IR 05000445/1990040

Safety Insp Repts 50-445/90-40 & 50-446/90-40.Violations Noted But Not Cited.Major Areas Inspected:Plant Status, Operational Safety Verification,Onsite Followup of Events, Followup on Fastener Testing for Bulletin 87-02

ML20058H554
Person / Time
Site:	Comanche Peak
Issue date:	11/13/1990
From:	Chamberlain D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20058H552	List: IR 05000445/1990040 ML20058H550
References
TASK-***, TASK-TM 50-445-90-40, 50-446-90-40, IEB-87-002, IEB-87-2, NUDOCS 9011210015
Download: ML20058H554 (20)

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APPENDIX V.S. NUCLEAR REGULATORY COMMISSION Region IV NRC Inspection Report: 50-445/90-40 50-446/90-40 Dockets: 50-445 Unit 1 Operating License: NPF-87 50-446 Unit 2 Construction Permit: CPPR-127 Expires: August 1, 1992 Licensec: TU Eiedrir: ,

Skyway Tower

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400 North Olive Street Lock Box 81 Dallas, Texas 75201 Facility Name: Comanche Peak Steam Electric Station (CPSES), Units 1 and 2 Inspection At: Glen Rose, Texas Inspection Conducted: Septc.9er 18 through 9ctober 30, 1990  :

Inspectors: W. D. Johnson, Senior Re:ident Inspector R. M. Latta, Senior Residei:t Inspector S. D. Bitter, Resident Inspector D. N. Graves, Resident Inspector W. M. McNeill, Reactor Inspector L. D. Gilbert, Reactor Inspector D. L. Garrison, Reactor Inspector

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ll13-90 D. D.(Chamberlain, Chief, Project Section 8 Date Division of Reactor Projects <

Inspection Summary In_syection Conducted September 18 through October 30, 1990 (Report 50-445/90-40; E0 446/90-40)

Areas Inspected: Unannounced resident safety inspection of plant status, operational safety verification, onsite followup of events, followup on fastener 'esting for Bulletin 87-02, maintenance observation, surveillance observation, followup on TMI action items, licensee event report followup, followup on previously identified items, and Unit 2 activities, h1210015901114

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4 4 g-2-t Results: Plant performance was improved during this inspection period compared ,

to the previous inspection period. Maintenance and surveillance activities '

observed on Unit I were properly cenducted. However, as discussed in paragraph 7, it was observed that improved coordination was needed in order to complete a complex surveillance test within the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> allowed by TS for having one channel bypassed. The licensee's control of the Unit 2 Train A emergency diesel generator overhaul was considered to be a strength. Licensee response to TMI action items and previously identified inspection items was ,

appropriate. Two noncited violations were identified in paragraphs 3.a and The first of these involved improper storage of equipment in the plant and the second was a licensee identified finding that the setpoints for the containment air Cooler condensate flow rate instruments were nonconservative. One inspector followup ite.n was identified for further review in paragraph 12.a regarding

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licensee disposition of connecting rod blue check results on Unit 2 EDG .

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DETAILS persons Contacted

'J. L. Barker, Manager, Independent Safety Evaluation Group (ISEG)

• J. W. Beck, Vice President, Nuclear Engineering O. Bhatty, Issue Interface Coordinator

• M. R. Blevins, Manager of Nuclear Operations Support

• H. D. Bruner, Senior Vice President

• R. C. Byrd, Manager, Quality Control (QC)

• W. J. Cahill, Executive Vice President, Nuclear

• C B. Corbin, Licensing Engineer

C. K. Feist, Design Basis Engineering Consultant Engineer

• S. P. Frantz, Newman and Holtzinger

• G. Guldemond, Manager of Site Licensing

• J. C. Hicks, Unit 2 Licensing Manager

• C, B. Hogg, Chief Engineer C. W. Killough, Procurement Quality Assurance (QA) Manager R. M. Kissinger, Manager Civil Structural R. N. L111eston, Project Engineering Administration

• D. M. McAfee, Manager, Quality Assurance (QA)

• J. W. Muffett, Manager of Project Engineering

• E. F. Ottney, Project Manager, Citizens Association for Sound Energy (CASE)

• S. S. Palmer, Stipulation Manager

• D. E. Pendleton, Unit 2 Assistant Project Manager F. L. Powers, Procurement Engineering Manager

