IR 05000302/1981021
| ML20042B730 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 11/13/1981 |
| From: | Brownlee V, Beverly Smith, Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20042B668 | List: |
| References | |
| 50-302-81-21, NUDOCS 8203250593 | |
| Download: ML20042B730 (17) | |
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'o,, UNITED STATES ,0 g NUCLEAR RE2ULATTIRY COMMISSION d a REGION 11 g / 101 MARIETTA ST., N.W., SUITE 3100 g*****g ATLANTA, GEORGIA 30303 Report No. 50-302/81-21 Licensee: Florida Power Corporation 320134th Street, South St. Petersburg, FL 33733 Facility Name: Crystal River Unit 3 Nuclear Generating Plant Docket No. 50-302 License No. DPR-72 Inspection at Cyrstal River site near Crystal River, Florida Inspectors: [[h.
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TgF. Stetka, S6rd6r ReHdent Inspector Date Signed h? e%-~' ////J/b"/ n BTW. Smith 'Res'ident* Inspector Ddte Signed I I //3 [f / Approved by: ytetIn A m . V. L. 8,fownlee, Sec~ tion Chief, Division of Date S'igned ~ Resident and Reactor Project Inspection SUMMARY Inspection on September 23 - October 23, 1981 Areas Inspected Routine inspection by the resident inspectors of plant operations, security, radiological controls, Licensee Event Report (LER's) and Nonconfonning Operations Report (NCOR's), nonroutine events, refueling activities, and licensee action on previous inspection items. Numerous facility tours were conducted and facility operations observed. Some of these tours and observations were conducted on back shifts. The inspection involved 232 hours onsite by two resident inspectors.
Results Five violations were identified (Failure to follow procedures for the performance of plant maintenance, paragraph 5.a; Failure to establish a fire watch following loss of a penetration fire barrier, paragraph 5.b; Failure to adhere to radiation protection procedures protective clothing requirements, paragraph 5.b(5); Failure to escort individuals that are required to be escorted while within the protected area, paragraph 5.b(7); Failure to follow procedures during refueling activities, paragraph 5.b(12).
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DETAILS 1.
Persons Contacted Licensee Employees
- G. Boldt, Technical Services Superintendent
- C. Brown, Nuclear Compliance Supervisor D. Brock, Nuclear Maintenance Specialist
- J. Bufe, Compliance Auditor M. Collins, Reactor Specialist
- J. Cooper, QA/QC Compliance Manager B. Crane, Plant Training Manager C. Goering, Acting Planning Engineer
- J. Henrikson, Plant Engineer W. Herbert, Technical Specification Coordinator V. Hernandez, Compliance Auditor C. Hunter, Acting Relay Department Foreman
- L. Kelly, Maintenance Specialist
- B. Komara, Compliance Auditor T. Lutkeahus, Technical Assistant to the Nuclear Plant Manager
- J. Martin, Plant Engineer P. McKee, Operations Superintendent
- G. Perkins, Health Physics Supervisor
- D. Poole, Nuclear Plant Manager
- H. Reeder, Nuclear Operational Technical Advisor G. Ruszala, Chemistry / Radiation Protection Manager D. Smith, Technical Services Superintendent
- G. Williams, QC Supervisor K. Wilson, Licensing Specialist Other personnel contacted included office, operations, engineering, maintenance, chem / rad, and corporate personnel.
- Present at the Exit Interview conducted on October 23.
2.
Exit Interview The inspectors met with licensee representatives (denoted in paragraph 1) on numerous occasions during and at the conclusion of the inspection on October 23.
During this meeting the inspectors summarized the scope and findings of the inspection as they are detailed in this report. During this meeting, the violations, and inspector followup items were discussed.
3.
Licensee Action on Previous Inspection Items (Closed) Inspector Followup Item (302/81-11-11): The licensee revised surveillance procedures SP-405 and SP-603 on September 25 so that the procedures' acceptance criteria is consistent with that stated in the facility Technical Specification F .
