IR 05000219/1986002

Insp Rept 50-219/86-02 on 860106-0202.Violation Noted:During Full Power Operations,Operable Snubber Made Inoperable to Remove Paddle for Use on nonsafety-related Snubber Violating Tech Spec

ML20154D406
Person / Time
Site:	Oyster Creek
Issue date:	02/25/1986
From:	Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20154D396	List: IR 05000219/1986002 ML20154D392 ML20154D400
References
50-219-86-02, 50-219-86-2, IEIN-83-69, NUDOCS 8603060074
Download: ML20154D406 (16)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /86-02 Docket N , License N DPR-16 Priori ty Category _C i

Licensee

GPU Nuclear Corporation 100 Interpace Parkway Parsippany, New Jersey 07054 Facility Name: Oyster Creek Nuclear Generating Station Inspection At: Forked River, New Jersey Inspection Conducted: January 6 - February 2, 1986 Participating Inspectors:

W. H. Bateman, Senior Resident Inspector J. F. Wechselberger, Resident Inspector R. L. Nimitz, Sr. Radiation Specialist Approved by:

A. R. BlougkT Chief 2~DateGd Reactor Projects Section IA Inspection Summaryl Routine and special onsite inspections were conducted by the resident inspec-tors and one region based inspector (215 hours0.00249 days <br />0.0597 hours <br />3.554894e-4 weeks <br />8.18075e-5 months <br />) of activities in progress in-cluding plant operations, physical security, radiation control, housekeeping,  !

fire protection, spent fuel pool repair, and receipt, handling, and storage of

, new fuel. The inspectors also met with various members of management to dis-

cuss recent events and changes, followed up on concerns that arose from the

vibroflotation activities associated with the proposed ESSF structure, attended a HVAC briefing, and continued a review of the Intermediate Range 10

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modificatio !

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2 Results:

i One violation of Technical Specification (Tech Spec) requirements was identi-fied involving hydraulic snubber operability (discussed in paragraph 6). Two unresolved items were identified involving concerns with the Intermediate Range 10 modification completed during the cycle 10 R outage. One inspe: tor follow-up item was identified involving radcon concerns related to the dive into the spent fuel poo Few significant plant problems occurred during this report period and the i

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plant continued operation near full power. Problems persisted with airborne contamination in the Augmented Of fgas ( A0G) building due to leaks from the A0G system; 9 people were slightly contaminated during one event when the 'B' recom-biner head was remove Licensee efforts to correct A0G problems continued. A spent fuel pool leak was located by vacuum box leak testing with helium and repaired by a diver without incident. The HVAC briefing was informative as to

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the progress made in restoring major HVAC systems to a functional status,

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DETAILS 1. Plant Operation Review 1.1 Routine tours of the control room were conducted by the inspectors during which time the following documents were reviewed:

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Control Room and Group Shift Supervisor's Logs;

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Technical Specification Log;

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Control Room and Shift Supervisor's Turnover Check Lists; 1 --

Reactor Building and Turbine Building Tour Sheets;

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Equipment Control Logs; i

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Standing Orders; and,

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Operational Memos and Directive The reviews indicated that the logs were generally complet .2 Routine tours of the facility were conducted by the inspectors to

make an assessment of the equipment conditions, safety, and adherence to operating procedures and regulatory requirements. The following areas were among those inspected

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Turbine Building;

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Vital Switchgear Rooms;

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Cable Spreading Room;

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Diesel Generator Building;

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Reactor Building; and,

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Battery Room The following items were observed or verified:

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Randomly selected fire extinguishers and hose stations were accessible and inspected on schedule.

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Fire doors were unobstructed and in their proper positio Ignition sources and combustible materials were controlled in accordance with the licensee's approved proceduras.

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Appropriate fire watches or fire patrols were stationed when equipment was out of servic Fire retardant wood was used for scaffolding.

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Jumper and equipment mark-ups did not conflict with Techni-cal Specification requirement Conditions requiring the use of jumpers received prompt licensee attentio ,

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Administrative controls for the use of jumpers and equip-ment mark-ups were properly implemented.

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Breakers for electrical equipment being worked were proper-ly tagged out.