• C, W. Rau, Unit 2 Project Manager

• A. B. Scott, Vice President, Nuclear Operations

• J. C. Smith, Plant Operations Staff

• P. B. Stevens, Manager of Operations Support Engineering J. A. Taylor, Procurement Evaluat ion Supervisor i J. E. Thompson, Civil / Structural Engineer Unit 2

• J. R. Waters Site Licensing

• Present at the exit intervie In addition to the above personnel, the inspectors held discussions with various operations, engineering, technical support, maintenance, and administrative members of the licensee's staf . Plant Status - Unit 1 (71707) ,

At the beginning of_this inspection period, the plant was in Mode 1 l proceeding to 100 percent power, which was attained on September 18, 199 On September 22, a heater drain pump discharge valve, FCV-2492, failed closed. The loss of heater drain flow caused the operators to reduce power and the plant was stabilized at approximately 80 percent reactor :

power. On September 23, the plant was placed in Mode 2 to perform repairs to the Nos. 2 and 3 main feedwater (MFW) flow control valves (FCVs). The

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-4-repairs were completed and the plant entered Mode 1 later on September 2 The main generator was placed on line using only one generator output bro ker (CB-8010). A main generator and subsequent turbine trip occurred while transferring electrical loads from the station transformer to the unit auxiliary transformer. The generator trip was caused by an indicating switch failure on a generator output breaker. A reactor trip did not occur because reactor power was only 37 percent (below P-9 setpoint of 50 percent). Reactor power was reduced to approximately 5 percent while repairing the failed limit switch on the generator output breaker. The turbine generator was synchronized to the grid on September 24 and reactor power was increased with a 45 percent administrative power limitation due to continuing repair work on the second generator output breaker. On September 25, the generator output breaker repair was complete and power

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was increased to approximately 100 percent. Reactor power was reduced to approximately 78 percent on September 26 as a result of excessive vibration / oscillation of the Nos. 2 and 4 MFW FCVs. On October 1, the load dispatcher requested that the plant be maintained at the current load (approximately 880 MW). On October 15, the plant load was reduced to approximately 67 percent reactor power at the load dispatcher's reques On October 25, the plant load was increased to approximately 79 percent reactor power at the load dispatcher's request. Reactor power was increased to 100 percent on October 29 to collect data for the rod control reactor coolant system (RCS) average temperature control program. At the end of this reporting period, the reactor was at 100 percent powe . Operational Safety Verification (71707)

The objectives nf this inspection were to ensure that this facility was being operated safely and in conformance with regulatory requirements, to ensure-that the licensee's management controls were effcetively

discharging the licensee's responsibilities for continued safe operatio to assure that selected activities of the licensee's radiological protection programs are implemented in confirmance with plant policies and procedures and in compliance with regulator,i requirements, and to inspect the licensee's compliance with the approved physical security pla .The inspectors conducted control room observations and plant inspection tours and reviewed logs and licensee documentation of equipment problem Through in_ plant observations and attendance Of the licensee's plan-of-the-day meetings, the inspectors maintained cognizance over plant status and Technical Specifications (TS) action statements in effec The following paragraphs provide details on certain concerns and issues identified during this inspection perio _ Storage of Norplant Equipment During plant tours, the inspector noted two cases in which nonplant equipment storage did not comply with the requirements of Procedure STA-661, "Nonplant Equipment Storage and Use Inside

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Seismic Category I Structures." An argon cylin ter near an auxiliary feedwater valve (1-HV-2492B) was not secured in place and a ladder was propped near the reactor trip breakers. Ne ther of these items was in a designated safe zone. The licensee promptly corrected these e discrepancies upon notification. In addition, .he licensee performed an audit to t.ssess compliance with STA-661 in v.4rious plant area This auvit identified three additional violations. Technical support personnel conduct partial plant walkdowns week 1/ such that all areas outsioe containment are inspected for STA-661 compliance every 5 weeks. The licensee's failure to comply with procedural i recairements for proper storage of equipment in the plant is a v'olation of TS 6.8.1. However, in accordance with Section V.A of the NRC's Enforcement Policy, no citation is being issued for this '

violation, b. Separation of Untested Portion of System .

