IR 05000346/1997004
| ML20141D837 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/09/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20141D795 | List: |
| References | |
| 50-346-97-04, 50-346-97-4, NUDOCS 9705200212 | |
| Download: ML20141D837 (18) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION lli Docket No: 50 346 License No: NPF-3 Report No: 50-346/97004 F
Licensee: Toledo Edison Company Facility: Davis-Besse Nuclear Power Station Location: 5503 N. State Route 2 Oak Harbor, OH 43449 Dates: March 3 - April 14,1997 Inspectors: S. Stasek, Senior Resident inspector K. Zellers, Resident inspector Approved by: John M. Jacobson, Chief Reactor Projects Branch 4
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9705200212 970509 PDR ADOCK 05000346 C PDR
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EXECUTIVE SUMMARY Davis-Besse Nuclear Power Station NRC Inspection Report No. 50-346/97004 This inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a six-week period of resident inspactio Ooerations
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Good operator awareness of equipment and system status was noted. Shift turnovers and shift briefs were well conducted and effectively communicated important operational information (Section 01.1).
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Operator usage of and adherence to administrative, operating, and surveillance procedures was good (Section 01.1).
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Engineered safety features and important-to-safety systems were verified lined up in accordance with plant drawings and the updated safety analysis report (Section 0 2.1 ).
Maintenance
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Material condition of important standby equipment was excellent overall (Section O 2.1 ).
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Maintenance personnel performed observed work activities in accordance with maintenance work order (MWO) work instructions and approved maintenance procedures in all cases (Section M1.1).
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Surveillance testing activities observed during the ir.spection period provided adequate assurance that the tested equipment could perform as designed (Section M 1.1 ).
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The procedure for foreign material exclusion did not provide sufficient guidance to workers in some cases (Section M3.1).
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An ongoing failure to perform technical specification required monthly testing of undervoltage relays that initiate essential bus loadshed and emergency diesel generator start was considered a violation of NRC requirements (Section M3.2).
Enoineerina
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Severalissues relating to the installation of the shield building blowout panels were identified during the inspection. The inspectors questioned whether caulking used to seal the edges of the panels could adversely affect the pressure at which they were set to blow out. A number of blowout panel shear bolts were not installed 2 -
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per the installation drawings. There was no procedure guidance for shear bolt installation as well (Section E'.1).
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Information supporting the acceptability of installing armor plates on several station fire doors was not readily retrievable. As such, the licensee could not easily
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determine that use of arrnor plating was acceptable and declared the subject doors inoperable and established compensatory fire watches while completing their review j
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(Section E2.2).
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Plant engineering identified that relays in the control circuits for all three service water blowdown valves had experienced age related degradation, and
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subsequently, all were replaced. In addition, the inspectors identified that a design
, flaw in the same control circuitry could cause the blowdown valves to misoperate :
as well, resulting in a portion of service water flow to be redirected from the normal '
safety loads. Similar relays and control logic were also used in the design and
! installation of other safety related valves. Extent-of-condition reviews were
- ongoing at the end of the inspection period (Section E8.6),
i i Plant Suocort
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NRC observed station activities were conducted in accordance with radiation j protection (RP) program requirements (Section R1).
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Radiation, high radiation, and contaminated area controls and postings were in conformance with RP procedures (Section R1).
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An emergency plan drill was observed on April 2. The inspectors noted that drill i
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objectives appeared to have been met, good participation by station personnel was
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evident, and the drill scenario was well controlled and realistically implemented (Section P1).
