IR 05000155/1986007

Insp Rept 50-155/86-07 on 860409-0710.Violation Noted: Failure of Operator to Maintain Desired Control Rod Position & to Investigate Abnormal Behavior of Sluggish Control Rod Movement

ML20207J336
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	07/21/1986
From:	Boyd D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20207J306	List: IR 05000155/1986007 ML20207J303 ML20207J309
References
50-155-86-07, 50-155-86-7, IEB-86-001, IEB-86-1, IEB-86-3, IEIN-86-002, IEIN-86-003, IEIN-86-2, IEIN-86-3, NUDOCS 8607290020
Download: ML20207J336 (11)

Text

. _ - _- . _ -. _-

,

'

.

U.S. NUCLEAR REGULATORY COMMISSION

REGION III Report No. 50-155/86007(DRP) ,

t

!

Docket No. 50-155 License No. DPR-6 Licensee: Consumers Power Company 212 West Michigan Avenue Jackson, MI 49201 i Facility Name: Big Rock Point Nuclear Plant

Inspection At: Charlevoix, MI Inspection Conducted: April 9 - July 10, 1986

i Inspector: S. Guthrie Approved By: D. C. Bo ,

Projects Section 2D

'7/d /[/d Date

!

l Inspection Summary

]

j Inspection on April 9 - July 10, 1986 (Report No. 50-155/84007(DRP))

Areas Inspected

Routine, unannounced inspection conducted by the Senior i Resident Inspector of licensee actions on previous inspection findings,

! operational safety, maintenance operation, surveillance observation, reactor

! trips, IE Bulletins, TMI Action Items, Licensee Event Report Followup, licensing l actions, and trainin ; Results: One violation was identified (failure of the operator to maintain desired control rod position and to investigate the abnormal behavior of

) sluggish control rod movement - paragraph 3.f).

i

!

.

.

I l

8607290020 860718

{DR ADOCK 05000155 PDR i

-- _ _ -

.

.

DETAILS

. Persons Contacted

• D. Hoffman, Plant Superintendent

• Pettijean, Planning and Administrative Services Superintendent

• G. Withrow, Engineering Maintenance Superintendent

• R. Alexander, Technical Engineer i

R. Abel, Production and Plant Performance Superintendent

, *L. Monshor, Quality Assurance Superintendent R. Barnhart, Senior Quality Assurance Administrator P. Donnelly, Senior Review Supervisor, Nuclear Activities Department D. Swem, Senior Engineer G. Sonnenberg, Shift Supervisor

!

D. Staton, Shift Supervisor W. Trubilowicz, Operations Supervisor

'

J. Beer, Chemistry / Health Physics Superintendent E. Evans, Senior Engineer J. Tilton, General Engineer D. Kelly, Maintenance Supervisor D. Ball, Maintenance Supervisor W. Blosh, Maintenance Engineer L. Darrah, Shift Supervisor

.

J. Horan, Shift Supervisor

R. May, Shift Supervisor 4 R. Scheels, Shift Supervisor i

• J. Warner, Property Protection Supervisor

! T. Fisher, Senior Quality Assurance Administrator S. Bartosik, Gerieral Quality A3surance Consultant R. Krchmar, General Quality Assurance Analyst D. Hice, Technical Engineer (Acting)

R. Schrader, Engineering, Maintenance Superintendent (Acting)

1 The inspector also contacted other licensee personnel in the Operations, Maintenance, Radiation Protection and Technical Departments.

!

• Denotes those present at exit interview.

! Licensee Action on Previous Inspection Findings J

(CLOSED) Open Item 155/85014-08, which tracked the licensee's commitment

<' to perfora an analysis to address inspector concerns that the Limiting Condition for Operation on No.1 Control Rod Drive pump proposed for

the Alternate Shutdown Systems Technical Specification was excessively long. The analysis was presented July 3,1986, and is discussed in Section 3 of this repor . Operational Safety Verification

The inspector observed control room operations, reviewed applicable logs l and conducted discussions with control room operators during the I

. - - - . . . - - - - - - _ - _ - - - - . - - . - - . - . ..-. - - - - .

