IR 05000440/1989017
| ML20246G295 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 08/18/1989 |
| From: | Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20246G287 | List: |
| References | |
| 50-440-89-17, NUDOCS 8908310290 | |
| Download: ML20246G295 (22) | |
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. U. S. NUCLEAR REGULATORY COMMISSION REGION'III
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. Report No. 50-440/89017(DRP) ,
Docket No., 50-440 License No. NPF-58
Licensee: Cleveland Electric Illuminating Company ~
Post Office Box 5000 Cleveland, OH 44101 Facility Name: Perry Nuclear Power Plant, Unit 1
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Inspection At: Perry Site, Perry, Ohio Inspection Conducted: June 6 through August 11, 1959 Inspectors: P. L. Hiland .
G. F. O'Dwyer J. A. Hopkins C. D. Pederson D. C. Kosloff Approved By: M. A. Ring, Chief D r /9,697 Reactoi- Projects Section 3) Dge Inspection Summary Insoection on June 6 through Aucust 11, 1989 (Report No. 50-440/89017(DRP))
Areas Inspected: Routine, unannounced safety inspection by resident inspectors of licensee action on previous inspection items; allegation followup; I operational safety verification; surveillance observation; maintenance
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observation; onsite followup of events; and plant status meeting, j Results: Of the seven areas inspected, one " licensee-identified violation" l tor which a Notice of Violation was'not issued was identified in the area of
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allegation followup (paragraph 3.d). The licensee-identified violation concerned inadequate procedures to control material in the containment pool swell region. One violation with two examples was identified in the areas of previous inspection findings (paragraph 2.a) and allegation followup (paragraph 3.a). That violation concerned the licensee's failure to report events to the hRC within four hours as required by 10 CFR 50.72. All of the above items were receiving management attentio PDR ADOCK 050 g O a 1 l
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, DETAILS Persons Contacted Cleveland Electric Illuminating Company (CEI)
- A. Kaplan, Vice President, Nuclear Group
- M. Lyster, General Manager, Perry Plant Operations Department (PPOD)
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- W. Coleman, Manager,.0perations Quality Section (NQAD)
- G. Dunn, Compliance Engineer (NSD)
- M. Grqyrek, Manager, Operations Section (PPOD)
- V. Higaki, Manager, Outage Planning Section (PPOD)
W. Kanda, Manager, Instrumentation and Controls Section (FPTD)
- S. Kensicki, Director, Perry Plant Technical Department (PPTD)
R. Newkirk, Manager, Licensing and Comp 1fance Section (NSD)
K. Pech, Manager, Technical Section (PPTD)
E. Riley, Director, Nuclear Quality Assurance Department (NQAD)
GF. Stead, Director, Nuclear Support Department (NSD)
- R. Stiffler, Shift Supervisor R. Stratman, Director, Nuclear Engineering Department (NED)
D. Takacs, Nanager, Mechanical Maintenance Quality Section (NQAD) U. S. Nuclear Regulatory Commission
- T. Colburn, Licensing Project Manager, NRR
- E. Greenman, Director, Division of Reactor Projects, RIII
- P. Hiland, Senior Resident Inspector, RIII J. Hopkins, Region III Reactor Inspector D. Kosloff, Resident Inspector, RIII
- G. O'Dwyer, Resident Inspector, RIII C. Pederson, Region III Reactor Inspector
- M. Ring, Chief, Projects Section 3B, RIII
- Denotes those attending the July 17, 1989 plant status meetin ' Denotes those attending the exit meeting held on August 11, 198 . Licensee Action on Previous Inspection Findings (92701, 92702) (Closed) Unresolved Item (440/89007-03(DRP)): Deportability of overheated cables discovered in the drywell following plant shutdown for the first refueling outag As previcusly documented in Inspection Report 50-440/89007, Paragraph 5.a., the licensee identified a number of safety and non-safety related cables within the Drywell that had become overheated during the plant's first operating cycle. The licensee, after discussions with the inspectors, submitted Licensee Event Report 89010-00 dated April 28, 1989, pursuant to the requirements of 10 CFR 50.73(a)(2)(vii). However, the licensee concluded that the overheated drywell cables were not reportable under the immediate notification requirements of 10 CFR 50.72. This item remained unresolved pending further review by the staf L _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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. During the report period, the staff completed their review of this item and concluded that the event due to excessive drywell temperatures was reportable within four hours pursuant ~to the requirements of:10 CFR 50.72. The staff review was documented in
~hRC memorandum Stuart A.'Treby, Assistant General Counsel for Rulemaking and Fuel Cycle to Jack Heltemes, Deputy Director AE0D dated June'23, 1989. In that review the staff concluded that the overheated cable and snubber event was reportable in accordance with 10 CFR 50.72(b)(2)(1) which required reporting of any unanalyzed-condition which significantly compromised safety. In the Statement of Considerations issued et the time of adoption of the current version of 10 CFR 50.72, the Commission stated the following to help determine whether an unanalyzed condition exists:
It is not intended that this paragraph apply to minor variations in individual parameters, or to problems concerning single pieces of equipment. For example, at any time, one or more safety-related components may be out of service due to testing, maintenance, or a fault that has not yet been repaired. Any trivial single failure or minor error in performing surveillance tests could produce a situation in which two or more often unrelated,
-safety-grade components are out-of-service. Technically, this is an unanalyzed condition. .However,-these events should be reported only if they involve functionally related components or if they significantly compromise plant safet Ilhen applying engineering judgment, and there is a doubt regarding whether to report or not, the Commission's policy is that licensees should make the report. (emphasis added).