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Report Details Summary of Plant Status The unit operated at nominally full power throughout the inspection perio l. Operations 01 Conduct of Operations 01.1 General Comments (71707)
The inspectors observed control room operators during their conduct of shift activities, walked down control room panels, reviewed logs, and held discussions with operations personnel during the inspection period. Good operator awareness of equipment and system status was noted. Shift turr. overs and shift briefs were well conducted and effectively communicated important operational informatio Operator usage of and adherence to administrative, operr.cing, and surveillance procedures were goo O2 Operational Status of Facilities and Equipment 02.1 System Walkdowns (71707)
The inspectors walked down the accessible portions of the following engineered safety features (ESP) and important-to-safety systems during the inspection period:
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high pressure injection system - trains 1 and 2
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low pressure injection system - trains 1 and 2
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containment spray system - trains 1 and 2
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emergency diesel generators #1 and #2
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station essential batteries
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hydrogen dilution system - trains 1 and 2
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auxiliary feedwater - train 2
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motor driven feedpump No substantive concerns were identified as a result of the walkdowns. System lineups and major flowpaths were verified to be consistent with plant drawings and the updated safety analysis report (USAR). Equipment material condition was found to be excellent in all cases. Pump / motor fluid levels were within their normal bands. Only very minor oil and fluid leaks were noted on occasion. Local and remote controllers were properly positioned and attendant instrumentation appeared to be functioning correctly.
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08 Mincellaneous Operations issues (92700) (92901)
08.1 (Closed) LER 50-346/94-002-00: Anticipatory Reactor Trip System inoperable Due to Failure to Trip Main Feedpump Turbine. With the plant at approximately 55%
power, operators removed the #2 main feedpump (MFP) from service and isolated the pump to repair a minor steam leak on an associated vent line. Because the main feedpump turbine was not actually tripped when it was removed from service, the turbine governor hydraulic control oil remained at normal pressure, however, the anticipatory reactor trip system (ARTS) logic monitored control oil pressure as an input for ARTS initiation. With the MFP shutdown at the indicated power level, the ARTS system could not have responded as designe The safety consequence of this event was minimalin that the ARTS system was designed to anticipate a reactor high pressure condition upon a trip of the main turbine above 40% power and/or loss of normal feedwater flow. The reactor protection system remained operable during this timeframe and would have ;
provided adequate protection to the core in response to a loss of normal feedwater l or a main turbine trip above 40% power. No credit for ARTS was taken in the licensee's l>SAR accident analyse Following the event, additional operator training was provided to address proper MFP shatdown/ trip actions, and applicable procedures were revised to clarify MFP shutdown requirements and reduced power operation utilizing one MFP. In addition, the licensee prepared and implernented plant modification 95-0011, to install test toggle switches in the ARTS input logic. This allowed operators to input MFP availability status to ARTS, independent of whether a MFP was actually trippe .2 (Closed) Unresolved item (50-346/96005-01(DRP)): Shift manager proficiency watchstanding requirements were vague. The licensee subsequently upgraded the guidance for shift manager proficiency watches, preliminarily in the form of a memorandum from the operations manager to all shift managers. The guidelines were in process of being translated to the shift manager qualification journals at the end of the inspection perio .3 1Cigsed) Inspection Followun item (50-346/96005-02(DRP)): Operations shift work schedules not consistent with technical specifications (TS). TS 6.2.3 was subsequently revised to better reflect current shift work schedules.
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11. Maintenangs M1 Conduct of Maintenance z
M1.1 Maintenance and Surveillance Activities (61726) (62707)
The following maintenance and surveillance tasting activities were observed / reviewed during the inspection period:
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MWO 7-97-0235-02 Service Water Valve SW 1381 Troubleshooting
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MWO 3-97-4402-01 59% Undervoltage Device Testing
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M'NO 3-97-1702-01 AFW #2 Turbine Governor Preventive Maintenance
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MWO 3-97-0721-02 EDG #1 Quarterly Engine Checks j
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MWO 3-96-5321-02 ECAD Testing of AFW Train 2 Electrical Cables
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MWO 196-0905-00 AFW #2 Pump Bearing Drainline Leak Repair j -
DB-ME-03040 SFAS Sequencer C1 Bus Undervoltage Relay j Response Time Test l
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DB-MI-05254 Nuclear instrumentation NI-05 (RPS Ch 2) Power i Range Adjustment
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DB-SP-03136 - Decay Heat Pump 1 Quarterly Pump and Valve Test l - -
DB-SS-03042 Control Room Emergency Ventilation System j Train 2 Monthly Test i
j Maintenance activities observed during the inspection period were conducted in a j controlled, coordinated manner. Pra-evolution briefs were thorough and performed
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knowledgeable of equipment being worked on, and performed work in accordance l with maintenar'ce work order (MWO) instructions and maintenance procedures. In-l field troubleshooting activities were well controlled and adequately supervised.
