IR 05000302/1981023

IE Insp Rept 50-302/81-23 on 811024-1202.Noncompliance Noted:Failure to Follow Procedures During Plant Operations, Failure to Follow Radiation Protection Procedures for Contamination Control & Failure to Calibr Instrumentation

ML20054K366
Person / Time
Site:	Crystal River
Issue date:	01/12/1982
From:	Brownlee V, Beverly Smith, Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20054K339	List: IR 05000302/1981023 IR 05000302/1982023 ML20054K338 ML20054K343 ML20054K350 ML20054K354 ML20054K358 ML20054K359 ML20054K360 ML20054K362 ML20054K364
References
50-302-81-23, NUDOCS 8207010483
Download: ML20054K366 (19)

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o UNITED STATES

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!"- n NUCLEAR REGULATORY COMMISSION

$ ec REGION ll

[ 101 MARIETTA ST., N.W., SUITE 3100 ob ATLANTA, GEORGIA 30303

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Report No. 50-302-81/23 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 N OPR-72 Inspection at n ar Cr7stak ver, Florida Inspectors: b 1 Arm 1 e 'f/UA f/ / /2 [ 2-T. F. S,tjetk Ddte S'igned

& c ,enior h6& Resident (Inspector B. W. . 1)h, Resi nt Inspectori i/s/r'

Date Signed Approved by: [/D V. 'm W L /2- L Br,gIwnlee, Section Chief, Division of Date Sidned Resident and Reactor Project Inspection SUMf1ARY Inspection on October 24 - December 2,1981 l

Areas Inspected This routine inspection by the resident inspectors covered plant operations, security, radiological controls, Licensee Event Report (LER's) and Nonconfonning Operations Report (NCOR's), nonroutine events, refueling activities, environmental protection, and the licensee's action on previous inspection item Numerous facility tours were conducted and facility operations observed. Some of these tours and observations were conducted on back shifts. The inspection involved 249 hours0.00288 days <br />0.0692 hours <br />4.117063e-4 weeks <br />9.47445e-5 months <br /> onsite by two resident inspector Results Three violations were identified (Failure to follow procedures during plant operations, paragraph 5.a and 3.a; Failure to follow radiation protection procedures for contamination control, paragraph 5.b(4); Failure to calibrate reactor outlet temperature instrumentation, paragraph 5.b(8). One deviation was identified (Total reactor outlet temperature instrumentation error is in excess ofFSARanalyzederror, paragraph 5.b(8)).

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DETAILS Persons Contacted Licensee Employees P. Baynard, llanager, fluclear Support Services

• Boldt, Technical Services Superintendent

• • • C. Brown, fluclear Compliance Supervisor D. Brock, Nuclear liaintenance Specialist J. Bufe, Compliance Auditor 11. Collins, Reactor Specialist

• J. Cooper, QA/QC Compliance flanager W. Cross, Licensing lianager

• T. Fay, Plant Engineer V. Hernandez, Compliance Auditor

• • S. Johnson, liaintenance Staff Engineer

• B. Komara, Compliance Auditor T. Lutkehaus, Technical Assistant to the Nuclear Plant flanager

• • li.11 ann, Compliance Auditor J. 11asada, Plant Engineer

• • P. ficKee, Operations Superintendent G. Perkins, Health Physics Supervisor

• • • D. Poole,fluclear Plant flanager

• H. Reeder, Shift Technical Advisor

• • G. Ruszala, Chemical / Radiation Protection Manager

• • D. Smith, Security and Special Projects Superintendent

• • • K. Lancaster, Senior Quality Auditor J. Lander, liaintenance Superintendent G. Williams, QA/QC Supervisor

• K. Wilson, Licensing Specialist Other licensee employees contacted included office, operations, engineering, maintenance, chem / rad, and corporate personne U. S. Nuclear Regulatory Commission

• • V. Brownlee, Chief, Reactor Projects Section 2

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State of Florida C. Ness, Health Physicist, Department of Health and Rehabilitative Services

