IR 05000440/2001016

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IR 05000440/2001-016 on 01/1-02/17/2002; First Energy Nuclear Operating Company; Perry Nuclear Power Plant. Temporary Modifications
ML020770485
Person / Time
Site: Perry FirstEnergy icon.png
Issue date: 03/14/2002
From: Christine Lipa
NRC/RGN-III/DRP/RPB4
To: Campbell G
FirstEnergy Nuclear Operating Co
References
IR-01-016
Download: ML020770485 (24)
v • d • e


Text

rch 14, 2002

SUBJECT:

PERRY NUCLEAR POWER PLANT NRC INSPECTION REPORT 50-440/01-16

Dear Mr. Campbell:

On February 17, 2002, the NRC completed an inspection at your Perry Nuclear Power Plant. The enclosed report documents the inspection findings which were discussed on February 26, 2002, with you and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions 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 this inspection, the inspectors identified two issues of very low safety significance (Green) that were determined to involve a violation of NRC requirements.

However, because of their very low safety significance and because they were entered into your corrective action program, the NRC is treating these issues as Non-Cited Violations in accordance with Section VI.A.1 of the NRC s Enforcement Policy. If you deny these Non-Cited Violations, you should provide a response with a basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Perry Nuclear Power Plant. In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/NRC/ADAMS/index.html (the Public Electronic Reading Room).

Sincerely,

/RA/Christine A. Lipa Christine A. Lipa, Chief Branch 4 Division of Reactor Projects Docket No. 50-440 License No. NPF-58

Enclosure:

Inspection Report 50-440/01-16

REGION III==

Docket No: 50-440 License No: NPF-58 Report No: 50-440/01-16 Licensee: FirstEnergy Nuclear Operating Company (FENOC)

Facility: Perry Nuclear Power Plant, Unit 1 Location: P.O. Box 97 A200 Perry, OH 44081 Dates: January 1, 2002 through February 17, 2002 Inspectors: Ray Powell, Senior Resident Inspector John Ellegood, Resident Inspector Steve Campbell, Senior Resident Inspector, Fermi Robert Jickling, Emergency Preparedness Analyst Approved by: Christine A. Lipa, Chief Branch 4 Division of Reactor Projects

SUMMARY OF FINDINGS IR 05000440-01-16; on 01/1-02/17/2002; First Energy Nuclear Operating Company; Perry Nuclear Power Plant. Temporary Modifications.

This report covers a 7-week routine inspection. The inspection was conducted by resident inspectors and a regional inspector. Two findings of very low risk significance were identified during this inspection and were considered to be Non-Cited Violations. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, Significance Determination Process (SDP). The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described at its Reactor Oversight Process website at: http://www.nrc.gov/NRR/OVERSIGHT/index.html. Findings for which the SDP does not apply are indicated by No Color or by the severity level of the applicable violations.

A. Inspection Findings Cornerstone: Initiating Events GREEN. The inspectors identified a Non-Cited Violation of 10CFR50 Appendix B, Criterion III, for failure to remove temporary lighting from the reactor water cleanup room after one-cycle as required by Field Clarification Request. The lights eventually degraded and caught fire.

The finding was greater than minor because it had an actual impact of causing a small fire in a room containing plant operating, fire protection and safety-related equipment.

The event was of very low safety significance because, although the finding contributed to the likelihood of an external event initiator, no equipment was damaged from the event. (Section 1R23.1).

Cornerstone: Mitigating Systems GREEN. The inspectors identified a Non-Cited Violation of 10CFR50 Appendix B, Criterion V, for failing to follow plant procedures to maintain electrical separation between Class 1E and Non-class 1E cables and conduits.

The finding was greater than minor because if left uncorrected, routing the extension cords near safety-related power cables increased the likelihood of rendering multiple trains of safety-related equipment inoperable given a fire from those temporary cables.

Further, the multiple examples of violating the electrical separation criteria indicated a lack of plant personnel knowledge of the requirement. The finding was of low safety significance because an actual fire had not occurred that rendered the associated equipment unavailable. (Section 1R23.2).

B. Licensee Identified Violations None

Report Details Summary of Plant Status: The plant began the inspection period with Unit 1 at 100 percent power. The unit remained at 100 percent power until January 19, 2001 when power was reduced to 60 percent for condenser tube plugging. The unit was returned to 100 percent power on January 21. The unit remained at 100 percent power until January 27 when power was reduced to 80 percent for control rod alignment. The unit returned to 100 percent power later that same day. On February 10, the licensee reduced power to 90 percent in order to recover an inadvertently scrammed control rod. The rod scrammed because a fuse had blown on one of its reactor protection system scram solenoids and a second channel was tripped as part of surveillance testing. Following rod recovery, the unit returned to 100 percent power.

Power effectively remained at 100 percent for the remainder of the inspection period.

1. REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity 1R04 Equipment Alignment (71111.04Q)

a. Inspection Scope The inspectors conducted a partial alignment walkdown of the Division 2 Emergency Service Water (ESW), a risk important system, to evaluate its readiness while the Division 1 train was declared inoperable due to ESW pumphouse ventilation maintenance. The walkdown included selected switch and valve position checks, review of associated effective operating procedures, and verification of electrical power to critical components. The inspectors reviewed sections of the Updated Safety Analysis Report (USAR) and Technical Specifications (TS) as applicable to the walkdown. The documents used for the walkdown are listed in the attached List of Documents Reviewed.

b. Findings No findings of significance were identified.

1R05 Fire Protection (71111.05Q)

a. Inspection Scope The inspectors walked down the following areas to assess the overall readiness of fire protection equipment and barriers:

  • Fire Zone 1CC-4a, Unit 1, Division 2 Cable Spreading Area
  • Fire Zone 1CC-4e, Unit 1, Division 1 Cable Spreading Area

Emphasis was placed on the control of transient combustibles and ignition sources, the material condition of fire protection equipment, and the material condition and operational status of fire barriers used to prevent fire damage or propagation.

The inspectors looked at fire hoses, sprinklers, and portable fire extinguishers to verify that they were installed at their designated locations, were in satisfactory physical condition, and were unobstructed. The inspectors also evaluated the physical location and condition of fire detection devices. Additionally, passive features such as fire doors, fire dampers, and mechanical and electrical penetration seals were inspected to verify that they were in good physical condition. Finally, the inspectors toured the reactor water heat exchanger room to assess the extent of damage to components in the room following the January 7, 2002 lighting string fire. The documents listed at the end of the report were used by the inspectors during the assessment of this area.

b. Findings No findings of significance were identified. Circumstances that caused the lighting string fire in the reactor water cleanup heat exchanger room are discussed in Section 1R23 of this report.

1R11 Licensed Operator Requalification (71111.11)

a. Inspection Scope On February 5, 2002, the resident inspectors observed licensed operator performance in the plant simulator. The evaluated scenario included severe weather, a pressure regulator failure, and a loss of reactor water level indication.

The inspectors evaluated crew performance for clarity and formality of communication; the ability to take timely action in the safe direction; the prioritizing, interpreting, and verifying of alarms; the correct use and implementation of procedures, including alarm response procedures; timely control board operation and manipulation, including high-risk operator actions; and group dynamics. The inspectors also observed the licensees evaluation of crew performance to verify that the training staff had observed important performance deficiencies and specified appropriate remedial actions.

a. Findings No findings of significance were identified.

1R12 Maintenance Rule Implementation (71111.12Q)

a. Inspection Scope The inspectors reviewed the licensee's implementation of the maintenance rule requirements to verify that component and equipment failures were identified, entered, and scoped within the maintenance rule and that select structures, systems and components were properly categorized and classified as (a)(1) or (a)(2) in accordance with 10 CFR 50.65. The inspectors reviewed station logs, maintenance work orders,

selected surveillance test procedures, and a sample of Condition Reports (CRs) to verify that the licensee was identifying issues related to the maintenance rule at an appropriate threshold and that corrective actions were appropriate. Additionally, the inspectors reviewed the licensees performance criteria to verify that the criteria adequately monitored equipment performance and to verify that licensee changes to performance criteria were reflected in the licensees probabilistic risk assessment.

During this inspection period, the inspectors reviewed:

  • Feedwater System The problem identification and resolution condition reports (CR) reviewed are listed in the attached List of Documents Reviewed.

b. Findings No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation (71111.13)

a. Inspection Scope The inspectors reviewed the licensees evaluation of plant risk, scheduling, configuration control, and performance of maintenance associated with planned and emergent work activities, to verify that scheduled and emergent work activities were adequately managed. In particular, the inspectors reviewed the licensees program for conducting maintenance risk assessments to verify that the licensees planning, risk management tools, and the assessment and management of on-line risk were adequate. The inspectors also reviewed licensee actions to address increased on-line risk when equipment was out-of-service for maintenance, such as establishing compensatory actions, minimizing the duration of the activity, obtaining appropriate management approval, and informing appropriate plant staff, to verify that the actions were accomplished when on-line risk was increased due to maintenance on risk-significant structures, systems, and components. The following specific activities were reviewed:

  • The maintenance risk assessment for replacement of static inverter 1R41K0090A on January 23, 2002. The work was risk significant due to the resulting Division 1 Emergency Diesel Generator (EDG) unavailability.
  • The maintenance risk assessment for ESW pumphouse A train ventilation system rework which occurred from January 19 through February 3, 2002. The work was potentially risk significant due to potential impacts on ESW system operability.
  • The maintenance risk assessment for planned Division 3 EDG ventilation system work. The inspectors verified the impact of the work on EDG availability was appropriately characterized.

b. Findings No findings of significance were identified.

1R14 Personnel Performance During Non-Routine Plant Evolutions (71111.14)

.1 Personnel Response to Smoke in Containment a. Inspection Scope On January 7, 2002, the licensee identified smoke in containment. The resulting investigation discovered a smoldering lighting string in a locked high radiation area - the reactor water cleanup heat exchanger room. The inspectors reviewed personnel performance including fire brigade and operator response to determine if operators had entered off-normal instructions properly. The inspectors reviewed procedures to determine whether the condition was reportable and whether the event should have been classified as an unusual event.

b. Findings No findings of significance were identified. Circumstances that caused the lighting string fire in the reactor water cleanup heat exchanger room are discussed in Section 1R23 of this report.

