IR 05000155/1997008

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Insp Rept 50-155/97-08 on 970430-0616.Violations Noted. Major Areas Inspected:Operations,Maint,Engineering & Plant Support
ML20151J787
Person / Time
Site: Big Rock Point File:Consumers Energy icon.png
Issue date: 07/22/1997
From:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20151J769 List:
References
50-155-97-08, 50-155-97-8, NUDOCS 9708050255
Download: ML20151J787 (14)
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U.S. NUCLEAR REGULATORY COMMISSION REGION lil Docket No: 50 155 License No: DPR-06 l Report No: 50-155/97008(DRP)

l Licensee: Consumers Energy Facility: Big Rock Point Nuclear Power Plant I

Location: 10269 U.S. 31 North l Charlevoix, MI 49720 Dates: April 30 - June 16,1997 Inspectors: R. J. Leemon, Senior Resident inspector L. N. Tran, Project Manager Approved by: Bruce L. Burgess, Chief Reactor Projects Branch 6 i

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PDR ADOCK 05000155 l G PDR

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EXECUTIVE SUMMARY Big Rock Nuclear Power Plant NRC Inspection Report 50-155/97008

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l This routine inspection covered aspects of licensee operations, engineering, maintenance, I and plant suppor Ooerations e The inspector concluded that the licensee did not follow procedure TV-07 when CRD insertion speeds were not adjusted to less than 28-seconds as required by j procedure. This was considered a violation. (Section 01.2)

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e Maintenance personnel left the containment equipment lock inner door open for l about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after bringing a pump into the containment. This was a violation of System Operating Procedure (SOP)-10 " Containment Vessel." Auxiliary operators toured the containment about every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, but did not recognize that the door should be closed. (Section 01.3)

l e Operability determination training was effective and resulted in operator discussions ;

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and a uniform understanding among the licensed operators of how the operability i I determination process is conducted. (Section 05.1) '

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l e Maintenance and surveillance activities were generally performed well and l l accurately documented. (Section M1.1) '

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e The inspector concluded that reactor engineers had good performance in monitoring nuclear instrumentation and in providing useful information to the nuclear control operators (NCOs) during plant startup. (Section E1.1)

Plant Sucoort e An Operations Support Center (OSC) training drill was effective in providing improved skills to experienced OSC team members, more experience to new OSC j team rnembers, and valuable feedback on command and control elements requiring !

i improvement to the OSC Director. (Section P5.1)

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Report Details Summarv of Plant Status

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The plant was shut down.at the beginning of the inspection period to refurbish the main generator breaker after it failed to close on April 28,1997. During a reactor startup on May 1,1997, control rod drive (CRD) F-3 double notched when being withdrawn from position "00." All rods were inserted into the core and time tested. On May 5, at 3:29 a.m., the reactor was taken critical and the main generator breaker was successfully closed at 6:32 p.m., ending a 64-day forced outage. On May 27, all rods were out of the core and plant coast down commenced. On June 16. the end of the inspection period, the plant output was 65 MWe,84 percent powe l l. Operations

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01 Conduct of Operations l

i 01.1 General Comments (71707) I j Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations. Specific events and findings are detailed in the sections belo ;

j 01.2 Control Rod Drive Insertion Soeed Time Testina l

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j Insoection Scoce The inspector reviewed procedure TV-07 "CRD Scram Test from Notch "23" and Coupling Integrity Checks," and technical specification (TS) 5.2.2.a, " Control Rod System." The inspector also discussed CRD insertion speeds with operations personnel and attended a plant review committee (PRC) meeting where a TV-07 procedure change was approve Observations and Findinas On May 1,1997. CRD F-3 double notched when being withdrawn from position

"00." The control operator had difficulty moving the CRD from position "00" and gave two withdrawal signals close together, causing the CRD to double notch. This was not an unexpected operator action when an operator has difficulty with a CRD moving from position "00." The licensee inserted all CRDs into the reactor and performed CRD time testing for insertion and withdrawal speeds to determine if there was a need to adjust the speeds. On May 2,1997, the inspector completed

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a review of TV-07 which revealed that five CRD rate set valves were fully open and their insertion speeds would not adjust to below 28 seconds. The inspector

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determined, through discussion with the licensee, that the 28 seconds was considered guidance and not a criteria for rod insertion. The inspector reviewed TS