This statement indicated that where an event involves only a
" single piece of equipment," a " trivial single failure," a " minor error," or a matter of like kind, it need not bt reported as an unanalyzed condition. But, conversely, if more than a single piece of equipment were involved or multiple failures of equipment were involved, then an unanalyzed condition likely existed. And if i engineering judgment was involved in determining whether an l unanalyzed condition existed, the licensee should err on the side
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of reportin In the overheated cable and snubbers event, the plant was operated for an unknown period of time with snubbers and a significant number of cables, both safety grade and nonsafety grade, in a harsh environment which was unanticipated. This environment was harsher than that for which the snubbers and cables were designed. Multipl pieces of equipment (cables and snubbers) were disabled as a result of being in this harsh environment which had not been previously recognized or analyzed. Consequently, the plant operated for an unknown period of time with only a single train of a number of safety ,
systems operable. Such an occurrence met the reporting criteria of
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10 CFR 50.72(b)(2)(1) and should have been reported to the tiRC within four hours. Failure of the licensee to report the overheated drywell cables and snubbers ev=nt discovered on March 29, 1989, within four hours is an example of a Violation (440/89017-OlA(DRP)) of
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. 10 CFR 50.72(b)(2). Unresolved Item 440/89007-03 is closed based on upgrading to a Violatio (Closed) Violation (440/86008-04(DRP)): Inadequate Controls hecessary to Establish the Operating Status of Safety Related Instrument This item was previously reviewed in Inspection Reports (IR) '
60-440/86011, Paragraph 2.j.; IR 50-440/86014, Paragraph 2.1.;
and IR 50-440/86018, Paragraph 2.e. At the conclusion of those reviews, this item remained open pending completion and issuance of controlled final as-built instrument drawings prior to startup after the first refueling outag ,
During the report period, the inspectors reviewed the licensee's I final report for Special Project Plan 1401, phase 3 dated April 14, 1988. As detailed in that report, the licensee issued final as-built Series 803 mechanical instrument drawings prior to startup after the first refueling. Completion of that activity satisfied the licensee's corrective action as detailed in licensee letter PY-CEI/0IE-0181L, dated March 14, 1966. This item is close (Closed) Open Item (440/87023-01(DRP)): Development of Area Ambient Temperature Monitoring Progra During the report period, the inspectors reviewed Periodic Test Ir.ctruction (PTI)-M99-P001, Revision 0, dated October 7, 198 That instruction established a semi-annual (184 days) frequency to record and evaluate ambient temperature conditions in the plan Ambient temperatures were to be taken in the Auxilitry Building, Fuel Building, Control Building, Turbine Building, Off Gas Building, Containment, and Drywell. Results of the temperature monitoring instruction were to be forwarded to the Nuclear Engineering Department lead HVAC Engineer. Based on the licensee establishing a program to raoniter area ambient temperatures, this itera is close (Closed) Open Item (440/88003-03(DRP)): Main Steam Isolation Valve Logic Modificatio In January 1986, the licensee identified that the control logic
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design for the main steam isolation valves (MSIVs) resulted in opening of the inboard or outboard MSIVs urder certain plant l conditions upon deenergization of a single P,eactor Protection System l
(RPS) bus. Licensee letter PY-CEI/NRR-0803L, dated February 12,
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1988, stated that a design modification performed during the first refueling outage would revise the MSIV control circuit logic to
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prevent the automatic opening of MSIVs on loss of a single RPS bu During this report period, the inspectors reviewed Design Change Package (DCP) 880057, " Main Steam Isolation Valve Control and Indication." A post modification test was successfully perfo'medr on June 23, 1989 and DCP 880057 was completed on July 5, 198 Based on the completion of the design change as committed, this item is close _ - - _- _ __ _
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, e.- (Closed) Violation (440/88003-01(DRP)): Failure to Declare Emergency Service Water (E5W) Loop "A" Radiation Monitor Inoperabl The licensee responded to the subject violation via letter PY-CEI/HRR-0831L, dated March 30, 1988, in a timely manner. As stated in that response, corrective action taken included declaring ESW loop "A" radiation monitor inoperable and complying with the applicable Technical Specification ACTION requirement. The. licensee attributed the sub 9ct violation to both design deficiency and personnel error. Corrective action for the personnel errors included counseling all control room operators on the careful consideration of Control Room alarms. In addition, the sequence of events that led to the violation was reviewed with operations personnel during Operator Requalification Training. . Corrective action for the design deficiency was to install lead brick shielding around the piping system that was providing the high radiation background ano causing the ESW Loop "A" radiation monitor to become inoperable. During the report period,.the inspectors noted tha Design Change Package'No. 670274 (lead brick installation) was completed on May 8, 1969. Based on the completion of corrective actions committed to by the licensee in their response to the subject violation, this item is closed, (Closed) 0:en Item (440/88011-01(DRS)): Repetitive Task to Verify TH~rque of iydraulic Control Unit hold Down Bolt This item discussed the inspectors' suggestion to the licensee during an inspection of licensee action in response to NRC Information Notice 87-56, " Improper Hydraulic Control Unit (HCU) Installation at BWR Plants." During the course of that review, the inspectors suggested that the licensee consider a 10 year cycle to verify torque of HCU hold down bolts. During this repcrt period, the inspectors reviewed the licensee's evaluation (contained in licensee memo Concel to Dunn, dated September 29, 1988) of that suggestion which concluded that a 10 year cycle (repetitive task) was not justified. Based on the fact that this item was a suggestion only-and the fact that the licensee performed an adequate evaluation to support their conclusion, this item is closed.
