IR 05000346/1997006
| ML20148U197 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/26/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20148U195 | List: |
| References | |
| 50-346-97-06, 50-346-97-6, NUDOCS 9707100086 | |
| Download: ML20148U197 (20) | |
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U. S. NUCLEAR REGULATORY COMMISSION l
REGION lli !
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l Docket No: 50 346 License No: NPF-3 i
i Report No: 50-346/97006 l l
Licensee: Toledo Edison Company !
Facility: Davis-Besse Nuclear Power Station
Location: 5503 N. State Route 2 l Oak Harbor, OH 43449 !
Dates: April 14 - May 27,1997
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inspectors: S. Stasek, Senior Resident inspector K. Zellers, Resident inspector D. Jones, Reactor Inspector Approved by: Christopher G. Miller, Acting Chief Reactor Projects Branch 4
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EXECUTIVE SUMMARY Davis-Besse Nuclear Power Station NRC Inspection Report 50-346/97006 This inspection included aspects of plant operations, maintenance, engineering, and plant support. The report covers a six-week period of resident inspection; in addition, it includes the results of an announced inspection by a regional specialis Ooerations
- Control room operators maintained suitable operational control of the unit throughout the inspection period. Unit cooldown and heatup evolutions were performed in a deliberate, controlled manner and in accordance with plant procedures (Section 01.1).
- Engineered safety features and important-to-safety standby systems were walked down and verified appropriately lined up and maintained in excellent material condition (Section 02.1).
- Equipment and systems responded to a May 4 plant trip as designed. Operator actions in response to the transient were timely and conducted in accordance with station procedures (Section 02.2).
- Control room personnel did not recognize that a post accident monitoring system hot leg temperature indicator was inoperable until identified by the NRC. The indicator appeared to have been inoperable for at least a week (Section 02.3).
- A pre-startup checklist verification step was signed as complete without engineering supervision recognizing that Technical Specification (TS) requirements mentioned in the procedure had been deleted from TS (Section 03.1).
Maintenance
- Surveillance testing and maintenance activities reviewed during the inspection period, were conducted in accordance with plant procedures. Clearances prepared to support maintenance work adequately protected both equipment and personnel (Section M1.1).
- During inspector review of emergency core cooling system (ECCS) leakage issues, the maximum allowed totalleakrate from each ECCS train was determined to be 20 gallons per hour. This leakrate appeared to not have been verified through periodic testing. As such, plant engineering and the NRC were evaluating the need to perform additional testing (Section M6.1).
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Ennineerina
- Requirements governing the preparation and implementation of action plans were not well delineated. The inspectors were concerned that action plans could potentially be used in lieu of approved plant procedures. Review of this matter was continuing at the end of the inspection period (Section E3.1).
- Good performance by the potential condition adverse to quality reporting (PCAOR)
review board was exhibited during the inspection period (Section E7.1).
Elant Suonort J
- The NRC identified outdated locally posted radiological survey maps in several areas of the radiologically restricted area (ftRA) on one occasion. Radiation protection (RP) personnel indicated the local survey maps had not been updated to reflect plant shutdown conditions since the plant trip approximately ten days earlier due to the associated increased work load on the RP technicians with the plant shut dow Plant personnel performing activities in the RRA had also failed to recognize the local survey maps were not current (Section R1.2).
- An April 30 emergency preparedness drill was satisfactorily conducted (Soction P1.1).
- The Army Corp of Engineers predicted potentiallocal area flooding during the upcoming summer and fall. As a result, the plant initiated a review of the site emergency plan evacuation routes during the inspection period to determine plan adequacy (Section P3.1). l
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Report Details Summarv of Plant Status The plant began the inspection period operating at about 100 percent power. On May 4, the main transformer deluge system inadvertently actuated and caused the transformer to trip, initiating a reactor trip. Subsequent troubleshooting of the main transformer identified a need to replace it. The unit remained shut down until May 26 when the reactor was again made critical. At the end of this inspection period, the reactor was in the final stages of plant startup, with the reactor in Mode 1, operating at low power levels, and plant personnel completing post maintenance testing activities on the main generator and main transforme . Operations 01 Conduct u Operations 01.1 General Comments (71707) l Control room operators maintained suitable control of the unit throughout the inspection period. Unit cooldown and heatup evolutions were performed in a deliberate, controlled manner and in accordance with plant procedures. Reactor operators were knowledgeable of the bases for alarming control room annunciators and were cognizant of in-progress plant evolution !