1- Surveillance testing activities observed during the inspection period provided
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adequate assurance that the tested equipment could perform as designed. The i inspectors independenti, .arified that the equipment functioned (under test
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conditions) per USAR descriptions. However, several undervoltage relays were i found not being tested within the frequency required by technical specifications j (reference Section M3.2).
M3 Maintenance Procedures and Documentation i!
l M3.1 Forelan Material Exclusion Controls l
l Insoection Scoce (62707)
The inspectors observed portions of an auxiliary feedwater (AFW) train 2 maintenance outage conducted on March 1 Observations and Findinas The inspectors observed that the in-field maintenance activities weie conducted in a controlled manner. Good coordination and communication were noted. in particular, the maintenance craft utilized appropriate foreign material exclusion (FME) controls during work on a drain line leak in the AFW pump inboard bearing oil syste '
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- However, the inspectors noted the MWO (196-0905-00) work package for the ,
bearin0 drain line leak did not provide for any FME controls during the job.
2 Followup discussions with maintenance personnel revealed that the maintenance
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administrative controls including procedure DB-DP-00005, " Foreign Material .
Exclusion." The planner had determined that a " Foreign Material Exclusion
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Requirements" form did not need to be included in the MWO package because <
l DB-DP-00005 allowed the work to be classified as not requiring FME controls. The '
, inspectors were concerned that the procedure did not appear to provide sufficient guidance to the maintenance craft to preclude possible introduction : f foreign _.
i materialinto the AFW bearing oil system. The maintenance crew working the job .
! also had recognized that omission of FME controls from the MWO package was l
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inappropriate and took additional actions to ensure FME was addressed during the I work. In addition, DB-DP-00005 was unclear in several areas, including the cleanliness i classifications of important systems. For instance, AFW was categorized as a Class C system. However, the procedure did not indicate whether this included ,
AFW turbine steam supply / exhaust, AFW condensate, es well as bearing oil.
I The licensee agreed that additional review of station FME procedural controls 1 i should be conducted. At the conclusion of the inspection period, the licensee was '
i evaluating additional areas and equipment for possible inclusion into DB-DP-00005.
i- The procedure it;, elf was also being reviewed in an attempt to better clarify pre-
- existing requirements. This matter is considered an inspection followup item i pending resolution of this matter and subsequent inspector followup review
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(50-346/97004-01(DRP)).
j' M3.2 Essential Bus Loadshed/Emeraency Diesel Generator Start Relavs No Tested in
} Accordance With Technical Soecifications 1 ' Insoection Scone (617261 i
< The inspectors observed the performance of MWO 3-97-4402-01,"59 %
! Undervoltage Device Testing" and reviewed associated procedures on March 25,
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d i Observations and Findinas l
The relay observed under test was designed to function in concert with other ;
similar relays in a logic scheme to sense a low voltage condition (59% of normal) :
on its assigned class E bus, initiate a bus loadshed, and send a start signal to the j associated emergency diesel generator (EDG). Monthly testing of these re!ays to !
ensure proper tripping at the specified voltage and time delay was required by Technical Specification (TS) 3.3.2.1, " Safety Features Actuation System", Yables 3.3-4 and 4.3 2. The inspectors identified that the licensee was not verifying the !