.2 Personnel Response to Scrammed Rod During Surveillance Testing a. Inspection Scope The inspectors evaluated operator response to a single scrammed rod which occurred during reactor protection system manual scram channel functionality testing on February 10, 2002. The rod scrammed when operators manually initiated the B channel scram signal while a fuse for the rods A channel scram solenoid was blown.

The inspectors evaluated operator performance to verify that actions were taken in a timely manner in accordance with Off-Normal Instruction (ONI) C51, Unplanned Change in Reactor Power or Reactivity, Rev. 8 and that rod recovery actions were appropriate. Additionally, the inspectors reviewed the associated condition report CR 0-0416, Rod 38-43 Inserted During SVI-C71-T0051. Finally, the inspectors reviewed procedures to determine whether the event was reportable b. Findings No findings of significance were identified.

1R15 Operability Evaluations (71111.15)

a. Inspection Scope The inspectors reviewed the operability determination associated with ESW pump house ventilation subsystem unavailability. The inspectors reviewed the licensees evaluation,

as documented in CR 02-0329, that for ambient conditions existing at the time, the ESW system remained operable during maintenance on the pump house ventilation subsystem.

The inspectors reviewed the operability determination performed for CR 02-0151. The CR described an issue involving potential problems associated with the computer program utilized to evaluate heat exchanger performance. The inspectors reviewed the licensees evaluation that the potential errors were bounded by analytical uncertainty compensation.

b. Findings No findings of significance were identified.

1R16 Operator Workarounds (OWAs)

.1 Nuclear Island Radiologically Restricted Area (RRA) Operator Rounds Accompaniment a. Inspection Scope The inspectors accompanied a plant operator, nuclear island RRA, during the performance of a normal rounds tour on February 25, 2002. The inspectors observed all log readings and equipment manipulations made by the operator. Any actions which indicated a potential problem that could increase initiating event frequencies, impact multiple mitigating systems, or affect the ability to respond to plant transients and accidents were considered as possible OWAs. Additionally, the inspectors discussed the effect of active OWAs with the operator.

b. Findings No findings of significance were identified.

.2 Nuclear Island Non-RRA Operator Rounds Accompaniment a. Inspection Scope The inspectors accompanied a plant operator, nuclear island non-RRA, during the performance of a normal rounds tour on February 24, 2002. The inspectors observed all log readings and equipment manipulations made by the operator. Any actions which indicated a potential problem that could increase initiating event frequencies, impact multiple mitigating systems, or affect the ability to respond to plant transients and accidents were considered as possible OWAs.

b. Findings No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19)

a. Inspection Scope The inspectors evaluated the following post-maintenance testing activities for risk significant systems to assess the following (as applicable): the effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed; acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate; tests were performed as written; and equipment was returned to its operational status following testing. The inspectors evaluated the activities against TS, the USAR, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications. In addition, the inspectors reviewed CRs associated with post-maintenance testing to determine if the licensee was identifying problems and entering them in the corrective action program. The specific procedures and CRs reviewed are listed in the attached List of Documents Reviewed. The specific post-maintenance activities evaluated included:

  • Diesel Fire Pump Operability Test following planned maintenance
  • ESW Pump House Ventilation System Train A Damper Stroking following planned maintenance b. Findings No findings of significance were identified.

1R22 Surveillance Testing (71111.22)

a. Inspection Scope The inspectors observed surveillance testing or reviewed test data for risk-significant systems or components to assess compliance with TS, 10 CFR Part 50 Appendix B, and licensee procedure requirements. The testing was also evaluated for consistency with the USAR. The inspectors verified that the testing demonstrated that the systems were ready to perform their intended safety functions. The inspectors reviewed whether test control was properly coordinated with the control room and performed in the sequence specified in the surveillance instruction, and if test equipment was properly calibrated and installed to support the surveillance tests. The procedures reviewed are listed in the attached List of Documents Reviewed. The specific surveillance activities assessed included:

  • Surveillance Instruction (SVI) E22-T1319, Diesel Generator Start and Load Division 3
  • Periodic Test Instruction (PTI) P54-P0045, Fire Detection Instrumentation Functional Test b. Findings No findings of significance were identified.

1R23 Temporary Modifications (71111.23)

.1 Temporary Lighting Overheats, Causing Smoke and Fire in Containment a. Inspection Scope The inspectors reviewed the engineering justification that allowed a string of temporary lights to be installed in the reactor water cleanup (RWCU) heat exchanger room within the containment. The power cord was part of the lighting system that smoked and caught fire on January 7, 2002. The inspectors reviewed engineering procedures, the licensees root cause determination that was documented in CR 02-0057, and conducted interviews with station personnel.

b. Findings GREEN. A Non-Cited Violation (NCV) of 10 CFR 50 Appendix B, Criterion III for not removing a string of temporary lights from the RWCU heat exchanger room after one cycle as specified on Field Clarification Request (FCR) 24726. A temporary 100-foot string of 10 lights was affixed to scaffolding using plastic tie-wraps and installed during Refueling Outage 6 (ended October 23, 1997) in the RWCU heat exchanger room. The scaffolding also supported lead blankets, used to reduce the radiation dose rate around the reactor water cleanup heat exchangers. Since the permanent lights had been disabled, site personnel requested that the temporary lights remain in the room for at least one cycle. An engineer wrote a design change via FCR 024726, "Temporary Lighting in the RWCU Heat Exchanger Room," to justify leaving the lights installed for one cycle.

