IR 05000219/1988028

Insp Rept 50-219/88-28 on 880911-1004 & 1014-29.Apparent Violation Noted.Major Areas Inspected:Activities in Progress,Including Plant Operations,Radiation Control, Physical Security,Maint & Forced Shutdown Activities

ML20206F953
Person / Time
Site:	Oyster Creek
Issue date:	11/15/1988
From:	Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20206F919	List: IR 05000219/1988028 ML20206F915 ML20206F929
References
50-219-88-28, IEB-80-08, IEB-80-09, IEB-80-22, IEB-80-3, IEB-80-4, IEB-80-8, IEB-80-9, IEB-81-2, IEB-84-03, IEB-84-3, IEC-77-13, IEC-79-12, IEC-80-03, IEC-80-04, IEC-80-22, IEC-80-3, IEC-80-4, IEC-81-02, IEC-81-2, NUDOCS 8811210412
Download: ML20206F953 (19)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /88-28 Docket N License N OPR-16 Priority -- Category C Licensee: GPU Nuclear Corporation 1 Upper pond Road Parsippany, New Jersey 07054 Facility Name: Oyster Creek Nuclear Generating Station Inspection Cor. ducted: September 11 - October 4 and October 14-29, 1988 Participating Inspectors: W. Baunack E. Collins D. Lew J. Wechselberger Approved By: 4A) '

1, C M //!/f C. Cowg f, Reactor Projects Section IA Date Inspection Summary:

Areas Inspected: Routine inspections were conducted by the resident inspectors rad one region based inspector (293 hours0.00339 days <br />0.0814 hours <br />4.844577e-4 weeks <br />1.114865e-4 months <br />) of activities in progress including plant operations, radiation control, physical security, maintenance, and forced shutdown work activities. In addition, inspectors reviewed 4160 V ground fault associated with the No. 2 emergency diesel cabling, isolation conden*2r system steaming phenomenon, drywell air lock testing, radiological contam, nation events, protected area breach and cathodic protection system installatio The inspectors also re-viewed licensea's actions in response to a 10 CFR Part 21 report on HFA relays, emergency service water pump failures and RE03A problem, TI 2515/98, Containment Temperature Profiles and TI 2515/66, Inspection Requirements for I.E. Bulletin 84-03, "Refueliag Cavity Water Seals."

Results: One apparent violation was addressed concerning drywell air lock testin YF 2Tf5/98 was complete TI 2515/66 remained open as well as I.E. Bulletin 84-0 Concerns were developed with timely reporting of security events and the practice of scramming the reactor to shutdown. The licensee agreed to evaluate the practice of inserting a scram to complete a normal reactor shutdow PDR ADOCK 03000219 O FDC

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TABLE OF CONTENTS PAGE 1.0 Summary of Plant Operations During Report Period (71707,71710,93702).............................................. 1 2.0 Orywell Ai rlock Te sting (71707, 61700, 42700) . . . . . . . . . . . . . . . . . . . . . . . . 1 3.0 Action on Previous Inspection Findings (71707) . . . . . . . . . . . . . . . . . . . . . . . 3 Closed Items: 77-CI-13 79-CI-12 80-22-01 80-BU-08 80-BU-09 80-CI-03 80-CI-04 80-CI-22 81-01-03 81-01-04 81-CI-02 83-08-02 83-08-03 84-01-01 65-01-01 4.0 Temporary Instruction 2515/98 (71707)........................ ....... 6 4.1 Description of Drywell Air Cooling System....................... 6 4.2 Orywell Bulk Temperature Calculations,. ........................ 6 4.3 Average Bulk Drywell Temperature Limit.......................... 7 4.4 Determination of Temperatures Utilized in the Remaining Environmentally Qualified Lifetime Calculations............... 7 Summary......................................................... 8 I.E. Bulletin 84-03, Refueling Cavity Water Seals (71707, 92703). . . . . 9 6.0 Monthly Surveillance Observation (61726, 37700)...................... 9 6.1 Standby Gas Treatment System Test............................... 9

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6.2 RE03 High Pressure Scram Instrument............................. 10 7.0 HFA Relay 10 CFR 21 Report ( 36100,71707)............................ 10 8.0 Co re Bo re D ri l l i n g ( 9 3 702 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

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Table of Contents PAGE 9.0 Reactor Cavity Epoxy Coating Removal (93702)......................... 11 10.0 Plant Operational Review (71707)..................................... 12 10.1 Normal Reactor Shutdown......................................... 12 10.2 Emergency Service Water Pump.................................... 12 10.3 Main Steam Line Maintenance Plug................................ 12 10.4 Control Room.................................................... 13 l 10.5 Facility Tours.................................................. 13 l

