IR 05000295/1987017

Safety Insp Repts 50-295/87-17 & 50-304/87-19 on 870731-0903.Violations Noted.Major Areas Inspected:Summary of Operations,Inadvertent Deenergizing of Two Redundant Heat Trace Circuits & Diesel Generator 1A Failure

ML20235B122
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	09/17/1987
From:	Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20235B109	List: IR 05000295/1987017 ML20235B107 ML20235B118
References
50-295-87-17, 50-304-87-19, NUDOCS 8709230551
Download: ML20235B122 (15)

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U.S. NUCLEAR REGULATORY COMMISSION l REGION III Report Nos. 50-295/87017(DRP);50-304/87.019(DRP)

Docket Nos. 50-295; 50-304 License Nos. DPR-39; DPR-48'

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-Licensee: Commonwealth Edison Company P. O. Box 767-Chicago, IL' 60690 l Facility Name: Zion Nuclear Power Station, Units 1 and 2 ,

Inspection At: Zion, IL Inspection Conducted: July 31 through September 3, 1987 Inspectors: M. M. Holzmer P. L. Eng N. R. Williamsen

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Approved : J h. Hinds, ef o9i7.tG eactor Projects ection IA Date Inspection Summary Inspection on July 31 through September 3, 1987 (Report No /87017(DRP); 50-304/87019(DRP))

Areas Inspected: Routine, unannounced safety inspection of licensee action on previous inspection findings; summary of operations; inadvertent de-energizing of two redundant heat trace circuits; 1A diesel generator failure; automatic starts of 2A auxiliary feedwater pump; operational safety verification and engineered safety feature (ESF) system walkdown;-surveillance observation; maintenance observation; licensee event reports (LERs); training; response to Region III requests; and site visit by Branch Chie Results: Of the 11 areas inspected, no violations or deviations were identified in 10 areas, and one violation was identified in the remaining area (failure to conduct Technical Specification required surveillance).

9709230551 870917 PDR ADOCK 05000295 G PDR v4

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DETAILS 1.- Persons Contacted i

• G. Plim1, Station Manager .

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E. Fuerst, Superintendent, Production

• T. Rieck, Superintendent, Services

• W. Kurth, Assistant Station Superintendent, Operations R. Johnson, Assistant Station Superintendent, Maintenance H J. Gilmore, Assistant Station Superintendent, Planning l

• R. Budowle, Assistant Station Superintendent, Technical Services -!

L. Pruett, Unit 1 Operating Engineer N.lValos Unit 2 Operating Engineer

.M. Carnahan, Training Supervisor R. Cascarano,. Technical. Staff Supervisor

• C. Schultz, Regulatory Assurance Administrator V. Williams, Station Health Physicist;

• J. Ballard, Quality Control Supervisor

• W. Stone, Quality. Assurance Supervisor

• T._Printz, Assistant Technical-Staff Supervisor

• T. Maiman, Vice President, PWR Nuclear Stations Operations o-

• Indicates persons present at exit intervie . Licensee Actions on Previous Inspection Findings (92701)

(Closed) Unresolved Item (295/87013-01): Safety injection' due to' all four main steam isolation valves opening due to personnel error. A j letter and notice of violation dated August 3, 1987, were issued for this event. The licensee's response, including corrective actions, {

is due to the NRC on September 4, 1987. Since the violation will remain l open to track the licensee's corrective actions, this Unresolved Item 1 is considered close l No violations or deviations were identifie . Summary of Operations Unit 1

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The unit operated for the entire inspection period at power levels up to 99%.

Unit 2 The unit began the inspection period in Mode 3 (Hot Shutdown). Mode 7 (Low Power Physics Test) was entered on August 1, 1987, and Mode 2 (Hot Standby) was entered on August 4. The unit was tied to the grid on August 8, 1987, and operated for the remainder of the inspection period J'

at power levels up to 99%.

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I.iAugus't.4,1987InadvertentDe-energjzatianofTwoTrainsofHeat' Tracing I on BAT (93702)

On August 4, 1987, at approximately 8:30 a.m. CDT, with Unit 1 at 95%;

power and Unit 2 critical at less than 1% power, an electrician de-energized both trains of heat tracing on the section of piping between the 0A boric acid (BA) tank and the suction side of the 1A BA transfer o pump. De-energization of both trains of BA heat tracing placed the

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licensee in Technical Specification 3.2.1.F.3, which requires borating

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to cold shutdown within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if the heat tracing cannot be restore

The condition was discovered by an equipment operator on routine round The operator immediately notified the shift engineer, who ordered the power restored to the heat trace circuits (51014). The heat tracing was de-energized for approximately 20 minute f The heat trace wiring consists, of a primary coil and a redundant coi I During the previous Unit 2 outage a portion of both coils had been c removed from the 1A BA pump suction line in order to facilitate repair ,

