and to verify that maintenance requests had been initiated for certain pieces of equipment in need of maintenance. The inspectors by observation and direct interview verified that the physical security plan was being implemented in accordance with the station security pla The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection control These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR, and administrative procedure Initial core loading was completed on April 24, 1986. Fuel handling equipment power supply problems experienced during the previous '
inspection period did not recur. While the IRM noise problems did recur during this inspection period, the contribution to fuel load delays was minima The inspectors have noted through control room observation and review of work schedules that as fuel load activities have progressed towards completion and as the Perry project organizations focus has increasingly shifted towards achieving initial criticality, the level of work activity requiring the cognizance and support of operating personnel has increased substantially over that during the first several weeks of fuel loading activities. The increased demand placed upon operating personnel may have contributed to the personnel errors that resulted in the events discussed in Paragraphs 7a, 7b, and 7d of this report.
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7. Onsite Followup of Events at Operating Reactors (93702) On April 13, 1986, at approximately 2:15 p.m., the Unit 1 and '
Unit 2 Plant Vent Radiation Monitor Sample Analysis System blowers were taken out of service by tag-out of a common supply breaker to permit preventive maintenance on the Unit 2 Plant Vent Radiation Monitor Sample Analysis System blower. At approximately 5:45 the same day the Sample Analysis System blowers were returned to l service. When the Unit Supervisor on the following shift reviewed i
the removed tags for closure of the tag-out, he determined that the tag-out had rendered the Unit 1 and Unit 2 Plant Vent Radiation Monitors inoperable and that applicable actions required by technical specification Limiting Condition for Operation (LCO)
3.3.7.10 had not been initiated. The requirement for immediate suspension of drywell/ containment purge with the Unit 1 Vent '
Radiation Monitor inoperable for reasons other than a nonconservative
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setpoint was violate Followup by the inspectors consisting of interviews of operating management personnel and reviews of the job traveller for the preventive maintenance task and Annunciator Response Instructions (ARIs) associated with the plant vent radiation monitors disclosed a number of personnel errors, including failures to follow administrative controls, which resulted in this even When the job traveller covering the preventive maintenance task was submitted for approval, operating personnel apparently misunderstood which sample blower associated with the Unit 2 Vent Radiation Monitor was to be worked on. The word description of the sample blower indicated that it was a Roots Rotary blowe By training, operating personnel believed this word description referred to the sample blower associated with the isokinetic probe which draws a sample off the Unit 2 plant vent. The Master Parts List (MPL) ide~ntification number, also contained on the traveller, indicated instead, that the Sample Analysis System blower associated with the Unit 2 vent radiation monitor was to be worked on. Though an apparent conflict existed between the MPL identification number
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and the word description of the item to be worked on, operating personnel assumed that the MPL identification number was incorrect and that the sample blower associated with the Unit 2 plant vent isokinetic probe was to be worked o In addition to the above described misunderstanding, operating
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personnel apparently did not recognize that the tag-out of the
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supply breaker for the Unit 2 sample blower would also result in a loss of power to the corresponding Unit 1 sample blowe Training of operating personnel on the Plant Vent Radiation Monitor included a technical position that rendering an isokinetic probe sample blower inoperable did not result in Plant Vent Radiation Monitor inoperability. The isokinetic probe sample blower is necessary to assure an isokinetic and autokinetic sample is drawn
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from the plant vent. The validity of the position taken regarding vent radiation monitor operability is an open item
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(440/86011-01(DRP)).
When the common supply breaker for the Unit 1 and Unit 2 Plant Vent Radiation Monitor Sample Analysis System blowers was opened to effect the tag-out, operating personnel received a control room i annunciator that should have alerted them to the fact that the l Sample Analysis System blowers were taken out of service and not the l
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isokinetic probe sample blowers. Based upon the earlier erroneous assumption that the isokinetic probe sample blowers would be taken out of service, operating personnel silenced the annunciator and took no further action. The Annunciator Response Instruction associated with the annunciator required that operating personnel initiate actions in order to comply with the applicable technical specification LC Operating personnel did not refer to that instructio As a result of the failure to resolve the apparent conflict in equipment identification on the job traveller, the erroneous assumption regarding the identity of the sample blower to be taken out of service, the failure to recognize that the Unit 1 vent monitor would be affected and, subsequent failure to follow annunciator response instructions, technical specification Limiting Condition for Operation 3.3.7.10 was exceeded. This is a violation (440/86011-02(DRP)).
