IR 05000155/1996006

Insp Rept 50-155/96-06 on 960727-0906.No Violations Noted. Major Areas Inspected:Operations,Engineering,Maint & Plant Support

ML20129H816
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	10/28/1996
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20129H808	List: IR 05000155/1996006 ML20129H804
References
50-155-96-06, 50-155-96-6, NUDOCS 9611060035
Download: ML20129H816 (15)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

l Docket No: 50-155 License No: DPR-06 Report No: 50-155/96006(DRP)

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l Licensee: Consumers Power Company l

Facility: Big Rock Point Nuclear Power Plant l

Location: 10269 U.S. 31 North l Charlevoix, MI 49720

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Dates: July 27 - September 06, 1996 Inspectors: R. J. Leemon, Senior Resident Inspector C. E. Brown, Resident Inspector l

! Approved by: W. J. Kropp, Chief, Projects Branch 3 Division of Reactor Projects l

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9611060035 961028 PDR ADOCK 05000155

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l f EXECUTIVE SUPMARY i

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Big Rock Nuclear Power Plant l NRC Inspection Report M-155/96006 l l

This routine 6-week inspection covered aspects of licensee operations, i engineering, maintenance, and plant support. Additionally, this report  !

includes the results of announced inspections by regional specialist i Operations

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. Operating crews performed well during on-the-job training and off-normal l

! situation communication. However, there was a weakness in the quality i l of turnovers by R0s leaving the control room for a short period of time (Section 01.2). ,

. An operating crew declared check valve, VPI-303, in the core spray l backup system operable even though the valve did not meet acceptance criteria during a surveillance test. Declaring the valve operable is ;

unresolved item pending further review of the operability determination l process by the NRC.

l i Maintenance j I

i . The documented operability determination for excessive leakage past '

l ' check valve, VPI-303, focused only on VPI-303 and did not fully document l

! the basis for the backup core spray system operability as affected by J l degraded VPI-3U3 operation. (Section M3.1).

Enaineerina

. The inspectors were concerned that the post incident system relief l

valves could stick open during an actual demand which would result in

! decrease flow in the reactor (Section El.1). The licensee has a plant life time exemption for a LOCA in one core spray train and a single failure in the remaining operable core spray train which could be a stuck open relief valve. Adequate core spray flow with a stuck open PI relief valve with a line break in the one of the core spray trains is considered an inspection follow-up item pending further NRC review (50-155/96006-02(DRP)).

i l . The investigation and analysis of the failed 86R pilot valve were l l technically sound, and the NRC staff had no concerns with the licensee's i

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operability determination for the four installed RDS-pilot valves {

(Section E2.1).

Plant Suncort

. The work activities to radiation profile of the spent fuel channels and

to clean the spent resin tank room were well supervised and completed as

planned without exceeding the dose estimate ('iection RI.1).

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Report Details Summary of Plant Status

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l The plant operated at full power throughout the period, and no events occurred which challenged the control-room operator I. Operations j 01 Conduct of Operations i

01.1 General Comments Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations. In general, the conduct of operations was professional and safety-conscious; specific events and findings are detailed in the sections belo .2 Shift Mannina and Formality a. Scope (71707)

l The inspectors observed the conduct of shift supervisors, nuclear control operators, and auxiliary operators during shift turnovers, plant rounds, and surveillance testing. These observations were compared to l the plant technical specification (TS) and final hazards summary report (FHSR) requirement l b. Observations and Findinos The minimum required shift manning, including two reactor operators (R0s), was maintained. However, during several shift turnover periods, one of the on-shift R0s would leave the control room for a short period of time without ensuring that the other R0 was fully cognizant of the plant status. The turnover consisted only of the statement, " Stepping out," and contained no information related to plant equipment or statu This statement was also routine for periods between shift turnover Pending further review by the licensee and the NRC, the adequacy of shift turnovers is considered an unresolved item (50-155/96006-01(DRP)).

Auxiliary operator (AO) performance during a diesel fire pump surveillance test was good. The A0 asked personnel to clear the area in-front of the diesel before starting the test in order to take l readings immediately after the diesel started. The A0 also frequently l checked the condition and parameters of the diesel during the 30 minute l diesel run.