• Present at 11/11/81 Exit interview

• • Present at 12/2/81 Exit interview

• • • Present at both Exit interviews Exit Interview The inspectors met with licensee representatives (denoted in paragraph 1) on numerous occasions during and at the conclusion of the inspection on

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December 2,198 During this meeting the inspectors summarized the scope and findings of the inspection as they are detailed in this repor During this meeting the violations and inspector followup items were discusse As a result of the reactor coolant system overflows that are discussed in

, paragraph 8 of this report, the licensee presented additional long term

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! corrective actions that are being taken to minimize recurrence of these events. These actions include:

-A procedural review is in progress to detemine if there are other procedures or conflicts between procedures that could result in system overflows or spill A procedural review is in progress to determine the feasibility of

developing valve check lists that are operational mode dependen This review is expected to be completed by March,198 The licensee is initiating a new computer system that will enable

operators to assess system status with respect to outstanding maintenance

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activities and equipment cle4rarces. This system is in partial operation now and is expected to be fully operational by the end of January,198 The inspectors acknowledged the licensee's presentation and stated that these activities would be reviewed in subsequent inspections.

l Licensee Action on Previous Inspection Findings (Closed) Inspector Followup Item (302/81-02-16): The licensee's review of Engineered Safety Features Actuation System (ESFAS) relay failures has identified the Clark type Pfl relay as the cause of these failures. These failures were subsequently discussed in NRC Inspection Report 50-302/81-05, paragraph 8.b (3). The licensee is modifying these relays to prevent sticking and this activity is being tracked in accordance with Inspector Followup Item (302/81-05-12). The licensee has also identified failure problems with solid-state triac relays manufactured by Hamlin. Since Hamlin no longer manufactures these relays, the licensee has had to replace these relays with a relay manufactured by Crydon. The licensee's activities in replacing Hamlin relays with Crydon relays has identified additional problems and these problems are identified in paragraph 6.b of this repor Items affecting both solid-state relays and Clark Pil relays are being tracked in accordance with items (302/81-05-12) and (302/81-23-08). This

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item is closed for record purpos (Closed) Unresolved Item (302/79-01-02): The licensee's leakrate procedure has been reviewed and independent calculations performed by the resident inspectors. In addition a reactor coolant system leakrate verification was conducted utilizing an independent calculation method (See NRC Inspection Report 50-302/81-15, paragraph 10) and the results of this verification compared favorably with the results of SP-31 , _ - - _ - __ . _ _ . .

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(Closed) Inspector Followup Item (302/81-15-11): Quality Operating Manual procedure Q0P 7.0 has been revised as revision 10 on October 15, 1981 to include a new section 4.1.3 This new section rt'uires the signing and dating of a QCI at the time the material is picked up or delivered. Action on this item is complet (Closed) Violation (302/81-13-01): The licensee has completed training of all necessary personnel by the November 15, 1981 date committed to in the revised response dated October 13, 1981. The inspector has reviewed records and interviewed personnel to determine the extent of this training and considers the licensee's actions to be complete (0 pen) Violation (302/81-19-05): The licensee has revised surveillance procedure SP-110, Reactor Protective System Functional Testing, to functionally exercise the reactor building pressure switches on a monthly basis. The resident inspector reviewed the revised procedure and observed the new testing. The licensee's response does not address whether a review of other systems was conducted to verify that all required channel sensors are exercised. The licensee will conduct this review and issue a revised response. This issue remains open pending completion of this review and issuance of a revised respons (Closed) Violation (302/81-11-07): The licensee has fomed three person subcommittees to perfom the safety evaluations required by 10CFR 50.5 These subcommittees then report their findings to the full Nuclear General Review Committee (NGRC) for review. Three person subcommittees are also used to review plant operating reports, operating abnormalities or deviations, deficiencies in design or operation, and corporate audit resal ts. The inspector reviewed NGRC meeting minutes for meetings conducted during the period of July thorough November 11, 1981 and the summary reports of subcommittee reviews. The inspector also attended the NGRC meeting conducted on November 12, 1981 to observe committee activities. Based upon these reviews and observations, the licensee's corrective actions are considered to be complet (Closed) Inspector Followup Item (302/81-21-11): The remote RC temperature instrument RC-4A-T12 has been replaced. The inspectors have no further question of this ite (Closed) Inspection Followup Item (302/81-21-06): The inspectors verified OP-406 revision adequately places the fuel transfer canal level alam in service during filling operation The inspectors have no futher questions on this ite (Closed) Inspection Followup Item (302/81-21-07): The inspectors verified the installation of replacement gages of a smaller range on the snubber tester. The inspectors have no further questions on this ite (Closed) Inspector Followup Item (302/81-21-08): The inspectors verified SP-605 revision to classify the checking / cleaning requirements on the EDG