I (Closed) Open Item (440/88012-03(DRP)): Plant Logs and Record During an Operational Safety Team Inspection (OSTI), a noted observation was the detail and accuracy of control room log In response to that observation, the licensee identified actions to be taken to improve the content and quality of control room logs. .As detailed in licensee memorandum Hegrat to Dunn, dated September 29, 1988, the improvement actions included a formet change and increased management attention. During the last several report periods, the inspectors have routinely reviewed plant operating logs and found the content to adequately detail plant evolutions and events. The inspectors will continue to review plant logs as part of their routine inspection effort; however, based on the current status of logkeeping, this item is close _
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, h. (Closed) Open Item (440/88020-03(DRP)): Loss of All Rod Position Indicatio In December 1988, the licensee experienced a loss of all control rod position indication and the ability to select and insert individual control rods except via a manual scram. The cause for the loss of control power was found to be a failed power supply. This item remained open pending the licensee's completion of their review and implementation of procedures to direct plant operators should a similar event occur. During this report period, the inspectors reviewed System Operating Instruction (501)-Cll(RCIS), " Rod Control and Information System," Revision 6, dated May 12, 1989. The inspectors noted that Revision 6 to S0I-Cll(RCIS), Section detailed the required steps for plant operators to determine control rod positions with the full core display out of service. Based on the licensee's action to provide instructions for determining control rod position during an event similar to the one occurring in December 1988, this item is close i. (Closed) Violation (440/87023-06(DRP)): Failure to Trip Reactor Protection System Log $c Due to Inoperable Intermediate Range Monitors (IRMs).
In July 1987, on three occasions, plant operators failed to place a trip system of the reactor protection system (RPS) in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> as required by Technical Specification 3.3.1. due to inoperable IRMs. The licensee reported this violation of Technical Specifications in Licensee Event Report (LER) 87052, dated July 31, 1987. The inspectors' review and closure of LER 87052 were documented in Inspection Report 50-440/87023, paragraph 8. The licensee respended to the subject viclation via letter PY-CEI/NRR-0809L, dated March 3, 1968, in a timely manne Corrective action taken included counseling all plant operators involved in the violations and training of all shift personnel on the correct interpretation of Technical Specification requirement Based on completion of the corrective action taken as stated in the licensee's response to the subject violation, this item is close j. (Closed) Violation (440/08015-03(DRP)): Failure to Report Loss of Control Room Emergency Recirculation System In October 1988, the licensee experienced a loss of both trains of the control room ventilation emergency recirculation. The licensee responded to the subject violation via letter PY-CEI/NRR-0945L, i dated December 9, 1988 in a timely manner. That response detailed l the licensees root cause of the failure to report the event within four hours as required by 10 CFR 50.72 (b)(2)(iii). As stated in their response, the licensee had concentrated their efforts on compliance to Technical Specification 3.0.3 and reporting the loss of a safety system function was not addressed. Based on the completion of corrective actions as stated in the licensee's response, which included notification to the NRC and reporting the event via LER 88040, this item is closed.
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. k.- (Closed) Open Item (440/88020-01(DRP))': Administrative Procedure ,
PAP-0402 Did Not Ensure Changes Made to QA Records Were Subject to
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an Adequate Review and Approval Process ~.
During a previous inspection, the inspectors noted that Perry Administrative Procedure (PAP)-0402, Revision 7, " Material Request ,
Processing," did not provide adequate procedural direction for /
changes to picking tickets and stores requisitions. . Concurrent with the inspectors' observations the licensee issued Action i, Request (AR) P000103-89000-0163. During this report period, the inspectors reviewed Material Control Instruction (MCI)-0413,
'" Material Request," Revision 0, dated March 7, 1989. The inspectors noted that Section.3.2 of the instruction required the Maintenance Manager to ensure that re-review of quality documents was obtained i when entries were revised by Maintenance personnel. Based on the satisfactory closure of AR P000103-89000-0163, this item is close One example of a. Violation was identifie . Review of Allegation (Closed) Allecation (RIII-88-A-0090): Probability assessment performed in Condition Report (CR)87-543 is a nonconservative hand wav Background To perform work in containment, scaffolding was installed at elevation 599', cuiside the drywell, from approximately Azimuth (AZ) 337 degrees to AZ 90 degrees. The scaffolding was seismically braced as required by procedures and was installed between November 18 and 23, 1987, during which time the plant was in operational conditions 1 and 2. On November 23, 1987, the licensee's Mechanical Design Section (MDS) Structural Group raised concerns that the scaffolding was a potential hydrodynamic missile hazard and Condition Report (CR)87-543 was generated in accordance with the licensee's procedures to track the concern and its disposition. The scaffolding was removed shortly after the condition-was identifie On June 30, 1988, an anonymous allegation was received by the Perry Resident Inspector concerning the disposition of CR 87-543. An after-the-fact assessment of safety significance was performed by the licensee as part of the disposition to CR 87-543. That assessment determined that multiple safety systems and components could have been adversely affected had the scaffolding become hydrodynamic missiles due to pool swell following a LOCA. Further, the assessment stated that the probability of such an event over the time the scaffolding was installed was low and that the scaffolding represented a negligible risk. Based on those assessments it was concluded that installation of scaffolding in the pool swell region could have been permitted for periods of up to seven days. The technical specification bases for 7-day Limiting Ccnditions for Operation were referenced in support of that conclusio The subject allegation concerned the technical adequacy of the disposition to CR 87-543. In particular, the application of the Perry
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. Plant Technical Specification bases was questioned due to the multiplicity of systems and components potentially rendered inoperable by the
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scaffolding in_the pool swell region after a LOCA.
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Region III Review On December 27, 1988, by letter from Edward G. Greenman, Director, Division of Reactor Projects, Region III, the NRC requested that the licensee review and followup on the subject allegation and submit the results of their review to Region III. The licensee provided their response via letter PY-CEI/NRR-0964L, dated January 26, 1969. During the report period the inspectors completed a review of the licensee's response as discussed below: Region III Question 1.a./ Regarding the dispositioning of the associated Condition Report:
Justify the acceptability of the disposition in view of the fact that it does not explicitly address all the elements of 10 CFR 50.59 for assessing the safety significance and the potential for the existence of an unreviewed safety qucstio Explain whether the dispositioning of the Condition Report was consistent with your Condition Report program / procedure Provide technical justification for why the condition for which the Condition Report was written should not be reviewed as an unreviewed safety quertior and explain how the review process failed to address the unreviewed safety question issu Licensee Response to Region III Quertion 1,a.#.