During the evaluation of the licensee's tag-out program, the inspector determined that although the chemical and volume control system (CVCS) had been accepted during the system turnover process, portions of this system for boric acid transfer had not been iydrostatically tested or certified in accordance with the ASME N-5

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certification process. These portions are not presently required to support Unit 1 operations. Subsequent to this determination, several meetings with the licensee's technical support staff and the CVCS system engineer were conducted to determine the operational status of the CVCS and the controls provided by the turnover program as defined in Station Procedure STA-802. As a result of these meetings and reviews of the applicable system clearances, equipment records, and Section 9.3.4 of the FSAR, it was determined that the CVCS system, which has some portions shared % L it I and Unit 2, was accepted by operations during the 802 process in October 1989. Furthermore, it was established that following this turnover process, one of the

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t Unit 2 boric acid transfer pumps had been operated, however, none of the untested portions of the CVCS had been used to support Unit 1 operations subsequent to the issuance of the low power license issued on February 8, 1990. Accordingly, no regulatory concern relative to the inappropriate use of this system exists at this tim The inspector walked down the CVCS to check the status of valves outside containment and required to be sealed in position by t Procedure ODA-403. All 36 valves were sealed in the proper positio The discharge isolation valve for boric acid transfer Pump 2-02 was closed under Clearance 2-89-00475. The locked valve list was found to be laid out well, grouping valves by location. The inspector noted that seals and tags were attached to the remote operators for valves and that the location of the reme'.t operators was generally a-different room than the one listed for valve location in the locked valve list. One typographical error was noted. Three valves listed

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1r.:ated on the 852-foot elevation of the auxiliary buildin Penetration Room 88 was noted to be warm and humid due to steam generator blowdown system leaks. These comments were given to licensee personnel who initiated corrective actio As determined by the inspectors, the licensee's administrative control system defined in Procedure ODA-403, " Locked Valve List," ,

appears to be adequate to prevent the inadvertent use of the untested portion of the CVCS, In addition, the licensee plans to certify the balance of the system prior to use for support of Unit 2 licensin ,

The inspectors concluded that, with the exception of certain nonplant equipment storage, the facility was being operated in accordance with plant procedures and regulatory requirement . Onsite Event Followup (93702)

The inspectors reviewed licensee action on NRC Information Notice 90-6 The licensee's review of NRC Information Notice 90-64 found that the centrifugal charging pump (CCP) suction line vent valves at CPSES could possibly unseat under differential pressure as described in the Notic Following a safety injection signal, volume control tank presNre could unseat the valves (1-HV-8220 and 1-HV-8221) and allow gas to flow to the suction of the CCPs. This would create a potential for gas oinding of the ,

pump Compensatory action was promptly taken to close man. sal isolation valves at the suction of each pump and venting the suction lines dail This action isolated the CCP vent lines from the valves it, question and removed the potential for CCP gas binding after safety irjection. The licensee is evaluating long-term corrective action for this proble The inspectors will complete review of this matter following receipt of Licensee Event Report 90-035-0 . Iollowup on Temporary Instruction (TI) 2500/27 Fastener Testing for NRC Dulletin 87-02 (25027)

The purpose of this inspection was to evaluate the adequacy of the licensee's root cause analysis and corrective actions associated with NRC i Bulletin 87-02. In response to the bulletin, the licensee tested 53-fasteners (33 safety-related [SR] and 20 nonsafety-related (NSR]). The testing established that four SR and five NSR fasteners deviated from some specification requirements. In regard to the SR fasteners, because the specifications allowed precedence of other requirements or arbitration of test results with othat testing, the fasteners were found to be acceptable to the specification In regard to the NSR fasteners, the slight deviatiors of chemistry requirements (one bolt and two nuts) and the hardnest, deviations (two nuts) were found to be acceptable by an enginerring evaluation. As noted in the TV response to the bulletin (TXX-88223 dated February 11,1988) the SR fasteners were found to be acceptable to the specification and the NSR fasteners were found to be acceptable by an engineering evaluation. The bulletin testing did not

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result in issuance of nonconformance documentation because the SR fasteners were found to meet the specification. A nonconformance document was not required for NSR material test results because the material performed no safety-related function and jeopardized no safety-related ,