Y inspection period. The inspector verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of the containment sphere and turbine building were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance. The inspector by observation and direct interview verified that the physical security plan was being implemented in accordance with the station security pla The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls. During the inspection period, the inspector walked down the accessible portions of the i.iquid Poison, Emergency Condenser, Reactor Depressurization, Post Incident, Core Spray and Containment Spray systems to verify operabilit The inspector also witnessed portions of the radioactive waste system controls cssociated with radwaste shipments and barreling, On April 25 the inspector observed an emergency exercise drill conducted as practice for the 1986 emergency exercise on May 2 The inspector observed activities throughout all areas of the plant and at the Emergency Operations Facility at Boyne City and noted the generally high level of knowledge among plant personnel of the requirements contained in the Site Emergency Plan and Emergency Plan Implementing Procedures. Interviews with participants, controllers, and observers indicated that the licensee's post drill critique was constructive and comprehensiv Fire brigade response during both the April 25 exercise and the May 20 exercise showed evidence of indecisiveness and lack of teamwork and organization among brigade members whenever the brigade leader was not present to direct brigade activitie On April 29 the inspector observed the response of the fire brigade to a drill conducted near the electrical equipment room. Training conducted by the brigade leader was instructive. During the drill the first aid kit located in the drill area was found to be missing following the emergency exercise drill on April 2 l On May 5 the inspector attended a briefing for the media and governmental leaders conducted by the licensee and the Michigan Department of State Polic On May 20 the inspector acted as an observer on the NRC team evaluating the 1986 exercise. The inspector's comments are incorporated into Inspection Report No. 155/86008(DRSS). During the inspection period the licensee cooperated with the Commission's request to monitor for and report on radioactivity levels in an attempt to assess the impact of the Chernobyl reactor incident. The information was requested by IE Information Notice 86-02. The licensee's augmented environmental sampling program

- - _ _ _ _ _ _ _

'

.

.

-

results were conveyed by letter dated May 14. While air samples at Big Rock did not indicate abnormal results, licensee sampling of dairy milk at local farms detected traces of Iodine-131, which is normally not detected. Iodine-131 levels ranging from 1.1 to 4 picocuries/ liter were detected. The U.S. Food and Drug Administration calls for preventive measures if the concentration of I-131 exceeds 15,000 picocuries per lite f. During a regular review of licensee Deviation Reports (DR) the inspector became aware of an incident of mispositioned control rods which occurred on February 3, 1986, in the previous reporting perio The event marked the third instance of mispositioned control rods in thirteen months. The first example occurred on January 28, 1985 and is reported in Inspection Report No. 155/85002(DRP), Section The second occurred May 27, 1985, and is discussed in Inspection Report No. 155/85007(DRP), Section The second instance was considered a Severity Level IV violation, but because it met the criteria of the Enforcement Policy for licensee self identified and corrected events (10 CFR 2, Appendix C, Part V.A) no violation was issued at that time. In each of the three cccurrences operator inattentiveness played a key role. Operator attention to rod position is particularly important since all rod movement and positioning is manually controlled; the unit has no electronic systems typical of newer plants to minimize the impact of human error by closely controlling the sequencing of control rod manipulatio In the February 3 occurrence control rod E-4 was inserted out of sequence from position 23 (fully withdrawn) to position 11 while operators were performing daily control rod exercises required by Technical Specification 5.2.2.(e). The daily rod exercise is not

.