Equipment operators adequately monitored the status of operating equipmen I Lubricating oil levels, motor and pump temperatures, fluid leaks, as well as general plant conditions were checked shiftly as required. Discrepant conditions were communicated to operations supervision in a timely manne Adherence to applicable programmatic and administrative controls was observe The status of inoperable equipment was effectively tracked and managed. Startup and operating mode restraint checklists were satisfactorily utilized with one exception (reference Section 03.1).
O2 Operational Status of Facilities and Equipment O 2.1 System Walkdowns (71707)
The inspectors walked down the accessible portions of the following engineered safety features (ESF) and important-to-safety systems during the inspection period:
- emergency diesel generators #1 and #2 e hydrogen dilution system - train 1
- auxiliary feedwater - trains 1 and 2
- low pressure injection system - trains 1 and 2 e containment spray system - trains 1 and 2
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No substantive concerns were identified as a result of the walkdowns. System lineups and major flow paths were verified to be consistent with the updated safety analysis report (USAR), plant drawings, and applicable procedures. Equipment material condition was excellent in all cases. Pump and motor fluid levels were within their normal bands. Very little oil and fluid leaks were noted. Auxiliary equipment necessary for system operability was properly aligned and functionin Local and remote controllers were appropriately positioned and attendant instrumentation appeared to be functioning correctl .2 Plant Trio ' Insoection Scope (93702. 71707)
The inspectors conducted a review of the plant trip that occurred on May 4 to assess plant equipment and personnel response. Alarm recorder and data acquisition and analysis system (DAAS) plots, and control room logs were reviewed. Additionally, the inspectors interviewed operations, maintenance and engineering personne Following the trip, the plant independent safety engineering (ISE) group performed a transient assessment to ensure that plant equipment responded to the trip as .
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designed. The inspectors subsequently reviewed the ISE post trip analysi Observations and Findinas i
The inspectors determined that at about 5:30 p.m., the main transformer deluge j system had inadvertently actuated, resulting in a large amount of low quality water i to be sprayed across the exterior surfaces of the main transformer. About five !
minutes later, an electrical fault caused the main transformer and main generator to trip, which in turn tripped the reacto !
I The inspectors verified that, with minor exceptions, post trip equipment response, '
including control rod drive system, turbine protection system, turbine bypass valves, and main steam safety and atmospheric dump systems all operated as ;
expected. All control rod groups inserted per design and within required response times. As expected,16 of 18 main steam safety valves (MSSVs) initially lifted to
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- control secondary side pressure. However, the lift setpoint of MSSV SP1787 i shifted and the valve subsequently began intermittently lifting at a lower set pressure. Operators decreased secondary side pressure to stop the valve from )
lifting. No adverse effect on cooldown rates resulted from the shift in lift setpoin !
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l The inspectors reviewed ISE's post trip analysis and determined that it accurately assessed the trip and post trip response. The ISE analysis appeared consistent with the NRC assessment of the even '
The plant determined that the most likely root cause for the deluge actuation involved an inadvertent trip of a temperature sensor providing input to the deluge )
system. The deluge initiation logic was arranged with 24 temperature sensors in
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the system, with any single sensor trip or failure able to actuate a transformer deluge. At the end of the inspection period, all 24 temperature sensors providing input to the main transformer deluge system were replaced and engineering was in process of evaluating the benefits of modifying the deluge actuation syste Conclusions Overall, plant equipment and systems responded to the trip as designed. Operator response to the transient was timely and conducted in accordance with station procedures. Plant post trip analysis adequately assessed the transient and was consistent with the inspectors' conclusions. Appropriate corrective actions were implemented to replace the main transformer and deluge system temperature sensor .3 Post Accident Monitorino Instrumentation Found Inocerable insoection Scoce (71707)
The inspectors conducted a routine walkdown of the control room on May 3, and reviewed instrumentation associated with the post accident monitoring system (PAMS), Observations and Findinos l'
The inspectors identified that PAMS instrument Tl RC3B6, channel 1 hot leg temperature indication for reactor coolant loop 1 was indicating about nine degrees lower than the other three hot leg temperature indications. The operations shift reviewed surveillance procedure DB-SC-03180, " Remote Shutdown, Post Accident Monitoring instrumentation Monthly Channel Check," and determined that the subject instrument was inoperable because the difference in temperature indications from Tl RC385 exceeded the procedure acceptance criteria of eight degrees. The licensee entered TS 3.3.3.6, Limiting Condition for Operation, and generated Potential Condition Adverse to Quality Report (PCAOR) 97-0578. This LCO required that the inoperable channel be returned to operable status within 30 days or to place the unit in hot shutdown within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ,
The inspectors thereafter conducted a review to determine how long the condition had existed. A computer point that obtained its signal from the same sensor, ,
T754, was found to be reading about 50 degrees below the value of the control I room indication. A review of the computer point value for the seven preceding i days revealed that the degraded situation had probably existed for at least seven days. Procedure DB-SC-03180 had been successfully performed two weeks prior to the date of inoperability identificatio Although no logs or surveillance tests had been performed on this instrument during the two-week period of time, several control room personnel, including reactor
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operators, senior reactor operators, and shift managers failed to identify this discrepancy during shift turnovers and during performance of their dutie Subsequent troubleshooting of this instrument string during the outage resulted in the instrument string returning to an expected indication value. Both computer point T754 and the PAMS localindication agreed in value. The exact failure mechanism could not be determined, but because the instrument string returned to normal indication during a terminal strip wiring and connection check, the root cause was suspected to have been a poor wiring connection.
