IR 05000263/2007006
| ML072960422 | |
| Person / Time | |
|---|---|
| Site: | Monticello, Perry  |
| Issue date: | 10/23/2007 |
| From: | Dave Hills NRC/RGN-III/DRS/EB1 |
| To: | Allen B FirstEnergy Nuclear Operating Co |
| References | |
| IR-07-006 | |
| Download: ML072960422 (19) | |
Text
October 23, 2007Mr. Barry S. AllenSite Vice President FirstEnergy Nuclear Operating Company Perry Nuclear Power Plant P. O. Box 97, 10 Center Road, A-PY-290 Perry, OH 44081-0097SUBJECT:PERRY NUCLEAR POWER PLANT NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000440/2007006(DRS)
Dear Mr. Allen:
On September 14, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed acombined baseline inspection of the Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications at the Perry Nuclear Power Plant. The enclosed report documents the results of the inspection, which were discussed with Mr. K. Krueger, and others of your staff at the completion of the inspection on September 14, 2007.The inspectors examined activities conducted under your license as they relate to safety andcompliance with the Commission's rules and regulations, and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Based on the results of the inspection, no findings of significance were identified.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter andits enclosure will be available electronically for public inspection in the NRC Public DocumentRoom or from the Publicly Available Records System (PARS) component of NRC's Agencywide B. Allen-2-Documents Access and Management System (ADAMS), accessible from the NRC Web site athttp://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/David E. Hills, ChiefEngineering Branch 1 Division of Reactor SafetyDocket No. 50-440License No. NPF-58Enclosure:Inspection Report 05000440/2007006(DRS) w/Attachment: Supplemental Informationcc w/encl:J. Hagan, President and Chief Nuclear Officer - FENOCJ. Lash, Senior Vice President of Operations and Chief Operating Officer - FENOC D. Pace, Senior Vice President, Fleet Engineering - FENOC J. Rinckel, Vice President, Fleet Oversight - FENOC R. Anderson, Vice President, Nuclear Support - FENOC Director, Fleet Regulatory Affairs - FENOC Manager, Fleet Licensing - FENOC Manager, Site Regulatory Compliance - FENOC D. Jenkins, Attorney, FirstEnergy Corp.
Public Utilities Commission of Ohio Ohio State Liaison Officer R. Owen, Ohio Department of Health
SUMMARY OF FINDINGS
IR 05000263/2007006(DRS); 08/27/2007 through 09/14/2007; Perry Nuclear Power Plant. Routine engineering baseline inspection.The inspection covered a 2-week announced baseline inspection on evaluations of changes,tests, or experiments and permanent plant modifications. The inspection was conducted by two regional based engineering inspectors. No violations were identified. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3; dated July 2000.A.Inspector-Identified and Self-Revealed FindingsNo findings of significance were identified.
B.Licensee-Identified Violations
No findings of significance were identified.
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REPORT DETAILS
1.REACTOR SAFETYCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R02Evaluations of Changes, Tests, or Experiments (71111.02).1Review of 10 CFR 50.59 Evaluations and Screenings
a. Inspection Scope
From August 27, 2007, through September 14, 2007, the inspectors reviewed sevensafety evaluations (SEs) performed pursuant to 10 CFR 50.59. The inspectors reviewed the evaluations to confirm that they were thorough and that prior NRC approval was obtained as appropriate. The inspectors also reviewed 15 screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. In addition, two applicability determinations (RADs) were reviewed to verify they did not meet the applicability requirements for a screening. The evaluations and screenings (24 samples) were chosen based on risk significance, safety significance, and complexity.
The list of documents reviewed by the inspectors are included as an attachment to this report.The inspectors used, in part, Nuclear Energy Institute (NEI) 96-07, "Guidelines for 10 CFR 50.59 Implementation," and Revision 1, to determine acceptability of the completed evaluations, and screenings. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments," dated November 2000. The inspectors also consulted Part 9900 of the NRC Inspection Manual, "10 CFR Guidance for 10 CFR 50.59, Changes, Tests, and Experiments."
b. Findings
b.1Concerns Regarding the Licensee's Evaluation and Actions to Address a ConditionAffecting the Emergency Service Water (ESW) Pump and Its Associated DischargeValveIntroduction: The inspectors identified an Unresolved Item (URI) involving the licensee'sevaluation and actions to address a condition that affected ESW pump A and its associated discharge valve. Specifically, the licensee did not to ensure that the ESW pump A discharge valve (1P45F0130A) would remain open when the pump was in operation during the loss of direct current (DC) Bus ED-1-A. In addition, the licensee did not to identify and evaluate the impact of this condition on the plant's safe shutdown equipment in the event of an Appendix R fire in the control room. This issue is unresolved pending further NRC review of the licensee's evaluation for the impact of this condition on the safe shutdown during a fire.
