IR 05000255/1999001
| ML18068A540 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 03/05/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML18068A539 | List: |
| References | |
| 50-255-99-01, 50-255-99-1, NUDOCS 9903160024 | |
| Download: ML18068A540 (18) | |
Text
I U.S. NUCLEAR REGULATORY COMMISSION REGION Ill Docket No:
License No:
Report No:
Licensee:
Facility:
Location:
50-255 DPR-20 50-255/99001 (DRP)
Consumers Energy Company 212 West Michigan Avenue Jackson, Ml 49201 Palisades Nuqlear Generating Plant 27780 Blue Star Memorial Highway Covert, Ml 49043-9530 Dates:
January 12, 19991 thr~-ugh February ZS, 1999: -. _*.
Inspectors:
Approved by:
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9903160024 990305 PDR ADOCK 05000255 G
PDR J. Lennartz, Senior Resident Inspector J. Maynen, D.C. Cook Resident Inspector 8. Kemker, Byron Resident Inspector Anton Vegel, Chief Reactor Projects Branch 6 Division of Reactor Projects
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EXECUTIVE SUMMARY Palisades Nuclear Generating Plant NRC Inspection Report 50-255/99001 This inspection included aspects of licensee operations, maintenance, engineering, and plant support. The report covers a 6-week period of resident inspection activitie Operations
In general, the conduct of operations continued to be professional and unnecessary distractions in the control room were minimized. Control room operators were aware of ongoing plant activities and plant equipment that was out of service. When questioned, operators were knowledgeable of the reasons that annunciators were in an alarm condition. No significant emergent equipment problems challenged plant operations during this inspection period. (Section 01.1)
Overall, crew performance during the observed dynamic simulator requalification examination was satisfactory. Crew teamwork, crew communications, event diagnosis, and implementation of emergency operating procedures were effective. Command and control by the Control Room Supervisor diminished at times but was adequate overal The inspector's overall evaluation of crew performance was consistent with the licensee evaluators. Also, the inspectors agreed with the licensee evaluator's grading of the competencies with a couple of noted exceptions. The noted exceptions resu.lted in.a higher grade than was warranted for the associated competency but would not have ch~-~ged the overall evaluation. (Section 05)_
Maintenance
The observed maintenance activities were completed in accordance with applicable procedures and the activities were frequently observed by plant supervisio *
Development of Technical Specification Surveillance Test Procedure R0-128-2,
"Emergency Diesel Generator 1-2 24-Hour Load Test," and the pre-evolution preparation activities that were conducted lacked rigor regarding attention fo detai Consequently, several editorial changes to the procedure were required after the test was commenced which delayed test completion. (Section M1)
The inspectors identified several material deficiencies on the Spent Fuel Pool Cooling System and an Emergency Diesel Generator. Plant personnel had missed prior opportunities to identify the material condition deficiencies. The active leak from the Spent Fuel Pool Cooling System component was easily accessible and the fuel oil line had wear indications which provided evidence that it had been rubbing on the adjacent
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*--------*component-during past Emergency* Diesel-Generator-1-2-operations. (Section.M2)---. ______ -*-*---
Th~ self-assessments that were conducted in the third and fourth quarters of 1998 regarding the 13-week work management process were self-critical and effectiv (Section M7)
The licensee identified, following the 1996 refueling outage, that the connectors for an environmentally qualified Core Exit Thermocouple and a non-environmentally qualified Core Exit Thermocouple were swapped during reactor vessel head installation.
. Inspector follow-up on the licensee's evaluation for this issue revealed that the Technical Specification requirements for incore detectors were met. However, the applicable procedure was determined to be inadequate. This non-repetitive, licensee-identified and corrected violation was treated as a Non-Cited Violation. (Section M8.4)
On December 26, 1998, Technical Specification Test Procedure Rl-47, "Rod Withdrawal.
Prohibit Interlock Matrix Check," Step 5.5.1, was not completed as written due to ineffective self checking and a lack of rigor regarding attention to details. In addition, procedure format weaknesses contributed to the occurrence of the error. This non-repetitive, licensee-identified and corrected violation was treated as a Non-Cited Violation. (Section M8.6)
Engineering
Reactor Engineering personnel thoroughly evaluated the failure of the Feedwater Loop "B" ultrasonic flow measuring probe and provided appropriate recommendations.
regarding operability and plant operations. (Section E1)
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. System Engineering personnel provided effective support to operations and
maintenance during performance of Technical Specification Test R0-128, "Emergency Diesel Generator 1-2 24-Hour Load Run." (Section E2)
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Unresolved Item 50-255/97018-02, "Use Of Heat Treated Steel Nuts and Bolts* On.