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Control room personnel did not recognize a post accident monitoring system hot leg temperature indicator was inoperable until identified by the NRC. Since the system had not been inoperable for greater than the 30-day TS limit, and the surveillance
, requirements were satisfied, no violation was cited for this situatio Because the exact root cause of the failure of the instrument string was not determined, followup inspection activities to track the performance of the instrument string is warranted, therefore this is an inspection followup item (50-346/97006-01(DRP)).
02.4 Containment Walkdowns (71707)
The inspectors performed several walkdowns of equipment and structures within the reactor building (containment) during the forced outage. No substantive equipment and material condition problems were identified. Paint and coatings appeared properly applied to equipment and structures in containment with no evidence of peeling or missing spot In addition, a walkdown of containment with the reactor coolant system at normal operating temperature and pressure was performed on May 25. No significant equipment leakage was identified. As-left material condition and housekeeping were satisfactor .
03 Operations Procedures and Documentation O3.1 Pre-Startuo Checklist Insoection Scone (71707)
On May 26, during evaluation of plant readiness to enter Mode 2 during plant restart, the inspectors reviewed procedure DB-OP-06911, " Pre-Startup Checklist." Observations and Findinos The inspectors noted that Step 7.14.1 of the procedure required that the plant engineering manager sign that surveillances conducted to meet a list of Technical
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two of the Technical Specification surveillance requirements, TS 4.6.1.6 and '
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TS 4.6.5.3 had been deleted from the Technical Specifications via license ;
{ amendment 205 issued February 22,199 l 4: .
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When this matter was brought to the attention of operations personnel, a temporar !
F alteration (TA) was initiated to correct the procedure.- When this was complete, j i engineering personnel re-completed Step 7.14.1 of the procedure prior to entry into
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The inspectors were concemed that engineering supervision had signed Step 7.1 !
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as complete without fully realizing that TS requirements had been deleted. In !
p _ addition, the license amendment change process should have identified that
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DB-OP-06911 referenced the subject Technical Specifications that were to be i
l deleted and should have ensured appropriate revisions were mad !
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Inspectors review of this matter was not complete at the end of the current !
!- inspection period. ' This matter is considered an unresolved item !
j (50 246/9700642(DRP)) pending completion of inspector revie l i
04 Operator Knowledge and Performance J
l t 04.1 Operator Resoonse to Auxiliary Boiler Trio i
- inanection Scone (71707) ,
!. The inspectors observed operator response to an inadvertent trip of the plant !
E auxiliary boiler on May 2 i l' :i Observations and Findinos 1: I
- - The auxiliary boiler tripped, as indicated by a control room annunciator, while the i plant was in hot shutdown (Mode 4) and in process of raising reactor temperature l to enter hot standby (Mode 3). The auxiliary boiler was in service to provide steam-
! to the main condenser air ejectors and main turbine gland seals to maintain main >
[ condenser vacuum.