3Description: The licensee initiated Condition Report (CR) 06-03087 in July of 2006,which identified a condition that affected ESW pump A and its associated discharge valve (1P45F0130A). Specifically, the licensee identified that if ESW pump A was in operation when a loss of DC Bus ED-1-A occurred, then the discharge valve 1P45F0130A would automatically close while the pump continued to run. The pump breaker would not trip because of the loss of it's DC control power. In addition, remote operation of the breaker from the control room would also be lost. Therefore, the ESW pump A would continue to run at shutoff head until operator action was taken to open the breaker locally. At the time of discovery, there was no procedural guidance in ONI-R42-1 "Loss of DC Bus ED-1-A," to direct the operators to shutdown the pump.The licensee reviewed the ESW pump and discharge valve electrical drawingsB-208-176, sheets 1 and 4, which indicated that a loss of DC control circuit for ESW pump A would result in de-energizing control relay 1P45-K8 that would consequently initiate a close signal to the discharge valve. The licensee determined that this specific condition/scenario did not exist for the ESW pump B as no close signal would be initiated for the discharge valve. The CR evaluation also indicated that if an open signal was present as it would be in the case of the division 1 emergency diesel generator (EDG)running or a division 1 loss-of-coolant-accident (LOCA) signal, the discharge valve wouldcycle open and closed until the operator took manual control of the valve. In the case of no open signal being present, the discharge valve would close and remain closed until manual control was taken to reopen the valve. However, due to the loss of DC power, once the valve was fully opened it would automatically cycle closed. As a result of the CR evaluation, procedure ONI-42-1 was revised to add steps that directed the operators to repeatedly open the ESW pump A discharge valve until an operator removed control relay 1P45-K8 at panel 1H13-P872. The procedure also directed the operators to trip the ESW pump A locally, if the pump was not required to support equipment operation.During the inspectors' review of Screening 06-03964 for the revision to procedureONI-R42-1, the inspectors questioned if there were any limitations on the number of times that the discharge valve could be cycled/stroked. Based on information specified in FTI-F0016, "Motor Operated Valve Diagnostic Testing," the licensee indicated that there was a limit of five times per every 5 minutes. Since test results indicated the valve strokes in approximately 30 seconds, the inspectors were concerned that the discharge valve motor could be damaged, while the valve was in the closed position because of the repeated cycling either by the operator action or automatically due to an EDG start or LOCA signal.The inspectors were also concerned that the licensee did not evaluate thiscondition/scenario impact on the Appendix R safe shutdown equipment and analysis.
Specifically, the inspectors were concerned regarding the operability and reliability of division 1 EDG and its associated ESW pump in the event of a fire in the control room.
The control room was designated as an alternate shutdown area where control room evacuation may be required and shutdown from outside the control room using the alternate shutdown panel may be required. Assuming a loss of offsite power (LOOP)that would result in starting the division 1 EDG and its associated ESW pump A, a subsequent loss of DC control power for the ESW pump due to fire-induced failures (i.e., loss of fuse) that could occur prior to control room isolation at the alternate shutdown panel, would have caused the ESW discharge valve to close. Operating 4under no flow conditions could result in damage to the ESW pump A and/or damage tothe division 1 EDG due to the loss of cooling provided by ESW. Following identification of this issue, the licensee entered this issue into their corrective action program as CR 07-26412. This issue is unresolved pending further NRC review of Perry's Appendix R evaluation for a control room fire and the affect of the above scenario on safe shutdown equipment. (URI 05000440/2007006-01 (DRS)) b.2Emergency Diesel Generators Non-Critical Trips Bypass Circuits ModificationIntroduction: The inspectors identified an Unresolved Item (URI) involving theadequacy of a 10 CFR 50.59 safety evaluation, which was completed for Engineering Change Package (ECP) 05-0229, "Division 1 and 2 Emergency Diesel Generator (EDG)
Bus Under/Degraded Voltage Start Logic Modification." Specifically, the inspectors questioned the adequacy of the licensee's basis for determining that changes implemented per ECP 05-0229, to bypass the non-critical trips for the EDGs on a bus under/degraded voltage (LOOP) signal initiation, did not require a license amendment.