Field Clarification Request 024726 evaluated the lights under electrical, environmental qualification, mechanical and fire protection considerations. Of particular note was that the environmental qualification evaluation focused on seismic concerns regarding decreased strength from radiation exposure of the plastic tie wraps supporting the lighting. The FCR did not consider radiation aging of the rubber conductor insulation, which, when subjected to high radiation fields, ages and degrades over time. The highest dose in the room exceeded 10 rads.

The engineer who wrote the FCR evaluated installation of these lights for a limited time and required removal of the lights after one cycle. The cycle ended when Refueling Outage 7 started March 27, 1999. Removal of the lights was not tracked in the design change process and the temporary lights remained in the room greater than one cycle.

The room contains the following safety related and shutdown components:

  • Power cables for reactor and remote shutdown panel pressure and level transmitters and indications
  • Power cables for drywell pressure transmitters
  • Power cables for suppression pool temperature indication
  • Division 1 Non regenerative heat exchanger tube outlet temperature element
  • Division 1 Leak Detection Temperature Elements
  • Division 2 Leak Detection Temperature Elements
  • Division 2 Hydrogen Igniters These components were in the overhead of the reactor water cleanup heat exchanger room and were not damaged by the fire.

On January 7, 2002, at about 3:00 p.m., a radiation protection technician plugged the lights into a receptacle to illuminate the room so the technician and an operator, who assisted, could conduct a leak inspection inside the heat exchanger room. After completing the inspection, the technician did not unplug the lights, rather, he told decontamination personnel in the area that the lights remained powered. Based on interviews, he assumed they would unplug the fixture. Unfortunately, the decontamination personnel thought the operator would request the oncoming crew to unplug the lights. Therefore, the lighting string remained energized and caught fire.

The cord smoldered and burned for several hours until the fire brigade unplugged the lights and discharged a fire extinguisher on the lighting string. Other than heat damage to temporary lead shielding and the lighting string, no equipment damage occurred from the fire.

The performance deficiency associated with this event was that the design control process to remove the temporary lighting after one-cycle as specified on FCR 024726 was inadequate. The finding was greater than minor because it had an actual impact of causing a small fire in a room containing plant operating, fire protection and safety-related equipment. Using the SDP phase 1 worksheet for the seismic, fire, flooding, and severe weather screening criteria, the finding was of very low safety significance because, although the finding contributed to the likelihood of an external event initiator, no equipment was damaged from the event. 10 CFR 50, Appendix B, Criterion III, states, in part, that measures shall be established for the identification and control of design interfaces and for coordination between participating organizations and that design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design. Contrary to 10 CFR 50 Appendix B, Criterion III, no design control process (coordination among organizations or established procedures) existed to remove the temporary lights from the reactor water cleanup heat exchanger room after one-cycle as required by FCR 024726. However, because of the very low safety significance and because the issue is in the licensees corrective action program, it is being treated as a Non-Cited Violation, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 50-440/01-16-01). This violation is in the licensee's corrective action program as CR 02-00057.

.2 Electrical Separation Criteria Involving Extension Cords a. Inspection Scope Following the temporary lighting fire in the reactor water cleanup heat exchanger room (Section 1R23.1), the inspectors reviewed industry events involving the routing of extension cords. The inspectors reviewed the licensees design specifications for

maintaining separation criteria between Class 1E and Non-Class 1E power cables. The inspectors conducted tours of the reactor building to determine if the licensee maintained electrical separation between safety-related power sources and extension cords.

b. Findings GREEN. A Non-Cited Violation of 10 CFR 50 Appendix B, Criterion V for failing to implement the following: 1) Plant Administrative Procedure (PAP)-0204, Housekeeping/Cleanliness Control, 2) Drawing D-214-004, Conduit and Cable Tray Separation Criteria, and, 3) Installation Specification 2250, Electrical Work and Equipment Specification, for installing temporary power cables in a manner to prevent violating the electrical separation criteria.

On January 8, 2002, the inspectors reviewed the NRC event database and found Event 33314 (dated November 26, 1997) that described a condition at the Pilgrim Nuclear Plant regarding extension cords being draped over or tie-wrapped to Class 1E Conduits in the reactor building, in violation of electrical separation criteria.

Subsequently, the inspectors reviewed the licensees investigation of the fire caused by a light string in the reactor water cleanup heat exchanger room (Section 1R23.1). The licensees report documented a walkdown of the plant to determine if other temporary power cables were overheating. This walkdown did not include evaluation of electrical separation criteria.