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Control room personnel were observed to perform immediate and supplemental p actions as directed by alarm response procedures. Control room personnel notified
- the assigned system engineer of the trip. The engineer accompanied the outside--
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assistant shift supervisor to investigate the cause. The plant manager and the -
1 outage director were involved in monitoring control room response and were kept
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apprised on the status of followup actions by the operations superintenden Throughout the auxiliary boiler trip recovery, the primary reactor operator i maintained an appropriate level of cognizance of decay heat removal means and did
{ not appear distracted by recovery actions associated with the auxiliary boiler tri J~
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The operations superintendent encouraged the shif t to consider planning for alternate primary to secondary heat transfer contingencies in the event that main condenser vacuum was lost. Additionally he advised the shift supervisor to maintain his oversight function after the shift supervisor had personally secured the mechanical vacuum pump at one poin When it was evident that the auxiliary boiler would not be immediately available to maintain main condenser vacuum, operators transferred decay heat loads from the main condenser to the decay heat removal system with the anticipation of losing condenser vacuu The root cause of the auxilia/y coiler trip could not be determined. However, the auxiliary boiler served no safety-related function nor provided any critical support function for normal power operation. The auxiliary boiler was subsequently started and placed back into servic ConclusionE Control room operators effectively responded to the auxiliary boiler trip and ensured suitable decay heat removal capabilities were maintained. Overall, good management oversight of control room trip recovery activities was note Miscellaneous Operations issues (92901)
08.1 (Closed) Violation (50-346/95009-01(DRP)): Short Shift Turnovers in response, operations management implemented more stringent supervisory reviews and observations of reactor operator turnover activities. The inspectors have_ since identified no further instances of operators performing short shift turnover .2 (Closed) Violation (50-346/96003-01(DRP)): Emergency Diesel Generator Placed in Standby Without Operators Completing All Procedural Verification Steps. In response, procedure adherence requirements were reinforced to all operations personnel during requalification training. Subsequently, additional procedure adherence concerns were identified, both by the NRC and the plant. .in response to those issues, plant management was in process of restructuring the administrative controls governing operations and surveillance procedure usag II. Maintenance M1 Conduct of Maintenance M1.1 Maintenance and Surveillance Activities (61726. 62707)
The following maintenance and surveillance testing activities were observed or ,
reviewed during the inspection period:
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DB-SC-03077 Emergency Diesel Generator 2184-Day Test
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DB-SC-03115 SFAS 18-month Interchannel Logic Test
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DB-SP-03152 AFW Interlock testing
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DB-SP-03150 AFP 1 Monthly Jog Test
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DB SC-03077 Emergency Diesel Generator 2184-Day Test
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DB-SC-10000 Main Transformer Backfeed Test
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MWO 1-96-0909-00 Repair / Replace PIM402 (SFAS Channel 4 RCS pressure indication)
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MWO 3-96-4938-01 Transformer XO1 Double Testing i
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MWO 1-97-0580-03 Replacement of Main Steam Safety Valve i SP17B7 l Surveillance testing activities observed during the inspection period were conducted in accordance with approved station procedures. Operator adnerence to test procedures was good, test deficiencies were written when required, and procedures were reviewed before use. Equipment problems, even those minor in nature, were brought to system engineers' attention, with appropriate corrective actions take The inspectors verified that tested equipment could perform as described in the updated safety analysis report (USAR).
l Maintenance activities reviewed during the inspection period were conducted in accordance with plant procedures and in a controlled manner. Clearances adequately protected equipment and personnel. Lifted wire logs, when required, were performed and satisfactorily tracked the status of lifted wire Instrumentation was observed to perform adequately after initial repair activities had been performed. Equipment was verified to have been declared inoperable when needed and limiting conditions for operation action statements were observe .
M3 Maintenance Procedures and Documentation M3.1 Inservice Insnection Insnection Scone (73052. 73755)
The inspectors reviewed documents related to nondestructive examination (NDE)
equipment, evaluations, and data associated with the examination of four high pressure injection (HPI) nozzles (50, 51, 58, and 59).
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I I Observations and Findinas i On April 21,'1997, Oconee Unit 2 experienced an event in which a through wall crack developed at the safe end to pipe weld on the HPI makeup line. The thermal ,
fatigue crack was attributed to a gap which developed between the safe end and '
the rolled area of the nozzle thermal sleeve which allowed alternate heating and cooling by hot reactor coolant which flowed through the gap and the cooler makeup -
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At Davis-Besse, HPi connections with thermal sleeves (nozzle 58 provides normal l makeup) were located on each of the four cold legs. The thermal sleeves in nozzles 58 and 59 were replaced in 1988 (reference Inspection Report 50-346/88009(DRS)) with redesigned sleeve !