I controlled by procedure and calls for the operator to select one rod at a time in no prescribed order and exercise that rod by moving it one notch. Rods fully inserted are not exercised, rods fully withdrawn are driven in one notch and returned to the fully withdrawn position, and rods partially withdrawn are, at the option of the operator, inserted or withdrawn one notch. Movement is verified by visually monitoring rod position indication, though some minute change in power indication might be observed if the rod is in a highly reactive position in the core. On February 3, the selector valve on Rod E-4 failed in the open position following the E-4 exercise, thus causing the E-4 rod to move in sympathy with drives subsequently exercised. The previous day E-4 had failed to move during performance of TT 03, Coupling Integrity Test, and an operator had verified the selectvi valve was exercising. The coupling integrity had been checked on E-4 and then the rod had been de-selected. On February 3, after exercising rods A-2, A-3, A-4, A-5 and B-1 the operator noticed io. 2 power level instrumentation dropping off rapidly. The operator had noticed the sluggish behavior of these rods while exercising them but did not stop to investigat Upon the discovery of the E-4 change in position from 23 to 11 the exercises were suspended and the Shift Supervisor notified. After

e

'

.

'

-

determining the selector valve was stuck (possibly from the previous M7) rod E-4 was returned to its correct, fully withdrawn position per Off Nonnal Procedure 2.i. Subsequent review by the Reactor Engineer determined the total power change was a decrease in reactor power of 9.8 MWt, with no thermal limits exceede ,

Licensee corrective action included performing the necessary repairs and testing on the E-4 selector valve and the publication of a letter to operators stressing the need for follow-up testing after minor maintenance performed by operators, including the need to submit a maintenance order if the equipment can't be made to function properly. Incomplete follow-up testing was the root problem

.

identified by the Corrective Action Review Board's evaluation, noting

! that no procedures were violated. The evaluation noted that the operator did not demonstrate complete attentiveness and that his activities were concentrated on assuring the proper movement of

! the sluggish rods instead of investigating possible explanations for

their abnormal behavior. Technical Specification 7.3.4, Normal Power

'

Operation, states that the principal function of operators during

.

'

power operation includes "the adjustment of the control rod pattern to maintain the desired power distribution" and "the evaluation of

'

abnormal conditions and the initiation of corrective action as required". Technical Specification 5.2.2.f, Abnormal Behavior of I the Control Rod System, states that "an immediate and thorough investigation shall be made of the occurrence of any abnormal

,

behavior. . . . of the control system to determine the cause and safety significance of the occurrence". Failure of the operator to i maintain desired control rod position and to investigate the abnormal

,

'

behavior of sluggish control rod movement is considered to be a violation (155/86007-01(DRP)).

Contributing to the event is an apparent deficiency in the preventive maintenance activities related to the selector valves on each of the 32 control rod drives. Records indicate that selector valves fail in approximately nine instances annually. The licensee conducts a weekly reduced pressure test to verify the selector valve will

, exercise, and conducts an inspection of the units every two years.

j Operators are familiar with the need to lubricate the shaft to i

, attempt to free stuck valves, but maintenance personnel are rarely l

involved unless the valve actually sticks and a maintenance order j is issued. The inspector requested the licensee evaluate appropriate j preventive maintenance measures to reduce selector valve problems.

j The licensee agreed, and at the close of the inspection period the j inspectors was informed of activities already underway.

g. On July 3 the licensee presented the results of their analysis conducted in response to inspector concerns that a 60-day operability

requirement on the No. 1 Control Rod Drive (CRD) Pump was excessively

long. The analysis also addressed the option of permitting the

! operation of No.2 CRD pump from the No. 1 pump's controls through a j series of switches during periods when the alternate shutdown (ASD)

system was called upon to be operable. The basis for the inspector's i I

,

i

. -- -, -, . _. , _ _ _ _

__ _____ _ _

,.

.