b. On April 23, 1986, at approximately 10:50 a.m. during surveillance testing of the A Standby Liquid Control (SLC) System, reactor water cleanup system supply valve 1G33-F004 failed to shut as required when the A SLC pump control switch was placed in the start positio After an unsuccessful attempt to remote manually close the valve, licensee personnel declared the valve inoperable and immediately shut and de-energized redundant containment isolation valve 1G33-F00 Inspector followup of this event which included interviews with licensee operating and technical department personnel disclosed that valve IG33-F004 had been rendered inoperable on April 14, 1986 when an erroneous and improperly processed modification to its control circuitry was acco:nplished. This modification was effected by a Work Request which had been written to resolve a discrepancy identified between the as-built control circuitry and an electrical design drawing which was being verified by as-built inspection. The design drawing had been obtained by the individual performing the as-built inspection significantly prior to the walkdow A design change reflected in the current controlled drawing had been incorporated into the control circuitry, but was not reflected on the drawing used for the as-built inspection. During the as-built inspection, the individual concluded that the field installed circuitry was wrong and should be modified because it did not agree with his drawing. That out-of-date design drawing was then
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I incorporated into the work package for the subject modification and was not recognized to be out of date by work planners preparing the work instructions. The individual initiating the work request and providing the erroneous drawing also performed the engineering review of the work package prior to its implementation. Again, the fact that the design drawing incorporated into the package was out-of-date was not recognized. Operating personnel authorized commencement of work with the understanding that the work package was for the removal of " spare" wiring and thus would have no effect on valve operability. Also no one in the review and approval cycle for the work order recognized that an improper administrative procedure was being used to verify an as-built drawing discrepanc Since valve 1G33-F004 was not recognized as inoperable following the modification on April 14, 1986, operating personnel did not require suspension of core alterations or the closure and de-energization of redundant automatic containment isolation valve 1G33-F001 as required by technical specification 3.6.4 Action Statement a. Core alterations were performed subsequent to the subject modification on April 14, 1986, and prior to discovery on April 23, 1986. Failure to comply with technical specification 3.6.4 is a violation (440/86011-03a(DRP)).
c. On April 30, 1986 at approximately 11:15 p.m. during the performance of Surveillance Instruction (SVI) T23-T1201, " Containment and Drywell Isolation Verification," licensee operating personnel discovered that redundant manual containment isolation valves IP54-F726 and IP54-F728 on a fire protection water supply line were both open thus violating Primary Containment Integrity. Immediate corrective action by operating personnel was to close the subject valve Inspector followup of this event included interviews with operating management personnel, review of SVI T23-T1201, review of Condition Report 86-408, and review of information provided by licensee personnel summarizing the documented history of these valves positions beginning prior to initial commencement of core alterations and ending with licensee discovery and corrective action on April 30, 198 The followup disclosed that the subject valves were documented to have been left open following the performance of Special Test GEN-M-0039 on March 12, 1986. Subsequently, on March 23, 1986, Periodic Test Instruction (PTI) P54-0033, " Fire Suppression Systems Valve Position Verification," was performed and documented the position of valve 1P54-F726 as open. On March 30, 1986, SVI T23-T1201 wrs performed and both valves were found open. Valve 1P54-F727 was closed by operating personnel the same day. Again on April 12, 1986, PTI-54-0033 was performed and valve 1P54-F726 was documented as open. On April 29, 1986, SVI T23-1201 was again performed and documented that both valves were documented as ope . -- . . . .- . . . . - . ~ - .~.- -. - .~. - ..
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During review of the SVI, it was recognized that at least one of the valves is required to be closed per Technical Specification i
. 3.6.1.1.2 for Primary Containment Integrity during core alter.g .
As previously stated, valve IP54 F727 was closed on April 30, 198 T h Each time the valve positions were checked and documented, the governing procedures required that each of the subject valves be i locked closed. SVI T23-T1201 indicated that the locked closed position was a required acceptance criteria per Technica Specifications. These discrepancies were noted in the completed ,
- procedures.by documented remarks which stated "...per P54 SOI" (Systes Operating Instruction) or "...per U.S." (Unit Supervisor).
