IR 05000346/1998009
| ML20236W233 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/28/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20236W226 | List: |
| References | |
| 50-346-98-09, 50-346-98-9, NUDOCS 9808050165 | |
| Download: ML20236W233 (15) | |
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U. S. NUCLEAR REGULATORY COMMISSION i REGION 111 l
l Docket No: 50-346
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License No: NPF-3 Report No: 50-346/98009(DRP)
Licensee: Toledo Edison Company Facility: Davis-Besse Nuclear Power Station Location: 5501 N. State Route 2 Oak Harbor, OH 43449 Dates: May 12 - June 23,1998 Inspectors: S. Campbell, Senior Resident inspector K. Zellers, Resident inspector Approved by: Thomas J. Kozak, Chief j
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Reactor Projects Branch 4 1 l
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9800050165 900728 PDR ADOCK 05000346 G PM ;
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EXECUTIVE SUMMARY Davis-Besse Nuclear Power Station NRC Inspection Report 50-346/98009(DRP)
This inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a 6-week period of resident inspectio Ooerations
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Effective communications and thorough control room briefs during plant restart following the refueling outage were noted. The activities were carefully controlled (Section O1.1).
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A power reduction to repair a main feed pump was well planned and controlled with effective management oversight (Section 01,1).
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Management's expectation to promptly generate a potential condition adverse to quality report (PCAQR) was not met following the inspectors' discovery of an improperly installed wafer check valve, a valve used to ensure effective emergency ventilation system operation. Although the check valve was promptly re-installed correctly, a PCAQR documenting this condition was not written until the inspectors brought the issue to
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The inspectors concluded that, on two separate occasions, operators performed actions without management approval when faced with unanticipated circumstances. In the first case, an operator did not follow an emergency diesel generator (EDG) operating procedure when he improperly opened a bus tie breaker. This deenergized a 4160-volt safety-related bus and momentarily overloaded EDG#2. In the second case, after an operator closed the wrong low pressure injection system suction valve, recovery actions were taken without control room supervisor approval. In both instances, operators did not l meet management's expectations (Section 04.1).
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The inspectors found several unscreened openings around the base of the emergency sump screen enclosure that could permit particles of sufficient size into the sump during recirculation following a loss-of-coolant accident to potentially plug containment spray nozzles. The design basis requirement of protecting the spray nozzles from plugging was not translated into the emergency sump design and is a design control violation (Section E4.1).
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The licensee implemented an effective and methodical approach to identify and comprehensively resolve deficiencies in the actuation logic surveillance test program in response to Generic Letter 96-01, " Testing of Safety-Related Logic Circuits" l (Section E8.1).
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The licensee performed a change to the protective relay design of a safety-related l
4160-volt breaker without properly establishing the suitability of the change, which is a non-cited violation of design control (Section E8.5).
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Report Details !
Summarv of Plant Status At the start of the inspection period, the plant was shutdown and Refueling Outage (RF)-11 was ongoing. After the outage was completed, a plant restart was commenced and the main generator output breaker was closed on May 23. Power was subsequently increased up to 98 percent and held at that level due to fouling in Steam Generator #1. On May 29, power was decreased to approximately 50 percent to allow repairs to Main Feed Pump #1. Repairs were accomplished and power was increased to and remained at 98 percent for the remainder of the inspection perio l. Operations 01 Conduct of Operations O1,1 General Comments (71707)
Operator Performance During Plant Restart The inspectors observed portions of the plant restart following RF-11. Control room briefs were thorough for the plant startup and operators appropriately followed plant procedures. The reactor startup was appropriately stopped as needed to evaluate abnormal control room indications or equipment deficiencies. Operators used clear and effective communications and control room supervisors maintained good command and control during the startup. The inspectors concluded that operators maintained good I
control of reactor coolant system temperature and pressure and core reactivity and that management provided effective oversight during the plant startu Operator Performance During Power Operations While the plant was at full power, the inspectors conducted routine tours of the control room and determined that operations personnel performed thorough shift turnovers and were cognizant of plant status when questioned. Operators were knowledgeable of plant conditions and appropriately responded to control room annunciators as they were received. Operators were effectively prepared by rehearsing on the simulator for a power reduction to 50 percent to allow repairs to Main Feed Pump #1. Operations management provided effective oversight and control during the downpower and the evolution was
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conducted without incident. The inspectors concluded that plant operations were conducted conscientiously with good management oversight.