The inspectors toured the plant and found two temporary power cords routed in violation of separation requirements. The first cord powered a portable airborne radiation monitor. This power cord and ran from a wall receptacle and routed around a safety-related electrical conduit in the reactor core isolation cooling system room. The power cord was routed within one-inch of conduits containing power cables for the open and close limit switches and solenoid for air operated steam supply drain pot drain line valve 1E51F026. During standby conditions, the valve remains open to drain condensate from a drain pot on the steam line entering the Reactor Core Isolation Cooling (RCIC) turbine. This fail-close valve closes on RCIC startup to prevent diverting steam from the RCIC turbine to the condenser.

The inspectors also identified a temporary power cord draped over safety related cable trays near the ceiling on the 620 foot level of the intermediate building. Safety-related Cable Trays A 660 and A 147 were closest to the power cord and contained power cables for the following systems:

  • Redundant Reactor Control System
  • Airborne Radiation Monitor System
  • Fuel Pool Cleanup and Cooling System
  • Containment Vessel and Drywell Purge
  • Annulus Exhaust Gas Treatment System
  • Fire Protection System
  • 480 Volt Alternating Current Electrical Distribution System
  • Uninterruptible Power Supply
  • Direct Current Electrical The licensee initiated CR 02-00091 to document the inspectors findings and promptly corrected the conditions.

The licensee conducted a walk down of the plant to identify other examples of violations of electrical separation criteria. Condition Report 02-00069 was initiated when the licensee found a power cord wrapped around safety-related conduit 1N27R189A, which supplied power to a pressure transmitter for the feed water leakage control transmitter.

The feed water leakage control system is used during a loss of feed water event, when the plant is shut down and when the main feed water isolation valves are closed. The transmitter provides a pressure permissive at 35 psig feed water pressure to allow operators to manually start the feed water leakage control system which provides a water seal to the main feed water isolation valve bonnet area. The water supply is from the residual heat removal keep fill system and provides a post Loss of Cooling Accident seal to prevent escape of airborne contaminants. An interlock signal to automatically stop the feed water control system occurs at 45 psig main feed water pressure. Upon loss of the transmitter, a zero pressure signal is generated, allowing manual operation of the feed water leakage control system. However, the ability to automatically stop the feed water control system is lost, requiring the operator to stop the system manually.

The performance deficiency associated with this event is failure to follow plant procedures which resulted in several examples of electrical separation criteria violations.

The finding was greater than minor because, if left uncorrected, routing the extension cords near safety-related power cables increased the likelihood of rendering multiple trains of safety-related equipment inoperable given a fire from those temporary cables.

Further, the multiple examples of violating the electrical separation criteria indicated a lack of plant personnel knowledge of the requirement. The finding was determined to be of very low safety significance using the phase 1 SDP screening criteria for seismic, fire, flooding, and severe weather because an actual fire had not occurred that would render the associated equipment inoperable. 10 CFR 50, Appendix B Criterion V, states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Procedure PAP-204, Housekeeping/Cleanliness Control, Step 6.1.1.8a states that those temporary power cords, including extension cords, are treated as Non-Class 1E Cables and shall conform to separation criteria of ISS-2250. Both Installation Specification 2250 and Drawing D-214-004 requires that the preferred minimum separation distance between Class 1E and Non- Class 1E conduit be one inch. The failure to follow PAP-204 is a violation. However, because of the low safety significance and because the issue is in the licensees corrective action program, it is being treated

as a Non-Cited Violation, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 50-440/01-16-02). This violation is in the licensee's corrective action program as CRs 02-00091 and 02-00069.

Cornerstone: Emergency Preparedness (EP)

1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)

a. Inspection Scope The inspectors reviewed Revisions 14 and 15 of the Perry Nuclear Power Plant emergency Plan to determine whether changes identified in Revision 15 reduced the effectiveness of the licensees emergency planning, pending onsite inspection of these changes.

b. Findings No findings of significance were identified.

4. OTHER ACTIVITIES (OA)

4OA1 Performance Indicator (PI) Verification (71151)

a. Inspection Scope The inspectors reviewed reported fourth quarter 2001 data for the Unplanned Scrams and Scrams with Loss of Normal Heat Removal PIs using the definitions and guidance contained in Nuclear Energy Institute 99-02, Regulatory Assessment Indicator Guideline, Revision 1. The inspectors reviewed station logs, event notification reports, condition reports, licensee cause analysis, and personnel statements for selected 2001 scrams to verify the accuracy of the licensees data submission.

b. Findings The inspectors reviewed the licensees determination that the December 15, 2001 scram did not involve a loss of normal heat removal. The inspectors did not reach the same conclusion as the licensee.

On December 15, 2001, a failure of the feedwater control system circuitry resulted in high reactor water level and generated a level 8 scram signal. The Reactor Feed Pump Turbines (RFPTs) tripped, as designed, at Level 8 and reactor water level dropped rapidly (less than 60 seconds) to level 2 due to loss of feedwater. As documented in personnel statements after the event, there was confusion during the initial stages as to what caused the transient. A Reactor Operator (RO) noted trips of both RFPT A&B, noted the Motor Feed Pump (MFP) failed to auto start, and noted that both the red and green indicating lights for the MFP were extinguished. The Unit Supervisor later

documented that it was announced in the control room that we had no feed pumps.