4 Vital Instrumentation:

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Selected instruments appeared functional and demonstrated i

parameters within Technical Specification (Tech Spec)

Limiting Conditions for Operatio Housekeeping 1 i

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Plant housekeeping and cleanliness were in accordance with approved licensee program No concerns were identified.

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1.3 A problem with the operability status of the fire pumps continued during this report period. The types of problems included out of specification low RPMs with the diesel and water leaks from the pump relief valve and piping connections. These are recurring problems and have resulted in Plant Engineering initiating a Tech Functions Work Request to address the Tech Functions was evaluating the

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problems at the end of the report period. The fire water pumps are required by Tech Specs; the licensee has followed appropriate Action Statement .4 The inspectors observed receipt, handling, QC inspection, and storage of new fuel assemblies. No concerns were identifie .5 The 'C' and 'D' Emergency Service Water (ESW) pumps were determined

' to be in the ASME Code Section XI Action Range for flow during per-formance of the routine monthly surveillance /IST. They were declared i *

inoperable. Troubleshooting concluded that the problem was not with pump performance but with the flow measuring instrumentation. As a result the "Controlatron," an ultrasonic flow measuring device, was repaired and the IST reperformed. The results indicated 'C' pump

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performance was within acceptable limits, but that the performance of

'D' pump was still in the Action Range. 'C' pump was declared oper-able and 'D' remained inoperable. Plant Engineering reviewed other monitored pump and system parameters and concluded there was nothing wrong with the performance of 'D' pump and subsequently re-established new baseline data and declared the pump operable. The inspectors expressed a concern with the practice of establishing new baseline data because it is done on a fairly frequent basis for the ESW pumps and appears inconsistent with the purpose of Section XI IS The inspectors will review the ESW IST program in a subsequent inspectio . Observation of Physical Security During daily tours, the inspectors verified access controls were in accor-dance with the security plan, security posts were properly manned, protected area gates were locked or guarded, and isolation zones were free of obstructions. The inspectors examined vital area access points to ver-ify that they were properly locked or guarded and that access control was in accordance with the security pla The licensee invited the inspectors to attend a vendor presentati0n in-volving state of the art motion detection systems. The inspectors attend-ed the presentation and found it informative. The licensee is continuing to pursue upgrading their present motion detection system to, in part, eliminate nuisance alarm No concerns were identifie . Radiation Protection

&aring entry to and exit from the RCA, the inspectors verified that proper warning sign < were posted, personnel entering were wearing proper dosime-try, personnel and materials leaving were properly monitored for radioac-tive contamination, and monitoring instruments were functional and in calibration. Posted extended Radiation Work Permits (RWPs) and survey status boards were reviewed to verify that they were current and accurat The inspector observed activities in the RCA to verify that personnel com-plied with the requirements of applicable RWPs and that workers were aware of the radiological conditions in the are On December 17, 1985 nine workers were slightly contaminated while working in the Augmented Offgas (A0G) building. The radionuclides involved were short-lived rubidium-88 and cesium-138. The airborne contaminants were apparently released into the A0G atmosphere when the end bell was removed from the 'B' recombiner blower during blower troubleshooting activitie Prior to removing the end bell, the 'B' recombiner subsystem (blower and palladium catalyst bed) had been isolated and purged with clean air. The licensee concluded the radioactive contaminants entered the 'B' recombiner subsystem through leaking valves. Corrective action was taken by the licensee to modify procedures and valve lineups when isolating a recombiner subsystem for maintenance to help preclude recurrence of this proble _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - - _ - _ _ -_

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At the end of this report period, the A0G system was back in service after completion of helium leak testing and repairs of identified leaks. One source of gas leaks from the system into the building atmosphere was de-termined to be through cracks in stainless steel hydrogen analyzer instru-ment piping apparently resulting from intergrannular stress corrosion cracking. The affected piping was replaced. Not all leaks were identi-fied as indicated by the presence of radionuclides in the A0G building during subsequent system operation. The indicated radiation levels, how-ever, have decreased from previous levels as a result of the repair . Expanded Safety System Facility (ESSF)

l Vibroflotation of the fill adjacent to the north side of the reactor building was in progress to prepare the area for the ESSF. As a result of this fill compaction activity, unexpected ground settlement occurred and the vibroflotation was stopped. The licensee and their consultants were I involved in analyzing the impact of the unexpected settlement on the ESSF at the end of the report perio The licensee agreed to make a technical presentation to NRC Licensing by the end of March 1986 to discuss the im-pact of this settlement on previously discussed issue . Status of Heating, Ventilating, and Air Conditioning (HVAC)