l O2 Operational Status of Facilities and Equipment O2.1 System Walkdowns (71707)
The inspectors walked down the accessible portions of the following engineered-safety-features and important to-safety systems during the inspection period:
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Emergency Core Cooling System Rooms
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Emergency Diesel Generators (EDG)
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High and Low Voltage Switchgear Rooms
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Auxiliary Feedwater Pump Rooms No substantive concerns were identified during the walkdowns. System lineups and major flowpaths were verified to be consistent with plant procedures / drawings and the Updated Safety Analysis Report (USAR). Pump / motor fluid levels were within their normal bands and equipment vibration and temperature was normal. Only very minor oil and fluid leaks were noted with the most observed around the EDG fuel oil syste Deficiency tags were used appropriately. Local and remote controllers were properly positioned, and instrumentation appeared to be functioning correctl O2.2 Shield Buildina Neaative Pressure Boundary fearaded Due to improper Installation of Qrain Wafer Check Valve Inspection Scope (71707)
The int.poctors conducted a routine inspection of Mechanical Penetration Room # Observations and Findinas On June 2, the inspectors discovered that a wafer check valve for a drain line in Mechanical Penetration Room #4 was installed incorrectly. The wafer check valve closes to isolate the drain line from areas outside the shield building area when the emergency ventilation system (EVS) starts. Isolation of areas outside the shield building ensures that the EVS can draw the required vacuum in the shield building area to minimize dose to the public after a postulated loss-of-coolant accident (LOCA).
The inspectors informed a radiation protection supervisor and a senior reactor operator of the improperly installed check valve, and the licensee promptly installed the check valve correctly. However, a week later, the inspectors determined that a PCAQR documenting the deficiency had not been written, which did not meet management's expectation of generating a corrective action document to evaluate the issue. After the inspectors informed plant management, PCAQR 98-1225 was written to document the improper installation of the wafer check valve. The operability evaluation for the PCAQR concluded that the drain line he!e was too small to impact EVS operability. The inspectors reviewed applicable procedures and verified that this evaluation was acceptable, Conclusions Management's expectation to promptly generate a PCAQR was not met following the inspectors' discovery of an improperly installed wafer check valve, a valve used to ensure effective EVS operation. Although the check valve was promptly re-installed correctly, a PCAOR documenting this condition was not written until the inspectors brought this issue to management's, attention a week later.
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04 Operator Knowledge and Performance 04.1 Operator Performance Durina Safety Features Actuation System (SFAS) Intearated Time Response and Channel 2 Functional Tests Inspection Scope (71707)
The inspectors reviewed the circumstances surrounding operator errors that occurred during SFAS test Observations and Findinas SFAS Integrated Time Response Test On May 12, equipment operators performed Procedure DB-SC-03113, "SFAS Integrated Time Response Test." This SFAS test required EDG #2 to be operating. During the test, a material problem caused continuous operation of the charging motor for the Service Water (SW) Pump #2 circuit breaker. Due to this problem, EDG #2 was secured to allow the SW pump breaker to be replaced. With Procedure DB-OP-06316, " Diesel Operating Procedure,"in hand, an equipment operator properly transferred the loads for essential Bus D1 to the nonessential Bus D2 and then opened the EDG output breaker. Once this was accomplished, the operator thought the electrical lineup was incorrect and, without communicating his intention to control room operators to deviate from the procedure, he opened tie Breaker AD 110 between Busses D1 and D2. This action resulted in an under voltage condition on Bus D1, caused the EDG #2 output breaker to shut, and resulted in the sequencing of the necessary electricalloads onto Bus D The equipment operator and another equipment operator, who helped in the EDG shutdown procedure, stepped outside the EDG room to discuss how to rectify the situation, but did not communicate to