'

'

concerns is presented in Section 7.c of Inspection Report No. 155/85014(DRP) and can be summarized by stating that the No. 1 CRD pump is the only pump capable of injecting water into the r'eactor vessel at any reactor pressure under ASD system operational conditions, and because of ADS system wiring, is capable of doing so if station power is unavailabl The licensee's analysis concluded that either rewiring to permit the No. 2 CRD pump to be powered by the emergency diesel generator (EDG)

or reducing the operability requirement from 60 to 14 days for No. 1 CRD pump each yielded a slight improvement in affected core damage probability. The licensee stated that the cost of the modification, estimated at $130,000, exceeded the value of the potential safety increase. While acknowledging that the change to the Technical Specification for the 14 day operability requirement represented no significant immediate cost, the licensee refused to take that action on the basis of reduced operational flexibility that could, at some point in the future, limit generating capacit h. During the inspection period the licensee's internal audit activities determined that the requirements of Technical Specification 11.4.5.3.A.1(h) to perform a test during each cycle to verify that the station batteries meet the loaded design time interval of eight hours was not being met. The licensee's test, TR655, " Station Battery Service Test", uses a postulated load profile of 61 minute The 61 minute test has been performed since the inception of battery testing at Big Rock in 1977. The licensee is unable to determine the origin of the eight hour test requirement, and attributes the 61 minute figure to an assumption made by the architectural engineering firm that constructed the facility. The licensee first identified the discrepancy in 1977, attempted to correct it in an aborted attempt to implement Standard Technical Specifications in 1977, and has taken no further action. The licensee's review indicates a load profile is available to support the 61 minute analysis, and intends to submit a Technical Specification Change request. The next regularly scheduled performance to TR655 is during the 1986 refueling outag i. During this inspection period the licensee experienced two instances of failure to meet the requirements of Technical Specification The first, involving the station battery load test required by Technical Specification 11.4.5.3.A is addressed in Section 3.h. of  ;

this report. The licensee's failure to conduct an eight hour test rather than a 61 minute test has no immediate impact on safety and l 1s being submitted to NRR as a technical issu )

The second example is a failure to perform the sampling requirements of Technical Specification 13.1.1.1.B and is discussed in Section 8 of this report. While the failure to sample resulted in no immediate impact on safety, this, like the first instance, represents a trend toward lack of administrative control over the requirements of Teclnical Specifications that is of concern to regulator l l

,

e i' .

'

-

The licensee stated their intentions to submit an LER describing the situatio . Monthly Maintenance Observation Station maintenance activities of safety related systems and components l listed below wera observed / reviewed to ascertain that they were conducted t l in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with technical specifications.

l The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemente Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performance, During the inspection period the licensee experienced two instances of Limiting Conditions for Operation on Reactor Depressurization System (RDS) battery cells. On April 10, Battery C, Cell No. I was declared inoperable based on low specific gravity per the requirements of Technical Specification 11.4.5.3.(3)(b). Appropriate testing was conducted to verify the operability of the other RDS channels while performing repairs. Following repair and testing, the RDS system was returned to full operability. On June 11 an identiccl situation occurred with Battery D, Cell No. On April 25 the inspector observed portions of maintenance performed on the turbine building crane bearings, work which involved the assistance of a vendor representativ During the inspection period the Diesel Fire Pump (DFP) was placed in Limiting Condition for Operation (LCO) status on three occasions (April 28, May 6 and 12). Problems included governor adjustment to place DFP rpm within procedural specifications and continued fuel '

,

system cleanings. The licensee increased the performance of Surveillance Test T7-20 from weekly to daily for several days to verify the accuracy of the governor setting. The licensee has committed to replace the DFP engine during the 1986 outag On May 30 the inspector reviewed licensee repairs to Control Rod Drive Pump No. I to remove a jammed nut and restore threads on the pump plunger. The pump was out of service for three day ,

. - _ _

.

.