At the request of the inspectors, the licensee explained to them the his-tory of HVAC problems at Oyster Creek and the efforts completed, in pro-gress, and planned to restore and upgrade plant HVAC systems. The inspec-tors requested this briefing because of the HVAC prcblems occurring at the ;

plant. Although most of these HVAC systems are not safety-related, their l proper functioning is important in providing plant habitability, maintaining equipment operating environments within design, and controlling the spread !

of radioactive contaminatio From the briefing, it was learned that many of the originally installed HVAC systems had seriously degraded to the point that they were non-functional. Additionally, there were cases of inadequate desig ,

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the early 1980's efforts were initiated to correct the problems and many systems have since been repaired and upgraded to a functional status in l the manual mode with restoration of automatic capability planned for the future. Degraded items that have been repaired or are in the process of being repaired include ductwork, insulation, supports, dampers, automatic l controls, rotating equipment, protective enclosures, piping, heating and cooling coils, flexible joints, filters, air operators, actuators, and control circuitr Preventive maintenance programs and periodic inspec-tions are now in place to maintain an operational status. The inspectors concluded from this briefing that the licensee is aware of the importance of HVAC and has spent a considerable amount of effort in upgrading and restoring degraded equipment. The number and type of HVAC problems that have occurred over the past year are indicative, however, that certain repairs were not sufficiently scoped and/or effected and that problems will continue to arise and require attentio i

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There appears to be sufficient management attention regarding HVAC to pro-vide confidence that the major HVAC systems will be totally restored and maintaine . Hydraulic Snubber Operability A review of Short Forms (work orders) involving work on two snubbers indi-cated a violation of Tech Spec requirements had occurred. Short Form 26521 l involved repairs to non-Tech Spec snubber 15/1. The repair of snubber l 15/1 involved replacement of a damaged paddle. When a spare paddle could l not be located, it was determined the paddle on snubber NQ-2-57 would be l suitable for use on 15/1 and that a spare paddle could be used on NQ-2-S7.

I This resulted in the issuance of SF 26522 to remove the attached front l paddle on NQ-2-S7 and to replace it with a spare. The Tech Specs state in

paragraph 3.5. A. Shock Suppressors (Snubbers)

l l During all modes of operation except cold shutdown and refuel, all safety related snubbers listed in Table 3.5.1 shall be operable ex-cept as noted in 3.5.A.8.b, c and d belo Paragraph 3.5.A.8.b states:

From and after the time that a snubber is determined to be inopera-ble, continued reactor operation is permissiblo only during the suc-ceeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless the snubber is sooner made operable or replace Snubber NQ-2-57 is included in Table 3.5.1 and 15/1 is not. Paragraphs [

3.5.A.8.c and d do not permit the licensee to vary from the requirements of 3.5.A.8.a and b. The action of disconnecting and exchanging parts on a l Tech Spec snubber to enable repair of a non-Tech Spec snubber rendered I

the Tech Spec srubber inoperable. This action violated the above l Technical Specification in that snubber NQ-2-57 was previously operable i and had not been determined inoperable prior to paddle replacement. This is a violation (219/86-02-01). The snubber was inoperable for less than 72 heurs, thus, despite the unacceptable action to degrade a fully operable :nubber, no plant shutdown was require A review of the particular circumstances that led to this violation indi-cat 3d SF 26522 was not clearly written and led the Group Shift Supervisor to believe the Tech Spec snubber was inoperable to begin with. As a re-sult, he approved the work activity. Additionally, Station Procedure 775.1.004, Rev. 9, Removal / Replacement of Bergen-Paterson Hydraulic Snub-bers, did not address control of snubber remova These issues should be addressed by the licensee in their response to this violatio During NRC review of this item, the licensee pointed out that a proposed Tech Spec Amendment, now under NRC review, will slightly revise the wording so that voluntarily rendering a snubber inoperable will be permissible, so long as the Action Statement time limit is not exceeded. The inspector l