control room personnel what had occurred. In response to the unanticipated open breaker indication in the control room, operations management tried but was unable to establish contact with the equipment operators. A reactor operator was dispatched to investigate the cause of the open breaker indicatio Before the reactor operator arrived, the equipment operators had reentered the room and closed Breaker AD 110 to parallel Busses D1 and D2. When the equipment operator paralleled Busses D1 and D2, the load on the EDG started to increase to a point where the EDG was becoming overloaded. The equipment operator did not understand what was occurring and took no actions; however, the other operator quickly opened Bus D1 to Bus D2 tie Breaker AD110 to stop the EDG overload condition. The cause of the load increase was that, unknown to the operators, the EDG govemor had switched from droop to isochronous mode because Bus D1 had experienced an undervoltage conditio Subsequently, the EDG was appropriately secure Following the event, operations management counseled the equipment operators and directed operator stand downs to emphasize self-checking protocols. The inspectors and the licensee reviewed associated procedures and interviewed the involved operators to l determine the factors that contributed to this event. The inspectors found that the operating procedure had already been marked up because of prior use and was less useful as a tool for self-checking. Further, the equipment operator sensed time pressure
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in completing the task and had the perception that this was a nonstandard shutdown of the EDG, which negatively affected his confidence level in operating the ED Engineering personnel determined that the EDG was motored during the event, which was also attributed to the EDG being in the isochronous mode. The reverse power relay sensed the condition, but did not cause the EDG output breaker to trip, because the protective feature was bypassed while the SFAS actuation signal was present. Further, because the EDG was loaded to about 3200 kilowatts for about 2-3 seconds, which was greater than the rating requirement of 30 minutes at 3035 kilowatts, the operability of the EDG was in question. The licensee subsequently performed electrical checks, testing and a vibration analysis of the EDG and verified the EDG was not damage Technical Specification 6.8.1.a states that applicable procedures recommended in !
Appendix A of Regulatory Guide 1.33, November,1972 shall be implemente l Appendix A of Regulatory Guide 1.33, November 1972, Section C.19.b requires 1 instructions for startup and shutdown of onsite safety-related power sources. The j instructions for shutting down the #2 EDG were in DB-OP-06316. The inspectors determined a TS 6.8.1.a violation occurred when the equipment operator took actions outside the bounds of Procedure DB-OP-06316. This non-repetitive, licensee-identified and corrected violation is being treated as a non-cited violation, consistent with Section VilB.1 of the NRC Enforcement Policy (NCV 50-346/98009-01(DRP)).
SFAS Channel 2 Functional Test i On June 1, three operators in different locations performed Procedure DB-SC-03111,
"SFAS Channel 2 Functional Test." As part of this test, low pressure injection (LPI)
suction valve DH 2734 was closed. An operator in the switchgear room removed the i close power fuses from the breaker that supplied power to LPI Pump #2 to prevent the pump from starting during the test. The next step was to close suction valve DH 2734 for LPI Pump #2. However, the control room operator erroneously closed suction valve DH 2733 for LPl Pump #1. This rendered both trains of the LPI system inoperable. This procedural violation constitutes a violation of minor significance and is not subject to formal enforcement actio In response to the error, the operators discussed actions that could be taken to rectify the situation without immediately notifying the control room supervisor of the error or of the proposed actions to restore the LPI system. Thirty-three seconds after the wrong suction valve was closed, and without approval from the control room supervisor, one of the operators restored LPI Pump #2 by reinstalling the close power fuses. Then, because the u suction valve for LPI Pump #1 was closed, the operator removed the close power fuses i
from the breaker for LPI Pump #1 to prevent operation of the pump without a water l source. Once these actions were completed, the control room supervisor was notified of the event and PCAQR 98-1201 was initiated to document the closing of the wrong suction valve.