59% undervoltage device allowable trip setpoints for any of the relays on n monthly ,
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Specifically, TS Tabie 3.3-4, under " Sequence Logic Channels" item b, included a note that indicated that allowable trip values should be verified during channel functional tests. TS Table 4.3-2 required that channel functional tests be performed monthly. However, the licensee was testing the voltage trip setpoint of the subject relays about every 4-6 months and testing the relay time delays every 18 month Following discussions with plant personnel, potential condition adverse to quality report (PCAOR) 97-0412 was promptly generated, and additional licensee review determined that the 90% undervoltage device relays had also not been tested at the appropriate frequency. The operating shift declared all of the 59% and 90%
relays inoperable. Asscciated action requirement 15 of TS Table 3.3-3 was then applied which consequently required both EDGs to also be declared inoperable. In lieu of taking the required actions associated with both EDGs being inoperable (which included initiation of a plant shutdown within two hours), TS 4.0.3 was invoked which permitted a 24-hour delay in carrying out the actions associated with two inoperable EDGs in order to complete the subject testin Within the er. suing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the plant conducted time response and voltage testing on all applicable 59% and 90% relays with no deficiencies noted. This allowed both EDGs to be declared operable. Subsequently, a monthly testing schedule for the relays was establishe Durir g f aeir review of this matter, the inspectors noted that engineering was in process of implementing a program to evaluate instrumentation testing associated with the technical specifications per NRC Generic Letter (GL) 96-01, " Testing of Safety-Related Logic Circuits." Although the individual reviews had just begun on a pilot basis, a charter had been prepared which included a number of review . riteria and a sequence to follow when performing the details of the review, it was unclear whether the pilot program would have identified the 59% relay test frequency problem. However, following NRC identification that the EDG relays had not been tested within the required frequency, the licensee revised the charter to specifically include applicable " lessons learned," and to incorporate additional detailed review criteria to address this area. As a result, it appeared that the current review program would be sufficient to identify similar problems if implemented as intende The GL 96-01 related reviews were anticipated to be completed before the summer of 199 ,
c. Conclusions The failure to perform setpoint verifications during monthly testing of the 59%
undervoltage relays is considered a violation of TSs 3.3.2.1 and 4.3.2.1.1 and associated TS tables (50-346/97004-02(DRP)).
, When tested, all relays met their required acceptance criteria, indicating that they l would have responded as designed if called upon.
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The engineering review program to evaluate the testing of safety related circuits in response to GL 96-01 appeared adequate to identify possible similar problem M8 Miscellaneous Maintenance issues (92700) (92902)
M8.1 (Closed) LER 50-346/94-004-00: Containment Hydrogen Purge /n/et Screen Not
/nsta//ed. This matter involved licensee identification during a refuel outage containment walkdown that the containment hydrogen purge (CHP) srstem inlet screen was not installed. The function of the screen was to prevent debris and foreign material from entering the inlet piping to the CHP system and impacting on operation of the containment isolation valves (CIVs), during and immediately following a loss of coolant accident (LOCA). During normal operation, the absence of the screen was not of concern in that no debris was postulated to be generate Debris could only enter the system and affect the functioning of the CIVs during times when the valves were opened in conjunction with elevated containment pressures. However, the valves were only opened during either normal power operation or to control containment hydrogen concentration several days following a LOCA. As such, the potential for debris entry into the line was minima The inspectors subsequently verified that the screen was reinstalle M8.2 (Closed) LER 50-346/94-005-00: RFS Channe/ 4 Response Times Exceeded. With the plant shut down and in mode 5, station testing activities identified that response times for reactor protection system (RPS) channel 4 functional units associated with reactor coolant (RC) low pressure and RC high pressure exceeded technical specification limits. The measured response times exceeded allowable .