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and only found a criteria of greater than 23 seconds for withdrawal of CRDs and no criteria for inserting CRDs. The inspector informed the licensee that procedure TV-

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07 presented the criteria for withdrawal and insertion of a CRD in a manner indicating that it was a requirement. The licensee then performed a safety evaluation which concluded that there was a safety concern in removing the 28 seconds as a criteria. Procedure TV-07 was changed to require that the insertion speed for a CRD to be greater than 26 second The inspector reviewed past TV-07s and found that in October 1994, all CRD ,

insertion times were less than 28 seconds. In April 1996, seven CRDs were left with insertion speeds greater than 28 seconds. In May 1997, five CRDs were left with insertion speeds greater than 28 seconds. The licensee accepted the CRD insertion speeds as had been the practice when the rate set valves were fully ope After the inspector questioned this practice, the licensee changed the procedure to reflect the required CRD timing criteri Procedure TV-07, " Control Rod Drive (CRD) Scram Test from Notch "23" and Coupling Integrity Checks," Revision 20, Step 5.2.h requires that when the CRD insertion speed is outside the range of 26 to 28 seconds that the Rate Set Valve, V-1, be adjusted as required to improve drive perforraance, and the CRD be re-timed. TS 6.8.1 required that written procedures are established, implemented, and maintained for all structures, systems, and components defined in the Big Rock Point Quality List. Control rod drives are defined in the Big Rock Point Quality Lis Leaving CRDs with insertion speeds greater than 28 seconds was a violation of TS 6.8.1 (VIO 50-155/97008-01(DRP)). Conclusion The inspector concluded that the licensee did not follow procedure TV-07 when CRD insertion speeds could not be adjusted to less than 28 seconds as required by procedure TV-0 .3 _Contynment Eauioment Lock inner Door Left Ooen Insoection Scooe l The inspector reviewed circumstances surrounding the containment equipment lock inner door being left open. The inspector reviewed station logs and condition report

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(CR) C-BRP-0339 " Equipment Lock inner Dcor Left Open." The inspector also I discussed this event with the operations manege: and a CR evaluator.

1 Observations and Findinas On May 30,1997, at 11:05 a.m., the containrnent equipment lock inner door was used to bring a refurbished containment heating and cooling pump into the containment to be reinsta!Ied. The outer door was closed and the pump was removed from the air lock by maintenance personnel, but the inner door was left open. On May 31 at 8:30 a.m., an auxiliary operator (AO) discovered that the containment equipment lock inner door was open for no apparent reason and closed ,

it. The licensee realized that the containment equipment lock inner door being open J was a violation of SOP-10, " Containment Vessel," which requires both doors of the j

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equipment lock to be closed at all times except for ingress or egress and indicated such in CR C-BRP-0339.

l The licensee identified the violatia t end the safety consequences were minor since ;

l the outer equipment lock door was uc3cd, as required, to maintain containment I integrity. However, enforcement discretion was not applied and a Notice of )

Violation was written in accordance with NUREG-1600 " General Statement of Policy and Procedures for NRC Enforcement Actions," Section Vil " Exercise of l l Discretion." The NRC had regulatory concerns that the containment equipment lock l inner (a very large door) was not discovered open for about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> when AOs tour the containment about every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Also, at the end of the inspection period the licensee had not proposed any corrective actions to prevent recurrence. Failure to close the containment inner door for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> was a violation of SOP 10, I l " Containment Vessel." (VIO 50-155/97008-02(DRP)). j 1 Conclusion Maintenance personnel left the containment equipment lock inner door open after bringing a pump into containment. AOs tour the containment about every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, but did not recognize for about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> that the door should be close Operator Training and Qualification 05.1 Operability Determination Trainina Insoection Scope On May 15,1997, the inspector attended an operator training session conducted by a shift supervisor (SS) for continued training on " Operable and Operability Determinations." .

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Licensee operability determinations have been identified as concerns at BRP in !

inspection reports 50-155/96010(DRP) and 50-155/97002(DRP). The objective of the operability determination training was to evaluate a Condition Report (CR) on a system, structure, or component and determine operability using all available resources, including TS,0-list, Final Hazards Summary Report (FHSR), on-call manager, operations management, engineers, and plant safety and licensing personnel. The resson for the training was to reaffirm the importance of making prompt and conscrvative operability determinations and to inform the operators of the NRC's positian on operability. During this training, the following items were discussed:

e GL 91-18, "Information to Licensee's Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on l Operability."