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of the 0A boric acid tark. -These coils also serve a section of pipe between the 1A and'1B BA pumps which was still in servic Consequently, both heat trace circuits remained energized. As the work on the tank neared completion, work request (WR) Z605E4 was written to replace the heat trace niring on the affected piping and component While working on WR Z60584, the electrician opened both circuit breakers feeding.the primary and redundant heat tracing for circuit 51014, de-energizing the required heat tracing on the section of pipe between the 1A and 1B BA pumps. Additional NRC review will focus on licensee controls for who may operate plant equipment, end on licensee policies for work which may proceed without out-of-service controls. This is considered an Unresolved Item pending review of the Licensee Event Report (LER), which is to be issued on September 3, 1987 (295/87017-01; 304/87019-01).

No violations or deviations were identifie '

One Unresolved Item eas '

identifie . Failure of 1A Diesel Generator During Surveillance Testing (93702)

On A' u gust 12, 1987, at approximately 1:30 P.M. CDT, with Unit 1 at 94%,

the licensee was performing PT-11. " Diesel Generator Loading Test," when an equipment operator noticed that the 3L fuel pump to the 1A diesel benerator (DG) was vibrating excessively. The operator proceeded to the

, control room and wrote a work request to investigate and tighten the bolts. The operator then returned to the 1A DG room, where he found that the fuel pump pedestal had broken and that lube oil was spilling onto the ;

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floor of the room. The operator called the control room, and control room J

operators immediately tripped the 1A DG. The licensee declared the 1A DG inoperable and initiated surveillance testing to demonstrate operability of the 1B and 0 DGs. The fuel injector associated with the 3L fuel pump was tested per surveillance procedure P/DG001/3-2N and found to be acceptable. A replacement fuel pump was obtained from the warehouse and l

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f installed in accordanca with WR 262031 and its associated travelle The inspector observed the injector test and portions of the installation .

of the new fuel pump on the 1A DG. The 1A DG was then mccessfully

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tested and returned to service within the 32-hour time constraint allowed by the Confirmatory Order dated February 29, 198 )

The licensee stated that the suspected root cause of the fuel pump failure was loosening of the bolts which hold the fuel pump assembly to the D Inspection of the broken parts revealed that three of the four bolts had sheared and that the pump pedestal plate had broken. The fourth bolt  ;

appeared bent but had not failed. During removal of the bolt material from the bolt L31es in the DG, it appeared that the bolts had partially backed out an ; consequently did not provide a seal capable of preventing a lube oil leak. The licensee checked bolt tightness for the fuel oil pumps for all five DGs. Three bolts were found which required minor retorquing. The licensee stated that the fuel pump and boltin beanalyzedbyitsSystemMaterialsAnalysisDepartment(SMAD)gwould , to s

i determine whether the cause for the event was a defect in either the fuel pump or bolt material. Receipt and review of the SMAD report and determination by the licensee of the cause of the event are considered an Open Item (295/87017-02).

No violations or deviations were identifie One Open Item was identifie . August 26 and August 28, 1987 Automatic Starts of the 2A Auxiliary Feedwater Pump During PT-5 Testing (93702)

On August 26, 1987, with Unit 2 in Mode 1 at full power, the 2A (steam driven) auxiliary fee (vater (AFW) pump automatically started during surveillance testing due to a relay failure. Operators were performing section 10 of test PT-5, " Reactor Protection Logic Tests, Reactor Normal or Shi.tdown," for Unit 2 reactor protection train B logi;. Section 10 verifies that the 2C (motor driven) AFW pump starts on Low-Low level on one of four steam generators (SGs). Because of the failure of one pair of contacts for relay AFP3/XB, the reactor protection logic for two of four SGs was satisfied (one due to the failed relay, and one from the test condition). Because the 2A AFW pump is designed to start on Low-Low level on two of four SGs, the 2A AFW pump automatically started. The 2A AFW pump was secured, and the AFP3/XB relay was replaced under WR 62372 to correct the problem. Following replacement of the relay, a continuity check was performed to verify proper installatio On August 28, 1987, with Unit 2 in mode 1 at 89% po,/er, the 2A AFW pump automatically started during post-maintenance testing for WR 62372 due to a relay failure. Operators were performing section 10 of test PT-5 for Unit 2 to close out the work request. Because of a misaligned contact arm on another relay, AFP1/XB, which is also in the twc of four logic for starting the 2A AFW pump, the reactor protection logic for Low-Low level l on two of four SGs was again satisfied. The 2A ffW pump was secured, and i i the AFP1/XB relay was replaced under WR 6242 N ___ _ ___- ---