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11.0 Radiation Protection (71707)......................................... 14 11.1 General..................................................... ... 14 j 11.2 Iodine Evolution Events......................................... 14 1 12.0 Review o f Periodic and Special Reports (90703) . . . . . . . . . . . . . . . . . . . . . . . 15 13.0 Observation of Physical Security (93702)............................. 15 13.1 General......................................................... 15 13.2 Protected Area Barrier.......................................... 15 14.0 Backshift Inspections................................................ 16 15.0 Exit Interview (30702, 30703)........................................ 16

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OETAILS 1.0 Summary of Plant Operations During Report Period During this report period the plant operated until 9/29/88, when the licensee elected to proceed to cold shutdown as a result of anomalous temperature con-ditions with both isolation condensers. The "B" isolation condenser started this steaming phenomenon on 8/28 (see Inspection Report 50-219/88-23) followed by the "A" isolation condenser steaming on 9/26. After some preliminary evaluation the licensee elected to shutdown on 9/29 reaching cold shutdown on 9/30. Based on the extent of the isolation condenser problem and other required maintenance to be completed prior to startup, the licensee chose to commence the 12R refueling outage two weeks early on 9/30. On 10/2 the lic-ensee experienced a ground fault on 4160 V emergency bus which was found to be associated with the No. 2 emergency diesel generator cabling. As a result of these events, an Augmented Inspection Team (AIT) was formed to review the inoperability of the isolation condensers and the loss of a vital electrical bus on October 2, 1988 (see Inspection Report 50-219/88-80). The inspector's preliminary review of these events, prior to the team inspection, are re-flected in the AIT repor .0 Orywell Airlock Testing Title 10 of the Code of Federal Regulations, Part 50, Appendix J, specifies three aspects to primary containment airlock testing:

(1) Periodic airlock testing at 6 month intervals, (2) Testing aftar periods when primary containment is not required and has been opened for access, and (3) Testing when the airlock is opened for access while primary containment is require The test pressure specified for situations (1) and (2) above is the pressure corresponding to the peak internal containment pressure ouring accident con-ditions, P For Oyster Creek, Pa=35 psi In a letter to the NRC dated November 22, 1978, the licensee described their program to meet the testing requirements of Appendix J. Where appropriate, exemptions to the testing requirements of Appendix J were requeste The licensee also specified the manner in which the drywell airlock was to be tested. The letter explicitly addressed situations (1) and (3) above, committing to the testing specified by Appendix J; however, it made no mention of the testing to be performed after situation (2) when the drywell is open fo- access when the primary containment is not require No exemption was requested for drywell airlock testin _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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In response to the Oyster Creek letter, the NRC responded in March 1982 with a safety evaluatio This evaluation found the licensee's proposed airlock testing acceptable and no exemption to the testing requirements of Appendix J was necessary. It should be noted that Oyster Creek Technical Specifica-tions do not specify the testing requirements of the licensee letter of 1978 or the testing requirements of Appendix J. Since November 5, 1971, plant  ;

Technical Specifications have specified that the drywell airlock be tested '

s at 10 psig every refueling outage. The licentee originally submitted a change  !

request to reflect the new testing requirements of the drywell airlock on July '

25, 1986. This request was revised and resubmitted on February, 15, 198 j A review of drywell airlock leak tests for 1987 and 1988 showed that the lic- ,

ensee has not been testing the drywell airlock in accordance with the require- -

ments of Appendix J after opening for accesi when the primary containment is *

not required. On five occasions, the reactor was started up and heated to  ;

greater than 212 degrees without the performance of Appendix J required dry- '

well air lock testing. On three of these occasions, the proper test was per-formed about 2 days later, after a drywel' entry. On the other 2 occasions, no retest was performed, and the reactor operated for about 12 and 46 days with an improperly tested airlock. Periods of reactor operations with an improper airlock test were as follows:

l 1/6/87 until 1/17/88: 12 days 1/19/87 until 1/21/87: 3 days 3/10/87 until 4/24/87: 46 days  !

5/15/87 until 5/16/87: 2 days 8/4/87 until 8/5/87: 2 days

Inspection report 50-219/88-23 addressed inspector review of the drywell air- '

lock testing associated with the startup in August 198 In addition, the review showed that the plant operated from 11/22/87 until 7/10/88, a period of 229 days without performing a drywell airlock leak tes Appendix J specifies that the airlock be tested every 6 month ,

Licensee review of drywell airlock leak testing, as reported in Licensee Event  !

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Report 88-23, concluded that improper testing had existed for several operat-

ing cycles.

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Discussions with the licensee indicated that even though the Oyster Creek .

letter of 1978 and the NRC safety evaluation of 1982 delineated the commitment  ;

to test in accordance with Appendix J, these testing requirements were not .

l incorporated into station procedures. Specifically, the test pressure (35  !