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system. The licensee did not perform a root cause analysis and establish corrective actions because it was determined, based on the above test results, that there was no proble The licensee has since established corrective actions in regard to procurement of SR and NSR (dedicated materials) because of three factors which came together at the same time. The three factors were the bulletin, the licensee experience with a vendor of suspect material, and industry initiative The licensee has established a procurement program that follows the direction given in " Guideline for the Utilization of Commercial Grade Items in Nuclear Safety Related Applications (NCIG-07)." ,

NSR materials which are to be commercial grade materials dedicated by TV Electric for SR applications are planned to be procured by " Method 1," +

which requires product inspection and testing upon receipt. In addition to the above, SR fasteners are to be procured from an approved vendor with certification of the material and sample product testing after receip This new procurement program was found te be defined in Materials '

Management Procedures MMO 6.02-01 through MMO 6.02-06 and Nuclear Quality Assurance Procedures 6.02, 3.14, and 3.09-11.0 The inspectors found that the corrective actions taken by the licensee associated with NRC Bulletin 87-02 appear to be adequat . Monthly Maintenance Observation (62703) .

Station maintenance activities for the safety-related and nonsafety systems and components listed below were observed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, and industry codes or standards, and in conformance with the TS.

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Repair of Valve 100-0087, a Train B emergency diesel generator (EDG)

lube oil strainer drain valve (Work Order C90-5067). i

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• Removal'of temporary strain gauge bolts and the installation of the permanent bolts on the Train B EDG turbocharger (Work Order C90-2137).

Repair of EDG air receiver check Valve 100-0060(WorkOrderC90-4361).

Replacement of the Train B EDG air starting Valve 100-0275(Work OrderC90-5064). The inspector observed that teflon tape was present on the fittings connecting the stainless steel tubing from the solenoid to the air start valve inlet. The valve being removed also had teflon tape in the same locations. This condition was pointed out to the mechanics and the QC representative presen ,

Inspection of the other air start valves on the Train B EDG by the inspector

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. indicated that teflon tape was present on 100-0276 and 100-0277. The work order was revised and the teflon tape was removed from all three valve *

Performance of a valve stroke length measurement on 1-PV-2325, a steam generator atmospheric relief valve (Work Order P90-6190). The technicians stroked the valve twice, measuring the stroke length with a dial indicator. Both measurements were read as 1.47 inches by the two technicians and the stroke length was declared satisfactory. The inspector observed the reading as 1.37 inches prior to the dial '

indicator being removed and questioned the technicians as to how the 1.47-inch reading was determined. One of the technicians explained that the indicated reading needed to be adjusted for the amount the zero position differed from true zero. The dial indicator was operated manually by the technicians while observing its respons The technicians decided that the dial indicator had been indicating properly and 1.37 inches was the correct reading and that the valve L stroke was incorrec It was determined later that the valve position limit switches affect the position of the platform from which the dial indicator is referenced. Moving the limit r, witch arms away from the valve stem connector plate and measuring the valve stroke indicated that the -

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Pressurization of the containment airlock for the leak rate test performed to test the integrity of 185-0016 -0017, -0037, -0038,

-0039, and -0040. The test was satisfactor *

Valve stroke length measurement of 1-PV-2327, a steam generator atmospheric relief valve (Work Order P90-6192). The recorded measurement was 1.509 inches (twice measured at 1.51 and once at1.507). The upper limit is 1.500 inches. The results were turned over to engineering for evaluation. Operations Notification and '

Evaluation Form (ONE) FX 90-2217 was written to document the l occurrence. Technical Evaluation WC-90-2570 was performed and the determination was made that the valve was within its acceptable flow l range even with the additional stroke length.

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Valve stroke length measurement of 1-PV-2326, a steam generator 4 atmospheric relief valve (Work Order P90-6191). The valve's stroke length was satisfactor *

Changing of the No. I and No, 2 steam generator blowdown filters (Work Orders C90-4944 and C90-4950, respectively). One retaining nut of one fastener on the the No. I filter was galled and could not be torqued properly. This condition was noted in the work order. No leakage was observed. An auxiliary operator was present to operate associated valves. QC and radiation protection personnel were also presen _ _

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Inspection, cleaning, and lubrication of positioners on feedwater FCV !

(1-FCV-0510,WorkOrder 90-6058).