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s The NRC review of the event revealed the followinfconcdns:~

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Following the Augwt.26 event, operators considered' test PT-5 to have s

ibeen satisfactorily (ompleted even though the unex,nectdstart of

the- 2A AFW pump occurre This appears contrary to IEEE Standard 338-1977, section 6.4.(2)(b). (Zion Station has no ccmmitment to

comply with IEE 338-1977.) Operators apparently felt that since

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the reactor protection system would have worked properly, the test

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, immediately initiated J,o correct the problem of the unexpected start of the 2A AFW pum '

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Discussion with shift operating personnel revealed that following

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the replacement of relay AFP3/XB, with a bench tested relay which

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passed a continuity check, they felt that the reactor protecticn system was operable, and that there was no reason to believe thet'

the 2A AFM pump would not start. They believed that any problem 7 with open contacts would result in a start of the 2A AFW pump on s Low-Low level in one of the remaining three SGs. Furthermore, the

, redundant train of safeguaids logic was operable. Consequently,'

they felt that no safety or operability problem existed and that s

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it would be acceotable to delay post-maintenance testing until the

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day shift when a ' fresh crew of operators would be available to conduct PT- (Tde crew conducting PT-5 on August 26 had been on l duty for 16 consecutive hours.)  !

As a reAyt of the foregoing decision 6nd the continuity check performed after AFP3/XB was replaced, the PT-5 which was required to close WR 62372 was not performed until two days later. The inspector commented that two days to perform post-maintenance testing on a safety-related relay was excessiv A continuity check may have verified the correctness of the installation of the new relay, but it was insufficient to verify the adequacy of the troubleshooting activity. The continuity ,

check was, therefore, inadequate for the purpose of verifying !

that the correct relay was replace l Since the AFP3/XB and AFP1/XB relays are Westinghouse (!!) NBFD relays, a potential generic issue existed regarding these relay In the first event, a variable resistance was measured across one pair of contacts; in the second event, one contact arm was cocked, causing the contacts not to touch when the relay actuated. The Commonwealth Edison Station Nuclear Engineering Department (SNED)

has reviewed these events for deportability under 10 CFR Part 21, and has concluded that the failures are not reportable. The failure modes of the relays will be included in the licensee's LE The 2A AFW pump was capable of starting throughout the period of interest. In addition, the reactor protection system would have functioned normally. Consequently, the safety significance of this '

event is low. The licensee stated that a letter to shift supervisors

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I would be . written to clarify post-maintenance testing concerns. The letter would also ensure that unacceptable results obtained during testing would include inadve nent starts of equipment other than that ;

which was supposed to operate, and that those results_would be documented '

on the test. This guidance is to be incorporated into station procedure This is considered an Unresolved Item pending NRC review of the ;

licensee's LER to verify the proposed corrective actions and to review 1 the licensee's analysis of'the failure modes of the W NBFD relays

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(304/87019-02). l

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No violations or deviations were identified. One Unresolved Item was ;

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identifie . Operational Safety Verification and Engineered Safety Features System Walkdowr, (71707 & 71710)

The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operators from July 31, 1987 through September 3, 1987. During these discussions and observations, the inspectors ascertained that the operators were alert, fully cognizant i of plant conditions, attentive to changes in those conditions, and that they took prompt action when appropriate. The inspectors verified the operability of selected emergency systems, reviewed tagout records and '

verified proper return to service of affected components. Tours of the auxiliary and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations, and to verify that maintenance requests had been initiated for equipment in need of maintenanc The inspectors, by observation and direct interview, verified that selected physical security activities were being implemented in accordance with the station security pla The inspectors observed plant housekeeping and cleanliness conditions and verified implementation of radiation protection controls. From July 31, 1987 through September 3, 1987, the inspectors walked down the accessibic portions of the 1A, IB and 0 diesel generator support systems to verify operabilit These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under Technical Specifications, 10 CFR, and administrative procedure The following comments were provided to the licensee:

Station housekeeping and cleanliness were improving steadily, and it appears that the licensee's efforts were beginning to produce result The PT-14 Equipment Surveillance Test Sheet No. 87-1-47, July 30, 1987, for a spurious high alarm in one channel of Containment Air Sampling Monitor (SPING) 1-RIA-PR40, correctly showed a Technical

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l Specification time limit of 14 days. The inspector reviewed i y Technical Specification Table 3.14-1, Action 25, which states l that-the inoperable channel must be restored to operable status within 14 days or the licensee must conduct a station review.