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I During turnovers, the inspectors observed that the A0s-in-training demonstrated the same good traits and acquired skills as the experienced A0s. The experienced Aos provided good on-the-job-traininq to these A0s in the following areas:

. Pre-job briefings

. Walk-throughs, including tour techniques, both inside and outside containment

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. Equipment tagouts for maintenance

. Communication techniques (repeat backs).

Newly licensed R0s were conscientious about the plant status and used good communication techniques when talking to either the other RO or to the A0s in the plan The shift supervisor also had good communication with management during the failure of core spray check valve VPI-303 to seat (Section M3.1).

Shift supervisor's communication with management during off-normal situations had been noted as a weakness by the resident inspectors in IR l 50-155/96002, Section E2.1, relating to a turbine trip. These recent l communications demonstrated an improvement in comparison to the turbine l trip event.

Conclusion Operating crews performed well during on-the-job training and off-normal situation communication. However, there was a weakness in the quality of turnovers by R0s leaving the control room for a short period of time.

! 08 Miscellaneous Operations Issues (92700)

l 0 (Closed) Violation 50-155/95006-07: Wide-Range Monitors (WRMs) High l- Neutron Flux Setting Exceeded TS Allowable Limits. Two of three neutron (

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WRMs were allowed to become out-of-calibration low during power  !

escalation following a forced outage. The high power reactor scram I would not have occurred until 132 percent of rated core power vice the

TS allowed 120 * 5 percent. This occurred because the WRMs were i

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recalibrated at 63 percent power during the power escalation and the follow-on crews didn't understand the potential for the WRM readings to ;

l diverge at higher power levels when the control rods were pulled further 1

! out of the core. Additionally, the operators failed to ensure that each l WRM channel indicated a: actual core power during the remainder of the

! escalation to full powe The inspectors have monitored several power reductions or escalations  ;

since this occurrence in early 1995. In each instance, the operators '

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were very attentive to the WRM calibration. Additionally, the WRMs ,

j indications were frequently verified by other control room indications ;

i of core power. The training given to the operating crews was effective

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l as verified by individual interviews with multiple operators. Shift

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turnovers during power changes have been controlled and formal. The

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inspectors concluded that the licensee's corrective actions had been appropriate. This item is closed.

l 08.2 (Closed) Violation 50-155/95006-Ole: See event description in Section 08.1. This item is closed.

! 08.3 (Closed) Violation 50-155/95006-04b: See event description in Section 08.1. This item is closed.

l 08.4 (Closed) Violation 50-155/95006-08: See event description in Section 08.1. This item is close .

l II. Maintenance M1.1 Conduci; of Maintenance l

M1.1 General Comments

! Inspection Scone (62703) (61726)

The inspectors observed all or portions of the following work activities:

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Maintenance Activities

. WO MSS 12611970 rcplace "B" turbine bypass hydraulic oil pump unloader l

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. WO SPS 12512030 test spare breaker

. WO SPS 12511884 functional inspection circuit breaker 052-lE21

. WO EPS 12610230 monthly battery readings (battery-2-C)

. WO PIS 12611813 rebuild RV-5082

. WO FPS 12611812 provide seal for fire barrier

. WO RDS 12610160 disassemble, inspect, repair, test SV-4987 pilot i valve  !

WO RDS 12610175 pilot valve leaks (S/N 86R-4)

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. WO RDS 12611864 adjust CV-4183 packing

. WO SPS 12610527 test breaker and contractor

. WO ASD 12610198 monthly battery readings (ASD)

. WO BLS 12611162 loading dock has deteriorating concrete

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. WO EPS 12610186 monthly battery readings (Battery-1) )

. WO BLS 12611334 repair fuel cask loading dock concrete piers j

. WO FPS 12611891 adjust engine speed switch i

. WO RDS 12412099 RDS pilot valve 73V-14 leaks through

. WO RWS 12611061 test sump pump for resin tank room repairs i

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as verified by individual interviews with multiple operator Shift turnovers during power changes have been controlled and formal. The inspectors concluded that the licensee's corrective actions had been appropriate. This item is closed.