l The inspectors reviewed the procedure and found it to be adequate. However, the . l control room operator relied on instructions from the operator at the SFAS cabinet since j the control room operator did not have the procedure in hand. Therefore, the control :
room operator was unable to verify that he was manipulating the correct valve. The l i
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control room operator who closed the wrong valve indicated to the inspectors that he was not distracted, fatigued, or rushe Although the three operators focused on placing the plant in a safe condition by restoring LPI Pump #2, they did not meet management's expectations to notify the control room supervisor of the error or to obtain approval on actions to rectify the error. During an experience review for the PCAQR, the licensee recognized similar characteristics with other recent operator events regarding communications, self-checking and procedure use ;
and wrote PCAQR 98-1284 to address the collective significance of these operational j issue . Conclusions on Operator Knowledae and Performance The inspectors concluded that, on two separate occasions, operators performed actions without management approval when faced with unanticipated circumstances. In the first case, an operator did not follow an emergency diesel generator (EDG) operating procedure when he improperly opened a bus tie breaker. This deenergized a 4160-volt safety-related bus and momentarily overloaded EDG#2. In the second case, after an operator closed the wrong low pressure injection system suction valve, recovery actions were taken without control room supervkor approval. In both instances, operators did not meet management's expectation Miscellaneous Operations issues (92901)
0 (Closed) Inspection Followap Item 50-346/97006-01(DRP): Inoperable Post Accident Monitoring Instrumentation. This involved the inspectors identification that Channel 1 Hot Leg Indicator TIRC3B was indicating nine degrees lower than the other three hot leg temperature indicators. In response to the finding, maintenance technicians performed thorough troubleshooting of the circuit that included consultations with a contractor who had expertise in resistance temperature detector (RTD) applications. The licensee determined that the RTD instrument string failure was due to high resistance in the wire connections. Consequently, the lug nuts for portions of the wiring were replaced. No degradation of the RTD had been noted since the high resistance condition was resolve . Maintenance M1 Conduct of Maintenance M1.1 Maintenance and Surveillance Activities (61726)(62707) i The following maintenance and surveillance testing activities were observed / reviewed l during the inspection period:
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MWO 3-98-05345-01 Calibrate Forebay Temperature Element TE738
- A,WO 3-98-0935-01 Inspect Containment Air Cooler #2 Service Water Inlet Isolation Valv i
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Activities observed were performed on schedule by knowledgeable personnel. Testing was performed in accordance with the test progra l!IJngineerina E4 Engineering Staff Knowledge and Performance E4.1 Containment Buildina Emeraency Sumo Desian Deficiency Inspection Scope (37551)
While inspecting the containment emergency sump during a closecut inspection of containment, the inspectors identified gaps at the base of the emergency sump screen enclosure. A followup inspection was performed to determine if this system configuration represented a design concem, Observations and Findinas On May 22, while the licensee increased power following RF-11, the inspectors noted a potential design inadequacy with the containment emergency sump while inspecting around the sump for debris. Specifically, the metal screen mesh enclosure installed over the sump had several 8-inch by 5/8-inch unscreened openings between the screen enclosure frame and the concrete floor. The sump screen is designed to remove debris from water supplied to the LPI and the containment spray systems during the recirculation phase following a LOCA. In response to the inspectors' finding, the licensee entered the containment building to inspect the unscreened openings. After examination, the licensee concluded that, based on engineering judgement, the openings were too small to permit passage of debris that could affect LPI and containment spray operation and that debris transport to the sump was unlikely during recirculatio The inspectors questioned the initial operability determinations because the smallest containment spray nozzle orifice size is 13/32-inch. The licensee wrote PCAQR 98-1151 to address the inspectors' concern that the unscreened openings could allow debris to bypass the screen enclosure and plug containment spray nozzle The Station Review Board convened to evaluate the potentially degraded condition and the impact on containment spray system operability. The board considered low transport velocities for material during recirculation, minimal material in containment available for transporting and entering the sump, and the use of containment air coolers as the primary means of removing heat from containment following a LOCA. From these considerations, the board determined that the gaps did result in the emergency sump screen being degraded from the intent of the USAR's description but that the sump would still perform its safety function as describe Consequently, the licensee implemented Temporary Modification (TM) 98-0022 to place 36 lead bricks around the base of the sump screen to cover the gaps. The licensee stated that a permanent modification would be considered for installation during RF-1 The inspectors reviewed the TM and determined that it was acceptabl _ _ _ - - _ _ _ . - - -
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The inspectors reviewed the sump design and determined that the licensee did not correctly translate the design basis requirement of preventing large particles from -
entering into the emergency sump. Davis-Besse USAR Section 6.2.2.6.2 stated that an intake screen with 1/4-inch openings is installed over the sump to prevent large particles from getting into the recirculating line and plugging the containment spray nozzle Drawing C-119, " Containment Vessel interior Concrete Fill," which depicted the installation details of the emergency sump, incorporated unscreened 8-inch by 5/8-inch
, openings that could have permitted large particles into the sump which could have l blocked the 13/32-inch spray nozzles. Criterion 111 of 10 CFR Part 50, Appendix B, wates, in part, that measures shall be established to assure that applicable regulatory requirements and design basis are correctly translated into specifications, drawings and ,
instructions. The failure by the licensee to translate the design basis requirements correctly, as set forth in the USAR, to the installed configuration, is considered a violation of 10 CFR Part 50, Appendix B, Criterion lll (VIO 50-346/98009-03(DRP)).