limits by 0.125 seconds and 0.165 seconds for RC high pressure and RC low pressure respectabl The licensee subsequently determined the cause for the increased response times was due to inadvertent installation of an incorrect buffer amplifier module in the RPS channel 4 logic cabinet. Safety consequences were determined to be minimal in that RPS channel 4 logic would still function to trip the unit at the appropriate setpoints, in addition, the other three RPS channels were verified to trip at both the appropriate setpoints and within the necessary response times. The RPS logic was configured to trip the unit upon tripping of any two out of four channel The incorrect buffer amplifier was replaced during the outage and RPS channel 4 was thereafter tested to ensure appropriate response timin M8.3 (Closed) Unresolved item (50-346/94005-02(DRP}l: Pilot operated relief valve (PORV) inadvertently opened during maintenance activities. During replacement of a reactor trip module (RTM) in RPS channel 1, the PORV inadvertently opened for approximately five seconds. RCS pressure decreased about 60 psi before the PORV reclosed. Troubleshooting determined that an electrical spike had occurred in the RTM causing a momentary erroneous high RCS pressure signal to the POR l
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Subsequently, the licensee determined tnat the RTM involved in this event was procured from the Sacramento Municipal Utility District (SMUD) along with several other spares. The subject RTM had been modified while at SMUD to an alternate configuration that caused the resultant problem. The licensee's receipt inspection had failed to identify the modified configuration. All similar SMUD RTMs were thereafter tested to ensure their proper configuration prior to returning them to store . Enaineerina E2 Engineering Support of Facilities and Equipment E Shield Buildino Blowout PanelInstallation i Insoection Scooe (37551) (717071 The inspectors performed a walkdown of the shield building blowout panels and reviewed associated installation requirements on April Observations and Findinos During the walkdown, it was noted that caulking was used to seal the edges of the blowout panels. The inspectors subsequently questioned whether the adhesive properties of the caulking could affect the pressure at which the blowout panels would release. The shear bolts installed in the panels were designed to release at a verv specific pressure range (0.65 to 0.67 psid). This difference amounted to approximately 100 lbs total pressure across the surface area of each panel. The inspectors were concemed that the addition of caulking could add greater than 100 lbs of additional holding strengt The inspectors also reviewed the installation requirements associated with the blowout panel shear bolts. The inspectors determined that the guidance provided to personnelinstalling shear bolts was minimal. No procedural requirements were in place, nor were the installation instructions provided in associated drawings easily translate A subsequent walkdown of the blowout panels revealed that several shear bolts were not installed per the specified drawing requirements. Several bolts were installad without all necessary washers (no washer installed on the head side of the bolt). Five shear bolts were also found installed without use of locktite on the nuts to ensure that they remained in a "fingertight" condition. In addition, during inspection of the shear bolts, one shear bolt failed at much less than normal release
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Pending completion of inspector review of the aforementioned issues, this matter is considered an unresolved item (50-346/97004-03(DRP)).
E2.2 Fire Door Armor Plates Insoection Scone (71750)
The inspectors performed a walkdown of several fire doors during the inspection
period. The inspectors reviewed door compliance with underwriter's laboratory (UL) certification requirements.
i Observations and Findiriga The inspectors noted that several fire doors installed in the plant utilized armor
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the standard indicated that larger kickplates should be tested and evaluated for i
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acceptability. The licensee used armor plates /kickplates of a size that covered approximately 40% of door surface area in several applications in the plant. The
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inspectors requested the associated testing and/or UL approval documentation to
ensure that the installed configurations were acceptable. At the end of the inspr' ion period, the licensee determined that the information supporting the accseptability of installing armor plates on station fire doors was not readily retrievable. As such, the station could not easily determine that use of armor
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plating was acceptable and declared the subject doors inoperable and established compensatory fire watches while completing their review. Pending the retrieval of i the subject information and subsequent inspector review, this matter is considered an unresolved item (50-346/97004-04(DRP)).
E8 Miscellaneous Engineering issues (92903)
E (Closed) Violation (50-346/94016-02(DRS)): Failure to adequately control a design
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change resulting in use of non-qualified fittings on RCS attached piping. In response, the licensee performed an engineering evaluation of the installation and concluded that the interim use of the non-qualified parts was acceptable. A
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maintenance work order (MWO 7-94-1288-01) was initiated to replace the subject pipe fittings with those of a qualified type. This MWO was completed during the tenth refuel outag E8.2 (Closed) Insoection Followuo item (50-346/94016-03(DRS)): 10 CFR 50.59 safety reviews (safety evaluation screening process) appeared to be limited and narrowly focused in certain applications. The screening process did not ensure that all seven questions used to determine whether an unreviewed safety question existed, were evaluated. The licensee subsequently revised plant procedure NG-EN-00304,
" Safety Review and Evaluation" including an attached safety review form (ED 7818-7) to better define and delineate screening criteria. In addition, training for applicable personnel was conducted to further define the screening requirement *
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Emphasis on the function of structures, systems, and components, in addition to a literal application of the USAH, was reinforce E8.3 (Closed) Unresolved item (50-346/95007-01(DRP)): Electrical equipment operating at greater than nameplate rating. Engineering subsequently performed a downstream load evaluatiori for busses YAU and YBU and determined the appropriate operating voltage ranges for each. The design criteria manual (DCCM-040) as well as the 120 VAC system description were revised to incorporate the modified output voltage limits. A check of rectifier YRF3 output voltage was added to the operators zone reading sheets with guidance incorporated to have electricians adjust the output voltage when needed to ensure conformance with nameplate rating. None of the subject equipment was determined to be inoperabl The inspectors since have identified no additional electrical equipment operating outside of nameplate rating limits. As such, this matter is considered closed. The inspectors will continue to monitor electrical equipment operating parameters as part of the routine inspection progra E8.4 (Ckped) Unresolved item (50-346/96003-06(DRP)): Appropriate corapensatory measures not taken for inoperable radiant energy shields. This matter was subsequently reviewed and concluded with issuance of a violation in inspection report 50-346/9600 E8.5 (Closed) Insoection Followuo (50-346/96005-06(DRP)): Spent fuel reconstitution activities not described in the updated safety analysis report (USAR).