This issue is unresolved pending further NRC review of Perry's licensing basis and theimpact of this modification on Perry's accident analysis.Description: A previous NRC Non-Cited Violation 05000440/2005003-14, "Failure toAdequately Address EDG Design Concern," identified an issue that involved the inabilityof the EDGs to re-start either manually or upon receipt of a bus under/degraded voltage signal within a 2-minute period following an EDG shutdown after surveillance testing.
The licensee implemented modification ECP 05-0229, which relocated the EDG bus under/degraded voltage start signal from the non-emergency start logic (manual start)to the LOCA start logic. Additionally, the licensee added seal-in relays to lock-in the bus under/degraded voltage EDG start signal. The seal-in relays were required due to the bus under/degraded voltage relays de-energizing after the bus was re-powered by the EDG.As a result of relocating the bus under/degraded voltage start signal, the EDG operationwas altered as follows:*During a bus under/degraded voltage start signal, the associated EDG startsignal will override testing activities, specifically the 2-minute period after EDG shutdown.*During a bus under/degraded voltage event, the associated EDG non-criticaltrips would be bypassed. These trips would continue to provide alarms on abnormal engine conditions to alert the operator of the engine abnormal condition, but would not trip the EDG. Prior to the implementation of this modification, the non-essential trips were not bypassed under these conditions.*If both EDG air start headers decreased to 150 psig, the associated EDGunder/degraded voltage start signal would be blocked. This would ensure that sufficient air remained for an additional manual start attempt. The under/degraded voltage EDG air start logic would be identical to the existing LOCA air start logic. Prior to the implementation of this modification, the EDGs circuit design included a 10 second time delay to allow for five cranks to start the 5EDG before it timed out and blocked the under/degraded voltage start signal.The licensee completed a 10 CFR 50.59 Applicability Review and 10 CFR 50.59SE 06-00185 and concluded that the modification did not affect the Technical Specifications, nor required a license amendment. The licensee evaluation based this conclusion on the following:*The EDGs would continue to be available to support the shutdown equipmenteven if a non-critical trip device spuriously actuated as the trip function was bypassed. *The alarm would still be activated in the main control room to alert the operatorsto manually shutdown the EDG, if desired. *The non-critical trips that were bypassed by this modification provided protectionof the commercial investment. *The EDG design basis accommodated the single failure criterion by theindependent diesel generator division. Once a EDG incurs a failure, the independent division functions to mitigate the accident/event. No credit for recovery of the failed EDG existed in the Perry's accident analysis. Therefore, the licensee concluded that overall probability of occurrence of an equipmentmalfunction had not been increased. During the inspectors' review of SE 06-00185, Revision 1, the inspectors questioned the adequacy of the basis for the licensee's conclusion and justification. Specifically, the inspectors were concerned that licenseedid not provide adequate justification to why the activity to bypass the non-critical trips on an under/degraded voltage start signal (non-accident condition), did not result in more than minimal increase in the likelihood of occurrence of a malfunction of the EDGs. The inspectors were concerned that the licensee relied on operator actions to secure the EDGs instead of automatic action. The inspectors were concerned that the EDG may be damaged prior to actions by the operators to secure it. The inspectors used the scenario in the previous section as an example/scenario during a discussion with the licensee to convey the concern. Specifically, prior to this modification the EDG would have automatically tripped due to high temperature in the event of lack of cooling for the EDG due to ESW discharge valve closing. However, after the modification, the EDG will continue to run until the operator manually stops it. This resulted in increasing the likelihood of damaging the EDG in a non-accident conditions. Furthermore, Technical Specification surveillance 3.8.1.13 required verifying that eachEDG's automatic trips were bypassed on actual or simulated emergency core coolant system initiation signal except for engine overspeed and generator differential current.