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0 (Closed) Violation 50-155/95006-Ole: See event description in Section 08.1. This item is close .3 (Closed) Violplin 50-155/95006-04b: See event description in l Section 0 This item is closed.

l 0 (Closed) Violation 50-155/95006-08: See event description in l Section 08.1. This item is close II. Maintenance M1.1 Conduct of Maintenance M1.1 General Comments a. Inspection Scone.162703) (61726)

The inspectors observed all or portions of the following work activities:

Maintenance Activities

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. WO MSS 12611970 replace "B" turbine bypass hydraulic oil pump !

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unloader l

. WO SPS 12512030 test spare breaker  !

. WO SPS 12511884 functional inspection circuit breaker 052-1E21 l

, . WO EPS 12610230 monthly battery readings (battery-2-C) l

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. WO PIS 12611813 rebuild RV-5082

. WO FPS 12611812 provide seal for fire barrier )

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. WO RDS 12610160 disassemble, inspect, repair, test SV-4987 pilot valve

. WO RDS 12610175 pilot valve leaks (S/N 86R-4)

. WO RDS 12611864 adjust CV-4183 packing t . WO SPS 12610527 test breaker and contractor l . WO ASD 12610198 monthly battery readings (ASD)

l . W0 BLS 12611162 loading dock has deteriorating concrete l . WO EPS 12610186 monthly battery readings (Battery-1)

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. WO BLS 12611334 repair fuel cask loading dock concrete piers

. WO FPS 12611891 adjust engine speed switch

. WO POS 12412099 RDS pilot valve 73V-14 leaks through

. WO RWS 12611061 test sump pump for resin tank room repairs

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Surveillance Activities I

. T1-02: primary system leakage test

. T1-09: heat balance calculation i

! . T7-04: reactor protection system logic test i

.- T7-20: diesel fire pump auto start

. T7-28: emergency diesel generator (EDG) auto test start

. T7-33: check of ASD equipment

. T30-22: ECCS valve test

. T30-44: containment Gamma monitor monthly channel check I

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. T90-07: RDS isolation valve test operate at power i

. T90-25: partial stroke test of MSIV (M0-7050)

. TR-65A: RDS "A" battery service test and discharging alarm operability verification l l . TR-65C: RDS-UPS "C" battery service test and discharging alarm l- operability verification l . TR-65D: RDS-UPS "D" battery service test and discharge alarm

operability verification l . TV-16C

RDS-UPS "C" battery discharge test

.. TV-16D: RDS-UPS "D" battery discharge test Observations and Findinas The mechanics and operators used the procedures at the job site, and job status was frequently monitored by visits from supervisors and system i engineers. On-the-job training (Section 01.2) and assistance to engineering (Section E2.1) were particularly noteworthy. Where applicable, proper radiation control measures were in us Conclusion The inspectors found the work done under these activities to be

correctly performed and accurately documented.

l M3 Maintenance Procedures and Documentation l M Inadeauate-Safety Evaluation for Leakaae throuah Tell-tale Durina l Testina of Backun Core Soray Check Valve (VPI-303) Inspection Scope On August 27,1996, at 8:33 p.m., the licensee while performing surveillance test T30-22, emergency core cooling system (ECCS) valve test, observed an approximate 4-foot plume of steam coming out of a tell-tale. This test included stroke timing the motor operated primary and backup (BU) core spray valves. The steam plume indicated that a check valve, VPI-303, was not fully seate The inspectors monitored the leakage through the tell-tale, attended licensee discussions and PRC meetings, reviewed surveillance test data, i and observed the gathering of test data related to valve position and

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tell-tale pressure.

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l b. Observations and Findinas Test procedure T30-22 required the A0 to verify that check valve VPI-303 was adequately seated by checking flow through the tell-tale between M0-7070 and M0-7071 while M0-7071 was open. The A0 documented that valve VPI-303 failed the acceptance criteria, "Less than a steady stream from the tell-tale," by marking the "N0" box in the procedure and notifying I

the shift supervisor (SS). A second attempt to seat VPI-303 by re-stroking M0-7071 was unsuccessfu The A0 notified the shift supervisor (SS) of VPI-303's failure to seat after M0-7071 was stroked open per T30-22. A second attempt to seat VPI-303 by re-stroking M0-7071 was also unsuccessful. At 9:07 p.m., the SS informed the on-call manager and the NRC resident inspector that the second attempt to seat VPI-303 had failed. The SS and on-call manager held discussions and concluded that the backup core spray would perform its intended safety function. The SS then held a discussion with the operation's manager who concurred with the conclusion and to the directions the SS had given his crew to increase monitoring the temperatures between M0-7070 and M0-7071 -- to ensure that the temperature remained less than 280 *F (about 50 psig). The SS did not declare VPI-303 inoperable even though the acceptance criteria was not met. The SS did initiate a condition report to document the problem with VPI-303. The ability to declare a component operable even though the component did not comply with the acceptance criteria defined in the i