Subsequently, USAR Section 9.6 was revised to include a description of those fuel reconstitution and fuel recaging activities anticipated to be performed in the spe.it fuel pno E8.6 (Ooen) Unresolved item (50-346/97003-04(DRP)): Service water strainer blowdown valves failed to close on several occasion Relay Failure: No specific root cause for two similar failures of the service water strainer control circuit latch and trip coil to operate correctly had been determined by plant engineering. Therefore, age degradation was determined to be the general root cause. Consequently, the control circuit latch and trip relays for all three service water pumps were replaced with equivalent relays. During the engineering extent-of-condition review, it was found that these type relays were also installed in safety features actuation system components control circuits. Therefore, since safety related, risk significant equipment may also be affected by the aging phenomena, further inspector followup is warranted in this area.
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Control Logic issue: The inspectors determined that a separate failure of the service water strainer blowdown valve to properly operate, where the valve was found partially open instead of fully open as in the first two failures, was due to a design flaw. Upon review of the controllogic drawings, the inspectors identified that the control logic design allowed for the service water strainer blowdown valves to deenergize at up to 25% open under certain conditions. When brought to the 12 -
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attention of plant engineering, they agreed that the valve had functioned as designed. Consequently, at the end of the inspection period, plant engineering initiated a request to perform a modification to the blowdown valve control circuit to eliminate the design fla Additionally, engineering determined that other control circuits in the plant may also have similar logic circuitry. Since this feature may be undesirable in other circuits as well, the inspectors will continue followup of this matte The inspectors investigated whether this design feature may have caused the service water system to have been inoperable in the past. The condition that was evaluated was for a design basis loss-of-coolant-accident, concurrent with a service water strainer blowdown valve 25% open, with ultimate heat sink temperatures at the TS limit of 85 degrees. Because of the relatively small diversion of service water flow (about 400 gpm), and because of design margins that were available in the containment air coolers, decay heat coolers, and other service water cooled equipment, the inspectors determined that the service water system would have still been able to perform its function as described in the updated safety analysis repor E8.7 (Closed) Unresolved item (50-346/97003-05(DRP)): Service water pump performance curve out-of-date. Thc inspectors found that the service water pump curve in the pump curves book did not match the curve attached to the service water pump quarterly surveillance test procedure. The curve in the pump curve book reflected pump performance for the originally installed pump. The curve in the surveillance test procedure reflected pump performance for a replacement pump that had been installed in 199 It was determined that the curve from the pump curve book was used as a basi.:ne curve for the performance of calculation C-NSA-011.01-003, " Allowable Service Water Flow Diversion During Cold Weather". This calculation was performed in order to provide a basis for being able to bypass service water flow through a standby component cooling water heat exchanger during cool weather in order to assist in maintaining service water header pressure below the service water relief valve setpoint. The original curve was provided by the manufacturer and contained performance data over a much wider range than that of the replacement pump curve. The new pump curve data range was restricted by the limitations of the in-plant installation and was not as useful as the performance data supplied by the manufacture The inspectors discussed this with the engineer who performed the calculation and reviewed the calculation itself to determine if the conclusions might have been adversely affected by use of the older curve. The inspectors determined that, due to the older pump curve performance data being more conservative than the newer pump curve, and because the old pump curve was purposely degraded to reflect
! worse case end-of-life and fouling characteristics, the use of the old pump curve
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instead of the new pump curve was conservative and had no adverse effect on the r