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A detailed description of the unscreened openings was not available in the USA Further, the licensee was unable to provide documentation that transport velocities or
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availability of material for transport in containment was considered in the original l
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emergency sump design to support the acceptability of having the unscreened opening l Nevertheless, the licensee concluded that the sump design translated the design basis requirement and that the engineering evaluation justifying equipment operability j supported this conclusion.
l Conclusions The inspectors found several unscreened openings around the base of the emergency sump screen enclosure that could permit particles of sufficient size into the sump to (
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potentially plug containment spray nozzles. The design basis requirement of protecting the spray nozzles from plugging was not translated into the emergency sump design j drawing and is a design control violation.
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E4.2 Missino Safety Evaluation for Fibrous Insulation Installed on Pressurizer Power Ooerated Relief Valve (PORV) Discharoe Line f
j While reviewing the deficient sump design described in Section E4.1, the inspectors l noted that a safety evaluation was not available to document the basis for permitting the use of fibrous insulation on the PORV discharge line. Fibrous piping insulation may dislodge, migrate to the sump and block flow through the emergency sump during the recirculation phase of a LOCA. This issue is an unresolved item (50-346/98009-04(DRP))
l pending the inspectors' review of the licensee's safety evaluation for insulation on the
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PORV discharge lin E8 Miscellaneous Engineerir g issues (92902)
E (Closed) Licensee Event Reoort 50-346/97-008-13(DRP); inadequate Testing of
_ -l Safety-Related Logic Circuits. The licensee established a program to review the logic l circuits at Davis-Besse in response to GL 96-01, " Testing of Safety-Related Logic Circuits." Senior engineers reviewed specific surveillance requirements using regulatory documents, surveillance procedures, and drawings to determine if actuation logic circuits !
were tested to literal compliance with TS requirements. Any noncompliance with TSs !
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identified by the engineers was reviewed by a multi-disciplined review board. Unresolved issues from the review board were elevated to the Station Review Board for resolutio Seventy-one surveillance associated with logic testing were identified in the review scope. Five thousanel man-hours were expended over a thirteen-month period to complete the revie Licensee Event Report 97-008-13 documented conditions where previous testing did not satisfy TS surveillance requirements. Surveillance procedures were changed, and subsequently, the procedures were implemented to test these logic circuits adequatel No inoperable circuits were identified during the test Technical Specification 6.8.1.c states, in part, that written procedures be established,
!mplemented and maintained covering surveillance and test activities of safety-related activities. Contrary to that requirement, the licensee identified conditions where
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surveillance test procedures did not test actuation logic circuits to literal compliance with TS surveillance requirements. The failure to satisfy the TS surveillance requirement for actuation logic circuits is a violation.- This non-repetitive, licensee-identified and corrected
, violation is being treated as a non-cited violation, consistent with Section Vll-B-1 of the l NRC Enforcement Policy (NCV 50-346/98009-05(DRP)).