Spent Fuel Pool Valves," will remain open because of some concerns that the
- inspectors ide_ntified while reviewing the' licensee's evaluation:. Specifically; the '; * * -: _;.:: ; '. :- *.
adequacy of the licensee's boric acid leak inspection and in-service inspection programs were a concern because they apparently did not identify or address subsequent boric acid accumulation on carbon steel bolting material on Spent Fuel Pool Cooling System manual valves. Also, the actual service application for the referenced Spent Fuel Cooling System valves apparently contradicted the service application specified in the lic~nsee's evaluation. (Section E8.1)
Plant Support
The fire brigade effectively used an industry event to conduct training for a hydrogen explosion in the main generator and resultant fire. (Section F.5)
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Report Details Summary of Plant Status The plant operated at full power during the entire inspection period. There were no significant emergent plant equipment problems that directly challenged plant operations. However, plant operators identified apparent oil leaks on Primary Coolant Pumps P-500 and P-508 as evidenced by slowly lowering trends on upper oil reservoir levels. System engineering and operations personnel were monitoring the trend I. Operations
Conduct of Operations 0 General Comments (71707)
The inspectors conducted frequent reviews of ongoing plant operations. In general, the conduct of operations continued to be professional and unnecessary distractions in the control room were minimized. Control room operators were aware of ongoing plant activities and.plant equipment that was out of service. When questioned, operators were knowledgeable of the reasons that annunciators Were in an alarm condition. No significant emergent equipment problems challenged plant operations during this inspection perio *
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Operator Training and Qualification
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The inspectors observed an operating crew during the annual requalification operating examinations conducted in the simulator on February 24, 1999. The inspectors also observed the licensee evaluator's critique of crew performance.and the associated documentation. In addition, the inspectors compared the dynamic simulator scenario quantitative attributes with the guidance specified in NUREG-1021, "Operator Licensing Examination Standards For Power Reactors." Observations and Findings Overall performance of the crew that was observed was satisfactory. Use of three way communications, for the most part, was extensive and effective. Crew teamwork, event diagnosis and emergency operating procedure usage were effective. The dynamic simulator exam scenario's quantitative attributes were consistent with the guidance contained in NUREG-102 Tile-inspectors noted-that, aUimes,_.cJ>IJ1111arid and control by the Control Room Supervisor diminished in that suggestions were provided* tcdhe Nlfclear*Control-- --
Operators instead of directives when actions needed to be taken. The diminished command and control unnecessarily delayed, but did not preclude, completion of some required action.:~ -~ ~.
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Also, the inspectors noted that extensive crew briefs and discussions amongst the Senior Reactor Operators regarding diagnosis and event mitigation strategy occasionally delayed implementation of emergency operating procedure actions. No
- adverse consequences regarding plant status resulted from the c;telay The licensee evaluators critique of crew performance was objective and critica Licensee evaluators had identified all of the crew and individual performance issues that the inspectors had observed as well as some additional items. The inspectors agreed with the licensee evaluator's overall evaluation of crew performance.. Also; the -
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inspectors agreed with the licensee evaluator's grading of the competencies with a couple of exceptions. The inspectors noted that the performance standard was not applied correctly by the licensee evaluators for some of the identified performance weaknesses. Consequently, a higher grade was applied by the licensee evaluators for a couple of the competencies than was warrante For example, Emergency Operating Procedure-5, "Steam Generator Tube Rupture,"
directed the crew to maintain the ruptured steam generator less than 940 psia. The basis for this step was to prevent lifting a main steam safety valve to preclude an uncontrolled release. However, the crew allowed the ruptured steam generator pressure to exceed 940 psia and did not make preparations to lower the ruptured steam generator pressure untilthe pressure was at 980 psia. The first main steam safety valve would open at 1000 psia +/- 3 percen *
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On a grading scale of 5 to 1, top performance to worst performance, the licensee evaluators applied a three to the applicable competency pertaining to.thi_s pertormanc~
weakness. Based on exceeding the pressure limit specified in the procedure and
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-reducing the margin_ to lifting a main steam safety valve in a ruptured steam generator,
. the *inspectcirs determined thafa grade of tWo was* warranted for this.. pEMorman'Ce'/.;:* *
weakness. The individual competency grading differences between the insp'ectors and licensee evaluators would not have changed the overall evaluation of the crew.. Conclusions The inspectors concluded that overall crew performance during the observed dynamic simulator requalification examination was satisfactory. Crew teamwork, crew communications, event diagnosis, and implementation of emergency operating procedures were effective. Command and control by the Control Room Supervisor diminished at times but was adequate overal The inspector's overall evaluation of crew performance was consistent with the licensee evaluators. Also, the inspectors agreed with the licensee evaluator's grading of the competencies with a couple of noted exceptions. The noted exceptions resulted in a higher grade than was warranted for the associated competency but would not have changed the overall evaluatio *...,,..'i..:.~_:.. __ ~ :..**. : ** ~ *. :~-+/-_~),*j