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i 07.1 Review of Institute of Nuclear Plant Ooerations (INPO) Reoort The inspectors and branch chief reviewed the latest INPO Report on the licensee's performance. No new safety concerns were noted that had not been noted by the NRC and no actions were required, 11. Maintenance M1 Conduct of Maintenance M1.1 General Comments

a. Insoection Scone (62703) (61726) l The inspectors observed all or portions of the following work activities:

Maintenance Activities e work order (WO) TGS-12710589: stop leak on moisture separator e WO PIS-12710614: LR-3111 containment water level trace erratic Surveillance Activities e TV-11: flux wire irradiation procedure i e T90-07: RDS isolation valve test operate at power e TR-01: control rod drive performance testing

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l e TV-07: CRD scram test from notch "23" and coupling integrity checks e T7-28: emergency diesel generator auto test start e T30-20: monthly stationary battery voltage and specific gravity readings Observations and Findinas Maintenance and surveillance activities were reviewed against the FHSR and were found to be satisfactorily performed. All observed work was performed with the work package present and in active use at the job site. Supervisors and system engineers monitored job progress, and appropriate radiation control measures were in plac Conclusion Maintenance and surveillance activities were generally performed well and accurately documente M1.2 Control Rod Drive Insertion Sneed Time Testina Procedure adherence to CRD insertion timing was weak (taference Section 01.1).

M1.3 RDS "B" Isolation Valve Slowlv Stroked Closed jnsoection Scooe After RDS ~B" isolation valve CV-4181 slowly stroked closed, the inspector reviewed surveillance test T90-07 "RDS isolation Valve Test Operate at Power";

condition report (CR) C-BRP-97-0316 "RDS isolation Valve Slow to Close"; TS 11.3.1.5.a; and station logs. The inspector also had discussions with operations personnel and the system engineer and attended a CARB meetin Observations and Findinas On May 19,1997, while performing surveillance test T90-07, RDS "B" isolation valve CV-4181 closed slowly (11.7 seconds, compared to procedure closing time criteria of < 9.0 seconds). The inspector verified that, according to TS, the safety function of CV-4181 was to open in <4.0-seconds. The valve opened in < seconds; therefore, the valve performed its safety function. The valve is also timed in the closed direction to monitor for degradation of the material condition of the valve. A review of records revealed that previous closing speeds were about 4 seconds. CV-4181 was cycled a second time and closed in 8.7 seconds. The licensee conservatively declared the valve inoperable, wrote CR C-BRP-97-0316, and at 5:30 a.m. began a 7 day LCO. TS 11.3.1.5.a states that should one RDS

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valve train become inoperable in the closed position, the reactor may remain in l power operation for a period not to exceed 7 days. CV-4181 was stroke tested j again and opened in 1.96-seconds and closed in 8.36-seconds, within the testing

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criteria. On May 19,1997, at 9:22 p.m., the licensee declared CV-4181 operable

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and exited the LCO. As a compensatory measure, the licensee increased the frequency of testing of isolation valve CV-4181 from 90 days to 30 day Conclusion RDS "B" isolation valve CV-4181 slowly stroked closed. The licensee conservatively declared the valve inoperable and entered TS LCO action requirements. CV-4181 was satisfactorily tested twice and the LCO was exite As a compensatory measure, the licensee increased the valve testing frequency from 90 days to 30 day M1.4 Main Generator Breaker Failure to Latch After Closina Insoection Scoce i

On April 28,1997, when attempting to synchronize the main generator to the grid, main generator breaker 116 OCB failed to latch. The inspector reviewed -

maintenance WOs and CRs related to the breaker latching problem:

e CR C-BRP-96-462 116 OCB Failed to Close e CR C-BRP-97-284 116 OCB Ouick Trip after Closure e WO SPS-12611278 Replace Seal-in-switch and Check Operation of Limit and Auxiliary Switches for 116 OCB e WO SUB-12710527 116 OCB Will Not Latch

The inspector also interviewed the system engineer and maintenance and operations personnel.