The inepectors concluded that the licensee implemented an effective and methodical GL 96-01 review program to comprehensively resolve deficiencies in the actuation logic surveillance test progra E8.2 (Closed) Unresolved item 50-346/98005-04(DRPP Peeling Paint on Containment Wall The concem involved the possibility that peeling paint, discovered during tours in the containment building, could dislodge, migrate to the emergency sump, and block recirculation flow to emergency core cooling pumps following a LOCA. Engineers performed an evaluation to determine the effect peeling paint had on sump blockage, and concluded that transport velocities were too low and the flowpath too torturous to carry paint chips to the sump. The inspectors reviewed the evaluation and found it acceptabl During a final closeout inspection of containment, the inspectors confirmed the deficiency was correcte E8.3 (Closed) Inspection Follow-uo item 50-346/96003-04(DRPk Displaced Fuel Assembly Spacer Grid. - This item involved the licensee's discovery in RF-10 that one spacer grid, used to provide lateral alignment of fuel asse nblies, was displaced se'veral inches lower than expected on one assembly. The contractor, Framatome Technologies, initially stated that a mispositioned spacer grid could adversely affect the fuel cootable geometry in a post-LOCA environment. Consequently, the contractor redesigned the fuel assemblies to prevent this condition and the licensee planned to replace the old design with the new design during subsequent refueling outage A letter from the B & W Owners Group, dated March 6,1997, regarding the potential safety concem on Mark B Grid Deformation, Framatome Technologies PSC 21-96-5, concluded that no degradation in safety nor any substantial safety hazard existed because of slipped spacer grids. The letter noted no permanent grid deformation was predicted, and the cootable geometry requirements were met for all fuel assemblies within the core because spacer impact loads for all faulted conditions were within the spacer grid elastic limit. Reactor engineers visually inspected the fuel assemblies during RF-11
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! and found or suspected nine fuel assemblies of having slipped spacer grid Consequently, one assembly in the core was replaced with a newly designed assembly and the remaining eight were placed in the spent fuel poo E (Closed) Unresolved item 50-346/97003-04(DRP): SW Strainer Blowdown Valve Failure This involved the inspectors' identification that SW strainer blowdown Valve SW-1380 was open without the necessary conditions present for it to be open. The licensee determined that a relay race in the control circuitry adversely affected the time logic which csused this condition.
! E (Closed) Unresolved item 50-346/98005-05(DRP): Potential Inadequate Modificatio During a SW pump start, an unexpected inrush current into a zero-sequence-type current transformer caused a newly modified ground fault relay (type ABB 50H) to actuate and l trip open the 4160-volt breaker for the pump. The modification involved replacing the l mechanical ground fault relay with a solid state ground fault relay to address seismic qualification concems.
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Electrical maintenance engineering staff determined, through testing, that the new type solid state ABB 50H relay had an actuation response time of about 0.04 milliseconds and that the old-style mechanical relay had an actuation response time of about 17 milliseconds. The fact that the response times would be significantly different and that transient voltages were present in the circuits were not known during the design change process, and therefore, this information could not be evaluated for acceptabilit Additionally, a reference text provided misleading information regarding the performance of zero-phase sequence type current transformers. However, vendor information was available that indicated a different model relay with a time-delay characteristic was frequently preferred over the ABB 50 H relay in order to avoid nuisance tripping due to i-
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' inrush currents. Therefore, the inspectors determined that the licensee had enough
' information to have selected the proper relay application.
l The inspectors determined that the licensee had performed appropriate remedisl corrective actions by adding a 0.2 second time delay feature to the modified relay Additionally, the root cause evaluation and corrective actions for this issue p ovided reasonable assurance that the event would not recur. No evidence was found that an inadequate change to a relay application was a recurring problem. Engineering management was evaluating whether this issue should be communicated as a generic issue to the industry.
l Criterion lli of 10 CFR Part 50, Appendix B, states that " Measures shall also be i
established for the selection and review for suitability of application of materials, parts, equipment, and processes that are esse..tial to the safety-related functions of the structures, systems and components.... Design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design." Contrary to this, the licensee performed a change to the design of a safety-related 4160-volt breaker without properly establishing the suitability of the chang This non-repetitive, licensee-identified and corrected violation is being treated as a non-cited violation, consistent with Section VilB.1 of the NRC Enforcement Policy (NCV 50-346/98009-06(DRP)).