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Miscellaneous Operations Issues (92901)
08.1,(Closed) Inspection Follow-up Item (IFI) 50-255/93029-01: "Uncontrolled Withdrawal of a Control Rod." The licensee investigated an uncontrolled withdrawal of a control rod event that occurred at the Fort Calhoun Nuclear Plant on November 13, 1993, with the plant in cold shutdown. The event occurred because two grounds developed
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simultaneously in the control rod drive system such that contacts used to control rod motion were bypassed. The control rod drive system was *an ungrounded system and therefore, presence of the grounds was unknow The licensee's investigation revealed that a similar event could occur at their facilit However, the event was considered unlikely and the adverse safety consequences were not significant in that several factors would either preclude or mitigate the event such as:
1) poor preventative maintenance, that resulted in excessive system degradation and contributed to the event at Fort Calhoun, was not an apparent problem at Palisades; *
2) the event was bounded by analyzed events in the Updated Final Safety Analysis Report; 3) control room operators had redundant indication of unexpected control rod movement; and 4) action could be taken to remove power from the control rod drive motors if uncontrolled movement occurre In addition, the licensee had developed Special Test, T-370, "Control Rod Drive Condition Monitoring," that was used to test for shorts and grounds in the control rod drive system circuitry. Special Test T-370 was tracked by the preventative maintenance program and scheduled to be performed prior to and after every refueling outage starting in 1998. The inspectors verified that Special Test T-370 was performed, as intended, during the 1998 refueling outage. This item is close (Closed) Licensee Event Report CLER) 50-255/98-013: "Transformer Tap Changer:
Failure Causes Inadvertent Diesel Generator Start." This event was documented in Inspection Report 50-255/98022. Investigation of the event revealedJhat the
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safeguards tap changer fllOtor contactor failed to open properly causing t~e tap.rnotor to. _..
drive transformer *output voltage to the minimum. position and, then lockout mech"anically".' *.. '.
Consequently, voltage on safety-related electrical busses decreased and _both *.
emergency diesel generators started as designed:",Also. the investlgati~)n*reveaieCI tha_t the tap changer had been "hunting" excessively for several shifts prior to the failur There was no adverse safety consequences associated with the event and all plant equipment responded as designe Both the lower and raise tap changer motor contactors were replaced to correct the problem. Post maintenance testing was subsequently completed satisfactorily and the safeguards transformer was returned to service. Also, operations initiated action to monitor safeguards transformer tap changer performance more closely. This* item is close (Closed) Unresolved Item CURI) 50-255/96002-02: Potentially inadequate 10 CFR 50.59 Evaluation for thermal power limit ~ 0 CFR 50, Appendix K, "ECCS Evaluation Models," assumes the reactor had been operating continuously at a power level of at least 1.02 times the licensed power level to allow for such uncertainties as instrumentation error. Also, the Final Safety Analysis Report (FSAR), Chapter 14,
"Safety Analysis," Section 14, 1.3, stated that the initial condition for transient analyses
- * *-- ** --- * --- - ---a-re-oas_e_d_cm-steady.:state-operations-at 2,530 *megawatts-thermal-(MWt-) with-an-applied _____ _
uncertainty factor for reactor power of +/- 2 percent to ensure conservative analysi The licensee had confirmed, through two engineering analysis (EA-HAR-91-01 and EA-HAR-91-10), that instrument uncertainties were within+/- 2 percent (actual+/- 1.90 percent and+/- *1.91 percent). However, General Operating
Procedure (GOP) - 12 was revised in December 1993, which incorporateP a +/-1 percent instrumentation uncertainty value for indicated reactor power; therefore, the licensee assumed that the reactor could be operated at up to 101 percent power and still be
- within the design basis if a +/- 1 percent instrumentation uncertainty was applie Consequently, GOP-12 inappropriately allowed the reactor to be operated *at greater than the licensed power level (100 percent, 2,530 MWt) but not to exceed 100.9 percent (2,555 MWt) provided the power, when averaged over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, did not exceed 100 percen The NRC staff had recognized brief power excursions above licensed thermal power limits could occur provided the average power level over any 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift is maintained no greater than the*100 percent limit. On February 7, 1996, indicated reactor power exceeded the license power limit of 2530 MWt when averaged over an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shif Peak power level recorded was 100.2 percent and the maximum 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift average was 100.1 percent. The inspectors did not identify any other instances where the facility had operated above 100 percent utilizing this operating practice. The procedure inadequacy resulted in a violation of requlatory requirements as documented in report 50-255/9700 A 10 CFR 50.59 screening was conducted in December 1993, for GOP-12, Revision 7, which incorporated the +/- 1 percent uncertainty. The questions to determine if an unreviewed safety question existed for the revision were all answered "no." However, the question, "Does the item involve a change to the facility as described in the FSAR?"