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reviewed the basis for the provision and fcund that it is to allow preven-tive maintenance, testing and minor corrective maintenance to improve equip-ment reliabilit Even under the proposed Tech Spec wording, scavenging parts from a fully operable snubber would be considered an abuse of the Tech Spec, and would not be considered acceptabl . Meetings with Management The inspectors m.t with senior management of Maintenance, Construction, and Facilities at which time they were updated on key personnel changes both onsite and in Parsippany. Additional subjects discussed included:

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Improving efforts to anticipate the need for Tech Spec changes during job planning;

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Expanding the job monitor program to use contractor personnel;

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SALP report and response; and

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Planning for the 11R outag The inspectors also met the the Chairman of the General Office keview Board who is also responsible for the Independent Onsite Safety Review Group. The topics discussed included the safety review, Tech Spec change, and unreviewed safety question processe . Review of Periodic and Special Reports Upon receipt, periodic and special reports submitted by the licensee pur-suant to Technical Specification requirements were reviewed by the inspec-tors. This review included the following considerations: the report includes the information required to be reported to the NRC; planned cor-rective actions are adequate for resolution of identified problems; and the reported information is vali The December, 1985 Monthly Operating Report was reviewed by the inspector In addition Special Report 85-03, involving deficiencies with 14 fire dampers, was reviewed during this report perio The inspectors were in-terested in this because the deficiencies were identified as a result of licensee followup of NRC Information Notice 83-69. The inspectors ex-pressed a concern regarding the two year delay in followup. The licensee stated efforts were in progress to catch up with the backlog. The inspec-tors have observed licensee activities that indicate they are aware of and j addressing the backlog proble No concerns were identifie _ .

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9 Reactor Water Level Instrumentation Low reactor water level sensors were replaced during the cycle IOM outage (October 18 to November 16,1985) to comply with the Environment Quali-fication Rule (10 CFR 50.49). The previous sensors, equipped with indi-cating gauges, were replaced with environmentally qualified switches without indicating gauges. This became significant when the Tech Spec requirement to perform a channel check on the instruments to verify hydraulic communication of the sensor with the reactor was considere The Office of Nuclear Reactor Regulation (NRR) required that an alternate means be devised to verify that the sensor was in hydraulic communication with the reactor vessel after performing routine surveillance action This would provide assurance that the instrument was correctly returned

, to service after performing the required surveillance actions, similar to the function the indicating switches previously served. The licensee, i to answer NRR's requirement, proposed a special valving sequence be performed to provide assurance that the instrument was in hydraulic com-munication with the reactor vessel. The manufacturer, Static-0-Ring, i

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Inc., concurred that this special valving sequence would pose no adverse affect on the switches nor instrument setpoint recover During a routine surveillance test on January 17, Reactor Water Low Level Scram Sensors RE 05/19A1, RE05A1, and RE05/1981 setpoints were found to

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I have drifted out of specifications. The instrument setpoints were cor-rected and the sensors returned to service. On January 20, the surveil-lance test was repeated to determine if the instrument drift experienced on the 17th may have been caused by the special valving sequence employed when returning the instruments to service following a surveillance. Dur-ing the repeat surveillance test on RE05A1, a half scram signal, initiated as part of the surveillance, could not be rese The licensee commenced a reactor shutdown and subsequently was able to clear the half scram by re-peating the surveillance on RE05A1, but not before the Tech Spec require-ment allowing an instrument to be inoperable for one hour per month for testing, was exceeded by eight minutes. The reactor shutdown was halted after the surveillance was completed until the licensee could make a de-termination regarding the operability of the instrumen Based on the erratic performance of RE05A1, the licensee decided to declare the instru-

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ment inoperable and recommenced a reactor shutdow Plans were immedi-ately formulated to replace the instrument during the reactor shutdow The new instrument was installed, calibrated, and returned to service and as a result the shutdown was halted at 500 MW !