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psig) to be used after periods when the drywell is open for access and primary  !

containment is not required, was not incorporated into the surveillance test procedure (665.5.005, "Orywell Airlock Leak Rate Test"). The surveillance  :

i specified the performance of the test at 10 psig. In addition, since the

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technical specifications specified performance of the airlock leak test at-each refueling outage, the master surveillance schedule also reflected this testing interva In order to correctly specify the required testing of .the drywell airlock, the licensee has submitted a revision to surveillance test 665.5.005 specify-

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ing testing at Pa, 35 psig, and a rev aion.to Station Procedure 312, "Reactor Containment Integrity and Atmosphere Control," specifying the testing require-ments of Appendix J. In addition, a revision to Station Procedure 201.1, ,

"Approach to Criticality," has been submitted adding a step to verify that '

the drywell airlock LLRT has been performed. The master surveillance schedule '

has been updated to reflect the 6 month testing interva The licensee currently has in place a Licensing Action item program which !

identifies and tracks commitments to ensure implementation. This program did not exist when the commitments to test the drywell airlock were made. The .

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licensee expressed confidence that had these commitments been made under the Licensing Action Item program, they would have been implemente The failure to perform drywell airlock testing at the required pressure, Pa, i

, and at the required frequency, every 6 months, constitutes a violation of the ,

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requirements of 10 CFR 50, Appendix J (50-219/88-28-01). The fact that the '

! licensee had submitted a technical specification change request delineating

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the testing requirements of Appendix J demonstrated that they understood what the testing requirements were. However, it was not until these requirements i were identified by the inspector that the licensee took action to implement

them, 3.0 Action on Previous Inspection Findings

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(Closed) Inspector Follow Item 77-CI-13, IE Circular 77-13, Reactor Safety Signals Negated During Testing. This circular described an event in which [

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test signals were simultaneously injected into several sensors which affected 6 i both protection and control systems. The licensee has in place Procedure N l 116, "Surveillance Test Program" and Procedure No. 107, "Procedure Control,"

which require each procedure have references, prerequisites, precautions, and

limitations and actions specified which assure that controls are identified

, which would avoid errors in the performance of surveillance testing,

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(Closed) Inspector Follow Item 79-CI-12. IE Circular 79-12 described a t

potential failure mode of EMD diesel generator turbochargers due to lube oil !

pressure problems under certain conditions. The circular also identified a I modification to improve the lube oil system. This modification was installed during the 11R outage and was inspected during Region I inspection 86-1 ,

i 1 (Closed) Unresolved Item 80-22-01. This item related to the acceptance cri- i"

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teria for snubber lock-up and bleed rat Specifically it questions if the acceptance criteria would account for the difference in temperature between

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operating and testing condition The licensea provided the inspector with a technical information bulletin for the GE SF 1154 silicone fluid which is the fluid used in snubbers at Oyster Creek. The bulletin indicates the SF 1154 fluid, in addition to other desirable properties has a useful service temperature range of -40 to 500 degrees The inspector had no further questions relative to this matte (Closed) Inspector Follow Item 80-BU-08. IE Bulletin 80-08, Examination of Containment Liner Penetration Welds, had been previously inspected in Inspec-tion Reports 85-29 and 86-34, and was left open until the liquid poison X6 containment penetration shop weld was volumetrically inspected. The licensee provided the inspector with an ultrasonic inspection data sheet which showed that the liquid poison X6 penetration had been inspected and dispositioned as acceptabl (Closed) Inspector Follow Item 80-BU-0 IE Bulletin 80-09, Hydramotor Actu-ator Deficiencies, describes certain deficiencies in ITT general controls models AH90 and NH90 series hydramotor actuator The licensee's response to this bulletin indicated that none of this series hydramotor valve actuators are used at Oyster Cree (Closed) Inspector Follow Item 80-CI-03. IE Circular 80-03, Protection from Toxic Gas Hazards, recommended that licensees evaluate their plants against toxic gas hazards, particularly chlorine gas. Oyster Creek uses sodium hypochlorite, a relatively stable chemical, in place of liquid chlorine for marine organism control. In addition the plant was evaluated in accordance with NUREG 0737, TMI Action Plan Requirements for Item 111.0.3.4, Control Room Habitability and found to be acceptabl (Closed) Inspector Follow Item 80-CI-04. IE Circular 80-04, Securing of Threaded Locking Devices on Safety-Related Equipment, described several in-stances of inoperability of safety-related equipment which were cause<' ly loosened threaded locking devices. Licensees were to review their ** salla-tion and maintenance procedures to determine that the securing of locking devices has been addressed. By internal memo the licensee indicated that maintenance procedures address manufactures recommended fastener preloads and/or the application of locking devices where required. The inspector had no further questions relative to this matte (Closed) Inspector Follow Item 80-CI-22. IE Circular 80-22, Confirmation of Employee Qualifications, recommends that employment procedures be reviewed to determine that the companies policies confirm the professional qualifica-tions of employees. The licensee provided the inspector with a copy of the company form associated with the employment of a new hir This form, among other controls, includes written statements from former employer The in-spector had no further questions relative to this matte (Closed) Unresolved Item 81- .-03. Licensee to review certain instrument drift problems. The lice s e has completed a trip uncertainty and setpoint analysis review of operav..g data base. Also, a probabilistic risk assessment