crankcase ejector and oil separator. The inspectors have no further question on this ite (Closed) Inspector Followup Item (302/81-15-06): The licensee has completed replacement of the aluminum bushings with stainless steel bushings in all of the inaccessible snubbers. The resident inspectors have followed these replacement activities by observing selected snubber rebuilding and testing and reviewing associated documentation. Due to additional functional test failures caused by unsatisfactory bleed rates, the licensee has committed to rebuild additional snubbers located in accessible areas. These committments are discussed in f4RC Inspection Report 50-302/81-29. The licensee's activities with respect to the inaccessible snubbers in considered to be comple te.

4. Unresolved Items There were no unresolved items identified during this report period.

5. Review of Plant Operations The plant continued in flode VI, refueling operations, and flode V, cold shutdown, for the majority of this inspection perio On December 2, 1981, the Plant entered flode IV, hot shutdown operations, in preparation for performing unit startup surveillance activitie Shift Logs and Facility Records The inspectors reviewed the records listed below and discussed various entries with operations personnel to verify compliance with TS and the Licensee's administrative procedures:

-Shift Supervisor's Log;

-Reactor Operator's Log;

-Equipment Out-of-Service Log;

-Shift Relief Checklist;

-Control Center Status Board;

-Short Term Instructions;

-Auxiliary Building Operators Log;

-Chemistry / Radiation Log;

-Daily Shutdown Surveillance Log;

-Refueling Supervisor's Log;

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-Work Requests;

-!!odification Approval Records; and,

-Outage Coordinator's Lo In addition to these record reviews, the inspectors independently verified selected clearance order tagout As a result of these reviews, the following violation was found:

During a review of logs on October 29, the inspector noted an eqtry in the Outage Coordinator's log discussing steam flashing in the reactor coolant system hot legs that occurred during the plant cooldown on October 1,1981. The inspector cross-checked this log entry with the Nuclear Shift Supervisor's log and the Nuclear Operator's log and noted that no such entry had been made in either of these logs. This finding was discussed with operations personnel and the licensed operators on a duty when the event occured. These personnel stated that a log entry i

was not made because they did not consider hot leg steam flashing and the resultant steam bubble formation in the hot leg to be an unusual occurrence or trend. The inspector disagreed with the licensee's conclusions and stated that voiding in the reactor coolant system of a pressurized water reactor in any location other than the pressurizer is an unusual occurrenc Procedure AI-500, Conduct of Operations, requires adherence to the Implementation llanual (OSIri). The OSIfi, in Section III.E.2.A and B, requires log entries in the lluclear Shift Supervisor's Log and the Nuclear Operator's Log for any unusual occurrences or unusual trends or conditions observed. Failure to adhere to these requirements is contrary to the requirements of Technical Specification 6.8.1 and is a violatio Violation (302/81-23-01): Failure to adhere to administrative and operational procedures.

l The licensee took immediate corrective action to prevent recurrence of this event. These actions included:

-Discussions with the operators involved and with other operating shifts to insure that all personnel understand that RCS flashing is an unusual occurrence;

-Revised the OSIll to clarify the logging instructions; and,

-Revised procedure OP-209, Plant Cooldown, to provide additional steps to minimize steam flashing in the RC The inspectors reviewed these corrective actions and have detennined that no additional response to this item is require . .

b. Facility Tours and Observations Throughout 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 The facility tours and observations encompassed the following areas:

-Security Perimeter Fence;

-Turbine Building;

-Control Room;

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-Emergency Diesel Generator Rooms;

-Auxiliary Building;

-Intermediate Building;

-Reactor Building;

-Battery Rooms; and,

-Electrical Switchgear Room During these tours, the following observations were made:

(1) lionitoring instrumentation - The following instrumentation was

, observed to verify that indicated parameters were in accordance i with the Technical Specifications for the current operational flode:

i -Equipment operating status;

-Area, atmospheric and liquid radiation monitors; l -Electrical system lineup;

, -Reactor operating parameters; and, I

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-Auxiliary equipment operating parameters.