The RO did not attempt to start the MFP. RCIC and High Pressure Core Spray (HPCS)

auto started as designed at level 2.

An incident investigator who interviewed the RO after the event told the inspectors that since RCIC and HPCS had auto started and were increasing reactor water level, the RO deferred pursuing immediate troubleshooting of the MFP. A plant operator was, however, dispatched to walkdown the MFP and the MFP breaker (which was accomplished prior to eventually starting the pump). As the transient response continued, a second level 8 trip was received approximately 4 minutes after the first. At some point after the second level 8 trip, the RO determined the indicating light problem was due to a bulb problem.

Control room logs indicate the pump and breaker walkdowns were completed approximately 16 minutes after the no feed pump announcement. After the second level 8 cleared, the trips were reset and several minutes later the MFP was started on the startup controller. Per the control room logs, the MFP was started 30 minutes after the start of the transient.

The licensee concluded that all systems functioned as designed and, as a result, there was no loss of normal heat removal. Licensee personnel, regulatory affairs, informed the inspectors that had the operators required the MFP they would have attempted to start it and it would have functioned as designed and therefore was always available.

The inspectors reviewed personnel statements, interviewed an incident investigator, and discussed the event with licensed operators. The inspectors concluded that during a transient such as the December 15, 2001 reactor scram, equipment is only available if operators consider it to be available. Based on personnel statements and control room logs, for some finite period of time, perhaps, fifteen minutes, control room operators believed they had no feed pumps available and took action accordingly based on the indications available to them. The inspectors concluded the uncertainty of the scram initiator combined with the lack of local indicating lights created sufficient doubt as to MFP availability. The operator did not attempt to start the MFP while level fell from level 8 to level 2.

In hindsight, with the transient fully understood, the inspectors agreed that, with the exception of a light bulb, equipment functioned as designed. The MFP did not start on loss of RFPTs because of the sealed-in initial level 8 scram signal; however, this was not initially recognized by the operators. The operators would need to recognize and reset the level 8 signal to allow the MFP to start. The fact remains, however, that the operators considered and treated the MFP as unavailable and utilized an alternate method of heat removal (RCIC).

NEI 99-02, Rev. 1 guidance stated that the indicator monitors that subset of unplanned and planned automatic and manual scrams that necessitate the use of mitigating systems and are therefore more risk-significant than uncomplicated scrams. The guidance also defined normal heat removal path as the path from the main condenser through the main feedwater system, steam generators (or reactor vessel), the main steam isolation valves, and back to the main condenser. Finally, the guidance stated

that complete loss of all main feedwater constitutes a loss of normal heat removal path condition if it cannot be easily recovered from the control room without the need for diagnosis or repair.

With respect to the NEI guidance, the inspectors noted that the December 15 event necessitated the use of mitigating systems in that RCIC and HPCS were used to restore reactor water level. Further, the operators announced the unavailability of the feedwater system and did not attempt to use it until the cause of the failure to automatically start on trip of the RFPTs was understood, the indicating light issue was resolved, and a field walkdown of the pump and breaker were completed. The inspectors concluded that diagnosis was required prior to recovering the normal heat removal path and that, therefore, this event should be counted as a scram with loss of normal heat removal.

Since this event, if counted as a scram with loss of normal heat removal, would result in the PI crossing the green to white threshold, the inspectors considered this issue an Unresolved Item (URI) (URI 50-440/01-16-03). The inspectors submitted a Reactor Oversight Process Feedback form in accordance with established procedures to document the disagreement with the licensee.

4OA3 Event Follow-up(71153)

(Closed) Licensee Event Report (LER) 50-440/2001-02: Oscillation Power Range Monitors Inoperable Due to Non-Conservative Setpoints. On June 27, 2001, General Electric notified the licensee that oscillation power range monitor (OPRM)

instrumentation scram setpoints were non-conservative due to non-conservative analysis. Upon notification, the licensee declared the system inoperable, completed TS required actions which required initiation of an alternate method to detect and suppress thermal hydraulic instability oscillations, and entered the issue into their corrective action program as CR 01-2582. The inspectors noted that operators were trained to monitor for potential instability (power-to-flow montioring) both prior to and after OPRM installation. The inspectors review identified no new issues. This is a minor violation not subject to formal enforcement.

4OA6 Meetings Exit Meeting The inspectors presented the inspection results to Mr. Guy Campbell, Site Vice President and other members of licensee management at the conclusion of the inspection on February 26, 2002. The licensee acknowledged the findings presented.

No proprietary information was identified.