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(Closed) Inspector Followup Item (302/81-15-05): The licensee revised surveillance. procedures SP-370 and SP-435 on September 22 to insure that decay heat valves DHV-4, 41, 42, and 43 are timed for stroking in the correct direction and that the stroking time for valve DHV-91 is included in the SP's. The licensee has also initiated a Technical Specification change to correct the errors in section.3.6.3.1.
(Closed). Inspector Followup Item (302/81-07-07): The licensee has revised and implemented training procedures TDP-202, Replacement Operator and TDP-203, Licensed Operator Requalification Training Program on.0ctober 2, 1981.
These procedures delineate 'the methods of incorporating operational feedback information into the operator training programs.
(Closed) Inspector Followup Item (302/81-15-09): The licensee has completed ! training for mitigating core damage for all required personnel. All but two I personnel completed the training by October 1, 1981.
The remaining two personnel completed their training on October 9,.1981.
(Closed) Unresolved Item (302/78-25-02): The licensee has issued and implemented training procedure TDP-203, Licensed Operator Requalification Training Program.
This procedure delineates the requirement and the forms-utilized to issue a semi-annual report of operator training results.
(Closed) Unresolved Item (302/81-19-04): The licensee has reviewed their surveillance procedures and determined that no other procedures utilize the Setpoint Document for safety-related instrument testing or calibration..The licensee has revised procedure SP-169, Plant Safety-Related Instrument - Calibration, on October 14, 1981, to provide all necessary setpoint infor-mation as an integral part of the procedure.
(0 pen) Inspector Followup Item (302/81-15-06): The licensee completed inspection of the accessible and inaccessible snubbers. No cracked bushings were identified on the accessible snubbers. A total of 9 cracked bushings were identified on the inaccessible snubbers. As a result of these inspec-tions, the licensee is planning to replace all inaccessible snubber aluminum byshings with stainless steel bushings during the present refueling outage.
The item will remain open pending completion of the inaccessible - snubber bushing replacement.
(Closed) Inspector Followup Item (302/81-02-17): The licensee has revised this issue and determined that the plant trip was caused by an unnecessary step in surveillance procedure SP-110 which required depressing of a " reset" button prior to closing the CRD breaker (The operator inadvertantly pressed the " trip" button to cause a trip while performing the SP-110 procedure step which required pressing of the " reset" button).
A plant modification completed in October 1979 makes it unnecessary to press the reset push-
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. button prior to closing the CRD breaker and the licensee has revised pro-cedure SP-110 to delete steps requiring depressing of.the " reset" push-button. The licensee has concluded that this action will minimize recur-rence of this event and the inspectors agree with the licensee's assessment.
(Closed) Inspector Followup Item (302/81-15-01): MUV-253 handwheel has been installed. The inspector has no further questions on this item.
(Closed) Unresolved Item (302/81-07-01): The inspector has looked at ' numerous clearance orders that have been issued and cleared and determined that this issue to be an isclated case.
The inspector has no further questions at this time.
(Closed) Inspecto.' Followup Item (302/81-13-05): The necessary procedure changes have been made to note the potential of disabling the permissive logic when transferring power supplies. The licensee's actions on this item is considered to be complete.
(0 pen) Inspector Followup Item (302/81-02-18): Revision 42 to AI-500 l provided -guidance for performing valve lineups on valves blue tagged to a ' department outside of operations, Revision 43 to AI-500 inadvertantly deleted this guidance.
This item will remain open until a new revision reinstates the valve lineup guidance.
4.
Unresolved Items
There were no unresolved items identified in this inspection report.
. 5.
Review of Plant Operations The plant continued with Mode,1 power operations until September 27, at which time a plant shutdown to Mode 6 refueling operations was commenced in preparat Mn for the start of a scheduled refueling outage.
The plant continued in Mode 6 for the duration of the inspection period.
l a.