- -Apparently on all of the above occasions, personnel involved in the review and approval of the test results failed to recognize that the valves were required to be closed in order to comply with technical specifications during core alterations. Core alterations did take
, place while both of the valves were open between March 12 and March 30, 1986. Both valvas were also open for an undetermined period of ,
time subsequent to March 30, 1986 and prior to April 30, 1986, Core l j alterations were also performed between these two dates. Failure to establish Primary Containment Integrity during core alterations is a
- violation (440/86011-03b(DRP)). - On May 7, 1986, while performing a tag-out on the Control Complex l and Diesel Generator Building CD, System, licensee personnel
i discovered incorrectly closed vaTve P54-F5622 on the pilot valve CO i
, supply header. Solenoid-operated pilot valves were utilized in the 2 '
system to admit CO, from this header to the main and selector
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control valve actuators. The main and selector control valves, in '
!. turn,-supply CO, to areas served by the system for fire suppressio The misposition8d valve, therefore, rendered the system inoperable i
in that automatic CO,, fire suppression capability for the Division i 1, 2, and 3 diesel g8nerator rooms was lost. Additionally, manual CO, fire suppression capability for the 4 KV switchgear rooms would .
have required additional operator actions to provide a CO 2 supply to !
the installed hose reel ,
Inspector followup of this event included interviews with licensee operating department management personnel, reviews of system tag-out records and fire impairment log entries, CO, system operating instructions (SOIs), and a walkdown of affetted portions of the ;
system and the areas protected by those portion f
I These followup activities disclosed that on April 7, 1986, Tag-Out
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No. 16-1242 was hung which closed valves P54-F1086, F-1087, and F-5622. Compensatory measures were established for affected areas
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in accordance with the licensee's fire protection progra i Subsequently, on April 12, 1986, this tag-out was cleared. Two of the valves, P54-F1086 and F1087 were part of a second tag-out and, as such, were required to remain closed pending clearance of this
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- second tag-out. Valve P54-F5622 was not part of the second tag-out
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the operator clearing the first tag-out' reasoned in some manner or another that since two of the three valves were required closed by the second tag-out, the third valve should also remain close ,
This was noted by the operator on the tag-out sheet and reportedly communicated through supervisory operating personnel to the Unit Superviso Further action was not taken to open the valve or include.it under the second tag-out. Subsequently, the second tag-out was cleared on April 26, 1986. Because valve P54-F5622 was not covered by the second tag-out it was not restored to its required open positio ,
Between April 26, 1986 and the time of discovery on May 7, 1986, -
compensatory measures required by the licensee's fire protection program were not implemente These required compensatory measures included hourly fire watches of the diesel generator rooms and deployment of CO, fire extinguishers within 15 ft. of the inoperable hose reels. This failure to implement the licensee's approved fire i protection program is a violation (440/86011-04(DRP)).
e. On April 22, 1986, during initial fuel foad activities, APRM D experienced a 148 millisecond (MS) spurious spike causing a high flux (15 percent reactor power setpoint) scram initiation signal from the APRM D circuitry. At the time of the event, fuel was being moved and the shorting links in the reactor protection system (RPS) 6 were removed as required allowing non-coincident scrams from the .
SRMs, IRMs and APRM The result of the scram initiation signal from APRM D was a half-scram on RPS D. The operators believed that .
they should have received a full scram due to the non-coincidence circuitry and initiated a manual scram by placing the mode switch in shutdow The half scram from APRM D through RPS D required two relays to drop out and the half-scram on RPS C (non-coincidence through the manual scram circuitry) required four relays to drop out. Licensee evaluation, therefore, concluded that due to the short duration of the initiation signal, the non-coincidence t i circuitry did not have sufficient time to actuate the full scra ;
i After licensee management was apprised of this finding, direction :
, was provided to perform a functional test to verify that the j relaying for the non-coincident trip of RPS D would occur following a simulated APRM D channel trip. The logic was observed to function
.' properly though time response data was not obtained or evaluate ! After determining the functional test to be satisfactory, licensee management personnel authorized recommencement of fuel loadin r The following morning the resident inspectors met with licensee personnel to review RPS logic electrical diagrams to verify that the
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identified differences between the RPS D and C non-coincident logic paths were as stated by the licensee. Details of the evaluation j
"; performed by the licensee were not immediately available for inspector review. The inspectors, therefore, expressed concern to the licensee that merely performing a functional test was not
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sufficient _to demonstrate that RPS logic components (ind.ividual
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i relays) were performing as intended by component riesign r l- specifications and that the observed behavior 'of the RPS during i the event was not indicative of incipient component failure. .