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i has shown that setpoint drift is of low importance to risk. The corrective action to resolve the problem consists of eventually replacing the instruments subject to drift. Certain instruments have been replaced and others are planned to be replaced during the next two refueling outages. This item is considered closed based on the licensee having established a course of actio (Closed) Unresolved Item 81-01-04. Licensee to evaluate waterproof integrity of safety related electrical components outside of the primary containmen The licensee has completed corrective action to protect reactor building safety related electrical equipment under budget activity 40230 (Closed) Inspector Follow Item 81-CI-02. IE Circular 81-02 recommended lic-ensees review and revise, as necessary, their administrative controls regard-ing operator performar.ce. The licensee has many administrative controls in place to assure satisfactory operator performance. Among the controls in place are, numerous memos to operators addressing conduct, performance, ad-herence to procedures, etc.; various required reading assignments; management off-shift tours; personnel responsibilities specified in a fitness for duty manual; an administrative procedure, standards of conduct; an annual command responsibility letter issued to group shift supervisors; as well as annual management interviews with operator The inspectors consider adequate con-trols have been established to assure satisfactory operator performanc ,

(Closed) 'nspector Follow Item 83-08-02. Review corrective actions taken to correct i design deficiency in the Standby Gas Treatment System which caused one train's inlet and outlet valve to open when the exhaust fan motor circuit breaker is racked out. A modification was made under Budget Activity 402526 which corrected thi2 deficienc (Closed) Inspector Fcilow Item 83-03-03. Review controls established for maintaining the trunnion room door closed. A sign has been posted stating the door must be kept closed except for passage. Also, the key for the lock on the door is being controlled by the group shift superviso (Closed) inspector Follow Item 84-01-01. Inspectors to followup on corrective actions taken to resolve limitorque problems reported in Licensee Event Report (LER) 83-24. The licensee, in a followup LER dated February 26, 1988, de-scribed the corrective actions taken and the results of these action (Closed) Violation 85-01-01. This violation dealt with the licensee not having established measures to promptly evaluate, take corrective action, and report a condition in which the containment spray water pressure was being maintained higher than the emergency service water pressure in the containment spray heat exchangers contrary to the FSAR system description. As part of corrective acticn this condition was reported in Licensee Event Report 84-026, the current Surveillance Procedure 607.4.003, "Containment Spray and Emergency Service Water Pump Operability and Inservice Test," requires the proper delta pressure be maintained as part of the acceptance criteria, and steps have been taken to assure open items on the commitment tracking system were properly resolved. This item is considered close _ _ .. ._

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4.0 Temporary Instriction 2515/98

'In response to 'femporary Instruction 2515/98, a review of Oyster Creek Nuclear Generating Station's drywell average ambient operating temperature profile was conducted. The review included the assessment of the adequacy of the average temperatere measurements to accurately reflect drywell conditions and the assessment of the consistency to use temperatures experienced by equipment in the calcult e of its remaining environmentally qualified lifetim .1 Description of Drywell Air Cooling System Oyster Creek Nuclear Generating Station has a General Electric Mark I type containment. The drywell air. cooling system consists of five re-

. circulation fans located on the 51' elevation. Each fan has a capacity rating of 20,000 efm. There are five air coolers which are located at the intake of the recirculation fans and are supplied by the Reactor Building Closed Cooling Water Syste Each fan takes a suction from a circular plenum located at the 91' 7" elevation of the drywell (6,400 cfm) and through a damper located at the L 47' elevation (13,400 cfm). The circular plenum at the 91' 7" elevation contains piping at three locations which draws air from above the re- i fueling bulkhead, and at five locations which draws air from the 89' L elevatio The fans discharge into a circular 91enum located at the 54' elevatio The circular piping discharges ah through nine ducting locations to the 54' elevation and through five duccing locations to the 37' 3" elevation. ;

Additionally, ducting from the circulum plenum provides cooling inside -

the biological shield at five locations and to the area ender the reactor vessel at one locatio .2 Drywell Bulk Temperature Calculation The Drywell Bulk Temperature is calculated by the licensee using the fol- [

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fB = 0.54 (T Avg Ia) + 0.347 (T Avg Ib) + 0.075 (T Avg II) + 0.036 (T i Avg III)  !