Condition Report 87-543 was processed and dispositioned in accordance with Plant Administrative Procedure (PAP-0606) Condition Reports and Immediate Notifications. This process provides for the evaluation of plant events or conditions with respect to deportability under 10 CFR 50.72, (Imediate Notification Requirements for Operating Nuclear Power Reactors) and 10 CFR 50.73 (l.icensee Event Report System). Additionally, the associated investigation, review, and approval procesc ensures that appropriate measures are taken to mitigate any consegeences of events which have already occurred, as well as to prevent recurrence of the event or condition. A review of CR 87-543 concluded that the evert vid not warrant 10 CFR 50.72 nctificaticn or a Licensee Event Report as required by 10 CFR 50.7 Extensive inter-disciplinary review resoited in recumFocatfurs for procedural changes to minimize the probatihths fcr evet recuacoch The Cor.dition Report process is not intended to provide for the j evaluatint of proposed chsnges, tests or experiments as discus &u in ,
10 CFR 50.59; moreover, the condition which existed frera Hovenber 16 1 through 23, 1987 was not a proposed chanc% test or experiment. As reI;uired by the condition report process, Lhe condition wJs reyieved with respect te requirements in the Safety Analysis Report end j
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. approved plant procedures. The condition was. determined to be L unacceptable with-respect,to plant procedures and was immediately
- corrected. The specific condition was also evaluated with respect j l to the need for conducting maintenance and surveillance testing in-order to maintain operability of all system functions during plant
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l operations. With respect to reporting requirements in 10 CFR 50.73,-
the' condition was determined to be non-reportable. The dispositioning of the Condition Report was therefore consistent with the approved procedure, and appropriate with respect to plant safet Inspector Review The inspectors discussed the above response with cognizant licensee personnel and informed the licensee that, contrary to the response provided, an event meeting the reporting requirements of 10 CFR 50.72 occurred between November 18 and.23, 1987. Specifically, 10 CFR 50.72(b)(2)(iii)(D) required a four hour report for any
.. condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. As detailed in the " event description" block of CR 87-543, scaffolding was installed in the pool swell region without analyzing the potential hydrodynamic missile hazard the scaffolding presented numerous safety.related components in the event of a design basis loss of coolant acciden Failure to report this event in accordance with 10 CFR 50.72 is an example of a Violation (50.440/6901701B(DRP)).
The inspectors also noteu that the response did not explicitly answer the unreviewed safety question issue; however, as discussed below in subparagraph c., the licensee agreed that multiple safety systems could be affected by hydrodynamic missiles generated from scaffolding in the pool swell region and revised administrative procedures to prevent introduction of scaffolding in the pool swell region during plant operation b, Region III Question Specify exactly what action you would have taken to determine the operability of equipment in the vicinity of temporary scaffolding should that scaffolding have been. required to remain in containment deyond the recommended limit of seven day Licensee Response to Region III Question As & corrective action for the Condition Report, a Temporary Change was approved for Plant Administrative Procedure (PAP-1402), " Control of Lifted Lecds, Jumpers, Temporary Electrical Devices and Mechanica? Foreign Items." In addition to' allowing limited amounts of scaffolding to be installed in restricted pool swell areas for periods up to 7 days without being identified as a Mechanical Foreign. Item (NFI), the procedure al:o specifically required all ';
scaffolding in these areas to be considered as an MFI at the end of i
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the seven day period. Since all NFIs require evaluation under 10 CFR 50.59 .the scaffolding would need to be removed if an Unreviewed Safety Question was presented. The Safety Evaluation performed would also consider the operability of systems and equipment present in the affected area. Each item would be considered on a case by case basis, and actions with respect to component operability would be guided by Technical Specification During the period these provisions were in effect, however, there were no instances in which scaffolding was erected in pool swell areas while in Operating Conditions 1, 2, or 3. The temporary change implementing these provisions has since been rescinded, precluding the possibility.of future occurrence Inspector Review The inspectors reviewed the above response with cognizant licensee personnel. As stated above, the licensee initially revised Perry Administrative Procedure (PAP) No. 1402 to allcw installation of scaffolding in the pool swell region of containment for a period of up to seven days after which time a safety evaluation would be performed. The justification for incorporating those administrative controls was detailed in Temporary Change Notice (TCN)-l to PAP-1402, Revision 6, dated October 7, 1968. The inspectors noted through review of that justification that the licensee based the seven day allowed scaffold installation time on a " probability assessment" and correlation to existing Technical Specification allcwed out of service time TCN-1 to PAP-1402 concluded that an unreviewed safety question did not exist. The inspectors noted that conclusion was not supported by a rigorous engineering evaluation of the affects scaffolding would have on safety related structures and components following the design basis loss of coolant accident. The " probability assessment" presented by the licensee might possibly have been used as an engineering judgment to allow installation of scaffold in the pool swell region for conditions or modes other than power operatio As a result of the licensee re-evaluating the use of scaffolding in the pool swell region following receipt of the subject allegation and discussions with the staff, the licensee removed the scaffolding provisions from PAP-1402 via TCN-3 dated January 12, 1989.