No discrepancies were identified in this area.

l (2) Shift Staffing - The inspectors verified by numerous checks that operating shift staffing was in accordance with Technical i Specification requirements. In addition, the inspectors observed

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shif t 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 are (3) 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 are (4) Radiation areas - Radiation control areas (RCA's) were observed to verify proper identification and implementation. These observ-ations included selected licensee-conducted surveys, review of step-off pad conditions, disposal of contaminated clothing, and area posting. Area posting was independently verified for occuracy through the use of the inspector's own monitoring instrument. The inspectors also reviewed selected radiation work permits and observed personnel use of protective clothing, respirators, and personnel monitoring devices to assure that the licensee's radition monitoring policies were being followe The following item was identified:

During a tour of the auxiliary building (AB) on December 1,1981, the inspector observed a person in a contaminated area wearing plastic shoe covers as the only means of radioactive contamination control. In addition, this worker was observed leaving the contaminated area partially carrying and partially dragging a bundle of rope without observing required contamination control measures (rope not bagged and/or gloves worn). The inspector questioned the worker as to the origin of the rope and lack of contamination controls. The worker

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indicated that the rope was probably contaminated and that he didn't think to bag it since he was taking it to another contaminated are The inspector had the individual stand fast and then notified a ChemRad technician of the situatio The ChemRad technician assumed control of the situation and determined that the individual had contaminated his hands to level of approximately 1000 disintergrations per minute (DPil) . A further check into the incident indicated the individual had entered the contaminated area without consulting ChanRad personnel for the clothing requirements as required by the standing radiation work

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permit (SRWP), and without the clothing that would have been specified for the area.

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Chemistry and Radiation Protection Procedure, RP-101, section 4. requires: An individual entering a contaminated area to wear proper clothing i

and devices necessary for contamination control, l

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8 Chem / Rad personnel to be contacted to specify the clothing requirements for entering a contaminated area covered by (SRWP.)

In addition, Section 4.4 of RP-101 requires work to be conducted in a manner consistent with maintaining a minimum of contamination sprea Failure to adhere to the requirements of RP-101 is contrary to the requirements of Technical Specification 6.11 and is a violatio Violation (302/81-23-02): Failure to adhere to the requirements of RP-101 for contamination contro (5) Radioactive waste controls - Selected liquid and gaseous rddioactive releases were observed to verify that approved procedures were utilized, th;c 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. flo problems were identified in this are (6) 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 pla 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. tio probelms were identified in this are (7) Operating Procedures - Operating Procedures (UP) use was observed to verify that:

-approved procedures were being used;

-qualified personnel were performing the operations; and,

-Technical Specification requirements were being followe The following procedures were observed:

-0P-202, Plant Heatup;

-0P-303, Draining and flitrogen Blanketing of the RC System;

-0P-404, Decay Heat Removal System;

-0P-608, OTSG Secondary Fill, Drain, and Layup; and-0P-209, Plant Cooldow fio problems were identified during these observation . .

(8) 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, was utilized; and,

-Technical Specification requirements were being followed.