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Later that morning, the inspectors were provided documented details . , of the evaluation performed by licensee personnel. The evaluation '
was incisive enough to identify a single relay (KISC) as having a longer than expected dropout time. The evaluation, however, did
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not utilize all available data from the event to bound the suspect i s
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relay's dropout' time and thereby draw any conclusion regarding the need to conduct further evaluation to determine whether or not K15C .
relay performance was, or could be expected to remain, acceptabl An evaluation by the resident inspectors in conjunction with an NRC
! regional office-based inspector dispatched to the site indicated '
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that the dropout time of the K15C relay could have been anywhere I
from approximately 90 to 150 msec. The manufacturer's specification ,
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for the relay indicated a maximum dropout time of 85 msec. The plant Technical Specifications do not contain response times for non-coincident trip of the RP '
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l Further review of this matter by the regional office-based inspector
, and subsequent licensee actions to address this matter will be .
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documented in NRC Inspection Report (440/86012(DRS)).
4 On May 4, 1986, startup test personnel in Unit 1 containment noted .
a missing holddown bolt on control rod drive hydraulic control unit
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(HCU) 34-27 and loose holddown bolts on four other HCOs. Ip i
response to this observation a Work Request was initiated tb have ,
j the remaining HCUs inspected for loose or missing bolts. The HCUs r i were subsequently declared inoperable on May 15, 1986.-
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! Inspector followup of this event included discussions with licensee operating and technical department management personnel, review of corrective action documentation, control room logs and LCO tracking
- documentation. The inspectors review disclosed that while the l licensee had not initially considered whether or not control rod drive scram accumulator operability was impaired, core alterations i had been suspended prior to the time of discovery on May 4, 1986 and May 15, 1986. Technical Specification LCOs specifically associated
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I with HCUs (scram accumulators) were, therefore, not violated over
this time fram The inspector expressed concern to the licensee that operating l
personnel be sensitive to the fact that operability of equipment
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is not only contingent upon meeting technical specification !
[ surveillance test requirements but also relies upon all design,
- manufacturing, and installation quality attributes necessary,to ;
, assure that the equipment is capable of performing its intended
! function. The adequacy of licensee actions to assure that
! identified deficiencies in any such attribu'tes are car.efully
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evaluated in a timely manner for impact on equipment operability is an unresolved itam (440/86011-05(DRP)).
The inspector was informed that initial investigation and evaluation by the licensee in consultation with General Electric Company (G.E.)
determined that quantified torque valves had been previously established by G.E. as necessary to support seismic qualification
', of the HCOs but that installation of the HCU holddown bolting at Perry was accomplished without quantified torquing requirement Instead, work proc 6dures called for the bolts to be " snug tight" or with the head of the bolts or the nuts making " firm contact" with mounting surfaces. Followup inspection to review the details surrounding this event will be tracked as an open item (440/86011-06(DRP)).
8. Licensee Event Reoorts FolloWop (90712)
Through direct observations, discussfor,s with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was acccmplished, cad corrective action to prevent recurrence had been accomplicied in accordance with technical specification LER 86-001-0 " Electrical Noise Caused Spikes on IRM's Resulting in RPS Aetustfons" LER 86-032-0 " GPS Actuation Oue to inadvertant Activation of Redundant Reactivity Control System" LER 86-003-0 "Cr6mped Work Location Causes Technician Error Resulting in RWCJ System Itoistion" Regarding LER C6-001-0, the hcensee indicated that a supplemental report will be submitted discussicig results of the licensee's ongoing investiga-tion end corrective actions. An inspection by the NRC Region III, Division of Reactor Safety to hs; decaented in NRC Inspection Report (440/86012) will provide an a6Ctioral independent assessment of licensee actions relative to the IRM aoise problem. Control rod position indication anomalies observed following the March 31, 1986, event discussed in the subject LER have been and will continue to be followed up by the inspector as discussed in Paragraph 2h of this repor Regarding LER 86-003-0, a similar event occurred on May 4, 1986, and will require submittal cf an LER. " Licensee a-tions specified in LER 86-003-0 to prevent recurrence were not inplemented prior to the second cccurrenc Inspector verification that corrective actions are implemented on an appropriate schedule will be accomplished during review cf the forthcoming LE No violations of regulatory requirements er deviction from commitments were identifie s