Maintenance activities observed de,ing this inspection period were acceptably performed by qualified personnel using adequate procedures and administrative controls, j Monthly Surveillance Observation (61726) l The inspectors observed the surveillance testing of safety-related systems and components listed below to verify that the activities were being performed in accordance with the TS. The applicable procedures were ,

reviewed for adequacy, test instrumentation was verified to be in l

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calibration, and test data was reviewed for accuracy and completenes The inspectors ascertained that any deficiencies identified were properly '

reviewed and resolve The inspector witnessed portions of the following surveillance test activities: j

Actuation test of Train A safeguards slave Relay K601 (Work Order $90-1891, Procedure OPT-463A).

Operability test of the turbine driven auxiliary feedwater pump (Work '

Order S90-2447, Procedure OPT-206A).

Digital channel operational test (DCOT) on the containment atmosphere particulate and gas radiation monitors in accordance with

Procedure INC-7096A (Work Order S90-2563).

Operability test of motor driven auxiliary feedwater pump (Work Order $90-2564, Procedure OPT-206A).

Train B residual heat removal (RHR) pump operability test (Work

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Order S90-2365).

EDG operability test (Work Order S90-2478, Procedure OPT-214A).

Train B solid state protection system (SSPS) logic actuation test of-October 26,1990_ (Work Order S90-2425). A number of limiting conditions for operation (LCOs) are entered during this procedure as a result of the entire Train B of SSPS being out of service. The most limiting action requirement requires that the pint be placed in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing. Two hours following entry ,

into the LCO action requirement, the surveillance was not. complet Plant management was notified that a power rec':ction would be performed to comply with the action requirement. Before any power i redtetion was performed, the procedure was terminated and SSPS was rettrned to an operable lineup and the LCO action requirements exited. The entire surveillance procedure was performed sat sfactorily on the following shif *W

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Emergency airlock leak test (Work Order 590-1854, Procedure EGT-716A)

performed on October 9, 1990. Du*ing the performance of ti.ose steps that called for pressurizing the airlork and establishing a stable airlock leak rate and pressure (Steps A4-2.1.2.7 and A4-2.1.?.8), a rush of air was heard and felc exiting the outer door equaliz og valve port. Immediately the shift supervisor was notified and the LCO for TSC 3.6.1.3 was entered. The leak rate was then quantifie Later, troubleshooting by mechanical maintenance personrel revealed that a combination of problems were the cause of the excessive leakage. Using corrective maintenance Work Order C90-6573, the flanges on the equalizing valve were tightened, and the linkage from the equalizing value to the outer door was adjusted. Subsequent retesting indicated that the airlock had a satisfactory leak rate.

The licensee initiated a ONE Form (FX-90-2293) which will address the loose flanges and misadjusted linkag The surveillance tests observed were conducted by qualified personnel using adequate procedures but the Train B SSPS logic test needed improved coordination in order to complete it within the allowed 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> . TMI Action Items (25565) As addrosed in various NRC inspection reports, the following TMI action items were previously closed on Unit 1. During the present inspection period, these items were eviewed for applicability to Unit 2. Based upon the review, and based upon similarities in design, programs, and procedures between Units 1 and 2, the following ;

items are closed for Unit 2: l Unit 1 Closing Item N Title Report N ,

t I.A.1. Shift Technical Advisor (STA) 50-445/89-17 on Duty 50-446/89-17 I.A.1. Shift Technical Advisor 50-445/89-37 Training 50-446/89-37 1.A. Shift Supervisor Administrative 50-a45/89-37 Duties 50-446/89-37 I.A.1. Limit Overtime for Operators 50-445/89-17 50-446/89-17 I.A.1.3. Minimum Shift Crew Complements 50-445/89-17 50-446/89-17

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I.A.2. Upgrade of Peactor Operator and 50-445/89-17 Senior Reactor Operator Training 50-446/89-17 Qualifications / Modify Training

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I. Shift Relief and Turnover 50-445/89-86 ,

Procedures 50-446/89-86 I. Shift Supervisor Responsibilities 50-445/89-02 50-446/89-02 I. Control Room Act ,' Establish 50-445/89-37 -

Authority to Lim'i ' Access 50-446/89-37 I. Procedures for Feedback of 50-445/89-37

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Operating Experience to Plant 50-446/89-37 Staff  ;