This should have been answered "yes", in that the assumed+/- 2 percent uncertainty factor, as described in the FSAR, was not being applied. Therefore, the failure.to,>-*. ~,::...
appropriately evaluate the revision resulted in a procedure that allowed the faCility to..
operate up to 100.99 percent of maximum licensed thermal power. _Consequently, a*
potential unreviewed safety question existed when ttie indicafed 'reaCto?pOwer;'when**
averaged over an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift, exce~ded the license power limit on February 7, 199 Also, the facility was potentially operated outside the design basi The inspectors concluded that an unreviewed safety question did not exist and that the facility was not operated outside the design basis. Subsequent tests and analysis showed that reactor power did not exceed 100 percent on February 7, 1996. A calorimet~ic uncertainty analysis was completed to more accurately reflect the calorimetric uncertainty. Also, an ultrasonic flow measurement (UFM) of the feedwater flow was performed which provided a more accurate indication of actual feedwater flow, independent of the installed feedwater venturie The results from the uncertainty analysis indicated that the actual calorimetric uncertainty was 1.01 percent. Therefore, when the uncertainty factor (1.01 percent) was added to the peak indicated power (100.4 percent}, the resulting value was within the Palisades design value (102 percent). In addition, results of the UFM testing revealed
* * ---that-actual-power-was-2-:-2-percent.less. than indicated power... _C.onsequently,_J.1s.ing 1!EM_ _________ _
results, the maximum power level that was achieved on February 7, 1996,. was 98.2 percent. Based on the uncertainty analysis and the UFM testing results, an unreviewed safety question did not exist and the facility was not operated outside the design basis. This item is close II. Maintenance M1 Conduct of Maintenance Inspection Scope (61726 and 62707) Portions of the following maintenance work orders and surveillance activities were observed or reviewed by the inspectors:
Work Order No:
24712421
24513531
24812527
24812528 Surveillance No:
DWT-08
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. R0-128-2.
- Observations and Findings ED-07 A Inverter No. 2A Load Transfer Diesel Generator 1-2 Ventilation Fan V-24C Engineered Safeguards Room Cooler Fan V-278 High Temperature Engineered Safeguards Room Cooler Fan V-278 Low Temperature Ultrasonic Flow Measurement Data Collection, Analysis, and Implementation Emergency Diesel G~~erator 1-2 24-Hour' L~ad Run.. ;-;.
- For the observed activities, the inspectors noted that the procedures were at the job site and that the workers were knowledgeable of the activities. Also, the inspectors noted, on several occasions, that supervisors were observing the maintenance activities that were in progress in the plan *
During performance of Technical Specification Surveillance Test R0-128-2, "Emergency Diesel Generator (EOG) 1-2 24-Hour Load Run," several procedural errors were identified after the test was commenced. For example, the auxiliary operator, while logging pre-start readings, identified that the starting air tank pressure reading incorrectly referenced the air tanks for EDG 1-1 vice EDG 1-2. Also, the Nuclear Control Operator, while performin*g Step 5.6, "Synchronize Generator," identified that the sequence of two steps had to be reversed to properly adjust and compare generator voltage with incoming bus voltage in order to complete the ste ____
The Rrocedural deficiencies were not considered sign'ificant. Also, the deficiencies were identified by opefratioiis *personnel during the-test-and were.appr.opri~t~ly_ corrected via editorial changes to the procedure. However, the procedure deficiencies -unne-cess-arily delayed the test and once when the EOG was running unloaded; a condition that the procedure specified to limit. Also, the number of errors indicated a lack of rigor regarding attention to detail during procedure development and review, and during the conduct of pre-evolution preparation activities by plant operator :
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Conclusions The inspectors concluded that the observed maintenance activities were completed in
. accordance with applicable procedures and that the activities were frequently observed
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Development of Technical Specification Surveillance Test Procedure R0-128-2,
"Emergency Diesel Generator 1-2 24-Hour Load Test,n and the pre-evolution preparation activities that were conducted lacked rigor regarding attention to detai Consequent!~. several editorial changes were required to the procedure after the test was commence M2 Maintenance and Material Condition of Facilities and Equipment (62707)
The inspectors evaluated the material condition of plant equipment during routine tour The inspectors identified several deficiencies that had not been previously identified by plant personnel. The deficiencies included: 1) boric acid build-up on manual valves in the Spent Fuel Pool Cooling System (see Section E8.1 for additional details); an active leak of approximately seven drops per minute from the discharge check valve hinge pin packing gland for Spent Fuel Cooling Pump P~518; 3) two fuel oil lines on EOG 1-2 had indicated wear because they were rubbing against adjacent components while the EOG was operating; and 4) a minor leak from Dilution Water Pump Discharge Drain Valve, MV-CW-552. Work requests were generated in a timely manner by 'plant personnel after the material condition deficiencies were brought to their attention. Plant personnel
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.had misse.d prior opportunities to identify the m~teria.1 conditic;m defici~ncies..,Jt:i.~ ~ctiv,.*:,,,.,. *=;=:.*
leak from the Spent Fuel Pool Cooling System component was easily accessible and the*
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_fuel oil iine had wear indications which provided evidence that it had t?eeri rubbing on the
'adjacent component during past EDG.1-2 a*perations. Though the imme.diate *s.afety-*' :: _;:,- *. :. * *, ; * 'i ';
consequence of these deficiencies was minor, if left uncorrected, these 'problems could*
potentially impact the ability of the systems to perform their required function. In addition, the inspectors were concerned that plant personnel, including operations *..