i Where T Avg Ia is the average of five thermocouples (thermocouples 100 [

l A-E) located at t'ie 33' elevation near the recirculation pump motors, ,

T Avg Ib is the a verage of five thermocouples (thermocouples 103 A-E)

I located at the 5( ' elevation near the safety valve area; T Avg II is the average of five thermocouples (thermocouples 104 A-E) l located at the 938 elevation near the reactor vessel bellows seal area; l and,

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T Avg III is the average of three thermocouples (thermocouples 105 A-C)

located at the 95' elevation near thr: reactor vessel flange are The T Avg terms in the Orywell Bulk Temperature calculation are weighed according to the difference in the "elative air volume which each set of thermocouples represent .3 Average Bulk Drywell Temperature Limit The licensee's Final Safety Analysis Report limit for Drywell Bulk Tem-perature is 150 degrees F. This limit was changed from 135 degrees F in June 1985 as a result of the difficulty in the Drywell Air Cooling System to maintain the Bulk Drywell Temperature within limit Presently there are no Technical Specification limits for Drywell Bulk Temperatur If the Bulk Temperature Limit is exceeded, however, required actions are addressed in the Emergency Operating Procedures and the Administrative Procedure 106.9, "Emergency Operating Procedure Supporting Calculations".

Additionally the NSSS Annunciator Response Procedure 2000-RAP 3024.01 provides guidance in the event the "0W TEMP HI" alarm annunciates. This alarm is triggered when the thermocouple at the recirculation fan cooler outlet reaches 115 degrees F which correlates to a Bulk Orywell Tempera-ture of 150 degrees .4 Determination of Temperatures Utilized in the Remaining Environmentally Qualified Lifetime Calculations Presently, the licensee is assigning temperatures for the purpose of environmentally qualified lifetime calculations based upon the zone in which the component is locate The zones and the correspondingly as-signed temperatures are:

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Bottom of Drywell to Elevation 74'. 139 degrees Elevation 74' to Elevation 89'..... 198 degrees Elevation 89' to Elevation 91.6'... 205 degrees Elevation 91'6" to Elevation 94'... 259 degrees Elevation 94' to Drywell Dome...... 252 degrees Orywell Dome....................... 270 degrees For the bottom of drywell to elevation 74' zone, the average of thermo-couples 100 A-E and 103 A-E were utilize It was concluded by the !!c-ensee that the large volume of the spherical portion of the sphere sup-ported good air circulation. The average of these thermocouples was plotted against drywell bulk temperature. A linear regression fit was then performed to project the temperature in this zone to that which would be expected if the drywell were operating at its maximum allowed bulk temperature of 150 degrees For the elevation 74' to Elevation 89' zone, thermocouples located in the circular supply plenum and a thermocouple in the circular supply plenum which monitors air temperature drawn from the drywell dome a

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utilized. From the above described ventilation ducting arrangement, the circular supply plenum draws air from the 89' elevation and the drywell dome. Calculations utilizing documented air flows compensated for the air temperature contribution from the drywell dome and thus determined the air inlet temperature at the 89' elevatio This calculated tempera-ture for this zone was extrapolated to the temperature which would be expected if the drywell were operating at its maximum allowed bulk tem-peratur For the elevation 89' to elevation 91.6' zone, thermocouple 106C (the highest of the 106 thermocouples) was utilized as the representative temperature based upon the air flow pattern. Above 91.6', the seal plate adds a significant amount of heat to the air. Thermocouples 104 A, B, and E were utilized to represent the elevation 91.6' to elevation 94'

zone. For the elevation 94' to drywell dome zone and the drywell dome zone, thermocouples 105 B-C and thermocouple 107 were utilized respec-tively. As with the previous zones, these temperatures were extrapolated to the temperature which would be expected if the drywell were operating at its maximum allowed bulk temperatur Currently, a revision to the temperatures utilized in remaining environ-mentally qualified lifetime calculations is being planned. The changes include the division of the drywell into seven temperature zones vice six zones. Furthermore, m:re recent data from the period August 6, 1987 through March 9, 1988 has been collected to determine the temperatures utilized in a zone. The mean thermocouple readings for this period will be determined and the highest mean projected to 145 degrees F (vice 150 4 degrees F in the presently utilized temperatures). Though the net effect of these results are less conservative frem the previous calculations, the temperatures still conservatively represent the temperatures which environmentally qualified components are subjected t .5 Summa ry The stratification of temperatures are evident in the Oyster Creek Nuc-lear Generating Station's containment. The temperature variation between the top of the drywell and the bottom normally exceeds 100 degrees The licensee, however, has compensated for this temperature stratifica-tien in their calculation of the bulk drywell temperature by utilizing a weighed average of eighteen thermocouple Furthermore, the licensee has utilized local thermocouple readings in determining the temperature used in the remaining environmentally quali-fied lifetime calculation Considerations of localized air flow and heat sources appear to be adequately addressed. Review of bulk drywell temperature data since April 1987 showed that the licensee's temperature extrapolations are reasonably conserva 'e. TI 2525/98 is considered close ( _ ,.