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.,- 1 conclusions of the calculation. Additionally, a revision to the calculation utihzed the 4 new pump curve and demonstrated that the original conclusions remained vali )
However, engineering personnel were planning as a corrective action, to add information to the pump curve book that indicated that the pump performance data j that it contained reflected originally installed equipment, and not necessarily I currently instal lcd equipment. Because the plant had initiated appropriate corrective actions to address this issue, and because no safety consequence was identified, this matter is considered close IV. Plant Suonort R1 Radiological Protection and Chemistry (RP&C) Controls (71707) (71750) j During the inspection period, the inspectors conducted frequent walkdowns of the radiological restricted area (RRA). Radiation, high radiation, and contaminated area controls and postings were verified to be in conformance with radiation protection (RP) procedures. In addition, the inspectors independently verified that actual area i radiation levels were consistent with current radiological surveys and postings. An increase in the quality and quantity of radiological postings was noted this inspection period. Personnelinvolved in the performance of activities observed by the inspectors, conducted those activities in accordance with RP program j requirement R2 Status of RP&C Facilities and Equipment (71707) (71750)
A sample of portable radiation survey instruments were inspected during the inspection period. The inspectors verified that they were functional and properly calibrated within required timeframes. Also, personnel contamination monitors (PCMs) located at the RRA exit point were verified to be functional and appropriately calibrate P1 Conduct of EP Activities (71750)
The inspectors observed portions of an emergency planning drill that was conducted on April 2,1997. Drill objectives appeared to be met, good participation by station personnel was evident, and the scenario was well controlled and realistically implemented. Operations personnel manning the control room simulator appropriately implemented the emergency plan, effectively utilized emergen::y procedures and effectively controlled simulator equipment. Good utilization of the operations support center and technical support center (TSC) resources was mad Although course of action recommendations were made from the TSC, operations
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personnel thoroughly evaluated the recommendations to ensure their appropriateness. An effective post drill critique was conducted.
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P2 Status of EP Facilities, Equipment, and Resources (71750)
l The inspectors walked down the emergency control center (ECC), technical support .
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center (TSC), and operations support center during the inspection period. All three l l emergency response facilities appeared to be well maintained and in a suitable state l
l of readiness. Associated equipment appeared functional and appropriately staged j to adequately support potential activation of the station emergency plan. Personnel
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access to the ECC and TSC was verified to be restricted as required by the licensee's progra S2 Status of Security Facilities and Equipment (71750)
I During the inspection period, the inspectors verified the integrity of protected area l (PA) fence barriers, that PA isolation zones were appropriately marked and l adequately free of foreign objects, and that PA lighting equipment and illumination levels were not noticeably degraded. Personnel processing facility ingress monitoring equipment was verified to be functional and appropriately maintaine V. Manaaement Meetinaa X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the l
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conclusion of the inspection on April 14,1997. The licensee acknowledged the findings presented.
l The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie X3 Management Meeting Summary On March 21,1997, a meeting was conducted onsite to present the results of the NRC's most recent systematic assessment of licensee performance (SALP). NRC Region 111 l
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management participating in the meeting included the Regional Administrator; Director, Division of Reactor Projects: Deputy Director, Division of Nuclear Materials Safety; and Chief, Reactor Projects Branch 4. The NRC Office of Nuclear Reactor Regulation was represented by the Project Manager for Davis-Besse. During the meeting, the ratings for
- each SALP functional area were discussed as well their bases and significant factors l considered by the SALP board during their deliberations.