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surveillance procedure is considered an unresolved item pending further NRC review (50-155/96006-02(DRP)).

On August 28, the plant review committee (PRC) reviewed the event and i

determined that the valve VPI-303 and the backup core spray system

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should have been declared inoperable as of the time that the A0 noticed i the excessive leakage past the tell-tale. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LC0 was then entered. A team formed to evaluate the VPI-303 leakage proposed a change to the acceptance criteria and a revision to T30-22 to install a pressure transducer to record tell-tale pressure. These proposed

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changes were approved by the PRC, and based on tell-tale pressure readings observed while developing the acceptance criteria change, valve

, VPI-303 and the backup core spray system were declared operabl Additionally, the PRC reviewed and approved the associated safety

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evaluation.

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The resident inspectors and NRC staff reviewed the safety evaluation and l concluded that the scope was too focused on VPI-303 allowable leakag The licensee did not initially provide the pressure rating of the piping, the type of valves, assurance that M0-7070 and M0-7071 would open with a plugged tell-tale, nor what routine monitoring was performed to ensure that the tell-tale was not plugged. If the backup core spray system was required to operate to provide cooling to the reactor core, water would flow from the fire system through the backup core spray

. system, controlled by motor-operated gate valves M0-7070 and M0-707 . 8

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! The NRC staff asked the licensee if valves M0-7071 and M0-7070 could- l l operate under design basis accident conditions if the spool piece i l between the two valves was pressurized to a pressure of 1335 psig. The )

! licensee determined that the high pressure piping included both the i spool piece and MD-7070. Additionally, the licensee concluded that the )

spool piece would not pressure lock at 1335 psig for the following <

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. Valve M0-7071.was tested monthly with 1335 psid across the disk and would open against a differential pressure of 1335 psi with a safety margin of 13 per cent using a calculation that included degraded voltage. Additionally, this valve, a flexi-wedge-gate l valve, was installed with a machined notch on the reactor side

disc to prevent pressure binding between the discs and allow any high pressure trapped in the spool piece to relieve to the

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reactor.

l . After M0-7071 opens, any residual pressure would be relieved and MO-7070 would be able to ope . Due to blockage, the tell-tale was modified in 1991 by moving the l hole from the bottom of a pipe plug to a drilled 1/8-inch hole in i the side of the piping, a location less likely to plu In l l accordance with procedure T30-22, the hole is checked clean i

! monthly. Additionally, the pipe plug is removed from the tell- )

i tale and any accumulated corrosion products are removed and sent l l to engineering for analysi l Conclusion l I

The documented operability determination for excessive' leakage past VPI-303 focused only on VPI-303 and did not fully document the basis for the i backup core spray system operability as affected by degraded VPI-303 I operation. As a result, the NRC staff had extensive communication with the licensee to fully understand the basis for operability in regards to the backup core spray system. The licensee subsequently documented the

operability evaluation to include the backup core spray system.

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El Conduct of Engineering l

El.1 Post Incident (PI) System Leakina and Liftina Relief Valves Inspection Scone The inspectors reviewed the recurrence of the post incident (PI) system

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relief valves lifting during surveillance testing for applicability to

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the licensing and design basis, i

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') . Observations and Findinas The initial pressure surge during diesel fire pump surveillance runs has usually caused one or more of the three PI relief valves, set at 155 psig and supplied from the fire system, to lift. Several of the plant staff expressed acceptance of this response even though it was not expected. A condition report (CR) written for a PI relief valve lifting and not reseating was initially rated as a level 4, trend onl However, the corrective action review board (CARB) reviewed the CR the following day and changed the classification to a level 3 which requires additional evaluatio The three PI relief valves and associated isolation valves were added during several modifications, and two of the relief valves were required to be in service. During a previous DFP test, one PI relief valve was isolated and caution tagged after leaking. On a subsequent DFP test, another PI relief valve lifted and leaked greater than the acceptance allowance. The licensee cleared the caution tag from the first relief valve and returned it to service, and the second relief valve was isolated. The first relief valve was no longer leaking. The licensee generated a condition report and work requests to repair both of the valves; however, only one of the relief valves was repaired by the licensee during the next few days.