I. Verify Correct Performance of 50-445/89-02 Operating Activities 50-446/89-02 I. Westinghouse Review of Low Power 50-445/89-72 Test Procedures, Power Ascension 50-446/89-72 Procedures, and Emergency Procedures II.B. Interim System for Post-Accident 50-445/89-09 Sampling 50-446/89-09 II.B.4. Develop Training Program for 50-445/89-67 Mitigation of Core Damage 50-446/89-67 II.B.4. Complete Initial Training for 50-445/90-07 Mitigation of Core Damage 50-446/90-07 111 D. Control Room Habitability / 50-445/90-02 Modification 50-446/90-02 b. (Closed) TMI Action Item I.D.1, " Control Room Design Review." This item had remained open pending resolution of three outstanding issues that were described in NRC Inspection Report 50-445/90-02; l 50-446/90-02. During the present inspection period, the NRC

, inspectors reviewed the licensee's actions taken to address these remaining issues. The issues, and the actions taken to address each,

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are addressed as follows:

(1) High ambient temperatures in the remote shutdown area. The l

licensee performed survey measurements of several environmental parameters (temperature, humidity, etc.) on June 5 and 6,199 The results were submitted to the NRC staff in TV Electric's letter (TXX-90251) dated July 17, 1990. The inspectors reviewed the results and were satisfied with the licensee's conclusion i

that this issue is closed.

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-12-(2) Noise level and lighting concerns in the control room. The l licensee evaluated the lighting glare problem. Several corrective actions were implemented to reduce the effects of the glare to an acceptable level. Furthermore, the licensee committed to assessing the source of the glare and evaluating conceptual design changes for replacing some of the flush mounted fluorescent lamp fixtures. In all cases, the glare was found to be minimal. The NRC inspectors found that the licensee's actions were satisfactory and sufficient to consider this issue close (3) Review of emergency operating procedures by the licensee to ?

verify that the parameters specified in the proceduras are consistent with actual display parameters (e.g., ter.ninology and labeling). The licensee has completed this review. The results were reported to the NRC via TV Electric's letter (TXX-90137)

dated April 12, 1990. The inspectors reviewed these results and were satisfied with the licensee's conclusion that this issue is close Because the licensee has completed all requirements, the inspectors considered this item closed (Unit 1 only).

c. (Closed) TMI Action Item I.D.2, " Safety Parameter Display System."

Supplement 22 (SSER 22) to the CPSES Safety Evaluation Report (NUREG 0797) documented the NRC staff's review of the licensee's actions taken to meet the-eight requirements for the safety parameter display system (SPDS). The review concluded that the licensee had not satisfied two requirements:

(1) Radiation status was not continuously displaye i (2) The licensee had not demonstrated that the SPDS could aid operators in rapid, ic11able determination of plant safety status because a 30-day vailability test had not yet been run on the SPD In SSER 22, the NRC staff accepted the licensee's commitments to address both of these issues. The staff concluded that, based on the licensee's commitments, all SPDS issues had been satisfactorily resolved.

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During the. present reporting period, resident inspectors reviewed the status of the two commitments. Specifically, the inspectors determined that the licensee's commitment tracking system contains a commitment (No. 23160) that calls for modifying the SPDS design to include " radiation control"~as a top-level continuous display. This is to be accomplished by the end of the first refueling outage. The inspectors also determined that the commitment to perform a 30-day availability test for the SPDS has been met. The licensee reported the results of the test in TV Electric's letter (TXX-90234) dated

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-13-l July 10,1990. The inspectors reviewed the test results and agreed with the licensee's conclusion that the SPDS has been demonstrated as being reliabl In c:,iiciusion, all requirements, except that of continwously rNnitoring radiation ievels, have been met. This exception is addressed via the licensee's commitment tracking syste Therefore, this action item is closed (Unit 1 only).

d (Closed) TMI Action Item III.A.I.2, " Upgrade Emergency Support Facility." This action item encompasses three a: tion items, I!!.A.1.U.1.A III.A.1.2.1.B. and III.A.1.2.3. Collectively, these three address the requirements to upgrade the emergency response

facilitie Supplemeat 22 (SSER 22) to the CPSES Safety Evaluation Report (NUREG 0797) documented the NRC staff's position ttat the licensee's emergency response facilities are adequate in accordan;e with NUREG 073) ret,uirements. The staff made this determination based on its review of tne CPSES emergency plan, the emergency pu n implementation appraisal, exercise observations, and onsite inspections. This item is closed (Unit I and Unit 2).