Shif t Logs and Facility Records The inspectors reviewed the records listed below and ai cussed various entries with operations persontc1 to verify compliance with TS and the licensee's administrative proo.. re a. " -Shi f t Supervisor's Loa- -Reactor Operator's Lop-Equipment Out-of-Servin Log, -Shift Relief Checklist; -Control Center Status Board; -Short Term Instructions; -Auxiliary Building Operators Log; , .~, . _ _, _ _ _ _ _,,_.._.--r__._ , _ _., _ _ _ _ _... -,. _ -, _, _ , _, _
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-Chemistry / Radiation Log; -Daily Shutdown Surveillance Log; -Refueling Supervisor's Log; -Work Requests; and-Modification Approval Records.
In addition to these record reviews, the inspectors independently? verified selected clearance order tagouts.
'On October 16, 1981, the inspectors reviewed a sampling of twenty-one - work requests (W/R's) which recorded completed work for the period of February 11 through October 5,.1981 on facility security systems. As-a result of this review, the. following eleven W/R's were found to have had the work completed either prior to signing of the W/R-.by the Nuclear Shift Supervisor (NSS) or without any NSS signature.
W/R No.
Date Work Completed NSS Signature Date 21652 2/11/81 2/12/81-22924 4/24/81 No Signature 22928 4/24/81 No Signature 23432 6/3/81 No Signature - 24966 6/16/81 8/9/81 19758 7/7/81 8/9/81 25202 7/14/81 No_ Signature 25201 8/4/81 No Signature 25281 8/4/81 8/9/81 . 22395 8/5/81 8/9/81 28555 10/5/81 No Signature Compliance procedure CP-113, Procedure for Handling and -Controlling, Work Requests, step 5.3.1.1 requires the NSS to sign Part III of the W/R before work is begun to insure that the NSS is cognizant of the work being performed. Failure to sign W/R's before work is begun is contrary to the requirements of Technical Specification 6.8.1 and Regulatory Guide 1.33 and is considered to be a violation.
Violation (302/81-21-01): Failure of NSS to sign work requests prior to commencing work as required by procedure CP-113.
b.
Facility Tours and Observations Througout the inspection period, facility tours were conducted to observe operations and maintenance activities in progress.
Some operations and maintenance activity observations were conducted during backshifts. Also, during this inspection period, numerous licensee meetings were attended by the inspectors to observe planning and management activitie I - . . . , , .
The facility tours and observations encompassed the following areas: -Security Perimeter Fence; -Turbine Building; -Control Room; -Emergency Diesel Generator Rooms; -Auxiliary Building; -Intermediate Building; -Reactor Building; -Battery Rooms; and-Electrical Switchgear Rooms.
As a result of these tours, the following was identified: At approximately 0645 un October 20, during a tour of the emergency diesel generator (EDG) rooms, the inspector discovered the fire door separating the "A" and "B" EDG motor control center rooms (D-203) and the fire door separating the "A" EDG motor control center from the "A" EDG (D-204) propped open without the presence of a fire watch.
The inspector notified the licensee and action was taken to restore the fire doors to functional status.
Technical Specification 3.7.12 requires all penetration fire barriers protecting safety-related areas to be functional. In addition a continuous fire watch must be estab-lished for a non-funcational penetration fire barrier.
Failure to establish a continuous fire watch for fire barrier doors D-203 and D-204 is contrary to Technical Specification 3.7.12 and is considered a violation.
. Violation (302/81-21-02): Failure to establish a continuous fire watch for fire barrier doors D-203 and 0-204 as required by T.S. 3.7.12.
During these tours, the following observations were made: (1) Monitoring instrumentation - The following instrumentation was observed to verify that indicated parameters were in accordance with the Technical Specifications for the current operational Mode: -Equipment operating status; -Area, atmospheric and liquid radiation monitors; -Electrical system lineup; -Reactor operating parameters; and-Auxiliary equipment operating parameters.
No discrepancies were identified in this are I -
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(2) Safety Systems Walkdowns The' inspectors conducted walkdowns of the following safety systems to verify lineups were in accordance with license requirements for system operability.
-Spent Fuel Cooling System; and, -Decay Heat Removal System.
No disc _repancies were identified in this area.