I The licensee has accepted the PI relief valves lifting and not fully reseating for many years. The inspectors were concerned that the relief valves could stick open during an actual demand which would result in decrease flow to the reactor. However, the licensee has a plant life time exemption for a LOCA in one core spray train and a single failure in the remaining operable core spray train which could be a stuck open relief valve. The licensee also had a engineering analysis that bounded a stuck open relief valve but was not part of the design basis of the plant. Adequate core spray flow with a stuck open PI relief valve with a line break in the one of the core spray trains is considered an i

inspection follow-up item pending further NRC review (50-155/96006-03(DRP)).

El.2 RDS Pilot Valve Internal Sleeve Corroded in Place Insnection Scone A spare reactor depressurization system (RDS) relief pilot valve failed '

to function during a test. Upon inspection of the inside of the 86R-4 pilot valve during disassembly, a cracked " tooth" was observed on the .

sleeve and corrosion was evident on the main disc and sleeve area. The I inspectors observed and assessed the licensee's actions in response to l this failed pilot valv j l Observations and Findinas

l On August 9, dimensions on spare RDS pilot valve 86R-4 were obtained in j preparation for future modifications. The valve had been taken off the o 1 i 10

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"B" train RDS header during the refueling outage in January 1996, i

successfully tested, and stored in the turbine building with the l openings taped. Upon disassembly, the system engineer determined that -

l the valve sleeve had rusted to the valve body and that corrosion was

! evident on the valve disc. This immediately raised concerns about the condition and operability of the model 86R pilot valves installed in the ;

RDS system. The licensee had four pilot valves of this model; two were

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spares and two were installed on the RDS header. On August 12, after discussion with the manufacturer, valve 86R-4 was sent to a laboratory for an inspection which revealed that mainly iron corrosion products l were present on the valve disc, the sleeve, and the seating area.

l l The spare RDS pilot valve 86R-1 was satisfactorily pop tested, disassembled, and inspected; no corrosion was found. The licensee sent the sleeve from the 86R-1 valve to the corporate lab for analysis. The analysis revealed that the corrosion in the 86R pilot valve contained i excessive sulphur which was most likely from foreign material in the l

valve. Also, the spare model 73V RDS pilot valves were disassembled and inspected for comparison to model 86R pilot valves. No corrosion or discrepancies were found.

l Based on the laboratory analysis results and the examination of the second spare 86R pilot valve, the licensee concluded that the two installed 86R pilot valves were operabl ;

c. Conclusion l The investigation and analysis of the failed 86R pilot valve were !

technically sound, and the NRC staff had no concerns with the licensee's operability determination for the four installed RDS-pilot valve E2 Engineering Support of Facilities and Equipment E2.1 Review of FHSR Commitments i j The inspectors compared plant operation and surveillance activities with

the Final Hazards Safety Report (FHSR). No deviations were noted during l

this report perio ;

IV. Plant Sunocr1 l R1 Radiological Protection and Chemistiy (RPhC) Controle i

R1.1 Radiation Profilina of Used Fuel Channels and Resin-Tank Room Cleanina Insoection Scone The inspectors observed the licensee preparation and performance of the first phase of the project to clean residual spilled resin from the room

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resin dump and concentrator tanks. Additional radiological work

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consisted of the underwater survey and dose characterization of channel racks and a representative sampling of used fuel channels for isotopic content to determine waste classification and packaging requirements.