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_ Onsite Review Committee (40700)
The inspectors reviewed the minutes _of the Plant Operations Review Committee (PORC) meetings No. 86-62 and 86-66 through.86-85 conductep ,
prior to and 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 activitie '
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No violations of, regulatory requirements or deviation from commitments were identifie ,-
10. Maintenance Program Imple:centation/ Monthly Mainten ce Observations (62700, 62703)
A partial review of the Perry Mechanical /Electrisal Preventive Maintenance (MEPM) program was documented in Inspection.keport 440/8600 During the review, the inspector found some disparities between what was required by the plant MEPM program and the vendor manuals for the safety-related large Emergency Service Water pum _" _
Subsequenttotheinspectionmentionedabcve,theinmectobm5twith maintenance personnel and discussed the above concern The maintenance personnel provided an Environmental Qualification Report for the large Emergency Service Water pump, that specified changes in gear drive x, lubrication every 18 months rather than the 12 months or 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> of ^ -
operation required by the vendors manual. The differences bstween the Control Rod Drive pump MEPM requirements and the . vendor manuals maintenance requirements was partially resolved when the licensee informed the inspector that Perry does not have a Lufkin gear reducer assembly. Most of the maintenance discrepancies were related to the Perry MEPH not specifying maintenance on the Lufkin gear reducer. The licensee also stated that they w?re performing a re-review of all vendor ,
maintenance. requirements to assure that the MEPM program addresses the vendor requirements. The licensee informed the inspector that the re-review was completed for safety-related equipment and ongoing for safety-related equipment. The inspector has no further concern with vendor required preventive maititenance not being incorporated into the preventive maintenance required by the plant MEPM progra The inspector reviewed several recently completed safety lated work package During this review, the inspector identified a concern with Work Order (W0) No. 86-5657. This WO'was initiated to repair a leaking constant oiler for the fuel poo,1 cooling and cleanup pump. During the performance of this W0, torque valnes for reassembly of bearing cap hold downs were taken from Perry General Maintenance Instruction (GMI)-002 Bolta that required 250 ft lbs. of torque per GMI-0021, were reported in the retest summary section of the WO as being torqued to 200 ft. lb due t.o' torquing equipment limitations. There had been no Field Change Request'(FCR) generated, requesting relief from the 250 ft. lb required. In addition, the corpleted W0 was reviewed and approved by
~ both Engineering and Quality Ass'urance without identifying the torquing deficienc y
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The inspector brought this deficiency to the licensee's attentio The licensee issued Nonconformance Report 0QCS-886 to document the torquing deficiency, and Action Requests (ARs) 0073, 0074, and 0075 to document the procedural errors that occurred by the Maintenance, Quality Control, and Engineering organizations, respectively. The ARs also solicited j corrective actions and steps to prevent recurrence. In addition, the i licensee reviewed approximately 25% of completed safety related Work i Orders issued since January 1, 1986. The licensee reported that they '
had found no other incidents where acceptance criteria, such as torque values, were not met without proper documentation of the deficienc The inspector reviewed the licensee's actions concerning the torquing error, including performing a sample review of the Work Orders re-reviewed by the licensee, and concluded that the incident was isolated and the licensee's corrective actions, thorough. The inspector has no further concerns with this issu The inspector observed portions of maintenance activities associated with WO 86-392 During this observation, the inspector was informed that maintenance personnel had inadvertently installed the wrong time delay relay in the Division 1 Standby Diesel Generator control circuitr A review by the licensee's Quality Assurance organization revealed that the W0 had been written to implement Revision 0 of DCP 86-171 which only required wiring changes. By the time work commenced, Revision 2 of the DCP was issued including both wiring and relay changes. The licensee determined that the maintenance planner and supervisor failed to revise the WO in conjunction with the DCP revisions causing confusion for the technicians and contributing to the installation of the wrong relay.
, Also contributing to the error was the improper control of salvage material While these discrepancies were identified by the licensee, the inspector intends to follow the licensee's corrective actions along with the other mairtenance concerns previously identified as Open Item 440/86008-02(DRP).
No violations of regulatory requirements or deviation from commitments were identifie . Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether it is an acceptable item, a violation or a deviation. An unresolved item is identified in Paragraph 7 . 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 7a and 7 _ _ _
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13. Exit Interviews (30703)
The inspectors met with the applicant representatives denoted in Paragraph 1 throughout the inspection period and on May 12, 198 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 natur .
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