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PARTIAL LIST OF PERSONS CONTACTED
! Licensee J. K. Wood, Vice President, Nuclear J. H. Lash, Plant Manager
R. E. Donnellon, Director, Engineering & Services T. J. Myers, Director, Nuclear Assurance L. W. Worley, Director, Quality Assurar.ce L. M. Dohrmann, Manager, Quality Services J. L. Michaelis, Manager, Maintenance J. L. Freels, Manager, Regulatory Affairs 1 M. C. Beier, Manager, Quality Assessment
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W. J., Molpus, Manager, Nuclear Training I L. D. Hughes, Manager, Davis-Besse Supply D. R. Converse, Manager, Davis-Besse Business Services R. J. Scott, Manager, Radiation Protection H. W. Stevens, Manager, Nuclear Safety and inspection J. W. Rogers, Manager, Plant Engineering G. A. Skeel, Manager, Security
. D. L. Eshelman, Manager, Operations j
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D. H. Lockwood, Supervisor, Compliance A. J. VanDenabeele, Supervisor, Quality Analysis F. L. Swanger, Supervisor, Nuclear Engineering
. D. R. Wuokko, Supervisor, Licensing L. A. Bonker, Supervisor, Radiation Protection i G. W. Gillespie, Superintendent, Chemistry T.- J. Chambers, Shift Manager, Operations i j D. L. Miller, Senior Engineer, Licensing
. G. M. Wolf, Engineer, Licensing i
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INSPECTION PROCEDURES USED
- lP 37551
- Onsite Engineering i IP 61726: Surveillance Observations l IP 62707: - Maintenance Observation j IP 71707: Plant Operations i IP 71750: Plant Support l- IP 92700: Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor l
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IP 92901: Followup - Plant Operations
- - IP 92902: Followup - Maintenance
{ IP 92903: Followup - Engineering
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ITEMS OPENED, CLOSED, AND DISCUSSED Ooened 50 346/97004-01(DRP) IFI FME procedure unclear 50-346/97004-02(DRP) VIO Failure to perform monthly technical specification testing of undervoltage relays 50-346/97004-03(DRP) URI Blowout panel shear bolt installation 50-346/97004-04(DRP) URI Armor plate certification questioned
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Closed 50-346/96005-01(DRP) URI Shift manager proficiency watchstanding requirements were vag'se 50-346/96005-02(DRP) IFl Operations shift work schedules not consistent with the technical specifications 50 346/94005-02(DRP) URI Pilot operated relief valve inadvertently opened during !
maintenance 50-346/94016-02(DRS) VIO Inadequate control of a design change 50-346/94016-03(DRS) IFl Safety reviews were limited and narrowly focused 50-346/95007-01(DRP) URI Electrical equipment operating at greater than nameplate rating 50-346/96003-06(DRP) URI Appropriate compensatory measures not taken for inoperable radiant energy shields 50-346/96005-06(DRP) IFl Spent fuel reconstitution not described in USAR 50-346/97003-05(DRP) URI Service water pump performance curve out-of-date 50-346/94-002-00 LER Anticipatory reactor trip system inoperable due to failure to trip main feedpump turbine 50-346/94-004-00 LER Containment Hydrogen Purge inlet Screen Not Installed 50-346/94-005-00 LER RPS Channel Four Response Times Exceeded l Discussed 50-346/97003-04(DRP) URI Service water strainer blowdown valve failures 17 ,
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, LIST OF ACRONYMS USED l- AFW Auxiliary Feedwater ARTS Anticipatory Reactor Trip System CFR Code of Federal Regulations CHP Containment Hydrogen Purge
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CIV Containment Isolation Valve ECAD Electronic Characterization and Diagnostics ECC Emergency Control Center EDG Emergency Diesel Generator
. ESF Engineered Safety Feature j FME Foreign Material Exclusion
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GL Generic Letter
- IFl Inspection Followup item i
IR inspection Report
, LER Licensee Event Report MFP Main Feed Pump
- MWO Maintenance Work Order i NFPA National Fire Protection Association NI Nuclear instrumentation NRC Nuclear Regulatory Commission NRR Nuclear Reactor Regulation PCAOR Potential Condition Adverse to Quality Report PCM Personnel Contamination Monitor PORV- Pilot Operated Relief Valve psi pounds per. square inch psid pounds per square inch differential RC Reactor Coolant RCS Reactor Cooiant System RP Radiation Protection RPS Reactor Protection System RRA Radiological Restricted Area RTM Reactor Trip Module SMUD Sacramento Municipal Utility District TS Technical Specification TSC Technical Support Center UL Underwriter's Laboratory VIO Violation
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