(3) Shift Staffing - The inspectors verified by numerous checks that operating shift staffing was in accordance with Technical Speci-fication requirements. In addition, the inspectors observed shift turnovers of major work groups, radiation protection personnel, refueling group personnel, and plant operations personnel, on different occasions to verify that information concerning plant status, operational problems, and other pertinent plant items were transmitted to oncoming shifts. No discrepancies were identified in this area.
(4) Plant housekeeping conditions - Storage of material and components and cleanliness conditions of various areas throughout - the facility were observed to determine whether safety and/or fire hazards exist. No discrepancies were identified in this area.
(5) ' Radiation areas - Radiation control areas (RCA's) were observed to verify proper identification and implementation.
These obser-vations included selected licensee-conducted surveys, review of step-off pad conditions, disposal of contaminated clothing, and area posting.
Area postings were independently verified for accuracy through the use of the inspector's own monitoring instrument. The inspectors also reiriewed selected radiation work permits and observed personnel use of protective clothing, respirators, and personnel monitoring devices to assure that the licensee's radiation monitoring policies were being followed. The
following item was identified: Ouring a tour of the turbine building on October 15, at approxi-mately 1415, the inspector noted two individuals performing maintenance on the main feedwater pumps which has been designated - as a contaminated area.
Review of the Standing Radiation Work Permit (SRWR) indicated that clothing requirements would be specified by the chem / rad technician and that the clothing re-i.
quirements, cotton gloves and shoe covers, was specified on the contaminated area posting sign.
The inspectors, noting that these personnel were not wearing cotton gloves, immediately notified the chem / rad technician. the technician took immediate corrective action by stopping the work,
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removing the personnel from the area, re-instructing them in the requirements to wear the proper clothing, and checking the indi-viduals for contamination (they were not contaminated).
The main feedwater pumps are slightly contaminated with no loose contamination and fixed contamination ranging from 6000 counts per minutes (cpm) to 100,000 cpm with the average fixed contamination range about 10,000 cpm.
Failure to adhere to the clothing requirements as specified by the' chem / rad technician is contrary to the requirements of procedure RP-101, Radiation Protection Manual, and 10 CFR Part 20. This.is considered to be a violation.
Violation (302/81-21-03): Failure to adhere to the protective clothing requirements of procedure RP-101.
' The licensee's corrective actions in this event included the actions discussed previously and the reporting of this event as a fir;t offense in accordance with the enforcement policy of procedure RP-101. Subsequent examinations of personnel inhabiting contaminated areas by the inspectors indicate that the licensee's controls are effective and. therefore no additional response to this item is considered necessary.
(6) Radioactive waste controls - Selected liquid 'and gaseous radio-active releases were observed to verify that approved procedures ' were utilized, that appropriate release approvals were obtained, that required samples were taken, and that appropriate release control instrumentation was operable.
Portions of radioactive waste shipment operations were observed to insure that the licensee's controls were implemented. These included witnessing portions of a drumming operation and verification that labeling, surveys, and records were properly completed.
(7) Security Controls - Security controls were observed to verify that security barriers are intact, guard forces'are on duty and access to the protected area (PA) is controlled in accordance with the facility security. plan. Personnel within the PA were observed to insure proper display of badges and that personnel requiring escort were properly escorted. Personnel within vital areas were observed to insure proper authorization for the area. The fol-lowing item was identified: - On October 14 at 0840 the inspector observed two contractor employees with escort-required badges in the protected area. The inspector did not observe any escort with these employees and upon questioning the employees determined that their escort had left j these employees to perform work in an adjoining area, thus placing - _ _ - . _ __, _.
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himself out of sight of the escorted individuals.
The escort returned soon afterward and, after assuring that the escort was aware of his escort responsibilities, the inspector notified licensee management of the event. Failure to provide an escort at all times while in the protected area for individuals not-autho-rized to enter this area without an escort is contrary to the requirements of 10 CFR 73.55(d)(6) and the Crystal River Nuclear Plant Security Plan.
Item of noncompliance: Failure to provide an escort at all times of individuals not authorized to enter the protected area without an escort. (302/81-21-04).