e. On June 12 the inspector reviewed with the licensee modifications to the fire protection sprinkler system in the exterior cable penetration room. The system had been modified to operate automatically as a compensatory fire protection measure associated with the requirements of 10 CFR 50, Appendix R. With the completion of the Alternate Shutdown modification in 1985, the need for the compensatory measure was eliminated and the sprinkler system with water supply isolation valves located in a different room. Work and procedural changes were processed, and a fire watch posted during the modification. The licensee performed inspections for piping integrity, nozzle aiming, and potential obstruction On June 18 the licensee informed the inspector that the containment sphere escape lock hac exhibited leakage well below technical specification and administrative limits but higher than that ordinarily found during leak testin Repairs were completed to the lock's equalizing valve and adjustments were performed on the lock's door operating mechanism. The lock was retested satisfactorily, On June 30 the inspector observed portions of maintenance on an area monitor recorder in the control room, involving gear replacemen On Jaly 2 the inspector observed licensee maintenance activities to replace the Reactor Depressurization System (RDS) depressurization valve top assemblies on the B, C, and D RDS trains and repack RDS isolation valve On July 2 the inspector reviewed the licensee's activities to diagnose and correct repeated failures of the turbine stop valve to operate on demand. The stop valve failed to operate in the reactor scram on July 1. The licensee determined that the linkage pulled by the solenoid to operate the mechanism was binding, a condition resulting from wear and possible misadjustment. Linkage parts from a spare unit were adapted and modified to minimize binding. The licensee sought vendor recommendations and conferred with other licensees using the same device. The unit was test operated repeatedly prior to reinstallation, and the master startup procedure was amended to provide for on line trip of the turbine from the control room during startu ' During the period July 3 - July 10 the inspector reviewed the licensee's corrective actions to address problems with the neutron instrumentation system that resulted in the spurious Reactor Protection System trip of July 1. Activities included inspection of the entire system from neutron detector to recorder and replacement of coaxial cables, the channel two range switch, and a picoamete The licensee discovered broken coaxial cable in the feedback circuitry between the picoammeter and range switch. The licensee also verified the operability of the event recorder which, on the July 1 trip, did not register the activation of RDS channel tw A major modification p'anned for the 1986 outage will replace aging equipment with modern nuclear instrumentation, a move expected to prevent the type of failure involved her .

'

No violations or deviations were identified in this are . Surveillance On July 10 the inspector observed performance of Surveillance T90-12, Reactor Depressurization System (RDS) depressurization valve test. The test involves lifting of the main disc off its seat on each of the four RDS depressurization valves using high pressure air. The test operates the valve using the solenoid and is activated by the control room operator. Successful depressurization is verified by use of a strip recorder. Earlier attempts to perform this test during the July forced outage had indicated valve performance that was marginally within performance specification of depressurizing from 1700 psig to 80 psig in under ten seconds, but was substantially longer that the typical performance of that valve which depressurizes fully in under one secon As a result of the valve's acceptable but unusual performance during the first surveillance attempt the licensee made modifications to the mechanical tolerances on top assembly parts using guidance from the valve manufacturer, causing further start-up delays but demonstrating an emphasis on safety over productio . Reactor Trips On July 2 at 10:22 a.m. the unit scrammed from approximately 90% power (68MWe) while technicians were Reactor Protection System (RPS) performing Logic Test. Themonthly surveillance test verifies the T30.01, operability of trip signals from the various plant parameters which activate the RPS system to shut down the reactor. The trip occurred while operators and technicians were testing the trip input signal on the No. 1 picoammeter and the No. 2 piccammeter also tripped upscale. The No. 2 picoammeter had been displaying erratic behavior for several days prior to the event. The RPS functioned properly, fully inserting all control rods in the reactor. However, the turbine stop valve failed to close, thus requiring manual turbine generator tripping by the operator. Fallin steam drum level activated the Reactor Depressurization System (RDS)g sphere evacuation alarm and evacuation of personnel working within containment was successfully accomplished. Steam Drum level low also signaled the successful start of the diesel fire pump (DFP).

In the post trip review the licensee identified the cause of the trip as the erratic behavior of the No. 2 picoammeter. Prior to restart the licensee committed to complete repairs to ensure the operability of the picoammster and the turbine stop valve. The unit entered a 10 day outage to perform these and other repairs, including turbine gland inspection and RDS depressurization valve top assemblies on three of four valve The licensee completed all notifications to state and federal agencie The licensee was particularly communicative and cooperative with the inspector in discussing the event and in identification of any inspector concerns that could be addressed during the brief outag n_ .

~

.