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The licensee's station review was performed on August 20, 1987, 21 ' days after PR-40 became inoperable. In a conversation with the NRR Licensing Project Manager (LPM), the inspectors learned that the intent of Action 25 is that fourteen days would.be the limit for either a return to operable status or completion of a station revie The licensee,- in the station review for this SPING inoperability, interpreted Action 25 as meaning that the station review only has to be done at some time after fourteen days of inoperability have passe 'j The NRR LPM indicated that in spite of the-intent, the licensee's actions met the requirements of Action 25. (See Appendix A for a complete discussion.) 4 No violations or deviations were identifie . Monthly Surveillance Observation (61726)  ;

The inspector observed and reviewed portions of the following test activities to ascertain whether testing was performed in accordance with .

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adequate procedures, whether test instrumentation was calibrated, whether limiting conditions for operation were met, whether removal and restoration of the affected components were accomplished, whether test results conformed with Technical Specifications and procedure require-ments and were reviewed by personnel other than the individual directing j the test, and whether any deficiencies identified during the testing '

were properly reviewed and resolved by appropriate management personne PT-10,SafeguardsActuationTest(Unit 1) .

- PT-11, Diesel Generator Loading Test (Diesel Generator 1A)

- PT-5, Reactor Protection Logic Tests, Reactor Normal or Shutdown (Unit 2)

- TSS 15.6.26, Control Rod System Checkout (Unit 2)

- TSS 15.6.52, Initial Criticality after Refueling and Nuclear Heating Level (Unit 2)

.- TSS 15.6,0, Flux Map Data Acquisition, Power Distribution and Incore/

Excore Axial Imbalance Checks (Unh. 2)

The inspector reviewed the completed test documentation for the following test activities:

- TSS 15.6.72, RTD (resistance temperature detector) Cross Calibration (Unit 2)

- TSS 15.6.57, Rod Drop and Timing Test (Unit 2)

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--Regarding TSS 15.6.52, " Initial Criticality after Refueling and Nuclear Heating Level," the inspector made the following comments:

Step 6.7 of the test procedure instructs the test personnel to

" verify" that the reactivity computer is ready.for use. However, the procedure for readying the reactivity computer is not controlled, in that there is no revision number, no date, and no signature showing approva *

The procedure for readying the reactivity computer requires a stopwatch, but there are no instructions in the procedure to require that a traceable, calibrated stopwatch be use The licensee agreed to consider procedure revisions based on these comment No violations or deviaticns were identifie . Monthly Maintenance Observation (62703)

Station maintenance activities on safety related systems and components were observed or revsewed to ascertain whether they were. conducted in accordance with approved procedures, regulatory guides, and-industry codes or standards and in conformance with Technical Specification The following items were considered during this review: whether the limiting conditions for operation were met while components or systems were removed from service, whether approvals were obtained prior to initiating the work, whether activities were accomplished using approved procedures and were inspected as applicable, whether functional testing and/or calibrations were performed prior to returning components or systems to service, whether quality control records were maintained, whether activities were accomplished by qualified personnel; whether parts and materials used were properly certified, whether radiciogical controls were implemented, and whether fire prevention controls were implemente Work requests were reviewed to determine the status of outstanding jobs and to assure that priority is assigned to safety-related equipment maintenance which may affect system performanc The following maintenance activities were observed or reviewed:

- Replacement of primary and redundant heat trace coils for circuit 51014 (see paragraph 4)

- Replacement of 3L fuel oil pump for IA diesel generator (see paragraph

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- Replacement of Unit 2 ESF relays AFP3/XB and AFP1/XB (see paragraph 6) Repair of motor operated valve 1 MOV-RH 87008 Following completion of maintenance on 1A diesel generator the inspector verified that the generator had been returned to service properly.

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On August 20, 1987, at 11:15 a.m. CDT during Unit 1 safeguards actuation tests, valve 1 MOV-RH 87008 failed to complete an opening stroke as called for in step 10-C.2, which specified that residdal heat removal (RHR) pump IB should start and that the valve should open. The opening of the valve was subsequently completed by remote actuation from the control room. The RHR pump was tripped from the control room, valve 1 MOV-RH 87008 was closed from the control room, and step 22-C.2 was run with the same results in that the pump started satisfactorily but the valve failed to complete an opening sequence. The valve was then left in the fully open position, which is the normal position for the injection phase of ECCS operatio On August 25, 1987,'1 MOV-RH 87000 was taken out of service. On August 26, 1987, the licensee realized that the out-of-service rendered the valve and its associated RHR system inoperable. Surveillance were commenced and the repair was expedited. The valve was returned to service later that same day after repairs and testing according to PT-28,

" Verification of Containment Recirculation Sump Valve Stroke, ECCS Continuity, and PORV [ power operated relief valve] Block Valve Test."