A similar violation of 10 CFR.73.55(d)(6)-for failure to provide an escort at all times of individuals not authorized to enter the protected area without an escort was identified during the in-spection of October 4 through November 3,1980 (NRC Report 50-302/80-38) and therefore this - violation is considered to _ be recurrent.
(8) Operating _ Procedures - Operating Procedures (0P) use was observed to verify that: -approved procedures were being used; qualified personnel.were performing the operations; and,- -Technical Specification requirements were being followed.
The following procedures were observed: -0P-406, Spent Fuel Cooling System (Section performing fuel transfer canal filling operation); -0P-209, Plant Cooldown; -0P-303, Draining and Nitrogen Blanketing of the RC System; -0P-404, Decay Heat Removal System; and, -0P-413, Waste Drumming System.
As a result of these observations, the following items were identified: During observation of fuel transfer canal filling operations per OP-406, the inspector noted that the Fuel Transfer Canal (FTC) low level ' alarm was not energized on the control room annunication panel even though the actual level was considerably below the low level alarm setpoint. The inspector questioned the operators as to why this alarm was not in an alarm status. The operators were not sure why and generated a work request to investigate the
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problem. As a result, it was discovered that a-local "on, off" alarm switch in the reactor building was in the "off" position to . silence the local horn. A review of OP-406 indicated-t at this
- switch was not directed to be returned to the "on" position-after the fuel transfer canal had been filled.to the normal refueling level. ~In addition,.a check of the electrical schematic of the alarm revealed a possible design problem in that the "off" position of the local alarm switch will cutout the status of the alarm in the control room thus defeating the control room alarm.
These problems were discussed with the_ licensee.
The licensee said they would evaluate a modification to the alarm circuit to
prevent the local alarm switch from defeating 'the control room alarm status. In addition OP-406 would be changed to ensure that the alarm is properly placed in service during and subsequent to filling the fuel transfer canal.
Inspector Followup Item (302/81-21-05): Review evaluation of FTC modificition to prevent local switch from defeating control room alarm status.
Inspector Followup Item (302/81-21-06): Verify OP-406 revision to adequately place FTC level alarm in operation during and sub-sequent to. filling operations.
(9) Surveillance Testing - Surveillance testing was observed to verify that: . -approved procedures were being used; qualified personnel were conducting the tests; -testing was adequate to verify equipment operability; -calibrated equipment, as required, were utilized; and-Technical Specification requirements were being followed.
The following tests were observed: !' -SP-179, Containment Leakage. Tests Types B & C;
-SP-170, Pressurizer Level Instrument Calibration (portions of transmitter calibration) -SP-354, Emergency Diesel Fuel Oil Quality and Diesel Generator Monthly Test (Data review for "A" EDG and observation on "B" EDG); -SP-405, Core Flood System Valve Operation Demonstration (data review); -SP-603, Decay Heat System Check Valve Leak Testing (data review);
-SP-421, Reactivity Balance Calculation (independent verification); -SP-435, Valve Testing During Cold Shutdown (data review); -SP-901 and 902, 4.160 KV ES BUS "A" and "B" Undervoltage Trip - Test;
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--SP-200, Hydraulic Snubber Functional Testing; and, -SP-201, Inaccessible Hydraulic Snubber Visual Inspection (data review only).
As a result of these ' observations, the following item was identified: During observation of SP-200, Hydraulic Snubber. Functional Testing, the inspector noted that the pressure gages used on the ITT. Grinnel snubber tester have a 0-300 psi range whereas the highest-pressure required during snubber functional testing.is about 800 psi and the lowest pressure is 50 psi.
The pressure gages range versus required pressures during testing indicate that lower ranged gages would be more appropriate for the circum-stances.
This. issue was discussed with the licensee and the inspector's comments were acknowledged. The licensee states that lower ranged gages would be obtained for the snubber tester.
Inspector Followup Item (302/81-21-07): Verify installation of replacement gages on the ITT snubber tester.
(10) Maintenance. Activities - The inspector observed maintenance - activities to verify that: -correct equipment clearances were in effect; -Work Requests (W/R's), Radiation Work Permits (RWP's), and Fire Prevention Work Permits, as required, were issued and being , followed; -Quality Control personnel were available for inspection activities as required; and, -Technical Specification requirements were being observed.