The licensee decided that the conservative approach in response to the Oconee event would be to inspect all four HPI nozzles and the 16 welds connecting the dissimilar metal welds to the check valves rather than just the one nozzle and four ,
welds subjected to makeup flows. Radiography testing (RT) was performed to identify any gap between the thermal sleeve and nozzle and for rotational '
movement of the thermal sleeve within the safe end. Weld buttons were located at ;
the~ ends of the thermal sleeves to prevent movement of the sleeve. If the sleeve '
rotated,it would smear the weld button which would be visible in the radiograp Ultrasonic testing (UT) was used to inspect the 16 welds between the check valves and the dissimilar metal welds of the nozzles for service induced flaws. The UT ,
was performed using Electric Power Research Institute (EPRI) performance ;
demonstrated ultrasonic equipment and personne The examination data was found to be in accordance with the applicable -
nondestructive evaluation procedures and American Society of Mechanical '
Engineers (ASME) Code requirements. The RT of the four thermal sleeves showed no indication of rotational movement or gap between the thermal sleeve and safe end. The UT of 16 welds showed no indications of service induced flaw Conclusions .
For the areas observed, ASME Code requirements were met and no violations or i deviations were identified. The licensee's approach in the insrection implementation appeared to be proactive and conservativ M6 Maintenance Organization and Administration M6.1 Leakaae Rate Testina of Emeraency Core Coolina System Insoection Scone (37551)
The inspectors reviewed a licensee identified issue where categorization of decay
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heat valves DH18 and DH19 and high pressure injection valves HP31 and HP32 in the plant inservice testing (IST) program required further evaluatio . .
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l Observations and Findinas
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The inspectors noted that engineering attempted to take timely and appropriate !
corrective actions to the self-identified issue. Engineering indicated that DH18 and !
DH19 would be classified as Category C tested valves and HP31 and HP32 as !
Category B tested valves under the plant IST progra [t The inspectors noted that Category B valves, although requiring verification of full closure, did not require quantification of seat leakage. Engineering indicated that ;
several other valves in the emergency core cooling system (ECCS) were similarly classified. The inspectors questioned whether an integrated ECCS design maximum ;
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leakage rate was specified that would encompass allindividual component leakage. . i Engineering performed a review of initial licensing documents and identified that the j maximum allowed ECCS leakage rate per train was 20 gallons per hour (GPH). The ;
inspectors questioned whether the ECCS leakage rate was verified to ensure that it ;
was less than the maximum allowed valu !
l The inspectors also noted that TS 6.8.4.a required implementation of a program to, f in part, address potential primary coolant sources outside of containment, including i leakage from the low pressure injection and high pressure injection systems. At the I conclusion of the inspection period the licensee was evaluating the need to perform 1 an integrated ECCS leakage rate test and whether the TS 6.8.4.a specified program !
was being properly implemented. Pending completion of licensee evaluation and l
subsequent inspector review, this matter is considered an unresolved item l
(50-346/97006-03(DRP)). -l i
M8 Miscellaneous Maintenance lasues (92700, 92902) :
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- M8.1 (Closed) LER 50-346/96-001-00: Inoperable Emergency Core Cooling Syste '
This event involved the plant failure to perform a Technical Specification required j surveillance to pedodically vent all of the high pressure injection system high point i In response, the licensee requested and received a license amendment which - l allowed the facility to be in compliance with the surveillance requirement of the _
Technical Specification until the next refuel outage. During the tenth refuel outage i completed in' June 1996, a plant modification to install a manual vent valve on the '
high point of the associated HPl piping was completed. The NRC had previously evaluated this issue as documented in Inspection Report 50-346/96002. That .