The safeguards actuation test (wherein the pump must start and the valve must open, simultaneously) was successfully repeated one day later, August 27, 198 Technical Specification 4.8.3.C requires that when one of the two RHR pump systems becomes inoperable, that the remaining RHR system, both centrifugal charging pump systems, both safety injection pump systems and the associated standby AC and DC power supplies be demonstrated operable immediately and daily thereafter. Loss of the auto-open feature of 1 MOV-RH 8700B rendered the valve and its RHR system inoperable as of August 20, 1987. The failure to perform immediate and daily surveillance in the period from August 20 to August 27, 1987, is a violation of Technical Specification 4.8.3.C (295/87017-03).

The following additional NRC concerns were identified: 1 Even though the licensee considered the valve operable following the August 20, 1987 failure, 1 MOV-RH 87008 was taken out of service for repairs on Tuesday, August 25, 1987, but the licensee did not recognize until approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later, on August 26, that Technical Specification surveillance were required for the )

out-of-service which removed power to the valve's motor operato *

Form PT-14," Inoperable Equipment Surveillance Tests " tracks Technical Specification (TS) required testing when TS equipment malfunctions. PT-14 No. 87-1-66 was closed on August 26, 1987, but WR Z-62185 shows that the Quality Control release was not given until & gust 28, 1987. This was identified in a Quality Assurance Audit finding dated September 4, 1987, as contrary to ZAP 3-51-1,

" Origination and Routing of Work Requests."

Several minor documentation discrepancies were also identified:

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The TS reference on the WR was different than from the TS reference on PT-14 No 87-1-66 (3.8.3.A,'versus 3.4).

Furthermore, the TS reference of 3.4 was too genera The upper. left corner of WR No. Z-62185 for valve 1 MOV-RH 8700B did not have the notation "PT-14" written in as required by ZAP 3-51-1, " Origination and Routing of Work

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The PT-14 No. 87-1-66 for August 20, 1987, has several blanks, i such as " Reason for 00S" and " Required Redundant Equipment."

The surveillance requirement for PT-14 No. 87-1-66 (August 20, 1987) was incorrectly shown as "None," and the " Time Limit" for both PT-14's (No. 87-1-66 and No. 87-1-70, August 25, 1987) was-incorrectly shown as "None."

PT-28 was not an adequate test for the purpose of declaring the-valve operable, since the test conditions for.PT-2B are not as restrictive as those for PT-10 when considering that the failure was due to the torque switch limiter plate being out of adjustmen PT-28 strokes valve 1 MOV-RH 8700B (RHR_ pump suction valve) with the RHR pump secured, while PT-10 strokes the valve open at the same-time the RHR pump receives a start signal. To adequately verify the correctness of the maintenance and to prove system operability, the appropriate test would have been with the RHR pump starting at the-same time the valve is stroking open (PT-10). This would have assured that the valve would not stop on high torque while a differential pressure exists across the valve dis The licensee's position was that, because the valve was in its safeguards position (open), that the valve should be considered operable. The inspector stated that since the valve could not perform one of its intended functions, that it was therefore inoperable. The inspector noted that the basis for this interpretation of operability is found in a previous enforcement action (see Appendix B).

The only circumstance for which 1 MOV-RH 87008 would not have been able to perform its design function would have been if it were called upon to open while the IB RHR pump was running. Since the valve is normally open ,

and was left in its normal position from August 20 through August 27, the ,

safety significance of the event is lo One violation and no deviations were identifie . Licensee Event Reports (LERs) Followup (92700)

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

action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications. The LER listed below is considered closed:

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UNIT 2- LER N0 DESCRIPTION 87001 Reactor Shutdown Due to Leakage of 0-Ring Level on Main Steam Isolation Valve Trip Solenoid Valve No violations or deviations were identifie . ' Training (41400)

During the inspection period, the inspectors reviewed abnormal events and unusual occurrences which may have resulted, in part, from training deficiencies. Selected events were evaluated to determine whether th classroom, simulator, or on-the-job training received before the event was sufficient to have either prevented the occurrence or to have miti-gated.its effects by recognition and proper operator action. Personnel qualifications were also evaluated. In addition, the inspectors determined whether lessons learned from the events were incorporated  ;

into-thetrainingl progra Events reviewed included the events discussed in this report. In addition, LERs were routinely evaluated for training impact. No events  !

reviewed this period were found to have significant training deficiencies i as contributors.- l Two events involving NRC inspector site access revealed several concerns regarding nuclear general employee training (NGET) and NGET card administration. In one event, an NRC Region III inspector had two NGET cards with conflicting information (one card from Zion and one from Dresden). In the other event, the Zion resident inspector was notified i that her site access for Byron Station (another Commonwealth Edison site) .