The licensee is conducting an investigation to determine the cause of the instrument failure, including having the manufacturer conduct a failure analysis of the instrument. Preliminary findings indicate that the fail-ure was a result of high differential pressure across the instrument, as a result of the special valving sequence performed during surveillance test-in The licensee has developed another method to ensure the instruments are in hydraulic communication with the reactor vessel which involves the installation of test isolation valves on the high and low pressure vents

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on each instrument. These valves were installed on RE05A1 when the switch l was replaced. This new method will significantly reduce the differential q pressure that the sensor diaphragm experiences during surveillance test-2 ing. The licensee has developed a modification package to install the test isolation valves on the remaining low level instruments as they are

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1 IRM Number 10 Safety Evaluation

! This concern was previously addressed in item 219/85-23-07 as a result of the MSIV closure scram on June 12, 1985. During the scram, the operators were unable to reset the trip until the reactor was depressurized to the

600 psig bypass setpoint for MSIV closure and the low condenser vacuum scram. The 600 psig bypass setpoint allows the operator to establish a

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main condenser vacuum and to supply steam to the secondary plant. After the 600 psig setpoint is exceeded, the reactor protection circuitry places the MSIV closure and low condenser vacuum scram in service. Therefore,

, decreasing plant pressure to 600 psig disables the MSIV scram and allows the trip to be rese The inability to reset a scram until the plant is depressurized to 600 psig, when a valid scram conditien no longer exists,

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is a concern. During the June 12, 1985 scram, this inability coupled with

the failure of the scram discharge volume (SDV) drain valves to seat pro-perly, resulted in a steam release to the reactor buildin An analysis conducted by General Electric for BWR 5 and 6 series reactors l permitted raising the bypass setpoint from 600 psig to 800 psia. This was i

submitted by the licensee and approved as a Tech Spec change. Due to

other considerations that follow, the plant elected to continue to operate

with the bypass setpoint at 600 psig. If the licensee elects to change this setpoint to 800 psia and operate the plant in accordance with the change as authorized by Tech Specs, then an analysis or documentation should be reviewed to ensure the General Electric BWR 5/6 analysis is suitable for the Oyster Creek facilit In addition, the basis for Tech j

i Spec Section 3.1, Protective Instrumentation (page 3.1-3a) uses 600 psig l

' while Table 3.1.1 Note B (page 3.1-12) uses 800 psia; this should be clarifie The licensee should adequately address the MSIV closure scram

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reset point (600 psig) in their analysis, clarify the Tech Spec with regard to the bypass setpoint, and provide documentation to support the

{ Technical Specification analysis tnat changed the setpoint from 600 psig to 800 psia. These items are unresolved. (219/86-02-02)

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The 600 psig bypass setpoint was not changed to the 800 psia setpoint

as authorized by Tech Specs due to a concern the licensee developed as a

result of the Intermediate Range Monitor (IRM) System Range 10 modifica-l' tion. The IRM Range 10 modification was installed during the cycle 10 i

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refueling outage to facilitate the clearing of LPRM downscales during re- l actor startups. The safety evaluation conducted to support their modifi- I j cation was completed in May 1982, but additional concerns were raised by i the licensee in June 1984, after NRC:NRR acceptance of the licensee's j

safety analysis for the modification, concerning the acceptability of i

this modification. The licensee's concern involved a postulated l

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l l reactivity addition accident (excessive feedwater addition, idle recir-l culation loop startup, etc.) occurring during a reactor startup while in

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the 25% safety limit associated with the GEXL correlation and possibly

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causing core damage. The scenario would have the operator, when less than l

600 psig, range the IRMs to range 10 (in violation of plant procedures)

when power escalated with the reactivity additio Operator upranging to l IRM range 10 would place a scram setpoint of 38.4% in effect and, coupled with plant pressure below 600 psig (which bypasses the MSIV closura scram initiated when switching to range 10 from below 850 psig), could poten-The IRM 10 safety evaluation

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tially lead to exceeding a safety limi discussed the successful transfer to IRM Range 10 when reactor pressure was greater than 850 psig and the protection afforded by an liSIV closure scram if the transfer was attempted below 850 psig. This is not entirely correct as the safety evaluation did not consider transferring to IRM range 10 from below 600 psig which mcy be conducted without an MSIV closure scra Corporate engineering wrote an internal memorandum (August 1984)

addressing this concern stating that no fuel damage would occur, but did not properly address exceeding the safety limit. The memorandum also ,

offered to further analyze the event and explore solutions to eliminate inadvertent entry into IRM range 10. The licensee had taken no action in this area until a recent inadvertent entry into IRM range 10 caused an MSIV closure scram to occur during a scram recovery. The licensee is currently contemplating hardware and circuitry changes as a possible solu-tion if their analysis supports a modificatio Pending the licensee's resolution, the potential to exceed the 25% safety limit as a result of the IRM Range 10 modification is an unresolved ite (219/86-02-03)