i Observations and Findinas On January 31,1996, maintenance performed on the 116 OCB to resolve  ;

ratcheting problems included the mechanism latch spring being replaced with a J vendor approved part. The breaker mechanism was inspected, lubricated, and successfully tested using local controls. Breaker testing from the control room was not possible because of tagging on the breaker controls for other reason On April 6,1996, when attempting to synchronize the main generator to the grid, the 116 OCB failed to close. On a next attempt, the 116 OCB's local counter showed that the 116 OCB had closed twice. It had closed but did not latch. After, opening and re-closing it latched closed. The original latch spring was reinstalled, and the latch mechanism was re lubricated. The mechanisrn seal-in-switch was

found to have a problem and replacement of the switch was scheduled for the next l outage. The 116 OCB was then closed satisfactorily. The licensee determined that

! the apparent cause of the 116 OCB closing problem was aging of the seal-in-switch (CR C-BRP-96-462).

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- On September 12,1996, WO SPS 12611278, to replace the seal-in-switch, check l operation of alllimit and auxiliary switches, and check operation of the 116 OCB

from the control room, was completed. The latch bearings were also replaced. The i

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i licensee considered the above work to be a general overhaul of the operating mechanism of the 116 OC ;

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On April 28,1997, the 116 OCB again failed to latch closed. On April 29,1997, I l the breaker mechanical mechanism was inspected by two industry breaker experts l employed by the licensee. Activities performed under WO SUB-12710527 were:

l 1) the solenoid cutoff switch was adjusted to allow air to stay on the closing piston longer to ensure latching the breaker; 2) two 1/8-inch thick hardened stop spacers were replaced with new neoprene gasket material to reduce bounce when latching ( the breaker; 3) an auxiliary spring was installed on the latch lever to help the '

t mechanism latch properly; 4) the breaker was test tripped at reduced voltages to ensure the breaker would still trip with the added spring: 5) the anti-pump micro- l switch was adjusted and tested three times and 6) the breaker was tested five !

times from the control room contro On April 30,1997, the main generator breaker was returned to service and on May 5,1997, at 6:32 p.m., main generator breaker 116 OCB was successfully )

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closed synchronizing the main generator to the gri l Conclusion During plant startups, main generator breaker 116 OCB failed to closa and latc After licensee maintenance personnel failed to solve the problem, the licensee enlisted the assistance of two industry breaker experts who resolved the breaker 1

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M8 Miscellaneous Maintenance issues (92902)

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M8.1 (Open) Follow-Un Itern 50-155/97002-05: material condition concerns with the control rod drive system. In inspection report (IR) 50-155/97002, an inspection l followup item (IFI) and a violation were both given item number 97001-01. To 1 resolve this inadvertent duplication of item numbers, the IFl was renumbered to 50- i 155/97002 05, 1 M8.2 (Closed) Follow-Uo item 50-155/95012-01: licensee oversight of site contractors. In November 1995, a backhoe operator cut three power cables during digging activities. Administrative Procedure 4.2.7 " Contracted Services,"

Revision 10 was revised to clearly define the duties for the contractor coordinator and to ensure that suppliers or contractors follow Consumers Energy procedures and policies.

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lLIJinaineering E1 Conduct of Engineering E Reactor Enaineer Plant Startuo Activities Insoection Scooq l The inspector observed reactor engineers perform plant startup activities to assist the NCOs in the safe startup of the plan Observations and Findinos The inspector observed the reactor engineers perform the following activities during plant startup: 1) monitor nuclear instrumentation for rate of power change and verification of doubling times,2) monitor CRDs which in the past had proble.ms moving from position "00." 3) provide the NCOs with the expected reactivity worth when an upcoming notch had high reactivity worth, and 4) use three way communication with the NCO l Conclusion l

The inspector concluded that reactor engineers' performance in monitoring nuclear '

instrumentation and in providing useful information to the NCOs during plant startup was goo IV. Plant Suonort R1 Radiological Protection and Chemistry (RP&C) Controls R 1.1 General Comrnents ,