had been terminated. Investigation revealed the following concerns: (

. Training records were not readily accessible for the NRC Region III inspector who needed to resolve conflicting information on two NGET cards. Hard copy records had been removed from the training  ;

building (for microfilming), and the NGET trainer on duty did not  ;

have access to computer records. This was resolved by havino the i NRC inspector repeat training unnecessaril The licensee stated that it would keep the duty NGET instructor informed regarding which training staff personnel have access to  ;

the computer records. In addition, the licensee stated that all )

training staff members would be trained to use the computer record system and that this training is expected to be complete by the end of October, 198 '

NGET cards specify whether " restricted" or " unrestricted" access is authorized, but the Commonwealth Edison Production Training Department Instructor's Guide for NGET (4/86) does not explain the difference between restricted and unrestricted acces )

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In a phone conversation with Mr. R. Greger of the Region III staff, the inspector was informed that no distinction is required as stated in Region III's access authorization letters, which are provided to the company on a monthly basis. This information was discussed with Mr. S. Bell of Commonwealth Edison's Production Training Center (PTC) in order for the utility to make any changes to its company-wide NGET program which might be necessar *

NGET trainers do not have specific instructions in the company's NGET procedure issued by PTC regarding what to do with old NGET cards when giving site specific NGET trainin In addition, it is not unusual for persons who frequently travel to many Commonwealth Edison sites to have been issued NGET cards from more than one site. At Zion, an informal checklist is used to remind NGET instructors to ask all persons receiving NGET training to produce their old NGET card (s), if they have any. These are destroyed when new cards are issue In a conversation with Mr. S. Bell (PTC), the inspector was informed that the licensee would consider providing guidance to all NGET trainers which would require the destruction of old NGET cards when issuing new one Administration of site specific NGET differs between site For example, when the basic NGET and site specific NGET training dates are not the same, site specific retraining dates (by which, if retraining is not performed, access is terminated) are computed differentl One site would have required retraining within a year of the last site specific training, whereas another site would have required site specific retraining within a year of the 4 last basic NGET trainin The licensee stated that this policy was revised on April 30, 1987, in a report of a meeting of NGET instructors. The current policy is that site specific NGET expires with the basic NGET and must be 4 renewed following the renewal of basic NGET. The report also stated q that the instructor's guide would be revised to reflect this chang i Resolution of these concerns is considered an Open Item pending determination by the licensee of the proper NGET practices (295/87017-04; 304/87019-03).

No violation or deviations were identified. One Open Item was identifie . August 26, 1987 Site Visit by Branch Chief (30702)

On August 26, 1987, a site visit and tour was conducted by the resident inspector for Messrs. W. Forney, Chief, Reactor Projects Branch 1, and J. Hinds, Chief, Reactor Projects Section 1A. Following discussions with

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the resident inspectors and the tour of the plant, a meeting was held ,

l with the station management and staff. Mr. Forney noted that there was !

I significant improvement in plant condition, as indicated by progress in I

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the licensee's painting program and a reduction in fluid leak Mr. Forney also noted that-the licensee still had much work to d Mr. P11m1 stated.that he agreed with that assessment, after which he reviewed several accomplishments made by the station this year, as well as performance data that is tracked by the station. Mr. P11m1 noted that several of the data indicated improved performanc Mr. Forney stated that the NRC would continue to monitor licensee performance, and would focus on personnel errors, equipment and plant performance, resolution of old regulatory issues such as open items and unresolved items, and work request backlo No violations or deviations were identifie . Followup of Region III Requests (92701)

The inspectors received a request from Region III by memo dated June 26, 1987, to review the licensee's program for monitoring ambient tempera-tures of areas which contain vital electrical equipment and instrumentatio The inspector determined that the only area which has a Technical Specification designated maximum ambient temperature limit is the con-tainment. Containment ambient temperatures are limited by Technical Specification 3.10.6 to greater than 65 degrees and less than 120 degrees and are required to be recorded shiftly by surveillance PT-0, Appendix Other areas which contain vital electrical equipment which may be-sensitive to high ambient temperatures include the auxiliary equipment rooms (AERs), 4KV vital switchgear rooms, unit battery rooms, unit computer room and diesel generator room Room temperatures for each of these areas, except for the battery rooms, are indicated in the control room; however, high temperatures are not alarmed or annunciate Licensee surveillance, PT-0, Appendix S, is performed on a shiftly basis and requires the equipment operator to note whether room temperatures deviate from " normal." The inspector noted that on two recent occasions, the shift engineer deemed it necessary to provide additional ventilation by means of a portable blower to the AERs because excessive room temperatures resulted in reactor protection or ESF bistables trippin The inspector reviewed PT-0, Appendix S, data sheets and noted that the temperatures of the Unit 1 and 2 auxiliary equipment rooms have been above the normal temperature range of 65 to 75 degrees since at least July 1, 198 No surveillance procedures were found which required recording the room temperatures in the diesel generator rooms, 4KV vital switchgear rooms or computer rooms. The inspector noted that the temperature of the unit battery rooms was recorded monthly during the monthly battery surveillance. No evidence of temperature trending was found.