personnel and system engineers, did not demonstrate a pro-active questioning attitude by not identifying these issues and initiating work requests until prompted by the inspectors.,
M7 Quality Assurance in Maintenance Activities (40500)
The inspectors reviewed the third and fourth quarter self-assessments that were conducted by the licensee's Production Team regarding the 13-week work management process. The assessments were considered self-critical in that several areas of concern were identified. Specific improvement actions were developed for the identified concerns and performance in the associated areas was being trended. Also, the assessments appeared effective in that targeted concerns showed improvement in the
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For example, forecast for completion of engineering work for associated work orders was an area of concern. Historically engineering forecast reliability had been around 50 percent accurate. Engineering forecast reliability in the fourth quarter was approximately 85 percent accurate. Aiso, maintenance activity completion rate, which averaged 84 percent for several past quarters, remained at or above 90 percent for the
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last two quarters. Increased activity completion rate was attributed to the improvement actions that were implemented to stabilize the maintenance schedul * The inspectors concluded that the self-assessments that were conducted regarding the 13-week work management process were self-critical and effectiv MB Miscellaneous Maintenance Issues (92902)
M (Closed) Inspection Follow-up Item 50-255/95004-06: "Cylinder Leakage Testing." lri 1995, a follow-up inspection to an NRC Diagnostic Evaluation Team inspection identified a weakness in the licensee's EOG testing program. Specifically, the Surveillance Procedure, M0-7A[B], "Emergency Diesel Generator 1-1[2]," did not provide explicit acceptance criteria for the quantity of fluid which, if ejected out of a cylinder, would constitute an inoperable engine. The licensee revised both of the surveillance procedures and the Operating Procedure, SOP-22, "Emergency Diesel Generators," to prevent engine operation and further evaluate any quantity of fluid greater than a fine mist. The inspectors concluded that the procedures provided adequate guidance fbr the operators to determine if engine operation should be stopped. This item is close M (Closed} Inspection Follow-up Item 50-255/95009-02: "Engine Exhaust Recirculation Issue." In 19§5, during a major disassembly of the 1-1 EOG, the licensee identified evidence of engine exhaust recirculation into the air intak.e. Carbon buildup was noted on the air intake of the aftercooler, the air inlet side of the cylinder heads, and the turbocharger. The most likely cause of the carbon buildup was running the EOG
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Surveillance Procedure M0-7 A[B]. "Emerg.ency Diesel Generator 1-1 [2]," and Standard.
- Operating Procedure SOP-22, "Emergency Diesel Generators," to limit the amount of time that the diesels are run.unloaded. Additionaily, the intake fans and-aftercooiEj('. *:.
tubes are inspected for carbon buildup and cleaned on an 18 month frequency. No excessive carbon buildup has been noted during these inspections. This item is close M {Closed} Inspection Follow-up Item 50-255/95009-03: "Water In Cylinder 7R." In 1995, during a major disassembly of the 1-1 EOG, the licensee identified evidence of water leakage into the 7R cylinder. The water intrusion was most likely due to a small leak in the afterc9oler. As part of the EOG overhaul, all cylinder heads were replaced, and the aftercooler was repaired. Since the EOG overhaul, the licensee has periodically conducted boroscopic inspections of the EOG cylinder head area to identify water leakage or other damage. The boroscopic inspections have not revealed any additional instances of water intrusion into the cylinder head area. This item is close M (Closed} Unresolved Item 50-255/96017-02: "Core Exit Thermocouples Swapped During Reactor Head Installation." On December 29, 1996, during the performance of a
- - ------------- ___ Palisades lncore Detector Algorithm run, the licensee identified that the Core Exit Thermo-coU-pleqce-T-) #-1 e was not.meeting j_t_s_ ~-~rveillance acceptance criteri Subsequent investigation by the licensee determmedthaftne*connectors for- - - -------
environmentally qualified CET #10 and non-environmentally qualified CET #13 were swapped during reactor vessel head installation. A review of the procurement and installation of CET #13 concluded that CET #13 met all of the environmental qualifications; therefore, the Technical Specification requirements for operable incore detectors were me._-.