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' I.E. Bulletin 84-03, Refueling Cavity Water Seals This bulletin was previously reviewed in Inspection Report 50-219/85-19 which indicated that the bulletin would remain open, pending a supplemental response from the licensee. The licensee's original response indicated that the re-fueling cavity water seals at Oyster Creek could not experience the cata-strophic failure that occurred at Haddam Neck and that the concerns indicated in .the bulletin are not applicable to Oyster Creek. The licensee's response did not address bulletin considerations. I.E. Information Notice 84-93, Potential for loss of Water from the Refueling Cavity, describes additional events involving problems with the cavity seals and provides a sample licensee response. The licensee plans to review their bulletin response to determine if additional information is warrante Temporary Instruction 2515/66, Inspection Requirement for I.E. Bulletin 84-03,

"Refueling Cavity Water Seals" was issued to provide guidance to the inspector in reviewing Bulletin 84-03. The inspector questioned what actions the lic-ensee had taken to determine the leak tightness of their cavity seal. Ap-parently some effort was expended in determining the integrity of the refuel-ing cavity seal at Oyster Creek and as a result three reports were written to describe this effort and the results. These reports were not made avail-able to the inspector before the end of the inspection period, but will be reviewed during the next inspection period. The licensee stated that the tests indicated little to no leakage past the seals and that from monitoring the seal leakage detection system and drains no water has been observed past the seals during the present refueling outag In reviewing the annun.iator alarm response procedure for the refueling cavity

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seal leak alarm, the ir.spector noted that the manual corrective actions sec-

, tion directs the operator to use the core spray system to recover level if i necessary per 2000-ABN3200.3 This abnormal procedure had not been issued l yet and was unavailable for the operator to use in the control room. When

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presented to operations management, a decision was made to simply delete the

! abnormal procedure from the alarm annunciator response procedure without de-

, termining what guidance should be provided to the operator until the abnormal l procedure was issued. Further discussions with operations indicated that i appropriate steps would be taken to ensure proper direction was provided by

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the annunciator alarm response procedure. The licensee indicated that cor-rective action would be taken to ensure that when procedure changes are made that other procedures would also be revised concurrently.

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TI2515/66 will remain open pending further inspector review.

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6.0 Monthly Surveillance _ Observation 6.1 Standby _ Gas Treatment System Test On September 26, 1988, the inspector observed the complete performance of Surveillance Procedure 651.4.001, "Standby Gas Treatment System Test."

The observed surveillance tested the automatic and proper operation of

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the Standby Gas Treatment System (SGTS), and satisfied the Technical Specification requirements for the testing of the reactor building high radiation isolation and SGTS initiation. Prior to the execution of the procedure, it was noted that proper administrative approval was obtaine The inspector while observing the surveillance verified that the proce-dure adequately directed the control roum operator to return the system to its initial configuration. After completion of the surveillance, a review of the test data results was conducted and verified to meet the acceptance criteria. A verification that the appropriate management and engineering personnel conducted a review of the completed procedure was checked. No unacceptable conditions were identifie .2 RE03A, High Pressure Scram Instrument The inspector reviewed the licensee's action in response to a false trip on RE03A, reactor high pressure scram instrumen RE03A has exhibited a number of problems in its recent operating history (see Inspection Re-ports 88-04 and 88-09) and the licensee plans to replace this instrument as part of the analog trip system replacement for RE03/RE15 during the current refueling outage, 12R. The false trip of RE03A occurred while testing the pressure sensor (IA83A) for NR108A ("A" electromatic relief valve), which is located on the same instrument rack with RE03A. As a result of the false trip, the licensee conducted a surveillance of RE03A to determine instrument performance. During the surveillance the in-strument on six successive trips fell within the required technical specification requirements. The plant shutdown two weeks later for the 12R outag The inspector developed no concern .0 HFA Relay 10 CFR 21 Report The inspector reviewed the licensee action to address General Electric Service Advice Letter (SAL) 192.1. The SAL described actions to be taken to determine if "HFA" relays experienced moving contact finger binding reported by 10 CFR 21 report. In the report the vendor described moving contact binding due to

manufacturing of contacts wider than the prescribed tolerance. The report i

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recommended testing of HFA relays manufactured before September 198 The licensee identified approximately 200 HFA relays in the warehouse for vendor testin This was accomplished with no contact binding discrepancies identi-fie The inspector questioned the condition of plant installed relay The licensee explained that Station procedure 732.2.006, "Relay Replacement Pro-