I The following tests were observed:

-SP-455, Functional Test of Vital Bus Redundant Transformers (completed procedure review);

-SP-112, Calibration of the Reactor Protection System, (reactor coolant pump power monitor (RCCPil) data review and procedure review);

-SP-532, Reactor Building (RB) and Auxiliary Building (AB) Fuel Handling Bridge Electrical Interlock Checks (RB main bridge only);

-SP-220, Source Range Functional Test During Refueling Operations (data review);

-PT-132, Leak Testing 0TSG "A" with Helium or Nitrogen (completed procedure review);

-SP-415 Teansfer From Preferred Off-Site Power Source to Alternate Off-Site Power Source and Return to Preferred Off-Site Source (observed entire procedure);

-SP-179, Containment Leakage Tests - Type B and C (AHV-1C and ID observed);

-SP-416, Emergency Feedwater Automatic Actuation (turbine driven pump only);and,

-SP-422, Reactor Coolant System Heatup Surveillanc As a result of these observations, the following items were identified:

During a review of SP-112, Calibration of the Reactor Protective System, (RPS), for the purpose of verifying compliance with Technical Specifications, the inspector discovered that the RPS hot leg resistance temperature detectors (RTD's) were not included in the calibration procedure for the reactor outlet temperature channel. This issue was discussed the licensee and the inspector's findings were acknowledged.

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l Technical Specification 4.3.1 requires the RPS reactor outlet channel to be calibrated at least once every 18 months. Technical Specifi-cation 1.9 requires a channel calibration to encompass the entire channel including the senso Contrary to Technical Specifications, the RPS reactor outlet RTD's were not included in the reactor outlet temperature channel calibratio Violation (302/81-23-03): Failure to include the reactor outlet RTD's in RPS outlet temperature channel calibratio A further review of SP-112 indicated that the existing reactor outlet channel calibration allows acceptance of a total error of + 3.065* This total includes a sensor error of + 0.065 F, a process error of +

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2.5*F, and a bistable error of +0.5*F, in addition, the + 3.065*F does not include the readability error of the signal converter meter used to calibrate the process equipment. This issue was discussed with the licensee and the inspector's comments were acknowledge Final Safety Analysis Report (FSAR) section 7.1.2.2.3 specifies a worst case assumption for the total error from sensor, processing equipment and trip bistable of the reactor outlet trip channel to be +1 Contrary to the above, allowable total channel error of the reactor outlet temperature trip channel exceeds the + 1*F worst case assumptions of the FSA Deviation (302/81-23-04): Failure to meet the worst case assumptions for allowable reactor outlet temperature trip channel total error of +-

1"F, as required by the FSA (9) flaintenance 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 Preventive 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 observe The following maintenance activities were observed:

-Pfl-145, Reactor Coolant Pump Motor Disassembly and Inspection (work package review); i l

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-11P-150, Disassembly and Reassembly of fluclear Service Seawater and Decay Heat Seawater Pumps (work package review of RWP-1A work);

-ftP-119, Relief Valve Removal, Repair and Reinstallation (D0V-33 work);

-!!P-130, Pipe Snubber liaintenance (testing, and rebuilding of numerous hydraulic snubbers);

-f1P-123, Disassembly and Reassembly of fluclear Services Closed Cycle Cooling Water Pumps (coupling alignment of SWP-1A);

-Pft-133, Equipment Lubrication Procedure (oil addition on SWP-1A);

-itP-166, Reactor Coolant Pump Seal Package Refurbishment and Testing (testing of RCP-3D seal);

-MP-147, Adjustment of Containment Purge Valves (AliV-1D seat adjustment work package review);

-itAR-79-10-86, Installation of Safety Grade Anticipatory Reactor Trip (work package review and observation of wiring modifications in RPS channel "B");

-f1AR 81-3-87, Reactor Building Purge Duct Supply Heater (review of work package);

-11AR 78-8-15, Power Level Upgraue Isolation System (portions of reactor coolant pump power monitor (RCPPM) installation);

-!!AR 78-8-159 Installation of Testing of RCPPM (time response testing observation;) and,

-itAR 81-5-28, CAV-1, 3, 4, 5 and 126 Relocation (review of work package and observation of CAV-1, 3, 4 and 5 relocation).

fio problems were identified in this are (10) Refueling Activiites - Refueling activities were observed and procedures were reviewed to verify that:

-approved procedures were being used;

-qualified personnel were perfoming the activities;

-procedures were adequate to accomplish the required activities;

-calibrated equipment (as required) were utilized; and,

-Technical Specification requirements were being followe _ _ _ _ _ - - _ _-

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The following activities were observed:

-FP-203, Defueling and Refueling Operations;

-SP-406, Refueling Operations Daily Data Requirements; and-SP-532, Reactor Building Fuel Handling Bridge Electrical Interlock Checks (main bridge only).