The licensee's actions on these items was responsive to the requ rements of the TW. -+ ion Plan. Only three items (I .C.1.1. I .C.1. and I. remain open for Unit 1. These items are expected to be closed at end of the first refueling outag . Onsite Fo. .owup of Written Report s of Nonroutine Events (92700)

The inspector reviewed the licensee event reports (LERs) listed below to determine whether corrective actions were adequate and whether response to the event was adequate and met regulatory requirements, license conditions, and licensee commitments, (Closed) LER 90-001-00, Source Range Flux Doubling Actuatio No reproducible cause of the event could be identified. The licensee believed the cause to be a momentary spike on a source range channel when its drawer was withdrawn from the cabinet during surveillance testing. The surveillance procedure was revised to require that the flux doubling signal be blocked prior to withdrawing the drawer. The related procedure for the other source range channel was similarly revisej. Other nuclear instrumentation surveillance procedures were re.iewed. This resulted in revisions to four other trocedures to ensure that the source range flux doubling switch was placed in the block position prior to withdrawing the instrument drawer. Other procedures involving reactor protective functions were reviewed to verify that the trip bistables are placed in " test" cr " block" in the proper _ sequence. This LER is close I

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i-14- (Closed) LER 90-016-00, Steam Generator Atmospheric Relief Valves Inoperable Due to Insufficient Stroke Length Setting }

The inspector reviewed the licensee's corrective actions related to this event. These included implementation of a design modification to increase valve stroke length and review of other calibration data sheets to ensure consistency with design requirement The inspector reviewed the stroke length measurement work orders (P90-6190, 6191, 6192, and 6193) performed in September 1990. Stroke length was acceptable for three of the valves but was slightly above '

the maximum of 1.5 inches for the fourth valve. ONE Form FX-90-2217 evaluation resulted in the maximum stroke length being increased to

1.513 inches. This LER is closed, (Closed) LER 90-017-00, Reactor Trip Due to FCV Solenoid Failure. A manual reactor trip was initiated due to low steam generator level following closure of an FCV. The licensee concluded that this valve failed closed due to rainwater intrusion causing failure of the associated solenoid coil. A junction box cover removed during maintenance permitted rainwater to enter a conduit leading to the solenoid. Corrective actions included replacement of the solenoid coil, inspection of solenoid assemblies on the other feedwater control valves, inspection for possible means of water intrusion,

. review of other outdoor components having the potential for a similar failure, review of the event with site personnel, and review and revision of work control programs to ensure consideration of detrimental environmental conditions during maintenance. This LER is closed, d, (0 pen) LER-90-031-00, Failure to Comply with TS Action Statement Due to Non-Conservative Alarm Setpoint. .This LER reported failure to comply with TS Action Statement 3.4,5.1 on August 22, 1990, when the i containment gaseous radioactivity monitoring nystem (CAGRM) was inoperable and the containment air cooler system (CARCS) condensate' '

flow rate instruments were inoperable. The licensee's review determined that the CARCS condensate flow rate instruments had been inoperable since initial licensing on February 8, 1990, due to nonconservative alarm setpoints. This issue was discussed in NRC Inspection Report 50-445/89-31. Compliance with TS 3.4.5.1 was achieved on August 24, 1990, and the proper CARCS alarm setpoints were installed on August 27, 1990. The LER stated root causes and identified corrective actions to prevent recurrence. These actions will be reviewed further during followp NRC review of the LER '

r The licensee's identification of the violation, reporting it to the NRC, and implementation of corrective actions meet the criteria of Section V.G.1 of the NRC Enforcement Policy (10 CFR Part 2, Appendix C) for non-citin Therefore, no citation for this violation will be issue >

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10. Action on previous Inspection Findings (92701)  ;

i t. . (Closed) Open Item (445/89200-0-01): TV Electric committed to augment facility management positions and operating shifts with experienced advisors and duty managers until the completior, of the power ascension test progra After completing the power ascension test program, the shift advisor and duty manager on shift programs were terminated on August 15 and September 1, 1990, respectively. The licensee determined that the objectives of these programs had been met and notified the NRC of their termination. The duty manager program was continued in a modified form. Duty managers serve for a week but do not stay on