KEY POINTS OF CONTACT Licensee G. Campbell, Vice President-Nuclear B. Boles, Operations Manager G. Dunn, Manager, Regulatory Affairs D. Gudger, Supervisor, Compliance T. Lentz, Manager, Design Engineering K. Ostrowski, Director, Nuclear Services Department D. Phillips, Manager, Plant Engineering T. Rausch, Director, Nuclear Maintenance Department W. Kanda, General Manager, Nuclear Power Plant Department R. Strohl, Superintendent, Plant Operations LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-440/01-16-01 NCV Failure to Remove Temporary Lights From the Reactor Water Cleanup Heat Exchanger Room After One Cycle 50-440/01-16-02 NCV Failure to Follow Procedures for Maintaining Electrical Separation Criteria 50-440/01-16-03 URI Scrams With Loss of Normal Heat Removal Reporting Criteria Closed 50-440/2001-02 LER Oscillation Power Range Monitors Inoperable Due to Non-Conservative Setpoints 50-440/01-16-01 NCV Failure to Remove Temporary Lights From the Reactor Water Cleanup Heat Exchanger Room After One Cycle 50-440/01-16-02 NCV Failure to Follow Procedures for Maintaining Electrical Separation Criteria

LIST OF ACRONYMS USED CFR Code of Federal Regulations CR Condition Report EDG Emergency Diesel Generator ESW Emergency Service Water FCR Field Clarification Request FENOC FirstEnergy Nuclear Operating Company HPCS High Pressure Core Spray LER Licensee Event Report MFP Motor Feed Pump NCV Non-Cited Violation NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation ONI Off-Normal Instruction OPRM Oscillation Power Range Monitor OWA Operator Workarounds PAP Plant Administrative Procedure PARS Publicly Available Records PI Performance Indicator PTI Periodic Test Instruction RCIC Reactor Core Isolation Cooling RFPT Reactor Feed Pump Turbines RO Reactor Operator RRA Radiologically Restricted Area SDP Significance Determination Process SVI Surveillance Instruction TS Technical Specifications USAR Updated Safety Analysis Report

LIST OF DOCUMENTS REVIEWED 1R04 Equipment Alignment SOI-P45/49 Emergency Service Water and Screen Wash September 19, Systems, Rev. 2 1995 VLI-P45 Emergency Service Water System, Rev. 4 August 22, 1989 Drawing Emergency Service Water System July 25, 2001 D-302-791 Drawing Emergency Service Water System April 17, 2000 D-302-792 Drawing Emergency Service Water Pumphouse April 17, 2000 D-912-630 Ventilation System USAR Emergency Service Water System Section 9.2.1 USAR Engineered Safety Features Ventilation System Section 9.4.5 TS 3.7.1 Emergency Service Water (ESW) System -

Divisions 1 and 2 1R05 Fire Protection Drawing Fire Protection Evaluation - Control Complex and September 2001 E-023-015 Diesel Generator Roof Plan, El. 638'-6" and 646'-

6" Drawing Fire Protection Evaluation - Unit 1 Reactor March 1991 E-023-018 Building Plan, El. 654'-0" Drawing Fire Protection Evaluation - Unit 1 Auxiliary and September 2001 E-023-002 Reactor Building Plan, El. 574'-10" USAR Section Fire Zone 1RB-1b 9A.4.2.1.1.2 USAR Section Fire Zone 1AB-1a 9A.4.2.1.1 USAR Section Fire Zone 1AB-1f 9A.4.2.1.6 USAR Section Fire Zone 1CC-4a 9A.4.4.4.1.1

USAR Section Fire Zone 1CC-4e 9A.4.4.4.1.5 National Fire Fire Protection Handbook, Edition 15.

Protection Association Condition Report Extension Cord Overheats Causing Smoke in January 7, 2002 02-0057 Containment 1R11 Licensed Operator Requalification ONI C51 Unplanned Change in Reactor Power or March 14, 2001 Reactivity, Rev. 8 ONI C71-1 Reactor Scram, Rev. 3 May 21, 2001 ONI C85-2 Pressure Regulator Failure - Open, Rev. 3 May 22, 1989 ONI ZZZ-1 Tornado or High Wind, Rev. 2 June 30, 1995 1R12 Maintenance Rule Implementation CR 01-0060 RFPT B Control Power Fuse Blown January 7, 2001 CR 01-0440 RFPT B Work February 7, 2001 CR 01-0864 Motor Feedwater Pump Cracks February 26, 2001 CR 01-1113 FM Found in Valve 1N27F160B March 5, 2001 CR 01-1228 Motor Feed Pump Did Not Trip as Expected March 8, 2001 CR 01-1586 Maintenance Rule Evaluation Required on March 21, 2001 Reactor Feedwater Booster Pump CR 01-1606 RPV Level Control March 22, 2001 CR 01-1983 Motor Feedpump Run With Minimum Flow Of April 29, 2001 Approximately 800GPM CR 01-2081 Repeat Failure of Motor Feed Pump Recirc Flow May 4, 2001 Control Valve to Stroke CR 01-2112 Maint Rule Evaluation is Required on Reactor May 7, 2001 Feed Pump A Min Flow Valve CR 01-2779 1N27F0170 [Motor Feed Pump Recirc Valve], July 17, 2001 Failed to Stroke Properly CR 01-2827 Damaged 2nd Stage Diffuser Vanes July 21, 2001