Flo problems were identified in this are (11) Pipe hangers seismic restraints - 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 bindin The inspector was notified of possible snubber installation deficiencies associated with the major refurbishment job being performed on all of the' reactor building (RB) hydraulic snubber The specific deficiencies cited were associated with possible missing adapter bushings on itSH-164 and 165 hydraulic snubber Based on this information, the inspector verified the installation of 16 snubbers that were rebuilt and reinstalled in the RB, including itSH-164 and 16 flo missing adapter bushings were found. The licensee was also notified of the alleged deficiencies in the snubber installation The licensee's inspection of the two snubbers revealed no missing adapter bushings but did indicate a missing spacer on f1SH-165. To assure an adequate level of confidence in the mechanic and the inspector associated with the f1SH-165 installation, the licensee

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reinspected 10 additional snubber installations in which these

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individuals participate Flo discrepancies were noted. Based on the inspector's review and the licensee's action to this issue,

, the allegation of missing adapter bushings is considered

unsubstantiated and the inspectors have no further questions on this issue at this time.

Review of Licensee Event Reports and ilonconforming Operations Reports The inspector reviewed Licensee Event Reports (LERs) to verify that

-The reports accurately describe the events;

, -The safety significance is as reported;

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-The report satisfies requirements with respect to information provided and timing of submittal;

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-Corrective action is appropriate; and,

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-Action has been take Ler's 81-63, 81-68, 81-69, 81-71, and 81-72 were reviewed. This review identified the following items:

(1) LER 81-63 reported that various hydraulic snubbers were discovered to be inoperable during surveillance testing. The resident and regional inspectors have followed the licensee's snubber rebuilding activities including observing selected snubber rebailding and testing and reviewing associated documentation. As discussed in paragraph 3 of this report in inspector followup item (302/81-15-06), the licensee has completed rebuilding snubbers located in inaccessible areas and has committed to rebuild snubbers in accessible areas. These canmitments are discussed in NRC Inspection Report 50-302/81-29. Two items remain to be completed, rebuilding of accessible area snubbers and completion of an engineering study to evaluate snubber failure Inspector Followup Item (302/81-23-05): Review accessible snubber rebuilding activities and the results of the snubber failure engineering evaluatio (2) LER 81-68 reported an omission in the accident analysis concerning the High Energy Line Break Outside Containment (HELG0C) due to a change in the operation of the turbine driven emergency feedwater pump (EFP-2). An engineering design evaluation is in progress to determine what modifications will be necessary to ensure HELB0C protection for these steam line '

Inspector Followup Item (302/81-23-06): Review the engineering evaluation and modification progress for the EFP-2 HELB0C concer (3) LER 81-71 reported an error in the accident analysis for the steam generator tube rupture accident. The licensee has commenced additional analysis to verify that the release limits of 10 CFR Part 100 would not be exceede The licensee expects to complete this analysis by April, 198 Inspector Followup Item (302/81-23-07): Review the revised accident analysis for the steam generator tube rupture acciden b. The inspector reviewed NCOR's to verify the following:

-Compliance withthe Technical Specifications;

-Corrective actions as identified in the reports or during subsequent .

reviews have been accomplished or are being pursued for completion; l-Generic items are identified and reported as required by 10 CFR Part 21; and, l

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l-Items are reported as required by the Technical Specification r 81-326 81-359 81-393 81-414 81-437 81-327 81-360 31-394 81-416 81-457 81-342 81-362 81-395 81-417 81-459 81-343 81-366 81-397 81-419 81-458 81-344 81-369 81-420 81-420 81-464 81-345 81-370 81-400 81-421 81-346 81-372 81-404 81-422 81-347 81-373 81-402F 81-427 81-348 81-374 81-405 81-428 81-350 81-375 81-406 81-429 81-352 81-377 81-407 81-432 81-353 81-385 81-409 81-431 81-354 81-386 81-410 81-433 81-355 81-387 81-411 81-434 81-357 81-388 81-412F 81-435 As the result of this review, the following item was identified:

NCOR 81-409 reported the failure of a bypass relay to de-energize during performance of surveillance test SP-132, Engineered Safeguards Channel Calibration. The failure of the relay to de-energize was traced to an induced voltage being generated in the control circuitry wire causing a Crydon solid-state triac relay to fail to turn off. The licensee has been replacing Hamlin solid-state triac relays with Crydon solid-state triac relays due to the unavilability of the Hamlin relay The licensee has identified additional problems with the Crydon relays (involving failure to pass receipt inspection and testing) and has presently stopped installation of these relays. The licensee is re-evaluating the use of the Crydon relays based upon these recent finding Inspector Followup Item (302/81-23-08): Review the licensee's evaluation of Crydon solid-state triac relays in ESFA . Environmental Protection On November 16, the inspector accompanied a health physicist from Florida's Department of Health and Rehabilitative Services (DHRS) during the weekly tour to check and collect samples fran the six sampling stations located around the Crystal River plant. The following sampling stations were checked:

-(C-26) located near the Florida Power Corporation (FPC) substation in Beverly Hills;

-(C-07) located near the water tower in Crystal River;

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-(C-46) located on FPC property near the fossil Unit No. 4;

-(C-04) located at the hydrodam near Yankeetown;

-(C-18) located near the Yankeetown water tower; and,

-(C-40) located on FPC property near the quarr The inspector discussed the sampling collection techniques with the health physicist and reviewed the sampling documentatio No inadequacies were identifie . Non-Routine Events Reactor Coolant System (RCS) Overflow From Borated Water Storage Tank On November 6,1981, at 2013, with the plant in 11 ode 6 (Refueling)

operations, approximately 1650 gallons of reactor coolant system (RCS)

water overflowed the partially drained RCS and flowed into the reactor building (RB). The water was contained in the RB with most of the water flowing into the RB sump. There were no releases or personnel contamination as a result of this even The cause of this overflow was failure of the operator to follow operations procedure OP-404, Decay Heat Removal System, during switching from the "B" Decay Heat train to the "A" decay heut trai Procedure OP-404 requires the closing of the borated water tank (BWST)

suction valve DHV-34 prior to starting the "A" decay heat train. Since the operator did not follow the procedure, DHV-34 was left open when the "A" train was started. This allowed water from the BWST to be pumped into the RCS thus raising the level in the partially drained system. Though the reactor vessel head had been placed upon the vessel, the head was unbolte the increasing level caused water to flow out around the reactor vessel flange, the control rod drive nozzles, the incore closure flanges and the once through steam generator (OTSG) upper manway When the operators received a " Low BWST Level" alarm, they immediately shut DHV-34, thus stopping the overflow. The excess water was then pumped back to the BWS Failure to follow procedures is contrary to the requirements of Technical Specification 6.8.1 and is considered to be another example of the violation identified in paragraph 5.a of this repor The licensee took immediate corrective action to prevent recurrence of this event. This corrective action included:

-Revised surveillance procedure SP-320, Operability of Boron Injection Sources and Pumps, to make the return to operation valve lineup mode

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dependent therefore assuring the BUST suction valves DHV-34 and DHV-35 ',

are shut following completion of the surveillance procedure while in '

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(The leaving of DHV-34 open following completion of

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11 odes 5 and SP-320 contributed to the overflow event); and, - 7-Discussion with the operator involved and with other operating shifts to emphasize the importance of following procedure ,

The inspectors reviewed the corrective action and have# determined that no additional response to this item is require ,

- l, b. 11ake-up System (fiUS) Overflow /

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/ I On flovember 15, 1981, at approximately 1930, with the plant in' hode 5 ' -

(Cold Shutdown) operations, approximately 100 gallons of reacts coolant system (RCS) make-up water flowed from the flus to the auxiliary'

building (AB) via an open drain valve (t1UV-339). 'The watcr.was '