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During forced outage 97-002 in January 1997. the Cables for CET #1 O and CET #13 were removed and properly reconnected. The Westinghouse Refueling Manual, CPAL-RFM-003, was revised to include a caution that the incore cable connectors are
. similar at the incore instrument flanges. Additionally, a requirement was added to the procedure for an independent verification of the cable connection at the patch panel and at the incore instrument flanges. These corrective actions appeared adequate to prevent recurrenc *
10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires that activities effecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawing Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, the Westinghouse Refueling Manual, CPAL-RFM-003, did not include appropriate qualitative acceptance criteria to determine that the CETs had been connected properly. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation (50-255/99001-01 (DRP))
consistent with Section Vll.B.1 of the NRC Enforcement Polic M (Closed) Inspection Follow-up Item 50-255/95004-,01: "Surveillance Acceptance Criteria." The inspectors had identified a concern with the acceptance criteria for as-left component cooling water (CCW) flow to the charging pumps. Technical Specification *
Surveillance Procedure Q0-17, "lnservice Test Procedure - Charging Pumps," required the as-left CCW.flow to Charging Pump P:-55C to be between 5.0 and 5.2 gp..,.. *
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Component cooling flow requirements to Charging Pumps P-55C and P~558 were * *
based on vendor requirements and acceptance criteria of Special Test T-223, *
"Component Cooling Water Flow Verification." Special Test T-223 balanced tio~i'< < *. * ** *-."..,,, ** *. *._**i;-:"' *
through the critical components that would be in service during accident conditions. *
However, the minimum required CCW flow to the charging pumps specified in the Final Safety Analysis was 5.0 gpm. Consequently, the required flow band appeared too tight to accommodate system fluctuations and remain above the minimum require System engineering personnel subsequently determined that the CCW flow band specified in Q0-17 was too narrow for Charging Pumps P-55C and P-558 to allow for system fluctuations. Special Test T-223 was revised to set CCW to the charging pumps at 9 - 10 gpm to increase the allowed CCW flow band to charging pumps P-558 and P-55C. In addition, the operator round sheets were.revised to record CCW flow to the charging pumps daily and the round sheets specified the minimum required CCW flow to the charging pumps was 6.0 gpm. A statement to notify system engineering if the flowrate was less than the minimum was also added to the operator round sheets. The inspectors concluded that the actions taken were adequate to ensure that CCW flow to the charging pumps would be maintained greater than the minimum required following
system-fluctuations._ Tb.is item is close ~ -----
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M (Closed) Escalated Enforcement Item (EEi) 50-255/98022-02: "Failure To Follow---------:---
Surveillance Procedure." On December 26, 1998, during performance of Technical Specification Test Rl-47, "Rod Withdrawal Prohibit Interlock Matrix Check," Step 5. required operations personnel to bypass both the variable high power trip and the*
thermal margin/low pressure trip of the "A" channel. The maintenance technician was required to independently verify that the actions had been performe ~*
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However, at a subsequent Step, 5. 7. 7, that required removal of the bypasses that were installed in Step 5.5.1, an operator noted that only the variable high power trip was bypassed. Therefore,* the operator failed to correctly perform procedure Rl-47,
. Step 5.5.1, and the maintenance technician failed to recognize the error while performing the independent verification. Licensee personnel generated a *Level 2" Condition Report (C-PAL-98-1977) to document and conduct a root cause evaluation for this issu The licensee's evaluation was completed on February 5, 1999, and concluded that ineffective self-checking and procedure format weaknesses were the root cause Procedure Rl-47, Step 5.5.1, format was considered weak in that the single step contained two actions. It was the second action of that particular step that did not get accomplished. The inspectors agreed with the licensees conclusion that ineffective self-checking was a root cause. In addition, the licensee identified that the procedure sponsor inappropriately used an editorial change to revise the procedure in the pas The inspectors also determined that the lack of rigor regarding attention to detail was a second root cause in that the second required action directed by Step 5.5.1 was missed
. by the personnel during performance of the procedure. The inspectors determined that the procedure format weakness contributed to the erro There were no adverse consequences regarding nuclear safety because the failure to bypass the thermal margin/low pressure trip was considered conservative. A thermal margin/low pressure reactor trip was available during the surveillance, as performed,.
.... which would not have been available if the trip was-bypassed as directed by the.. ~._,,_;="- /,_ *
procedur *
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. The' corrective actions* to prev.ent recurrence, 'as docum~hted in the'evaluatiorf df
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C-PAL-98-1977, included: 1) revise Rl~7; 2) review the event with the operating crews to use as a tool to re-enforce the standards of self-checking; and 3) counseling for
- procedure sponsor regarding the inappropriate* use of an editorial change to add the step to bypass the thermal margin/low pressure trip in the procedure. The inspectors *
concluded that the corrective actions, as documented, were adequate to prevent recurrenc CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires that activities affecting quality shall be accomplished in accordance with prescribed procedures. Contrary to the above, Technical Specification Surveillance Test Rl-47,
"Rod Withdrawal Prohibit Interlock Matrix Check," was not accomplished in accordance with the prescribed procedure. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation (NCV) consistent with Section Vll. of the NRC Enforcement Policy (50-255/99001-02(DRP)).