'

Theough the end of the reporting period the inspector reviewed the licensee's corrective actions prior to startup. Repairs to the turbine stop valve and neutron instrumentation were completed, as were repairs and testing on Reactor Depressurization System Valve No violations or deviations were identified in this are . Licensee Action on IE Bulletins By letter dated May 6 the licensee responded to IE Bulletin 86-03 of November 15, 1985. The Bulletin required a response within 180 days on the subject of motor operated valve common mode failures during transients due to improper switch settings. The Bulletin's concern over high differential pressures across safety related motor operated valves is not applicable to Big Rock because the plant does not utilize high pressure coolant injection / core spray and emergency

! feedwater systems.

l During the period the inspector reviewed the licensee's response to IE Bulletin 86-01, Minimum Flow Logic Problems That Could Disable RHR

'

Pumps. The Bulletin describes the potential for single failure in the RHR minimum flow logic and has the potential to disable all RHR I pumps. The inspector concurs with the licensee's conclusion that the l

' concerns are not applicable to the Big Rock Point facility because of significant design differences between Big Rock and newer vintage plants with RHR systems. Big Rock uses a core spray system for LOCA makeup and containment spray and a shutdown cooling system for l decay heat removal when shutdown, none of which contain minimum flow recirculation lines with minimum flow valves and bypasses.

-

Beyond l the licensee's response to the Bulletin no further action is l

i anticipated, No violations or deviations were identified in this are . Licensee Event Report Followup

' Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective

-

I action was accomplished, and corrective action to prevent recurrence had

.

l been accomplished in accordance with technical specifications.

l (Closed) LER 86-003, Inadequate Procedure Resulting in Failure to meet i

Technical Soecification 13.1.1.1.b. Technical Specification 13.1.1.1.b l

requires daily analysis of circulating water samples for radioactivity when the Canal Sample / Circulating Water Discharge Monitor is out of service. Although daily samples were obtained as required, the licensee failed to recognize the need to increase the analysis frequency from weekly to daily. The canal sample pump failed on May 21, daily sampling was commenced, and the failure to analyze was discovered on May 27, whereupon daily analysis was commenced. The pump was returned to service May 28. The licensee attributed the cause of the incident to inadequate

L

r

.

.

procedural guidance to address the recently implemented Radiological Effluent Technical Specifications. The licensee overlooked the Technical Specification action statement that requires the increased sampling frequency. The safety significance of this event is minimal. Licensee corrective action included revisions to both operations and chemistry procedures and a memorandum to personnel in both groups describing Technical Specification requirement . Licensing Activities By letter dated May 6 the Commission issued Amendment No. 84 to Facility Operating Licensee No. DRP-6 for the Big Rock Point Plant. The amendment adds a new condition to Control Rod Drive Surveillance testing to require all testing prior to a startup following an outage greater than 120 days in lengt By letter dated May 8 the Commission approved the licensee's request pursuant to 10 CFR 20.302 to retain contaminated soil in place at the Big Rock facility. The staff concluded that radiation exposures to station workers are small compared to routine occupational exposures at the j facility and that possible radiation risks to the general public resulting I from the disposal are well below regulatory limits and small in comparison i to exposure received annually from natural background radiation. The land in question is located beneath the turbine building's concrete floo By letter dated May 12 the Commission issued Amendment No. 85 to Facility Operating Licensee No. DRP-6 for Big Rock Point. The Amendment deletes certain reporting requirements now included under 10 CFR 50.72 and 50.73 and adds a definition for Reportable Even . Training During the week of April 28 the INP0 Training Accreditation Team reviewed the licensee's training progra During the week of May 5 the inspector observed the administration of oral and written examination to four licensee candidates for the Senior Reactor Operator (SR0) license by an NRC operator licensing contractor. All four SR0 candidates successfully passed the examinatio . Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1)

throughout the month and at the conclusion of the inspection period and summarized the scope and findings of the inspection activities. The licensee acknowledged these findings. The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspectio The licensee did not identify any such documents or processes as proprietar .