11. Reactor Building 23' Elevation Sprinkler The automatic sprinkler system on the Reactor Building 23' elevation was actuated on January 14, 1986 by truck exhaust fumes. The truck was located in the reactor building access airlock, unloading new fuel canisters. The exhaust fumes caused an ionization detector in the fire protection system to initiate delug? system No. 8 (reactor building system south - cable trays at elevation 23'). The licensee had previously installed a ventila-tion exhaust trunk to safely conduct the exhaust fumes to the outside en- i vironment during return of spent fuel from West Valley. This arrangement '

was removed after the West Valley spent fuel shipments were completed and was not reinstalled for the new fuel shipments. The licensee has refabri-cated an exhaust trunk to remove the truck fumes and plans to reflect this in a procedure chang The deluge system actuation wet safety related equipment and as a result Core Spray System II was declared inoperable by the licensee. Core Spray i System II is equipped with splash shields to protect the vital equipment '

from this type of even The pump motors were meggared and examined to determine if there was any degradation. The test results indicated that

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Core Spray System II was unaffected by the event and it was, therefore, returned to service. The licensee performed additional walkdowns of the 23' elevation to determine if any equipment had suffered damage. No problems were identified. A fire watch was established until the deluge system vas returned to operation.

j No inspector concerns were identified.

1 Fuel pool Liner Repair The licensee discoverad leakage from the spent fuel pool liner coming from the

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tell-tale drain lines in the Shutdown Cooling room in December. The

! initial leak rate was approximately three gallons per hour. The licensee j j

employed a helium leak test to locate the leak in the spent fuel pool next i to a swing bolt used to fasten previous spent fuel racks. The licensee

speculated that the cracks in the liner resulted from relieving the j stresses placed on the liner by the torqued swing bolt and the weight of the spent fuel rack. It was at this time when the old fuel rack was being removed from the pool that the licensee noticed the leak in the l liner as indicated by the flow of water from the tell-tale drain.

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i Repair procedure were developed and a diving contractor was selected to l perform the underwater weld repair. The repair involved fillet welding a i 10" diameter piece of pipe, approximately 8" long with a cover plate on a one end, to the liner floo This pipe section covered the swing bolt and

the cracks, thus isolating the leaks. The licensee took special precau-

tions to minimize the diver's exposure, including relocating items in the ,

pool. The inspector met with onsite Radiological Controls personnel to i discuss licensee planning and preparation for the repair of the Spent Fuel Pool liner, i

The following matters were discussed by a Region I based Radiation Specialist on January 30, 1980:

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planning and preparation,

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establishmert and approval.of procedures (as necessary) for:

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diving operations

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emergency response (e.g. loss of breathing air, loss of pool l i water, damage of diving equipment and suits)  !

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exposure control including source checking radiation survey instrumentation j

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pool decontamination,

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radioactive source control (e.g. incore instrumentation),

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control of access to fuel,

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licensing requirements (if necessary),

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training of personnel in changes to procedures,

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water clarity,

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control of diver approach to spent fuel,

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dose mapping of pool including gamma and neutron radiation,

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personnel dosimetry and its calibration (whole body, skin, and

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use of multiple survey instrumentation, its calibration and periodic verification of operability,

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use of survey meters and alarming dosimeters during underwater work,

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contamination centrol including control of possible point sources (e.g. small chips).