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Ill. Engineering E1 Conduct of Engineering Inspection Scope (37551)
The inspectors reviewed the root cause analysis and the corrective actions regarding the failure of the Feedwater Loop "B" ultrasonic flow measurement (UFM) device Observations and Findings The UFM devices were used to provide a more accurate indication of actual feedwater flow than the installed feedwater venturies. Data from the UFMs was utilized to calculate a correction factor for heat balance power and provided a more accurate indication of actual reactor powe Feedwater Loop "B" UFM probe was not providing a satisfactory signal and had to replaced on February 5, 1999. Following probe replacement, licensee personnel measured the probe spacing in accordance with UFM installation analysis (EA-BWB-96-01) and noted that the probes were approximately 60 mils further apart than before being replaced and 40 mils beyond the original installation measurement Reactor Engineering personnel generated Condition Report C-PAL-150 to document the issue.
. The inspectors noted that Reactor Engineering personnel rigorously evaluated the prob_e
- spacing differences and vendor representatives were brought to the site as consultants.* **
... A detail~d plan that ide_n!ified required ac.tions, resources needed; and pla.ntppwer"
.. recommendations was provided to plant management. Also, Reactor Engineeiin~f ~
provided a thorough operability recommendatiqn to operations personne * Conclusions *
The inspectors concluded that Reactor Engineering personnel thoroughly evaluated the failure of Feedwater Loop "B" UFM probe and provided appropriate recommendations regarding _operability and plant operation E2 Engineering Support of Facilities and Equipment (37551)
The inspectors observed portions of Technical Specification Test, R0-128, "Emergency Diesel Generator 1-2 24-Hour Load Run." The inspectors noted that System Engineering personnel were frequently present during the test to monitor Emergency Diesel Generator performance and installed test equipment. Also, System Engineering
- --personnel pr5>vided assistance to plant operators to ensure that the Diesel Generator power factor specified in the*test procedure was established and maintained. The inspectors concluded that" System Engineering personnel provided-ample.support during performance of the surveillance tes.-'.*,.*
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ES Miscellaneous Engineering Issues (92700, 92903)
E Review of Unresolved Item 50-255/97018-02: "Use Of Heat Treated Steel Nuts and
. Bolts On Spent Fuel Pool Valves*: The inspectors noted that heat treated steel nuts and bolts were used on the body to bonnet connections for Spent Fuel Pool Valves*MV-131 *
and MV-132 following maintenance to repair body to bonnet leaks. However, the spent fuel pool contained borated water and therefore, use of stainless steel bolting materials would have been consistent with Generic Letter 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants, a regarding this issu The licensee's evaluation concluded that using heat treated steel nuts and bolts on MV-131 and MV-132 was acceptable which was based on: 1) Specification Change SC-87-031, which was conducted to evaluate replacing carbon steel bolting with
stainless steel bolting in systems that contained boric acid, concluded that using carbo steel bolting in MV-131 and MV-132 was acceptable; 2) future body to bonnet leakage from the valves would be quickly identified because the valves were in high traffic areas, not insulated, and easily visible; 3) the valves would be inspected in accordance with Engineering Procedure EM-26, "Engineering Boric Acid Leak Inspection," which required boric acid walkdowns at every refueling outage as a minimum; 4) design/repair specification, M-260, piping class sheet class HC (150 psig, austenitic stainless steel),
allows for use of either carbon steel or stainless steel; and 5) the valves service application was in a low pressure (less than 50 psig) and low temperature (less than 100.°F) syste *
- _ The inspectors noted some concerns, 9uring _thei~ review :of the license~*s e_v~luation_,
and a walkdown of the system. Specifically, the inspectors identified a visible leak as. *' "...
evidenced by boric acid accumulation on the body to bonnet area for MV-132. Also bo*ric acid was evident on the packing gland forMV-131. *There we*re*no wo*rk requests.... * **. *"' *,
written for these valves which indicated that the leaks had not been identified. ' * *
Consequently,'the licensee's evaluation regarding quickly identifying any future leaks*
was a. concern. The inspectors verifi~d that the boric acid leak inspection *of these valves had been completed during the 1998 refueling outage. In 1998, the licensee documented that boric acid accumulation was not observed on MV-132 and only minor boric acid accumulation was observed on the packing gland for MV-131 during the inspectio Also, boric acid accumulation on the carbon steel bolting material for MV-132 was a concern in that the licensee's in-service inspection program and boric acid leak inspection program apparently had not identified the issue. The inspectors discussed the issue with System Engineering personnel. Subsequently, a work request was generated to clean up the boric acid and to increase the torque on the body to bonnet bolts. The inspectors were concerned that the bolts were apparently not inspected for
-- **degradatio In addition, the inspectors noted that MV-131 and MV-132 were the.cross connect valves between Spent Fuel Pool Cooling System and the Shutdown Cooling Syste Standard Operating Procedure-27, "Fuel Pool System," Revision 33, controlled operation of MV-131 and MV-132 when the Spent Fuel Pool was aligned to be cooled by the Shutdown Cooling System. During that evolution, MV-132 was throttled to maintain Low Pressure Safety Injection Pump discharge pressure greater than 150 psi Therefore, the actual service application for MV-131 and MV-132 apparently
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contradicted the service application specified in the licensee's evaluation regarding low pressure, (less than 50 psig).