No violations or deviations were identifie . Open Items Open Items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action I

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.e on the part of'the NRC or' licensee or both. .Two Open Items disclosed during this inspection are discussed in paragraphs 5 and 1 . Unresolved Items Unresolved ite'ms are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncompliance or deviations. Two Unresolved Items disclosed during this inspection are discussed'in paragraphs 4 and . Exit Interview (30703).

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The inspectors met with licensee representatives'(denoted in Paragraph 1)

throughout the inspection period and at the conclusion of the inspection on September 3, 1987, to summarize the scope and findings:of th inspection activities. The licensee acknowledged the inspectors'

comments. The inspectors also discussed the likely informational content'

of.the inspection report with regard to documents or processes reviewed by the inspectors during the. inspection. The licensee did not identify any such documents or processes as proprietar I l

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e APPENDIX A Interpretation of Technical Specifications Table 3.14-1, Action 25 Action 25 reads as follows:

"With the number of OPERABLE channels less than the minimum number required, restore the inoperable channel to OPERABLE status within 14 days or conduct a station review to determine a plan of action to restore the channel to operability."

In a phone conversation between Mr. J. Norris, NRR, and the resident inspectors, the following interpretation was provided:

The intent of Action 25 is that 14 days is the time limit for either restoration to operable status or completion of a station revie However, since Action 25 is not explicit enough to require the station review by the 14th day, the review may be done at any tim i s

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.f#"%~ APPENDgifED STATES 8" n NUCLEAR REGULATORY COMMISSION 4' E WASHINGTON, D. C. 20555

f Pill!CIPAL STAFF l g*****,# M l DRF i {

JUN 2 91983 ges i i

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$n!? I QJat 4 MEMORANDUM FOR: Charles E. Norelius, Director Dan i lac I f 5@n j j j Division of Project' and Resident Programs 51 ,

i 1 Region III [ [ l /

OL lF i LE! A FROM: Darrell G. Eisenhut, Director '

Division of Licensing, NRR

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, , SUBJECT: TECHNICAL ASSISTANCE TO DELINEATE APPROPRIATE LICENSEE ACTIONS FOLLOWING FAILURE OF CERTAIN ECCS/ CONTAINMENT ISOLATION VALVES REFERENCE: Memorandum from C. E. Norelius to D. G. Eisenhut, Dated May 2,1983, Subject: " Request for Technical Assistance to Delineate Appropriate 1icensee Actions Following Failure of Certain LCCS/

Containment Isolation Yalves."

The referenced memorandum requests NRR's position concerning licensee actions following the failure of certain ECCS/ containment isolation valves in regard to incidents which had recently occurred at Dresden Unit 3. In your memorandum, you raised five questions and offered the regional position on each of these five questions. Your questions and the regional positions are restated below and are followed by NRR's positio Question 1 .

The LPCI (1501-5D) valve has an automatic open signal which opens the valve if closed when an ECCS automatic initiation signal is received. The automatic open portion of the valve was not operable because when the valve closed it was incapable of being automatically opened. Would the valve be considered. inoperable or operable with cnly the automatic open capability inoperable, and the valve was open with the capability of closing?

The regional position is that if any automatic function of an ECCS cr Containment Isolation becomes inoperable the valve is inoperabl The valve must be considered inoperable even if the automatic open function is not presently needed (e.g., valve open and automatic open fanction not operable).

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[ Charles E. Norelius, Directo '

M 2 91983 Response * - ' ~~ *

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We concur with the regional position on this matter. Such a position is consistent our intent as. expressed in our Standard Technical Specifi-cation definition of Operable-Operability which all power reactor licensees were requested to adopt via our generic letter of April 10, 198 It is our position that if any function of the LPCI valve is inoperable, the valve must be declared inoperable. Furthermore, since opening of the valve is required for the LPCI system to perform its

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. intended function, the LPCI system also must be declared inoperabl Question 2 When an ECCS/ Containment Isolation valve is declared inopera'ble, what is the correct action to be taken by the licensee? Should the valve be placed in the open position thereby maintaining the operable status of that ECCS loop but violating the Technical Specification definition of primary containment? Or, should the valve be placed in the closed position, and in that way maintain primary containment, but by the same action, enter a Limiting Condition for Operation _(LCO) for the affected system?