i Region III Question In the disposition of the subject Condition Report you concluded that a seven day allowed time period to have temporary scaffolding erected in the pool swell region of containment is consistent with Technical Specification Bases for Limiting Conditions of Operation and associated Action Statements for equipment out of service. Ve would note, however, that such equipment allowed out of service times apply to single safety systems / trains that have design backup
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systems or trains. The missiles generated from scaffolding failure due to the impact of hydrodynamic forces in the pool swell region would potentially lead to the simultaneous degradation /inoperability of multiple safety systems. Explain your rationale for applying the seven day logic given the potential impact on the operability of multiple safety system '
Licensee Response to Region III Question The procedure change and 10 CFR 50.59 Applicability Check to PAP-1402 was approved upon review and acceptance of the engineering evaluation which recommended the seven day limitation on scaffolding installed without MFI tags. The engineering analysis assumed that only one of the redundant safety systems would be impacted by y missile hazards from a proposed LOCA. Based on that assumption, and the low probability of a LOCA occurring anytime during the life of the plant, the seven day allowance was considered to be acceptable and consistent with a single ECCS out of service tir;.e. After discussions with members of your staff and the staff of Nuclear Reactor Regulation, CEI has re-evaluated the condition and agreed that multiple safety systems could, in fact, be impacted. PAP-1402 has been modified to remove the scaffolding provisions previously established. At no time while the procedure change was in effect was scaffolding installed in pool swell areas in Operational Condition 1, 2 or Inspector Review As discussed above, the inspector noted that PAP-1402, Revision 6, was revised via TCh-3, dated January 12, 1989, to require all temporary equiprmat installed within the pool swell region to be processed as a Mechanical Foreign Item which required a safety evaluation in accordance with 10 CFR 50.59 prior to installatio Based on the inspectors discussions with the licensee, the controls established were adequate to prevent installation of scaffolding in the containment pool swell regio d. Region III Question Fxplain what you perceive to be the root cause of the event and what corrective action you have implemented or will implement to prevent recurrenc Licensee Response to Region III Question The root cause of the event addressed in Condition Report 87-543
, is procedural inadequacy. As established by the CR investigation.
l working in various procedures with regard to scaffolding did not clearly consider the potential missile hazards of scaffolding located within pool swell / weir swell regions. Subsequent procedure il m_ - _ _ - - _
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changes have made it clear that scaffolding in the pool swell areas is tu be controlled under requirements for Mechanical Foreign Item These procedural requirements include 10 CFR 50.59 evaluation of all MFI's installe Inspector Review The inspectors reviewed the licensees root cause investigation summary contained in CR 87-543. As discussed in that summary, the administrative procedures used to install scaffold in containment 4 on November 18, 1987, were not clear. Since the scaffold was i seismically braced in accordance with PAP-0508, "PNPP Operating Rules and Practices," the requirement to identify the scaffold as a Mechanical Foreign Item and perform a safety evaluation were not imposed as allowed by PAP-1402. An additional procedure, PAP-0204,
" Housekeeping / Cleanliness Control Program," required a Mechanical Foreign Item tag and safety evaluation for items left in containment during operation that could have presented a missile hazard to safety systems; however, PAP-0204 did not identify the pool swell i event as a potential motive force for generating hydrodynamic j missile !
i The failure of the licensee to provide adequate instructions for installation of scaffolding in containment during, plant operation is a Violation (50-440/89017-02(DRP)) of 10 CFR 50 Appendix B which requires that activities affecting quality shall be prescribed by documented instructions of a type appropriate to the circumstance The licensee identified the installed scaffolding as a potential hydrodynamic missile concern on November 23, 1987. Condition Report 67-543 was initiated and prompt corrective action was taken by the shift supervisor by directing its immediate removal. Corrective action to prevent recurrence was completed by revising administrative procedures to preclude installation of scaffold in the containment pool swell region. Based on the above, the inspector considered the failure to have adequate procedures for installation of scaffolding in the containment acol swell region to be a " Licensee Identified Violation" for which a Notice of Violation will not be issue Conclusion The allegation was partially substantiated in that the licensee's bases )
for allowing installation of scaffolding in the containment pool swell l region during plant operation was not supported by a factual assessmant -
of impact on safety systems the postulated hydrodynamic missiles presented. After re-evaluating the administrative centrols put into I place following the initial " probability assessment," the licensee agreed ;
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that multiple safety systems could have been impacted by the postulated hydrodynamic missiles ar,d revised the applicable procedures to preclude the introduction of scaffolding in the pool swell region during plant
- operation. This allegation is close )
One example of a Violation and one " Licensee Identified Violation" were I identifie i I
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4. Operational Safety Verification (71707, 71711)
The inspectors observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during the inspection period. The inspectors verified the operability of selected emergency systems, reviewed tag-out records and verified tracking of 1 Limiting Conditions for Operation associated with affected component '
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 i in need of maintenance. The inspectors observed plant housekeeping / l cleanliness conditions and verified implementation of radiation I protection control l During the report period, the licensee completed their first refueling outage. The inspectors observed the licensee's rfistart activities which .
included initial plant startup performed on July 24, plant shutdown on l July 31 to repair steam leaks identified during the initial plant heatup, and plant restart on August 3, 1989. On August 5, the main generator was synchronized to the grid ending the licensee's first refueling outag The inspectors noted that the licensee's approach to plant startup following the extended outage was cautious and well controlle Observations of control room activities indicated that plant operators were well briefed on planned events and were provided sufficient time to conduct startup testing and surveillance activitie Report of a Fire in Containment On June 20, 1989 at about 1:45 p.m., control room operators received a verbal report of a fire in the containment. The inspectors observed activities in the control room following the initial repor At about 1:54 p.m., the control room received word that the rcported fire was in fact a 20 pound dry chemical fire extinguisher that discharged its contents after being dropped curing transport. The inspectors observed control room communications with the fire brigade which had bacn assembled at the Operations Suppcrt Cente Communicatitr.s appeared to be good. Fol'iowing the identification of the cause for the reported fire, the fire brigade was secured and a critique of the event was conducted to identify weaknesses in the 1 response. The details of that critique were contained in Condition Report (CR)89-255. The off-site fire department from Perry township responded to the site and made plans to assist the onsite fire brigade. Follcwing the termination of the fire response, the Perry township Fire Chief, Mr. Sitts, briefed the inspectors on the plans prepared to combat the fire. The inspectors noted that -
1 Mr. Sitts was knowledgeable in the plant layout and hrd made plans with the onsite fire brigade leadcr to combat the suspected fir In addition to the licensee's critique of their response to this l event, the inspectors noted that followup inspections of electrical equipment subjected to the dry chemical spray were performe !