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contained in the AB with most of the water flowing into the AB sum There were no releases or personnel contamination aso ' f , result of'this s

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The 11US had baen " tagged out" in accordance with a system clearance (10-365) to .tllow maintenance on the system. .Following completion of maintenance, clearance 10-365 was released thereby allowing the 110S to be refilled. Though the operator released the clearance in accordance with procedures, the operator failed to check the clearance log to insure that there were no additional outstanding clearance.t on the MU An additional clearance (10-310) was still outstanding on de i1VS and left drain valve i1VV-339 open. Since the licensee's method of operation includes a walk-down of a system when it is returned to

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service for refilling, the open drain valve was quickly identified and

! shut thus stopping the flo The licensee took immediate corrective action to prevent recurrence of-this event. This corrective action included: /' ,

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-Issued a memorandum to all Nuclear Shift Supervisors (NSS) >dascribing this event and the "practive actions that need to be implemente '

The fiSS were dirr to to review this memorandum with their shift; s

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personnel; aM-Revised the herot: _n Section Implementation fianual (OSIri) as revision 6 to require operations to review all outstanding clearances on a system prior to releasing the clearanc The inspectors reviewed these corrective actions and have -no further questions on this item at this tim c. Reactor Coolant System (RCS) Overflow Frc.1 Spent Fuel Cooling' System

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On flovember 17, 1981, at approximately 1230, approximately 3000 gallons of spent fuel pool (SFP) cooling water was pumped into the RCS via the

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. 17 9 decay heat removal (OHR) syste This water caused the RCS, which was partially drained and in fiode 5 (Cold Shutdown) at the time, to fill the reactor. vessel and dump water into the drained coolant loops which then ran out~of.the open loop drain valves to the reactor building (RB)

sump. The amount of water flowing from the open drain valves was in excess,of'the RB drain capacity therefore allowing some water to flow onto the RB floor. There were no releases or personnel contamination

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as a result of this even The cause of this event was an open valve (SFV-54) which cross

, cennected the decay heat removal system with the spent fuel cooling system. The, licensee was in progress of lining up the DHR and SFP systems to allow use of the SFP cooling purification system to purify

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the RCS water. The proper valve lineups were conducted with operators verifying correct valve positions by initialling valve lineup sheet Some time during the completion of the valve lineup (which shut SFV-54)

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and the commencing of RCS purification via the SFP system, valve SFV-54 was opene <

The licensee investigated this event and could not identify a reason for the reopening of valve SFV-54. The vtIve lineup was completed at about 0700 and the SFP purification flow started about 112 The inspector's review of this event included a review of associated documentation, interviews of on-duty operators, and a walk through of

, the valve lineup procedure for SFV-54 with the operator involved. The inspectors were also unable to determine cause for the even To prevent recurrence the licensee has taken the following actions:

-SFV-54 is a manually operated valve utilizing a chain operator. The liensee has locked all chain operated valves for systems that contain radioactive fluids or are required for reactor safety; and,

-Issued Short Term Instruction (STI) 81-92 directing operators to perform double valve verification on all critical systems for the remainder of the outag The inspectors have reviewed the licensee's actions including a walk

through the verify all specified chain operated valves were locked and have no further questions on this item at this tim d. Inadvertant Engineered Safeguards (ES) Actuation

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On November 29, 1981, at approximately 0715 a spurious ES actuation occurred. The plant was in flode V operations (cold shutdown) with "A"

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decay heat train in servic "A" vital power bus was being returned to service subsequent to maintenance when a spurious trip was received on-

"A." ES system resulting in the actuation of "A" ES trai Due to plant

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conditions at the time of the actuation (equipment secured and/or tagged out for outage work) "A" Emergency Diesel Generator was the only ES equipment actuated. The ES train was reset immediately by the operator The problem was attributed to a voltage surge caused when reenergizing "A" vital bus, flaintenance on "A" vital bus was completed and returned to service with no additional ES problems occurring. The inspectors reviewed this event and have no further question i

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