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cedure," incorporated the recommendation of General Electric Service Informa-tion Letter Number 44, Supplement 4, which provides guidance for checking for contact binding. According to the licensee, this procedure was performed for those HFA relays replaced in accordance with I.E. Bulletin 84-02, "Failure of General Electric Type HFA Relays in use in Class 1E Safety Systems." The

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licensee plans to complete the HFA relay replacement required by I.E. Bulletin 84-02 this outage. In addition, the licensee stated that their surveillance program would detect any problems with those relays installed in the plan The inspector found the licensee's response to this issue to be acceptabl t

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8.0 Core Bore Drilling On October 13, 1988, the licensee recommenced cors bore drilling to install drywell cathodic protection system. The drywell cathodic protection system is a modification to reduce the general corrosion rate of the drywell which is postulated to have been caused from the leakage of water past the reactor cavity seal. Core bore drilling was suspended on August 8, 1988, when the licensee confirmed that the drywell had been damaged by core bore drillin Prior to recommencing core bore drilling, the licensee completed a final critique and implemented corrective actions. The critique identified the root cause of the drywell damage event to the "unreasonable reliance" that the core bore drill would encounter the sandpocket. Furthermore, the critique identi-fled that either the assessment of the location of the sandpocket or the ade-quacy in laying out the job was inadequate. The licensee's permanent correc-tive actions to this event include:

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the evaluation of the absolute errors involved in core bore drilling,

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the assessment of the adequacy of the survey methodology,

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the implementation of administrative controls and required authorizations to proceed beyond drill stop points,

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the requirement to evaluate each planned hole for "hole specific" design work,

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the requirement to issue a "construction hold" when any anomaly is en-countere Currently, the licensee has suc:essfully completed drilling six hele The inspector has no further questions regarding this matte .0 Reactor Cavity Epoxy Coating Renoval

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As part of the corrective actions to prevent water leakage from the refueling cavity, a permanent epoxy coating was applied to the trough area associated with the drywell-reactor building seal bellows. Shortly after the epoxy ap-plications, the reactor vessel dryer assembly was removed. As a planned part

, of this evolution, the dryer assembly was sprayed with water as it was moved from the vessel to its storage area, As a result of this spraying, the epoxy coat was wet. Shortly afterwards, it was observed by the licensee that the

epoxy coating was not adherin The licensee evaluated the problem and decided to remove the original appli-cation of epoxy by using a high pressure water lance. A new coating would be applied, a removable latex coating. This was the same coating that was being applied on the reactor cavity walls. The licensee concluded that the wetting of the epoxy coating prior to the completion of curing resulted in the inability of the coating to adhere.

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The licensee prepared a safety evaluation for the new activities. This was reviewed by the inspector, and the inspector concluded that the licensee ade-quately addressed the concerns associated with the removal of the old epoxy and the application of the new coating. No unacceptable ccnditions were identifie In addition, the licensee performed an analysis on a sample of the improperly cured epoxy coating in order to determine if chloride compounds were suscept-ible to "breaking down". This condition could result in chloride ions leach-ing out and creating an adverse environment for plant materials. The test results demonstrated that the compounds remained intact and that chloride ions from the improperly cured epoxy coating did not create a hazar .0 Plant Operational Review The inspector reviewed details associated with key operational events that occurred during the report period. A summary of these inspection activities follow .1 Normal Reactor Shutdown

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During this report period the licensee performed a normal plant shutdown by scramming the reactor from 20 percent power. Apparently the licensee intends to scram the reactor as a normal method of plant shutdown. The inspector questioned whether the routine scramming of the reactor had been fully evaluated from operations and engineering perspective. In response, the licensee is evaluating this practice. The inspector will review this evaluation in a future inspectio .2 Emergency Service Water Pump On September 16, 1988, the Emergency Service Water (ESW) pump 520 faile The ESW pump 52C had failed the month before. As a result of the two failures, the inspector reviewed the material history records for all the ESW pumps to determine whether a pattern of increased ESW pump fail-ure rate was occurring. Due to the small amount of failure date avail-able, no conclusion could be made. The licensee, however, has shipped the two failed pumps to Byron-Jackson in California for failure analysis and repair. The licensee's effort to determine the failure mechanism for the ESW pumps is positive. The inspector will follow up on the re-sults of the Byron-Jackson analysi .3 Main Steam Line Maintenance Plug The inspectors reviewed the design of the main steam line maintenance plug, which is used by the licensee to conduct maintenance activities on the MSIV's while the reactor vessel is flooded up above the main steam lines. The inspector was satisfied that the dec.ign was not strictly pneumatic, but questioned the licensee with regard to the maximum leak-rate possible if the pneumatic portion faile The licensee was not

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aware of the estimated leakrate if a pneumatic failure occurred and con-tacted the vendor to determine the leakrate. The inspector identified no unacceptable condition .4 Control Room 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;

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Reactor Building and Turbine Building Tour Sheets;

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Equipment Control Logs;

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Standing Orders; and,

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Operational Memos and Directive No unacceptable conditions were identifie .5 Facility Tours 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 are 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 The following additional items were observed or verified: Fire Protection:

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Randomly selected fire extinguishers were accessible and in-spected on schedul Fire doors were unobstructed and in their proper position.