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b. Sleeping Health Physics Technician On June 29, 1989, at about 12:30 p.m. the Senior Resident Inspector (5kI) observed an on duty health physics (HP) technician sleeping in the drywell. Upon entering the drywell to monitor a test being conducted by the licensee, the SRI observed twc EP technicians, 1 assigned to monitor the drywell radiological conditions, laying down '
insice the drywell equipment hatch. One of the HP technicians was alert and spoke with licensee personnel who were accompanying the SRI. The second HP technician was not alert as evidenced by closed eyes and loud " snoring." The SRI requested the first HP technician to awake the individual and inform licensee supervision. At the time of the observation, the drywell was controlled as a contaminated radiation area. No other personnel were in the drywell since it was the lunch hour. The licensee advised the SRI that the two HP technicians involved were contractors employed during the current refueling outage and that their employment was terminated on June 29. The licensee also stated that it was not their policy to allow personnel to sleep within the radiological controlled are c. Drywell and Containment Closecut The iicensee conducted "closcout" inspections of the Drywell and Containment in preparation for the July 24, 1989 refueling starta These inspections were conducted using Integrated Operating Instruction (101)-1, Revision 4, " Cold Startup." The inspectors accompanied one Supervising Operator (a licensed Reactor Operator)
and one Quality Assurance representative on the "closecut" inspectio During the inspection, several deficiencies were identified including loose debris and trash on the floor and in the Suppression Pool, components with water leaks, temporary instrumentation cabling improperly restrained and equipment hoists stowed near Drywell ccmponents. The licensee corrected all of the identified deficiencies prior to the reactor startu The inspectors were concerned with the apparent lack of management attention that the "closecut" inspections received. The licensee stated that the Drywell and Containment had been inspected and implied that the "closecut" inspection was mostly for the inspectors benefit. Based on the volume and nature of the deficiencies identified, the licensee was encouraged to devote more resources to future "closecut" inspections with increased management involvement in the actual walkdow Additionally, the inspectors noted that most of the trash and loose debris (i.e. tools, tie-wraps, tape and tubing) were a result of inadequate post-maintenance restoration practices as opposed to poor housekeeping. Greater attention to post-maintenance restoration activities was required to prevent future cleanup efforts in High Radiation areas such as the Drywell and Containmen ,
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I e ' Loo'se Fasteners on Electrical Switchgear' Panel Doors On July.19, 1989 during routine plant tours, the inspectors identified several electrical switchgear panel. doors with loose
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fasteners.. Some of.the panels housed Emergency Core Cooling System (ECCS) equipment-including.the Division I Diesel Generator Breaker, both Emergency Service _ Water pumps and.the "A" Residual Heat Removal Pump. . The licensee was informed of the condition and. asked to evaluate the seismic qualification and operability of the ECCS equipment. This issue had.been previously identified in Inspection keport 50-440/86018. The monthly exit notes documented for Report 50-440/86025 stated that the licensee'had addressed the issu The. licensee conducted a complete physical walkdown of the safety related switchgear panels to identify, document, and repair any loose fasteners. All fasteners were either immediately tightened or a Work Order was initia.ted to repair the fastene >
On July 24, 1989 the licensee completed the evaluation of the-seismic qualification and operability of the ECCS equipment with loose. fasteners on the switchgear. panel doors. The evaluation determined that only two fasteners were required to hold the door secure. In all cases the panels had at least two fasteners holding the doors in place. . Additionally, the evaluation stated that a loose door could not generate enough force to damage the equipment housed by the panel or cause the equipment to operate improperl The evaluation then focused on panels which had electro-mechanical relays mounted on the doors. The licensee stated that with at least two fasteners in place, the electro-mechanical relays mounted on the doors would not be affected by a seismic even The inspectors noted that incomplete post-maintenance restoration
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activities contributed to the problem. This issue was another example (see drywell and containment closecut) of the need for both the maintenance staff and their management to pay more attention to post-maintenance activitie No Violations or Deviations were identifie . Monthly Surveillance Observation (61726)
For the surveillance activities listed below, the inspectors verified one or more of the following: testing was performed in accordance with procedures; test instrumentation was calibrated; limiting conditions for operation were met; removal and restoration of the affected components were properly accomplished; test results conformed with technical specifications and procedure requirements and were reviewed by personnel 15 l m i
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. ~other thanithe individual directing the test; and that.any deficiencie : identified during the testing were properly reviewed and resolved by appropriate management personne Surveillance Test N Activity-SVI-E31-T0074-A'. Isolation Actuation Instrumentation / Reactor Water Cleanup System Isolation Differential Flow High and Differential Flow. Timer Channel Calibration for lE31-N076B" SVI-C51-T0234-B Source Range Monitors B and D Channels Functional Test"
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SVI-M51-T0321-A' Accident Mo'nitoring - Primary Containment and Drywell Hydrogen-Concentration Analyzer and Monitor Division 1 Calibration SVI-C51-T0022-A IRH A and E Neutron Flux Trips Channel Functional Test No Violations or Deviations were identifie . . Monthly Maintenance Observation (62703)
Station maintenaroe activities of safety related systems and components listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes er standards and in conformance with technical specification F'
The following items were considered during this review: the limiting *
conditions for operation were met while components or systems'were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved precedures and were inspected.as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality i control records were maintained; activities were. accomplished by 1 qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemente Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performanc !