- The inspectors discussed their concerns regarding this unresolved item With li~nsee management. Pending further review of the licensee's evaluation, this item will remain ope E (Closed) LER 50-255/98-011: "Inadequate Lube Oil Collection System For Primary Coolant Pumps. a This licensee identified and corrected event was discussed in detail in Inspection Report 50-255/98022 and was a Non-Cited Violation. No new issues were revealed by the LER. This item is close *
E (Closed) LER 50-255/98-012: "Failure of Main Steam Isolation Valves To Fully Close Under No-Flow Conditions." This licensee identified and corrected event was discussed.
in detail in Inspection Report 50-255/98022 and was a Non-Cited Violation. No new issues were revealed by the LER. This item is close E (Closed) IFI 50-255/96007-0HDRP): "Completion of Cable Ampacity Reviews." The licensee determined that a number of cables did not meet the ampacity design basis stated in 'Section 8.5.2 of the Palisades FSAR. Specifically, power cables in overfilled cable trays (trays with greater than 30 percent physical fill) had not been analyzed or * *
dispositioned to confirm that existing ampacity limits were acceptable. The inspectors also discussed this issue in NRG Inspection Report 50-255/96017 and requested that the Office of Nuclear Reactor Regulation review the licensee's. alternate ampacity methodology to determine if it complies with the Palisades FSAR. Inspector FollO\\\\'.".\\.IP *
Item 50-255/96017-05 will remain operi to track this issue pending Office of Nuclear
. Reactor Regulation review of the licensee's methodology. This item is closed.
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IV. Plant Support F5 Fire Protection Staff Training and Qualification (71750)
The inspectors observed a fire drill that was conducted on January 28, 1999. The training included a walkdown in the turbine building, led by the fire brigade leader, to identify plant equipment that could potentially be affected by a hydrogen explosion in the main generator and a subsequent lube oil fire. Also, the location of specific plant equipment that could be used to isolate hydrogen to the main generator was identifie In addition, "table top" training was led by the Shift Supervisor to discuss the location of extinguishing equipment that could be used for the fire and the potential difficulties that would be encountered. An industry event regarding a hydrogen explosion and resultant fire was discussed during the training which was effective in that it demonstrated the actual probability for the event. The inspectors concluded that the fire brigade effectively used an industry event to conduct training for _a hydrogen explosion in the main generator and resultant fir *
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V. Management Meetings X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on February 25, 1999. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie _..,..
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PARTIAL LIST OF PERSONS CONTACTED Licensee G. R. *Boss, Operations Manager E. J. Grindahl, System Engineer D. G. Malone, Licensing R. L. Massa, Shift Operations Supervisor T. J. Palmisano, Site Vice President D. W. Rogers, General Manager, Plant Operations J. Schwan, System Engineer M. S. Holmberg, Reactor Engineer, Rll I R. G. Schaaf, Project Manager, NRR IP 71707:
IP 62707:
IP 61726:
IP 37551:
IP40500:
- 1p 71750:
IP 92901:
. IP92902:
IP 92903:
INSPECTION PROCEDURES USED Plant Operations Maintenance Observations Surveillance Observations Onsite Engineering Effectiveness of Licensee Controls Plant Support Activities Followup - Operations
- Followup - Maintenance Followup - Engineering
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Opened 50-255/99001-01 50-255/99001-02 Closed ITEMS OPENED, CLOSED, AND DISCUSSED NCV Core Exit Thermocouples Swapped During Reactor Head Installation NCV Failure To Follow Surveillance Procedure 50-255/93029-01 IFI Uncontrolled Withdrawal of a Control Rod 50-255/98-013 LER Safeguards Transformer Tap Changer Failure Causes Inadvertent Diesel Generator Start 50-255/96002-02 URI 10 CFR 50.59 Evaluation 50-255/95004-06 IFI Cylinder Leakage Testing 50-255/95009-02 IFI Engine Exhaust Recirculation Issue 50-255/95009-03 IFI Water in Cylinder 7R
- 50-255/96017-02 URI Gore Exit Thermoco'uples Swapped During Reactor Head Installation
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50-255/99001-01 NCV Core Exit Thermocouples Swapped During Reactor Head Installation 50-255/95004-01 IFI Surveillance Acceptance Criteria 50-255/98022-02 EEi Failure To Follow Surveillance Procedure 50-255/99001-02 NCV Failure To Follow Surveillance Procedure 50-255/98-011 LER Inadequate Lube Oil Collection System For Primary Coolant Pumps 50-255/98-012 LER Failure of Main Steam Isolation Valves to Fully Close Under No-Flow Conditions 5D=255i96007=0 l --
IFI-Completion of Cable Ampacity Reviews Discussed 50-255/97018-02 URI Use of Heat Treated Steel Nuts and Bolts on Spent Fuel Pool Valves
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