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control and verification of movement of spent fuel,

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bioassays of diving personnel,

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training and qualification of personnel on applicable procedures,

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applicable NRC guidance in this area (e.g. IE Information Notice N , " Overexposure of Diver in Pressurized Water Reactor (PWR) Re-fueling Cavity"),

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breathing air quality for divers,

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previous diving operations at reactor facilities in NRC Region Documents Reviewed

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Procedure A15A-51752, Rev. O, " Underwater Repair of Spent Fuel Pool Liners," January 23, 1986 1 1 --

Procedure MTH-80-0004, Rev. 6, " Procedure for Underwater Diving Work Associates," dated December 7, 1933 (Nuclear Utility Construction)

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ALARA Review 86-047, " Perform Weld Repair of Spent Fuel Pool Liner,"

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Various underwater radiation survey results

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personnel training records

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applicable Radiation Work Permit Findings Within the scope of this review, the following matters needing licensee attention were identified. The licensee satisfactorily resolved these issues prior to inspection completion of reviews in this are '

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Evaluate adequate calibration of personnel monitoring devices (e.g. TLDs, pocket dosimeters) used to quantify personnel exposure in the spent fuel pool. Ensure the calibration is appropriate for the radiological environment (e.g. , gamma, and beta radiation).

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The licensee evaluated the energy monitoring capabilities of his per- i i sonnel monitoring device and determined it to be acceptable. Algo-i rithms could properly quantify exposur Evaluate the capability of radiation survey instrumentation (me-ters, pocket dosimeters) to properly assess exposure conditions in the spent fuel poo Ensure instrument calibration is acceptable for j

the radiological environment being monitored.

I The licensee evaluated the capabilities of his instruments and deter-mined them to be acceptable.

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Provide / ensure clear guidance relative to personnel that are responsi-

ble for ensuring workers are trained (qualified in diving procedures).

The licensee revised procedures to address this matte Establish clear guidance on actions to take following potential identification of damage to diving equipment and suit .

The licensee revised procedures to address this matte '

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provide controls and establish a minimum program (as necessary) to bioassay and determine a potential intake of tritium and alpha

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The licensee performed an appropriate evaluation and reviewed controls j to ensure these matters were addresse The licensee agreed to address and resolve the following matters prior to diving operations (50-219/86-02-04).

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Evaluate potential neutron doses to divers, I

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Clearly define minimum requirements for source / response checking of l survey in:truments and alarming dosimeters )rior to each dive, l

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Provide specific guidance relative to minimum acceptable water clari-l ty for diving operations to start / continue,

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Establish diving suit contamination / radiation limits, and

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Provide controls (as necessary) to ensure the use of proper instrument-tation for performing underwater surveys by a dive The following positive attributes were noted:

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The licensee performed and documented an ALARA review for the diving operatio The licensee decontaminated the work area by u^se of underwater vacuumin Water clarity was very goo The licensee removed radioactive sources from the work area, includ-

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ing spent fue The licensee assigned and ledicated specific individuals to the tas As a result of the licensee's effert, the diver received relatively low ex-posures; a whole body dose of 29 mrom and an extremity dose of 94 mrem was received during the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 18 minute div . Surveillance Testing The inspector reviewed the followiig surveillance tests to determine if ,

the tests were included on the master surveillance schedule, were techni- '

cally adequate, and were performed at the required frequenc .3.013 -- Reactor Low Level Test and Calibration, Revision 17, 12/16/85 This surveillance was performed on January 20, 1986 and is discussed in paragraph 9. 'Short Form 33076 was written to replace RE05A l 665.3.021 -- Containment Electrical Pehetration Nitrogen Blanket Surveillance, Revision 2, 11/08/85

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636.4.003 -- Diesel Generator Load Ter.i., Revision 23, 12/02/85 i

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645.2.002 -- Fire Pump Diesel Battery Verification, Revision 12, 9/26/85 620.3.003 -- APRM Surveillance Test and Calibration, Revision 13, 9/20/85 Short Form 32063 was written to have the power test poten-tiometer cleaned or replace ; 1 Licensee Action on Previous Inspection Findings  ;

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(Closed) Inspector Followup Item (219/85-23-07): Adequacy of IRM Range 10 l modification to insure 25*. safety limit is not exceeded during a startup

< reactivity accident.

This item is updated in this report (paragraph 10) and was changed to two l' unresolved items. Licensee action will be tracked by the unresolved items; '

therefore, this item is closed.

I 1 Exit Interview

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A summary of the results of the inspection activities performed during j

this report period were made at a meeting with senior licensee management

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at the end of the inspection. The licensee stated that, of the subjects

! discussed at the exit interview, no proprietary information was included.

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