review concluded with issuance of an escalated enforcement actio M8.2 . (Closed) LER 50-346/96-005-01: Inadequate Control of Heavy Loads in the Containment Building. This matter involved an event where an eight ton reactor vessel head lifting tripod was inadvertently traversed over the open reactor vessel in violation of plant procedures as well as NRC requirements associated with the control of heavy loads. The NRC had previously reviewed this matter which
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111. Ennineerina E3 Enginsedng Procedures and Documentation E Control of Action Plans insoection Scone (71707)
The inspectors reviewed operations night order log sheets, one of which included two action plans, during a control room walkdown on May 2 Observations and Findinas Note 6 of the May 16,1997 operations night orders indicated that the water ,
between the containment isolation valves for the pressurizer auxiliary spray line, I and the water between the containment isolation valves for the reactor coolant drain line, should be partially drained per PCAOR 96-1199. The report provided interim resolution of NRC Generic Letter (GL) 96-06 concerns involving thermally induced containment penetration overpressure concerns. The night orders included two action plans that were to be used for accomplishing the partial draining evolutions. These lines had been partially drained of water prior to the forced outage that had started on May 4, but the lines had been used to support plant cooldown and required re-draining prior to startu The two action plans described the specific steps for the operators to take in order to partially drain the lines. The steps described obtaining shift supervisor permission, notification of radiation protection personnel, obtaining water collection bottles, verification of equipment lineups, manipulating plant equipment, draining -
water from the lines, and restoring plant equipment. Aiditionally, there were signature blocks for the preparer, approver and performer of the action plan. The action plans appeared to direct actions similar to procedures without having undergone the same review and approval process as procedure ,
i The only available documentation that prescribed the requirements for the generation, control and execution of action plans consisted of a plant engineering oolicy, PE-05. This policy provided guidance for plant engineering personnel for the peparation of plant engineering action plans to address administrative or hardware 4 related activities. However, the action plans for' addressing the interim resolution of l GL 96 06 concerns were generated, approved, and executed by operations
' personnel. No administrative controls to address operations use of action plans werre in plac At the end of the inspection period, the inspectors as well as plant personnel were reviewing policy PE-05 to determine if it was sufficient to control generation and use of action plans in general. - Additionally, inspector review to determine if the steps performed by the two GL 96-06 action plans should have been prescribed by
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E7 . Quality Assurance in Engineering Activities l'
E PCAQR Review Board 437551)
The inspectors observed a PCAOR review board convened to assess a number of PCAORs deemed ready for review. Good participant knowledge of corrective action
program requirements was demonstrated during the meeting. The PCA. ors which I had inco.nplete documentation and/or required followup were sent back to the evaluator for further action. Also, board members satisfactorily discussed the appropriateness of initial assessments, remedial actions, root causes and actions to prevent recurrence. A repetitive problem regarding the inadvertent blocking of fire doors was addressed by assigning a root cause and corrective action to prevent recurrence evaluation to be done.
E8 Miscellaneous Engineering issues (92700, 92903)
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E (Closed) Unresolved item (50-346/96002-05(DRP)): Potential Inadequate 2-Evaluation of Electrical Hot Shorts. The NRC had previously reviewed this issue as j documented in Inspection Report No. 50-346/96008 with the resultant being issuance of escalated enforcemen !
b E8.2 (Closed) Unresolved item (50-346/96003-07(DRP)): Reactor Coolant Pump Motor i Oil Piping Nt,t Configured as Required. The licensee subsequently submitted LER 50-346/97-004 00 that discussed the details of each of the individual issues. This matter will be reviewed as part of the inspectors' followup to the subject LE E8.3 (Closed) Insoection Followuo item (50-346/96010-06(DRP)): Effects of Mirror insulation Debris. The inspectors reviewed a pre-existing engineering evaluation that addressed the potential effects of debris generated from a high energy line break in containment in areas near the emergency sump. The inspectors also performed a walkdown of the containment adjacent to the emergency sump and determined the amounts of mirror insulation utilized were consistent with the engineering evaluation. An engineering' evaluation concluded that mirror insulation debris would not adversely affect proper functioning of the emergency sump screenin E8.4 (Closed) Deviation (50-346/95004-02(DRS)): Failure to Trend Low Pressure Transmitter Performance. This issue involved a commitment the licensee had made to trend low pressure rosemount transmitter performance via weekly reviews of
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computer data points. During the inspection it was identified that 12 of the transmitters were not being trended via the computer. Engineering subsequently determined that the subject transmitters did not have inputs to the compute Although the actions committed to in the licensee's response to NRC Bulletin 90-01, Supplement 1, were inaccurate for the subject 12 transmitters, performance was being trended using alternate mean ~