The following maintenance activities were observed: -MP-115, RC Pump Inspection and Replacement (portions of seal refurbishment for RCP's A and C); -MP-130, Pipe Snubber Maintenance; -MP-109, OTSG Relief Valve Removal and Replacement; -MP-402, Maintenance on "Limitorque" Valve Controls (in support of Modification Approval Records (MAR) 81-6-7, Grease Relief.for Limitorque Valves and 81-6-6, Lubrication Change of Limitorque Valves); -MAR 81-10-18. Replacement of Agastat Relay on "A" EDG; and, -SP-605, EUG Inspection / Maintenance (both "A" and "B" EDG's).
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As a result of these observations, the following items were found: a.
During the inspectors' observation of maintenance _ on the Emergency Diesel Generators (EDG's) in accordance with-procedure SP-605, the inspector, questioned whether procedural step 6.2.3 requiring the checking and cleaning of the crankcase ejector and oil separator should be completed prior to EDG test running. This step was added to procedure SP-605 af ter the licensee had evidence of high' crankcase pressure (reference NRC Report 50-302/81-05, paragraph 8.b) during EDG testing it was determined that the vendor's recommended maintenance frequency for these items was not sufficient for EDG standby service. It appears that cleaning of these items is not necessary - unless increasing crankcase pressure is evident.
The licensee will revise SP-605 to clarify the intent of step 6.2.3.
Inspector Followup Item (302/81-21-08): Review the revision to procedure SP-605, step 6.2.3, to revise the checking / cleaning requirements on the-EDG crankcase ejector and oil separator.
b.
During a review of the maintenance. and subsequent testing performed on Emergency Diesel Generator (EDG) B, the ~ . ' inspector noted that the diesel tachometer was defective and that work request (W/R) 25507 was written to replace same.
This W/R was initiated on July 17, 1981, and a s' of October 23, 1981, the tachometer had not been replaced due to long lead-time procurement problems with the vendor.
It is expected that the tachometer will be received by November 17, 1981.
Inspector Follow up Item (302/81-21-09): Review licensee's actions to replace defective tachometer on EDG B.
(11) Pipe hangers seismic restraints -_ Several pipe hangers and seismic restraints (snubbers) on safety-related systems were observed to insure that fluid levels were adequate and no leakage was evident (where appropriate), that restraint settings were. appropriate, and that anchoring points were not binding. See section 3.7 of this report for.information about the snubber cracked bushing issue.
No additional problems were noted in this area.
(12) Refueling Activities - Refueling activities were cbserved and procedures were reviewed to verify that: -approved procedures were being used; _ _ _ _ -
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qualified personnel vere performing the activities; procedures were adequate to accomplish the required activities; -calibrated equipment [as required) were utilized; and, -Technical Specification requirements were being followed.
The following activities were observed: -FP-302, New Fuel Assembly Unloading, Inspection,. Storage and Container Reclosing (Section 12); -FP-409, Reactor Vessel Head Removal; -FP-203, Defueling and Refueling Operations; and, -SP-406, Refueling Operations Daily Data Requirements (independent verification).
As a result of these reviews, the following item was identified.
On October 15, the inspectors observed the lifting and placement of the reactor vessel head.- This activity is conducted in -- accordance with refueling procedure' FP-409, Reactor Vessel Closure Haad Removal.
During the head lift, the inspector noted that.
procedure FP-409 was available but did not notice personnel signing off procedural ' steps as the activity progressed.
On October 21, the inspector reviewed procedure FP-409 and noted that the step signoffs for this procedure were all signed off as completed by the same person on October-15 except step 7.8 which involved calibration of the polar crane-load cell. Step 7.8 was not signed off.
Procedure FP-409 encompasses considerable activity performance including removal of equipment hatch missle shields (step 8.6), uncoupling of axial power sha~ ping rod leadscrews,' (step 8.8), draining of inlet and outlet headers on the fueling canal (step 8.9), disconnecting and logging of all position indication cables and power cables (step 8.10), uncoupling and parking of shim safety rods (step 8.15), withdrawing from core of incore instrument assemblies (step 8.16), etc.