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b. Observations and Findinas l In order to conserve dose, the licensee had hired an experienced

! contractor to clean the spent resin room. Additionally, the licensee l used radio head-sets for communications, tele-dosimetry, and television I

monitors to remotely monitor dose and job performance. This was the first time tele-dosimetry had been utilized for direct dose surveillance at this site. The total dose was 15.4 person-rem versus an ~ estimated

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17.788 person-rem for the project. An interview with the project l coordinator revealed that unexpected problems (rebuilding the transfer pump four times due to loss of suction and clearing a plugged spent-l resin tank room to radwaste sump drain line) cost 5.3 person-rem.

l The cleaning was completed with no spills and reduced the stored l radioactive waste on-site by 5-cubic feet. Also, the dose rate in the *

room was reduced by 200 mrem / hour.

l The channel profile work went well even with the use of contractor l workers. Controlling dose was a major focus point, and the work was

, completed under the dose goal for this phase of the project. The l licensee had built a temporary shield wall on the refueling deck to reduce the exposure when the supervisor and worker were not on the spent fuel pool (SFP) bridge. An interview with the ALARA coordinator

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revealed that the wall created a very low dose waiting area, reducing l the dose rate from 1.5 to 0.6 millirem per hour (15 to 6 micro-sieverts per hour).

Channel radiation profiles on 88 of 150 channels, now in the SFP, were

! completed. Dose rates ranged from 10 rem to 13,000 rem on the transition pieces and from 100 rem to 400 rem on the square portions of the channels. The round portions of the channels averaged 2 rem. The i licensee concluded that channels from the last core cycle may require some decay time prior to shipmen c. Conclusion The work activities to radiation profile of the spent fuel channels and to clean the spent resin tank room were well supervised and completed as planned without exceeding the dose estimate. However, several weak

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health physics practices identified during the work are discussed belo l l (Sections RI.2 and R4.1) i

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V. Manaaement Meetinas X1 Exit Meetina So-- ry l

l The inspectors presented the inspection results to members of licensee l management at the conclusion of the inspection on September 11, 199 !

The licensee acknowledged the findings presente The licensee did not identify any of the documents or processes reviewed by the inspectors as proprietar l i l

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PARTIAL LIST OF PERSONS CONTACTED l l l l Licensee l P. Donnelly, Plant Manager R. Addy, Assistant Plant Manager S. E>eachum, Systems and Project Engineering Manager i E. Bogue, Chemistry / Health Physics Manager i G. Boss, Operations Manager D. Hice, Maintenance Manager

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G. Withrow, Plant Safety and Licensing Director 1  !

INSPECTION PROCEDURES USED IP 37551: Engineering IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 61726: Surveillance Observations IP 62703: Maintenance Observation IP 64704: Fire Protection Program IP 71707: Plant Operations IP 71750: Plant Support Activities

! IP 73753: Inservice Inspection IP 83729: Occupational Exposure During Extended Outages l IP 83750: Occupational Exposure

! IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power l Reactor Facilities

! IP 92902: Followup - Engineering ,

IP 92903: Followup - Maintenance ,

ITEMS OPENED and CLOSED l Ooened i

155/96006-01 URI Weak Turnovers During Shifts l 155/96006-02 URI Declaring component operable that does not meet l acceptance criteria l 155/96006-03 IFI Sufficient core spray flow

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155/95006-07: VIO WRM high neutron flux scram setting exceeded TS allowable 155/95006-Olc: VIO inadequate communications within the shift concerning WRM recalibration 155/95006-08: VIO inadequate communication between shifts concerning WRM recalibration 155/95006-04b: VIO Inadequate training on WRM operations during power escalation

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l LIST OF ACRONYMS USED l ALARA As low As Reasonably Achievable A0 Auxiliary Operator CFR Code of Federal Regulations ESF Engineered Safety Feature FHSR Final Hazards Summary Report HP Health Physics IFI Inspection Followup Item IP Inspection Procedure

IPE Individual Plant Evaluation IPTE Infrequently Performed Test and Evolution IR Inspection Report LCO Limiting Condition for Operation LER Licensee Event Report NCV Non-Cited Violation NOV Notice of Violation NRC Nuclear Regulatory Commission R0 Reactor Operator RP Radiation Protection RPA Radiologically Protected Area SFP Spent Fuel Pool SS Shift Supervisor SV Solenoid Valve TS Technical Specification UE Unusual Event URI Unresolved Item VIO Violation ,

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