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E8.5 (Closed) LER 50-346/96-007-00: Control Room Emergency Ventilation System Design Bases Calculation Error. This issue involved an identification by engineering that control room emergency ventilation system inleakage limits were based upon inaccurate assumptions and calculations. Once identified, the licensee took appropriate corrective actions to further limit the operational inteakage rates to less than the newly analyzed limit The NRC previously reviewed this matter as documented in inspection Report No. 50-346/9601 IV. Plant Suncort R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 General Comments (71750)
During the inspection period the inspectors conducted frequent walkdowns of the radiologically restricted area (RRA). Radiation, high radiation and contaminated area controls and postings were verified to be in conformance with radiation protection (RP) procedures. Walkdowns of the containment were conducted during the most recent forced outage to verify appropriate radiological postings, up-to-date and accurate surveys were conducted, and personnel adherence to RP procedural requirements were adequately implemented in containmen R1.2 Outdated Locally Posted Radioloaical Surveys Insoection Scone (71750)
On May 15, the inspectors conducted a tour of the auxiliary building and reviewed selected locally posted radiological survey map Observations and Findinos The inspectors noted that severallocally posted radiological survey rnaps were dated April 28, predating the May 4 plant trip. The areas included ECCS pump rooms #1 and #2, and the decay heat cooler pit. Although the plant was currently in cold shutdown with decay heat train #1 (in ECCS pump room #1)in service for shutdown cooling, the survey maps still reflected full power operating condition As such, the survey maps did not accurately reflect current radiological conditions in the subject areas. The local postings were updated very shortly after RP supervision was notified. The licensee stated in discussions that locally posted radiological survey maps were not required by the site RP program, and that the local postings were for information only. When questioned why the locally posted survey maps had not been updated following shutdown of the unit, RP personnel indicated that because of the higher than anticipated workload associated with the outage, several routine activities, including the posting of the local survey maps, had not been completed within their normal time frame l l
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4 Conclusion k The inspectors noted that local radiological information provided to plant workers l did not accurately reflect conditions in some areas. In addition, personnel
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performing work activities in the RRA had also failed to recognize (or report) that the local survey maps were not current. At the end of the inspection period the
]_ inspectors had not completed the review of this matter. This is considered an y unresolved item (50-346/97006-05(DRP)) pending completion of inspector review, R2- Status of RP&C Facilities and Equipment j R2.1 Calibration of RP instrumentation (71750)
A sample of portable radiation survey instruments were inspected during the j l' inspection period. The inspectors verified that the instruments were functional and l
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properly calibrated within required time frames. Personnel contamination monitors (PCMs) located at the RRA exit point, as well as near the containment personnel hatch, were verified to be functional and appropriately calibrated. A sample of area radiation monitors (ARMS) were also evaluated and determined to be appropriately functional, with up-to-date calibrations and setpoints conservatively establishe P1 Conduct of EP Activities P Observation of Emeroency Prenaredness Drill (71750)
On April 30, the inspectors observed portions of an onsite emergency preparedness drill from the simulator control room (SCR), technical support center (TSC), and emergency control center (ECC). The' exercise scenario was aggressive, involved numerous equipment failures and was designed to eventually cause declaration of a general emergency. Emergency classification upgrades were performed in accordance with the emergency plan and were made in a timely manner. Good '
j communications between the TSC, Operations Support Center (OSC) and SCR were 1 observed. Control room shift briefs adequately communicated the status of the plant to control room personne .
P3 . EP Procedures and Documentation P Potential Floodina Conditions 1 Insoection Scope (71750) l The inspectors evaluated information that the licensee had received from the Army -l Corps of Engineers concerning potentiallocal area flooding predicted to occur during the upcoming summer and fal l l
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i Observations and Findinas i The inspectors noted that the projected Lake Erie water level conditions could
adversely affect the approved emergency plan evacuation routes during certain ;
extreme storm conditions. Based upon the information received from the Army i Corps of Engineers, the emergency planning (EP) organization initiated a review of ,
local area topography to evaluate whether revisions were necessary to the emergency plan evacuation routes during the times of potential flooding, in
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addition, EP was planning to update their site isolation procedure if conditions !
warranted that alternate routes be taken for relief personnel to travel to the sit !
Pending completion of the licensee review of this matter and determination as to .
whether evacuation routes could adversely be affected, this matter is considered an inspection followup item (50 346/97006-06(DRP)).
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V. Manaaement Meetinas -
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X1 Exit Meeting Summary i
The inspectors presented the inspection results to members of licensee management at the
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i conclusion ot the inspection on May 27,1997. The licensee acknowledged the findings i presente The inspectors asked the licensee whether any materials examined during the inspection i should be considered proprietary. No proprietary information was identified.