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. Portions of the fallowing maintenance activities were observed / reviewed:
, Work Order Activity 890004280 Repair to Inboard Containment Isolation Yalve E51-F063 880007186 Main Steam Isolation Valve 22A Regarding Work Order No. 880007186, the inspectors review was 1 prompted due to an event that occurred on June 27, 1989. As documented in Condition Report (CR)89-261, control room operators discovered three of the four inboard main steam isolation valves (MSIVs)Lclosed and the fourth, MSIV22A, in an intermediate positio At the time of discovery the inboard MSIVs were supposed to be open in support of.a "In Service Leak Test" being performed.- The apparent cause for MSIV closure was the isolation of Service Air to the drywell with a subsequent slow biced-off of air pressure i due to ongoing work on the drywell personnel air-loc !
On a loss of service air header pressure the MSIV valve design was such that each independent MSIV air accumulator provided the motive
' force to rapidly close the MSIV. However, in order for that closure sequence to occur without an existing closure signal, the HSIVs two and/or four way directional control valves must shift to exhaust air from the below-piston area of the valve actuator and direct the !
closure air supply to the above-piston area of the valve actuato The shifting of the two and four way directional control valves, <
without a closure signal present, was a function of decreasing air l pressure decrease in the isolated service air headcr allowing the l directional control valves' springs to shift the control valve l l
The licensee prepared a troubleshooting plan and attempted to recreate the plant conditions by slowly decreasing the service air header pressure to the drywell and observing MSIV movement. The i conclusion reached by the licensee was that for the HSIV closure on 'i June 27, existing plant conditions and work cctivities allowed the !
drywell service air header to be gradually and intermittently ]
depressurized over an unknown time period. As the air header ;
pressure was slowly depressurized, normal leGage past the independent MSIV air accumulator check valves would repressurize :
the air header and at the same time reduce the air pressure within i the accumulator. Partial shifting of the four way directional i control valve due to the slow decrease in air header pressure would 1 have allcwed air pressure to slowly decrease on the belcw-piston j area of the MSIY pressure valve actuater. Based on the j troubleshooting performed, the licensee replaced the air pack on
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HSIV 22A as a precautionary measur The inspectors witnessed the troubleshooting test and reviewed test data obtained by the licensee. In addition, the inspectors rcviewed in detail the maintenance work performed on MSIV 22A during the
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. refueling outage. The inspectors noted that the licensee had incorporated a " slow-closure" test in accordance with the recommendations of General Electric Service Information Letter fio, 477 dated December '3, 1988. In addition, the licensee had conducted accumulator cher.k valve leakage tests during the refueling outage with satisfactory results as documented in Surveillance Instruction (SVI)-B21-T2200, dated May 23, 198 The licensee provided a report of their root cause investigation into the above MSIV 22A anomaly by letter PY-CEI/01E-0355L, dated August 10, 1989 to Region III. Based on the inspet. tors observations of troubleshooting and review of test data, the licensee appeared to have adequately adaressed the MSIV 22A anomaly that occurred on June 27, 198 b. During plant startup from the refueling outage, additional MSIV failures to " Slow-Close" were identified. On July 25 with the plant in operatjenal condition 2 (STARTUP), at 10 psig reactor pressure and less that 1 percent power, control room operators attempted to slow-close the outboard MSIVs. Three of the four outboard MSIVs slow-closed as expected; however, MSIV 28C failed to stroke close on spring pressure alone. MSIV 28C was closed via the normal control switch which allowed the MFIV's air accumulator to direct the normal closure air pressure to the above-piston area of the valve actuato After establishing a troubleshooting plan, the vicensee attempted to recreate the failure of MSIV 28C to slow-close on July 2 However, the failure to slow-close could not be recreate :
The licensee briefed the inspectors and Region III management on the failure of MSIV 28C to slow-close on auly 25, and their unsuccessful attempt to recreate the failure on July 26 (at the time of troubleshooting on July 26, the plant was at 0 psig). As discussed with Region III management on July 26, the licensee performea a slow closure test on July 28, with the plant at about 5 percent power and normal temperature and pressure. The insaectors witnessed the July 28 test and noted that MSIV 28C slow stroced closed about 50 percent of its total stroke when it stopped. The licensee ceveloped a troubleshooting plan that performed a slow-closure test of each MSIV j at rated tem >erature and pressure. That troubleshooting effort 1 identified tie f act that five of eight MSIVs would not f ully l slow-stroke closed at rated temperature and pressure with low steam flow. The licensee's investigation determined that the cause for )
failure to slow-close was overtorqued spring guide bushings. Under the plant conditions tested (i.e. rated temperature and pressure with minimal steam flow), the spring guide bushings were binding on the valve yoke preventing full slow-closure. The licensee stated that the vendor instruction manual directed that the spring guide bushings were to be installed " snug-tight" and a maximum torque was not provided. Throughout the above troubleshooting effort the licensee verified the safety function of all MSIVs to close by the successful performance of their fast-closure (3-5 sec.) MSIV i surveillance test. Adjustments were made to the affected MSIV guide I bushings at rated temperature and pressure. Successful slow-closure 18 i
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.- tests were performed en all MSIVs prior to entry into Operational Condition 1 (POWER OPERATION) on August 5, 1989. The inspectors noted that MSIY troubleshooting was performed in accordance with an approved plan and the results were properly evaluated. The licensee provided a' report to Region III discussing the root cause evaluation for this-event concurrent with the report-discussing the. July 25 failure to slow-close of MSIV 22A discussed abov .No Violations or Deviations were identifie . Onsite Followup of Events at Operating Power Reactors-(93702)