The inspectors were concerned that design changes implemented per ECP 05-0229 have affected this surveillance requirement and a license amendment was required to revise the Technical Specification. Based on the licensee conclusion in the 10 CFR 50.59, the licensee implemented thedesign modification and planned to revise the affected sections of the Updated Final Safety Analysis Report and Technical Specification Bases without a license amendment to indicate that the EDG's non-critical trips were bypassed during design basis accidents and bus under/degraded voltage conditions. Following identification of this issue, the licensee entered the inspectors' concerns into their corrective action program as 6CR 07-26346. The inspectors also discussed this issue briefly with the Office of NuclearReactor Regulation (NRR), in which NRR staff initially supported the inspectors' positionthat a license amendment was required to revise the Technical Specification surveillance requirement, but indicated that further evaluation was required. This issue is unresolved pending further NRC review of Perry's licensing basis to determine if the activity implemented per modification ECP 05-0229 did result in more than minimal increase in the likelihood of occurrence of a malfunction of the EDGs and to evaluate if the activity has also affected the Technical Specification surveillance requirement 3.8.1.13. Specifically to evaluate if the design change that bypassed the non-critical trips on an under/degraded voltage start signal has not affected the accident analysis and that the existing Technical Specification surveillance assures that the surveillance for the EDGs still meet the requirement of 10 CFR 50.36. (URI 05000440/2007006-02(DRS))1R17Permanent Plant Modifications (71111.17B).1Review of Permanent Plant Modifications
a. Inspection Scope
From August 27, 2007, through September 14, 2007, the inspectors reviewed sixpermanent plant modifications (six samples) that had been installed in the plant during the last two years. The modifications were chosen based upon risk significance, safety significance, and complexity. As per inspection procedure 71111.17B, two modifications were chosen that affected the barrier integrity cornerstone. The inspectors reviewed the modifications to verify that the completed design changes were in accordance with the specified design requirements, and the licensing bases, and to confirm that the changes did not adversely affect any systems' safety function. Design and post-modification testing aspects were verified to ensure the functionality of the modification, its associated system, and any support systems. The inspectors also verified that the modifications performed did not place the plant in an increased risk configuration.The inspectors also used applicable industry standards to evaluate acceptability of themodifications. The list of modifications and other documents reviewed by the inspectors is included as an attachment to this report.
b. Findings
No findings of significance were identified.4.OTHER ACTIVITIES (OA)4OA2Identification and Resolution of Problems.1Routine Review of Condition Reports
a. Inspection Scope
From August 27, 2007, through September 14, 2007, the inspectors reviewed 9 7Corrective Action Process documents that identified or were related to 10 CFR 50.59evaluations and permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent plant modifications and evaluations for changes, tests, or experiments issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problems into the corrective action system. The specific corrective action documents that were sampled and reviewed by the inspectors are listed in the attachment to this report.
b. Findings
No findings of significance were identified.4OA6Meetings.1Exit MeetingThe inspectors presented the inspection results to Mr. K. Krueger and others of thelicensee's staff, on September 14, 2007. Licensee personnel acknowledged the inspection results presented. Licensee personnel were asked to identify any documents, materials, or information provided during the inspection that were considered proprietary. No proprietary information was identified.ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- K. Krueger, Plant General Manager
- D. Evans, Operations Manager
- B. Ferrell, FENOC Licensing
- D. Gartner, Design Engineer
- T. Hilston, Design Engineering Manager
- M. Koberling, Plant and Equipment Reliability Manager
- J. Lausberg, Regulatory Compliance Manager
- K. Russell, Regulatory Compliance
- J. Shaw, Engineering Director
- S. Seman, Plant Engineering
- A. Watkins, Design EngineerNuclear Regulatory Commission
- M. Franke, Senior Resident Inspector
- J. Robbins, Acting Resident Inspector
- M. Wilk, Resident Inspector
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened05000440/2007006-01URIFailure to Adequately Correct and Evaluate a ConditionAffected the ESW Pump and Its Associated Discharge
Valves (Section 1R21.3.b.1)05000440/2007006-02URIEmergency Diesel Generators Non-Critical Trips BypassCircuits Modification (Section 1R21.3.b.2)
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LIST OF DOCUMENTS REVIEWED
The following is a list of licensee documents reviewed during the inspection, includingdocuments prepared by others for the licensee.
- Inclusion on this list does not imply that NRC