CR 01-3966 Collective Maintenance Rule Evaluation of Motor November 14, 2001 Feed Pump Component Failures System Health Feedwater Control System Status Report 1st Quarter 2001 Report System Health Feedwater Control System Status Report 2nd Quarter 2001 Report System Health Feedwater Control System Status Report 3rd Quarter 2001 Report System Health Feedwater Control System Status Report 4th Quarter 2001 Report System Health Feedwater System Status Report 1st Quarter 2001 Report System Health Feedwater System Status Report 2nd Quarter 2001 Report System Health Feedwater System Status Report 3rd Quarter 2001 Report System Health Feedwater System Status Report 4th Quarter 2001 Report PAP-1125 Monitoring the Effectiveness of the Maintenance April 4, 2001 Program Plan, Rev. 6 Logs Control Room Logs 01/01/01 - 12/31/01 NUMARC 93-01, Nuclear Energy Institute Industry Guideline for Revision 2 Monitoring the Effectiveness of Maintenance at Nuclear Power Plants 1R13 Maintenance Risk Assessments and Emergent Work Evaluation Week 9, Period 4 Forecast Risk Profile January 21, 2002 Week 10, Period 4 Forecast Risk Profile January 28, 2002 Week 11, Period 4 Forecast Risk Profile February 4, 2002 WO 02-000122- Replace Static Inverter 1R41K00090A 000 WO 00-005257- Rework Damper IAW SMRF 00-5027 000 WO 01-16229- Replace the Auto and Manual Status Indicating 000 Lamps for Div. 3 DG Room Fan 2C

SOI-M43 Diesel Generator Building Ventilation System, May 24, 1990 Rev. 5 USAR 9.4.5.2.4 Diesel Generator Building Ventilation System PAP 1924 On-Line Safety Assessment and Configuration November 30, 2000 Risk Management, Rev. 2 1R14 Personnel Performance During Non-Routine Evolutions Plant Narrative Logs January 7, 2002 CR 02-0057 Extension Cord Overheats Causing Smoke in January 7, 2002 Containment NUREG 1022 Event Reporting Guidelines 10 CFR 50.72 and 50.73, Rev. 2 Plant Narrative Logs February 10, 2002 ONI C51 Unplanned Change in Reactor Power or March 14, 2001 Reactivity, Rev. 8 SVI-C71-T0051 Reactor Protection System Manual Scram April 27, 1988 Channel Functional, Rev. 2 CR 0-0416 Rod 38-43 Inserted During SVI-C71-T0051 February 10, 2002 1R19 Post-Maintenance Testing WO 01-17117- Repair Air Leak at Gasket on Intake Manifold November 29, 2001 000 Blanking Plate SOI-P54 (WTR) Fire Protection System Water, Rev. 0 August 14, 2001 USAR 9.5.1 Fire Protection System WO 00-005257- Rework Damper IAW SMRF 00-5027 000 PTI-M32-P0004 Emergency Service Water Pump House August 7, 1987 Ventilation System Train A Damper Stroking, Rev. 0 USAR Engineered Safety Features Ventilation System Section 9.4.5 NH90 Series Hydramotors Maintenance Manual, March 19, 1997 Rev. 9

1R22 Surveillance Testing SVI-E22-T1319 Diesel Generator Start and Load Div. 3, Rev. 10 December 14, 2000 SOI-E22B Division 3 Diesel Generator, Rev. 6 May 11, 1995 TS 3.8.1 AC Sources - Operating USAR Onsite Power Systems Section 8.3.1 PTI-P54-P0050 Unit One Fire Detection Instrumentation January 9, 2002 Functional Test for SDP-1H51-P929 PTI-P54-P0045C Fire Detection Instrumentation Functional test for January 8, 2002 SDP-H51-P219 NFPA 72E Standard for Automatic Fire Detectors 1974 NFPA 72 Chapter 7, Inspection, Testing and Maintenance 5-79 1R23 Temporary Modifications Field Clarification Temporary Lighting in the RWCU Heat October 13, 1997 Request 024726 Exchanger Room CR 02-0057 Extension Cord Overheats Causing Smoke in January 7, 2002 Containment Plant Narrative Logs January 7, 2002 NEI 0674 Specification Changes, Rev. 8 CR 02-0069 Temporary Power Cable Separation Violation January 8, 2002 CR 02-0091 Temporary Extension Cord Routing January 9, 2002 PAP 0204 Housekeeping/Cleanliness Control, Rev. 10 Drawing Conduit and Cable Tray Separation Criteria, D-214-004 Rev. T Installation Electrical Work and Equipment Specification, Specification 2250 Rev. 1 1EP4 Emergency Action Level and Emergency Plan Changes Perry Nuclear Power Plant Emergency Plan, Rev. 14 Perry Nuclear Power Plant Emergency Plan, Rev. 15

4OA3 Event Follow-up CR 01-2582 OPRM Operability - Potentially Non- June 27, 2001 Conservative Stability Reload Calculation IOI-3 Power Changes, Rev. 7 November 8, 1993 TS 3.3.1.3 Oscillation Power Range Monitor (OPRM)

Instrumentation 24

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