These activities were conducted over a period of time beginning about October 1 and extending through October 15, whereas the sign-off sheet indicates these activities as being completed on one day, October 15.
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addition, the signoff sheet did not indicate that step 7.8 was completed even though this step was a pre requisite for the head i lift. The inspector also noted that the cable disconnect / connect i sheets which are required to be used to log cable numbers and
cable locations to insure proper reconnection prior to the head " lift were not completed.
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Failure to complete the procedure sign-offs and cable disconnect / connect sheets as required by FP-409 is contrary to the require-ments of Technical Specification 6.8.11and is considered to be a violation.
Violation-(302/81-21-10): Failure to follow facility procedures during refueling activities.
A similar violation of Technical Specification 6.8.1 for failure to initial procedural steps as required by the maintenance procedure during the performance of maintenance was identified during the inspection period-of. June 27 through July 20,1981 (NRC Report.50-302/81-13) and therefore this violation is considered to be recurrent.
6.
Review of -Licensee Events Reports (LER's) and Nonconforming Operations Reports (NCOR's) a.
The inspector reviewed LER's to verify that: -the reports accurately describe the events; -the safety significance is as reported; -the report satisfies requirements with respect to information provided and timing of submittal; -corrective action is appropriate; and, -action has been taken.
, LER's 81-50, 81-57, 81-59, 81-60, and 81-61 were reviewed. As a result of this review, the following items were identified: 1.
LER's 81-50 and 81-60. stated as a part of the corrective actions that an evaluation is being conducted to determine if failed fuel is present. As discussed in NRC Inspection Report 50-302/81-15, paragraph 6.a.1, a test was run'on July 29, 1981 to provide data for this evaluation. On October 7, 1981, the licensee's Nuclear Support Services group completed their failed fuel analysis which.
is consistent with previous analysis results. This analysis was i compiled into a report that will be submitted to Nuclear Reactor Regulation (NRR) as justification for raising the Technical Specification 3.4.8 specific activity limit from 1.0 uci/gm.- The-inspector has reviewed this additional report and has no further questions on this item at this time.
(2) LER 81-59 reported the inoperability of a remote reactor coolant , (RC) temperature instrument (RC-4A-TI2). The licensee is planning to replace the indicater during the refueling outage.
Inspector Followup Item (302/81-21-11): Veri fy replacement of remote RC temperature instrument (RC-4A-7I2).
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(3) LER 81-61 reported the failure of Engineered Safety.7eature Actuation System (ESFAS) channel RB-3 to actuate during performance of a surveillance test. The failure was caused by a stuck Clark type PM SUK8-7-76 relay. The licensee has had similar failures of these relays and plans to modify the relays to prevent recurrence of these type events. This item is being tracked in accordance with Inspector Followup Item 50-302/81-05-12.
b.
The inspector reviewed NCOR's to verify the following: -compliance with Technical Specifications; -corrective actions as identified in the reports or during subsequent reviews have been accomplished or are being pursued for completion; -generic items are identified and reported as required by 10 CFR, part 21; and, -items are reported as required by the technical specifications.
The following NCOR's were reviewed: 81-94 81-322 81-340 81-145 81-323 81-341 81-149 81-325 81-349 81-164 81-328 81-351 81-171 81-329 81-356 81-174 81-330 81-361 81-179 81-332 81-363 81-234 81-333 81-364 81-261 81-334 81-365 81-271 81-335 81-367 81-309 81-336 81-368 81-315 81-337 81-371 81-316 81-338 81-376 81-317 81-339 81-378 81-379 No discrepancies were identified in this area.
7.