The regional position is th'at the licensee should take the most con-servative approach to the problem. The valve should be closed to maintain primary containment and the appropriate LCO should be entered. However, if only the automatic open function is inoperable the valve may be opened while keeping closure capability, declare the valve inoperable, and take the action of the LCO associated with an inoperable ECCS syste Response Our position on this question is that the licensee should follow the requirements of the applicable technical specification. In the event the applicable technical specification does not provide explicit criteria for positioning an inoperable ECCS/ Containment Isolation valve, we would generally concur with the regional position that the valve should be closed so as to maintain containment integrity. Furthermore, the ECCS loop should be declared inoperable and its action statement complied with. In the case where such a valve was inoperable solely as a result of being unable to automatically open, we would consider it acceptable to maintain the valve in an open position provided the ECCS loop wa:

declared inoperable, its action statement was complied with, and the valve was capable of being closed by an automatic containment isolation signa I

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o ' Charles E. Norelius, Director- -3-JUN 191983

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Question 3 Once an ECCS/ Containment Isolation valve is declared inoperable and

'the valve is. then placed in a' designated configuration (either open or. closed), should the valve be electrically deactivated to prevent a possible automatic-initiation from causing the valve configuration to change? . (An example would be closure of the 1501-5D valve and electrically deactivating it to prevent an automatic initiation of -

LPCI to. reopen the valve). Should an automatic initiation, in this case, be intentionally. byyassed?

The regional position is that the valve should remain open if remote closure is possible. If this is not the case, the valve should be _

closed to pruvide containment isolation and it should be electrically disarmed to prevent reopening. In either case, the appropriate LC0 for inoperable ECCS system should be entere .

Response If, as- discussed in our response to your Question 2, an inoperable valve is closed to maintain containment integrity, it is our position that this . valve should be electrically deactivated to preclude its subsequent

. inadvertent actuation. However, if a valve is inoperable and is being maintained in its open position in accordance with the criteria given in our response to your Question 2,-it is our position that this valve should not be electrically deactivated since it would then be incapable of closing to provide containment isolation. Furthermore, we do not ,

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believe that automatic initiation of the ECCS loop should be bypasse . Question 4  !

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Since the valve was not in the Technical Specifications, the associated action statement was not apparent to the licensee. Should ECCS pump suction valves (LPCI and Core Spray) be included in the Technical Specifications to identify the action to be taken when they become inoperable? If this is affirmative, Table 3.7.1 of the Dresden Unit 2 and 3 Technical Specifications, should be changed to include the valve The regional position is that these valves should be included in Technical Specifications and appropriate action statements should be adde _ _ _ _ _ - __

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O Charles E. Norelius, Director -4-p2 1333 p

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Response '

As part of the ECCS systems, these valves would not normally be listed separately since the definition of Operable-Operability (which all power reactor licensees were requested to adopt via our generic letter of April 10, 1980) would require these valves to be operable in order _for the ECCS systems to be operable. However, since these valves are considered as part of the boundary for containment isolation, it is our position that they should be included in Table 3.7.1, Primary Containment Isolation, of the Dresden Unit 2 and Unit 3 Technical Specifications and we will request the licensee to submit a licensee amendment to add them to Table 3. ,

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The SEP topic recommendation was that appropriate procedures for operator action should be provided. The licensee had not issued these procedure Under what conditions would these valves be required to be closed?

The regional position is that these valves should be closed in the event of excess leakage, or if the capability of closing the valve remotely becomes inoperabl ,

Response As previously stated in our response to Question 2, our position is that the licensee should follow the requirements of the applicable technical specificatio In the event the applicable technical specification does not provide explicit criteria for positioning an inoperable ECCS/Contain-ment Isolation valve, we concur with the regional position that, in general, the valve should be closed so as to maintain containment integrity. Further-more, the ECCS loop should be declared inoperable and its action statement complied with. In the case where such a valve was inoperable solely as a result of being unable to automatically open, we would consider it acceptable l to maintain the valve in an open position provided the ECCS loop was declared inoperable, its action statement was complied with, and the valve was capable of being closed by an automatic containment isolation signal. However, there may be situations which can arise which will dictate different actions be )

taken concerning the disposition of these valves and they should be addressed j on a plant specific basis by the license 'f y v en wuu Darrell G. E1senYut, Director Division of Licensing l

Office of Nuclear Reactor Regulation l l

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