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Ignition sources and combustible materials were controlled in accordance with the licensee's approved procedure Appropriate fire watches or fire patrols were stationed when equipment was out of servic Equipment Control:

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Jumper and equipment mark-ups did not conflict with Technical Specification requirement !

Conditions requiring the use of jumpers received prompt licen-

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see attentio Administrative controls for the use of jumpers and equipment mark-ups were properly implemente Vital Instrumentation:

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Selected instruments appeared functional and demonstrated parameters within Technical Specification Limiting Conditions for Operation, Housekeeping:

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Plant housekeeping and cleanliness were in accordance with approved licensee programs. In addition, the inspector made a tour of the drywell during a backshift to review housekeepin No unacceptable conditions were identifie .0 Radiation Protection 11.1 General During entry to and exit from the RCA, the inspectors verified that pro-per warning signs were posted, personnel entering were wearing proper dosimetry, personnel and materials leaving were properly monitored for radioactive contamination, and monitoring instruments were functional and in calibratio Posted extended Radiation Work permits (RWPs) and

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survey status boards were reviewed to verify that they were current and I accurate. The inspector ob:,erved activities in the RCA to verify that

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personnel complied with the requirements of applicable RWps and that l workers were aware of the radiological conditions in the are .2 Iodine Evolution Events

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In addition, the inspecto= reviewed aspects of iodine evolution events when removing the reactor vessel head which resulted in personnel iodine

! uptake, a radioactive spill outside the new radwaste building, and a

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personnel contamination inside the drywell. All three of tnese events were reviewed in detail by a radiation specialist and are discussed in Inspection Report 50-219/88-3 .0 Review of Periodic and Special Reports Upon receipt, periodic and special reports submitted by the licensee pursuant to Technical Specification requirements were examined by the inspectors. This review included the following considerations: the report includes the informa-tion required to be reported to the NRC; planned corrective actions are ade-quate for resolution of identified problems; and the reported information is valid. Monthly Operating Reports for August and September were reviewe No unacceptable conditions were identifie .0 Observation of Physical Security 13.1 General Ouring daily tours, the inspectors verified that access controls were in accordance with the Security Plan, security posts were properly manned, protected area gates were locked or guarded and that isolation zones were free of obstructions. The inspectors examined vital area access points to verify that they were properly locked or guarded and that access con-trol was in accordance with the security plan, 13.2 Protected Area Barrier On October 23, 1988, the licensee identified a degradation of a portion of the protected area barrier. This was initially discovered by an equipment operator at 1: 45 a.m. and repcrted to the group shift super-visor (GSS) in the control room. The GSS was concerned with possible equipment problems associated with this deficiency and did not consider protected area barrier aspects. At 10:30 a.m. plant security became aware of the deficiency and took appropriate actio Although the protective area boundary was somewhat degraded, compensatory measures were in effect for an unrelated matter in the area. Due to the existence of these compensatory measures and the difficulty and knowledge required to exploit the degradation, the inspector has no significant i safety concerns with this event.

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This event, however, uncovered a weakness in operations personnel's awareness of the security barrie Furthermore, weaknesses in the com-munications between operations and security and in the knowledge of 10

, CFR 73 Appendix G requirements appear to be factors in this event. The licensee determined an additional contributing factor to the late report

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, of this event to the NRC was the fact that the Administrative Procedure j 126, "Procedure for Notification of Station Events," did not reflect the i

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changes in the 1988 revision of 10 CFR 73 Appendix The licensee in-tends to make a revision to this procedure to reflect the 10 CFR 73 Appendix G reporting requirement The licensee is currently preparing a Licensee Event Report and formu-lating corrective actions to address this event to prevent its recurrenc The inspector had no further questions regarding this matte .0 Backshift Inspection NRC inspections of licensee activities on backshift were conducted on the fol-lowing dates:

Sunday, September 26, 1988 Wednesday, October 26, 1988 Areas of inspection included the observation of control room activities and surveillances and plant containment tour .0 Exit Interview A summary of the results of the inspection activities performed during this report period were made at meetings with senior licensee management at the end of this inspection. The licensee stated that, of the subjects discussed at the exit interview, no proprietary information was included, i

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