Using Inspection Procedures 71707 and 71750, the inspectors made frequent tours i of the radiologically protected area (RPA) and discussed specific radiological controls with the ALARA coordinator and various radiation protection (RP)

technicians. The inspectors observed plant conditions and licensee performance including radiation protection practice P5 Staff Training and Qualification in EP P Ooerations Sucoort Center Trainino Drill Insoection Scone On June 10,1997, the inspector attended Operations Support Center (OSC)

classroom training and observed an OSC training dril _ - - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ Observations and Findinas The inspector attended the OSC classroom training and discussed the training objectives with training personnel. The objectives were as follows: perform personnel accountability; set up the OSC in the machine shop, dispatch maintenance teams, track emergency priorities, provide the OSC Director with practice in command and control, improve drill skills of experienced OSC members, develop skills of the newer OSC members, and move the OSC to the machine shop when require During the training drill, the inspector observed that the maintenance team worked well, that HP provided good coverage, and the bridge phone communicator kept the status boards up-to-date. However, the inspector noted that the OSC Director did not use a procedure nor did he recognize that on a LOCA the OSC should be moved to the air compressor oom. Additionally, once the OCS members were in the air compressor room, the inspector had a concern with the members getting too close to electrical switches. Once this concern was raised to the OCS Director, a safety rope to protect the switches was promptly installed, Conclusion An OSC training drill was effective in providing improved skills to experienced OSC team members, more experience to new OSC team members, and valuable feedback on command and control elements requiring improvement to the OSC Director.

S1 Conduct of Security and Safeguards Activities S 1.1 Conduct of Security and Safeauards Activities The inspectors monitored the licensee's security program during routine activities and tours to ensure that the approved security plan was being implemented. The inspectors noted that personnel within the protected area displayed proper photo-identification badges and individuals requiring escorts were properly escorted. The inspectors also observed that personnel and packages entering the protected area were searched by appropriate equipment or by han V. Manaaement Meetinos X1 Exit Meetina Summarv The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on June 16,1997. The licensee acknowledged the findings presente The licensee did not identify any of the documents or processes reviewed by the inspectors as proprietar ,

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PARTIAL LIST OF PERSONS CONTACTED Licensee K. Powers, Plant General Manager R. Addy, Assistant Plant Manager S. Beachum, Systems and Project Engineering Manager K. Pallagi, Chemistry / Health Physics Manager G. Boss, Operations Manager D. Hice, Maintenance Manager G. Withrow, Plant Safety and Licensing Director J. Rang, Decommission Manager M. Bourassa, Licensing Senior Technical Analyst

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1-K. Wooster, Nuclear Emergency Planner i

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, INSPECTION PROCEDURES USED l

IP 37551: Engineering

! IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing

, Problems 1 IP 61726: Surveillance Observations IP 62703: Maintenance Observation

! IP 64704: Fire Protection Program j IP 71707: Plant Operations IP 71750: Plant Support Activities

, IP 73753: Inservice Inspection l

IP 83729: Occupational Exposure During Extended Outages  !

IP 8375 Occupational Exposure  !

IP 92700: Onsite Followep of' Written Reports of Nonroutine Events at Power Reactor l l Facilities i

j IP 92902: Followup - Engineering

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.lP 92903: Followup - Maintenance )

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155/97002-05 IFl Material Condition Concerns with the CRD System

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155/97008 01 VIO CRDs Left with High Insertion Speeds after Time Test l 155/97008-02 VIO Containment Equipment Lock Inner Door Open i

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155/95012-01 101 Licensee Oversight of Site Contractors i.

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' LIST OF ACRONYMS USED i r j ALARA- As Low As Reasonably Achievable l BRP Big Rock Point ,

t CARB Corrective Action Review Board *

CFR Code of Federal Regulations  !

CR Condition Report CRD Control Rod Drive  :

DRP- Division of Reactor Projects l FHSR Final Hazards Summary Report GL Generic Letter <

HP Health Physics IFl inspection Followup item

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l IP inspection Procedure

IR inspection Report

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LCO Limiting Condition for Operation r LOCA Loss of Coolant Accident  :

MWe Megawatt Electric .

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NCO Nuclear Control Operator NRC Nuclear Regulatory Commisr, ion OSC Operation Support Center ' '

PR Plant Review Committee RDS Reactor Depressurization System i RP Radiation Protection ,

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RP&C Radiological Protection & Chemistry *

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SS Shift Supervisor TS Technical Specification TV Test Variable VIO Violation WO Work Order i

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