a .~ General The inspectors performed onsite followup activities for events which occurred during the inspection period. Followup inspection included one or more of the following: reviews of operating logs, procedures, condition reports; direct observation of licensee actions; and interviews of licensee personnel. For each event, the inspectors reviewed one or more of the following: the sequence of actions; the functioning of safety systems required by plant conditions; licensee actions to verify consistency with plant procedures and license conditions; and verification of the nature of the event. Additionally, in some cases, the inspectors verified that licensee investigation had identified root causes of equipment malfunctions and/or personnel errors and were taking or had taken appropriate corrective action Details of the events and licensee corrective actions noted during the inspectors' followup are provided in Paragraph b. beic Details (1) . Reactor Protection System (RPS) Trip - No Rod Novement I At about 1:58 p.m. on June 13, 1989 with the plant in operational condition 5 (REFUELING), a Reactor Protection System actuation occurred with no rod movement due to a high Scram Discharge Volume (SDV) water level. The SDV filled because an Alternate Rod Insertion (ARI) was generated from Division 2 of the Redundant Reactivity Control System (RRCS)
when a High Reactor Dome Pressure signal was inserted on channel "A" as required by a surveillance test (SVI) being performed. The licensee was investigating why channel "B" of j p the ARI logic was tripped which was necessary to cause the j Division 2 ARI. It appeared that the "B" channel had been >
tripped during implementatieri of a design change package which had been worked a significant time period previcusly and the logic had not been reset. The licensee was seeking to quantify )
the time period that the "B" channel was improperly left not reset. The SVI had a prerequisite requiring that technicians performing the surveillance ensure that the other channel was not tripped; however, when the technician checked the local j panel the only annunciator that was lit was the ':RRCS Trouble" i annunciator. The technician assumed that it was lit because l
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- the "B" Reactor Recirculation Pum) breakers were tagged open which also causes it to light. T1e licensee was considering changing the surveillance to require that the technician reset the ARI logi (2) Loss of Shutdown Cooling At about 10:00 p.m. on July 17, 1989, Shutdown Cooling was lost due to personnel error by a licensed operator. The plant was in mode 4 at 120 degrees with both. recirculation pumps running at the time. While performing Surveillance Instruction (SVI)-G43-T2001, " Suppression Pool Makeup Valve Operability Test," the operator inadvertently closed the A Shutdown Cooling (SDC) inboard suction isolation valve, lE12-F009, resultin the tripping of the operating Residual Heat Removal (RHR) gmp. g in The operator was to close the Suppression Pool Makeup Yalve from the Separator Storage Pool, 1G43-F0308, which is immediately adjacent to lE12-F009 on the control board. SDC was restored within 4 minutes during which time there was no detectable temperature rise. The redundant B RHR loop was available at the time of this even (3) Reactor Water Cleanup Isolation At about 1:00 a.m. on July 26, 1989, the licensee experienced an unexpected Engineered Saf ety Feature (ESF) actuation due to an automatic isolation of the reactor water cleanup syste With the pla t in operational condition 2 (STARTUP) at zero power, plant operators attempted to establish reactor vessel level control via the reactor water cleanup system in accordance with the system operating instructions. During that evolution a large delta flow error was generated resulting in a system isolation. Plant operators verified an actual leak was not present and established the required system lineup about 30 minutes after the initial isolation. The licensee notified the NRC operations center of this event via the ENS at about 4:00 a.m. on July 26, 196 (4) Reactor Water Cleanup Isolation j At about 11:30 a.m. on July 31, 1989, while in operational condition 3 (HOT SHUTDOWN), the licensee experienced an unexpected ESF actuation due to an automatic isolation of the reactor water cleanup system. Just prior to the event, the licensee had performed a planned manual scram fron 10 percent i reactor power. Following the manual scram, both "A" and *b" delta flow instruments tripped resulting in the system isolation. Plant operators verified an actuel system leak was not present and restored the system to a normal lireup. The ,
licensee notified the KRC operations center of this event via 1 the ENS at about 1:00 p.m. cn July 31, 196 j
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. (5) Single Rod Scram At about 4:00 a.m. on August 7, 1989, while in operational condition 1 (POWER OPERATION) at 35 percent reactor power, the licensee experienced an unexpected single rod scram. During the performance of Main Turbine Cnntrol and Stop Valve surveillance testing, a one-half scram signal was initiated in accordance with the surveillance instruction. When the hcif scram signal was initiated, control rod 06-23 scrammed from position 12 to position 00. Plant operators identified that the "B" single rod insertion test switch for control rod 06-23 was left in test after an earlier rod scram time tes Therefore, when the Turbine Valve test inscrted a half scram on Channel "A," both the "A" and "B" scram solenoids on control rod 06-23 were deenergized. Plant operators returned th? "B" scram solencid test switch to normal, reset the half scram signal and returned control rod 06-23 to position 12. The licensee informed the NRC operations center of this event via the ENS at about 7:00 a.m. on August 7, 196 No violations or deviations were identifie . Plant Status Meetings (30702)
NRC management met with CEI management on July 17, 1969, at the Perry Power Plant, in order to discuss the current status of the Refueling Outage, recent events, and licensee initiatives to improve the quality of plant cperatit.g and maintenance activities. These meetings are being held on a periccic (initially monthly) basi . Violations For Which A " Notice of Violation" Will Not Be Issued The NRC uses the Notice of Violation as a standard method for formal' zing the existence of a violation of a legally binding requirement. However, because the NRC wants to encourage and support licensee's initiatives for self-identification and ccrrection of problems, the NRC will not generally issue a Notice of Violation for a violation that meets the tests of 10 CFR 2, Appendix C, Section V.A. These tests are: (1) the violetion was identified by the licensce; (2) the violation would be categorized as Severity Level IV or V; (3) the violation was reported to the NRC, if required; (4) the violation will be corrected, including measures to prevent recurrence, within a reasonabic time period; and (5)
it was nct a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previcus violatio A violation of regulatory requirements identified during the inspection for which a Nctice of Violation will not be issued is discussed in Paragraph l
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. '1 Exit Interviews (30703) l 1he inspectors met with the licensee representatives denoted in Paragraph 1. throughout the inspection period and on August 11, 198 ]
The inspectors sumarized the scope and results of the inspection and i discussed the likely content of the inspection report. The licensee did not indicate that any of the information disclosed during the inspection could be considered proprietary in natur I I
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