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PARTIAL LIST OF PERSONS CONTACTED
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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. M. Dohrmann, Manager, Quality Services J. L. Michaelis, Manager, Maintenance J. L. Freels, Manager, Regulatory Affairs M. C. Beier, Manager, Quality Assessment W. J. Molpus, Manager, Nuclear Training D. L. Eshelman, Manager, Operations R. J. Scott, Manager, Radiation Protection H. W. Stevens, Manager, NS&l L. Hughes, Manager, Davis-Besse Supply C. A. Price, Manager, Business Services J. W. Rogers, Manager, Plant Engineering F. L. Swanger, Manager, Davis-Besse Engineering i M. D. Shepherd, Supervisor, Performance Engineering D. R. Wuokko, Supervisor, Licensing D. W. Schreiner, Supervisor, ISE
- D. H. Lockwood, Supervisor, Compliance D. Imlay, Superintendent, Operations R. B. Coad, Superintendent, Radiation Protection G. W. Gillespie, Superintendent, Chemistry M. K. Leisure, Senior Engineer, Licensing l G. M. Wolf, Engineer, Licensing T. J. Chambers, Shift Manager T. Koslowski, Student, RA
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INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 61726: Surveillance Observations )
IP 62707: Maintenance Observation IP 71707: Plant Operations IP 71750: Plant Support Activities l IP 73052: Review of Procedures IP 73755: Inservice inspection Data Review and Evaluation IP 92700: Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92901: Followup - Plant Operations IP 92902: Followup - Maintenance l IP 92903: Followup - Engineering IP 93702: Prompt Onsite Response to Events at Operating Power Reactors ITEMS OPENED, CLOSED, AND DISCUSSED i
Ooened 50-346/97006-01(DRP) IFl Inoperable post accident monitoring hot leg temperature indicator 50-346/97006-02(DRP) URI Pre-startup checklist not appropriately completed 50-346/97006-03(DRP) URI ECCS leakage rate testing not performed 50-346/97006-04(DRP) URI Action plans not adequately controlled 50-346/97006-05(DRP) URI Outdated local radiological survey maps 50-346/97006-06(DRP) IFl EP evacuation routes could be affected by local flooding Closed 50-346/95009-01(DRP) VIO Short shif t tumovers !
50-346/96003-01(DRP) VIO Emergency diesel generator placed in standby without operators completing all procedural verification steps 50-346/96002-05(DRP) URI Potential MOV electrical hot short due to fire induced damage ,
50-346/96003-07(DRP) URI Reactor coolant pump Appendix R concerns 50-346/96010-06(DRP) IFl Mirror insulation debris following a postulated HELB in containment
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50 346/95004-02(DRS) DEV Failure to trend low pressure transmitter performance i 50 346/96-001-00 LER Inoperable Emergency Core Cooling System 50-346/96-005-01 LER Inadequate Control of Heavy Loads in Containment Building 50-346/96-007-00 LER Control Room Emergency Ventilation System Design Bases Calculation Error
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k, LIST OF ACRONYMS USED AFP Auxiliary Feedpump AFW Auxiliary Feedwater .
ARTS Anticipatory Reactor Trip Systern ASME American Society of Mechanical Engineers CF Code of Federal Regulations DAAS Data Acquisition and Analysis System DEV Deviation ECC Emergency Control Center ECCS- Emergency Core Cooling System EDG Emergency Diesel Generator EPRI Electric Power Reseerch Institute ESF Engineered Safety Fecture GPH Gallons Per Hour GL Generic Letter HELB High Energy Line Break HPI High Pressure Injection ISE Independent Safety Engineering LER Licensee Event Report MOV Motor Operated Valve MSSV Main Steam Safety Valve MWO Maintenance Work Order NC Non Cited Violation '
NDE Non. Destructive Examination i NRC Nuclear Regulatory Commission NRR ['
Nuclear Reactor Regulation OSC Operations Support Center !
PAMS Post Accident Monitoring System i PCAOR Potential Condition Adverse to Quality Report PCM Personnel Contamination Monitor RCS Reactor Coolant System i
RP Radiation Protection '
RRA Radiologically Restricted Area !
RT Radiography '
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SCR Simulator Control Room SFAS i Safety Features Actuation System J SRB Station Review Board TA Temporary Alteration 1 TI Temperature Instrument TS Technical Specifications TSC Technical Support Center USAR Updated Safety Analysis Report UT Ultrasonic Testing VIO Violation
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