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I IV. Plant Support R4 Staff Knowledge and Performance in RP&C R Radiation Protection Performance Durina Containment Entry The inspectors observed the performance of health physics technicians during j containment tours at the end of the refueling outage. The technicians exhibited good radiation and contamination control practices while providing the inspectors survey ]
coverage during entries into the containment. The technicians ensured that escorted personnel maintained as low as is reasonably achievable (ALARA) practices by surveying the areas during the entry. The inspectors concluded that radiation protection personnel provided effective information to escorted personnel to maintain dose ALAR .I V. Manaaement Meetinos X1 Exit Meeting Summary 1 The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on June 23,1998. The Director of Engineering Services disagreed with the design control violation regarding the emergency sump. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietar No proprietary information was identifie l
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PARTIAL LIST OF PERSONS CONTACTED Licensee R. E. Donnellon, Director, Engineering Services D. L. Eshelman, Manager, Plant Operations J. L. Freels, Manager, Regulatory Affairs J. L. Michaelis, Manager, Maintenance C. A. Price, Manager, Business Services W. J. Motpus, Manager, Nuclear Training H. W. Stevens, Manager, Nuclear Safety and inspections F. L. Swanger, Manager, Design Basis Engineering G. W. Gillespie, Superintendent, Chemistry R. A. Greenwood, Supervisor, Radiation Protection S. W. Roberts, Shift Supervisor, Plant Operations R. C. Hovland, Senior Engineer, Plant Engineering J. M. Vetter, Nuclear Auditor, Quality Assurance G. M. Wolf, Senior Engineer, Regulatory Affairs A. Conway, Student, Regulatory Affairs NRC i S. J. Campbell, Senior Resident inspector, Davis-Besse
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K. S. Zellers, Resident inspector, Davis-Besse
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! INSPECTION PROCEDURES USED L
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IP 37551: Onsite Engineering IP 61726: Surveillance Observations I i
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IP 62707: Maintenance Observation IP 71707: Plant Operations IP 92901: Followup - Plant Operations i
IP 92902: Followup - Engineering ITEMS OPENED AND CLOSED Opened 50-346/98009-01(DRP) NCV Improper Breaker Switching and Overloading of the Emergency Diesel Generator 50-346/98009-03(DRP) VIO Failure to Translate Emergency Sump Design Specifications into the USAR 50-346/98009-04(DRP) URI Missing Safety Evaluation for PORV Discharge Line insulation 50-346/98009-05(DRP) NCV Failure to Meet Technical Specification Surveillance Testing of Actuation Logic 50-346/98009-06(DRP) NCV Inadequate Modification of Electrical Relays Closed 50-346/98009-01(DRP) NCV Improper Breaker Switching and Overloading of the Emergency Diesel Generator 50-346/98009-05(DRP) NCV Failure to Meet Technical Specification Surveillance Testing of Actuation Logic 50-346/97006-01(DRP) IFl Inoperable Post Accident Monitoring instrumentation 50-346/98009-06(DRP) NCV inadequate Modification of Electrical Relays 50-346/97-008-13(DRP) LER Inadequate Testing of Safety-Related Logic 50-346/98005-04(DRP) URI Peeling Paint on Containment Walls 50-346/96003-04 (DRP) IFl Displaced Fuel Assembly Spacer Grid 50-346/97003-04(DRP) URI Service Water Blow Down Valve Failures
- 50-346/98005-05(DRP) URI PotentialInadequate Modification of Electrical Relays
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LIST OF ACRONYMS AND INITIALISMS USED ALARA As Low As is Reasonably Achievab'-
EDG Emergency Diesel Generator EVS Emergency Ventilation System GL Generic Letter LOCA Loss-of-Coolant Accident LPI Low Pressure injection IFl Inspection Followup item IR Inspection Report MWO Maintenance Work Order NCV Non-Cited Violation NRC Nuclear Regulatory Commission PCAQR Potential Condition Adverse to Quality Report PDR Public Document Room PORV Power Operated Relief Valve RF Refueling Outage RTD Resistance Temperature Detector SFAS Safety Features Actuation
, SW Service Water TM Temporary Modification TS Technical Specification URI Unresolved item
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USAR Updated Safety Analysis Report VIO Violation l
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