Visual Inspection of Reactor Vessel Internals During the refueling outage the licensee performed visual inspectior.s of the reactor vessel intenals using a TV camera. These inspections included the following items: -Approximately sixty thermal shield bolts; -The botton of the reactor vessel; -Fuel assembly hold-down springs; and, -Fuel assembly upper end fitting / " .- . . . . . ,
The thermal shield bolts and the bottom of the reactor vessel were inspected to determine if any thermal shield bolt failures were occurring or had occurred as has been indentified at the Oconee Unit.1 reactor. The inspec-tion did not indicate any bolt failures.
The fuel assembly hold-down springs were examined to determine if. spring cracking had occurred. Hold-down spring cracking was originally identified at the Davis-Besse Nuclear Plant in 1980 and one cracked spring was identified at this facility during the last refuel outage (Reference NRC.
~ Report 50-302/80-23). The inspection did not. identify any hold-down spring cracking.
Fuel assembly upper end fittings were examined to determine if the loose part found in the reactor coolant system on March 26 (Reference: NRC Report 50-302/81-05, paragraph 9.a) was an upper guide tube nut as postulated by the reactor vendor (B&W). On October 26, the licensee noted a missing upper guide tube nut on fuel assembly NJ 0204.
The end of the guide tube had sheared off.with the attached nut.
The reactor vendor performed a safety evaluation to assure continued operation with the missing -guide nut was acceptable until the'present refuel outage, however the safety evaluation did not include continued cperation through the next operating cycle. The licensee intends to use fuel assembly NJ 0204 in the next operating cycle.
The inspector. questioned licensee representatives about the use of this fuel assembly and stated that an additional safety evaluation should be performed if continued use of the assembly is planned. The licensee ' acknowledged the inspector's comments and is investigating what additional actions will be necessary to allow continued usage of this fuel assembly.
Inspector Followup Item (302/81-21-12): Review the licensee's actions with respect to continued use of the fuel assembly with the missing upper guide tube nut.
The inspectors reviewed the video tapes for the thermal shield bolts, reactor vessel bottom, and the upper guide tube missing nut.' With the exception of the inspector followup item the inspectors have no further questions at this time.
9.
Nonroutine Events Radioactive Spill from Radioactive Waste Liner On October 2, 1981, a Chem Nuclear Inc. radioactive waste liner with solidified radioactive waste was moved from the auxiliary building (AB) to the storage area outside of the AB and then capped. Temperature recordings of the liner indicated the liner temperature was a constant 159 F and had reached dynamic equilibrium in its vented condition. On October 3, at
f.
~ .
. . .. , .
approximately 2300, the lid blew off the liner, spilling approximately 10 gallons of radioactive water on the ground inside the Radiation Control Area (RCA).
Decontamination of the ground was performed successfully.
No release limits were exceeded.
An investigation by Chem-Nuclear Inc. into this issue indicated the major contributing factors to be: 1.
The liner was capped prior to completing the exothermic reaction and, 2.
a possible lower than optimum waste-to-additive ration was used resulting in the formation of local hot spots in the liner.
The combination of these two factors caused the temperature inside the liner to increase to a temperature greater than 175 F forcing the liner lid off and releasing the radioactive water to the ground. As a result of this event, the following corrective actions will be implemented to preclude future events of this nature: 1.
OP-413 will be modified to provide additional guidance specific parameters for use prior to capping a solidified liner. This modifi-cation will include maximum temperatures and repeatable cool-down rates for use by the operator.
2.
OP-413 modifications will be made to provide for increased mixing time following the final cement addition.
This mixing will be at minimum speed so that the vortex formation in the mix is prevented and additional air will not be pumped into the cement.
In addition, the
increased mixing time will further work the cement causing a reduced viscosity and aid the escape of residual trapped air.
. 3.
A control step will be added to OP-413 to requ ' verification of mixture viscosities using the installed hydraulic oil pressure guage.
Hydraulic pressure will be used to confirm correct cement addition and to insure that the system will not seize during the final mixing stage.
' 4.
An engineering design review of the mixing assembly will be conducted to ensure there is adequate mixing even under "high viscosity"
conditions.
The inspectors have reviewed this issue with the licensee and Chem Nuclear ,
Inc. and have determined that the corrective actions being taken are being taken are sufficient to preclude a recurrence of this event.
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