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site during back shifts unless plant events indicate a nee Administrative Procedure STA-104 has been revised to reflect the deletion of the shift advisor progra This item is closed, (Closed) Open Item (445/9020-01): NRC review of the basis for the feedwater isolation valve minimum accumulator pressure of 2040 psi The inspector reviewed Calculation ME-CA-0000-221 This calculation determined the stroke times of the feedwater isolation valves (1-HV-2134, -2135, -2136, and -2137) under maximum operating stem thrust requirements and minimum nitrogen gas pressure (2040 psig) in the actuator. The results were:

Valve Stroke Time (sec.)

1-HV-2134 5.17 1-HV-2135 4.10 1-HV-2136 4.64 1-HV-2137 5.28 The licensee performed tests on June 14 and 15,1990, to determine main feedwater isolation valve stroke times with accumulator nitroger pressures of 2580, 2250, and 2040 psig. The stroke times were found to be relatively insensitive to accumulator nitrogen pressur For 2040 psig the results were:

Valve Stroke Time (sec.)

L 'l-HV-2134 4.48 1-HV-2135 3.55 1-HV-2136 3.98 1-HV-2137 4.61 These tests indicated that the TS required stroke time of 5 seconds was satisfied with accumulator nitrogen pressure at 2040 psig. The calculation indicated that the valves would close under actual flow

,. and differential pressure conditions within the safety analysis l assumed time of 6.5 seconds after considering instrument response i time of 0.878 seconds. This item is close '

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-16-11. Followup on Corrective Actions for Violations ( 9'. 702 ) ,

The inspector reviewed the licensee's response *.o the below listed violations to determine whether corrective actions were taken as stated, and whether response to the events was adequate and met regulatory requirements, license conditions, and commitments, (Closed) Violation 445/8986-01: This violation was issued as a result of identifying a number of modifications made to systems without having been processed as design changes or documented and evaluated as temporary changes, i

Corrective actions taken were revising the temporary modification

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procedure to more clearly define what constitutes a modification, performing a walkdown of plant systems to identify and evaluate items that may be system modifications, issuance of a new engineering procedure on system walkdowns that includes temporary modification identification, discussions with the shift supervisors and crews to increase their awareness that actions taken to mitigate extreme temperature effects on equipment may constitute a modification, and a directive to all managers and suoervisors emphasizing the need to ensure that temporary modifications are processed in accordance-within plant procedural controls. These actions have been complete This violation is considered closed, (Closed) Violation (445/9019-02): Failure to perform reactor coolant !

system iodine analysis following power change as required by T The inspector reviewed the licensee's corrective actions related to this violatio These included:

Counseling the unit supervisor involve l

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Placing caution placards on the main control boards,

Revising the test procedure to include reference to the special condition surveillance requirement,

Reviewing other procedures for similar weaknesses and revising ,

them as necessary,

Revising administrative procedures, and

Reviewing LER 90-015 with licensed operators, f Licensee actions were considered acceptable in response to this

, violatio This item is closed, i

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-17- (Closed) Violation 445/9019-03: Inadequate test procedur I

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The review criteria of Test Proce: lure EGT-327A, " Steam Generator i Atmospheric Relief Valve Capacity Test," related to valve capacity I was in error. The licensee's review of this violation determined that the engineering review and approval of this test procedure j involved a unique sequence of events leading to the procedure being inconsistent with the design basis document (DBD). The test  ;

procedure was revised and the capacity of the valves was determined to meet the DBD requirement Licensee Event Report 90-016-00 '

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discussed valve stroke length problems which were resolved. Other test procedures were reviewed to verify appropriate review criteri No similar discrepancies were identified. This item is closed, (Closed) Violation 446/9022-01: Wrong component - opened Unit 2 ,

reactor coolant pump seal water return filter instead of the Unit 2 :

reactor coolant filter which was authorized by a work orde The inspector reviewed DNE Form FX-90-179e-and die resultant plant i incident resolutio Corrective actions included discussion of the incident with maintenance work crews and development of a training ;

course on attention to detail. This training was scheduled to be provided to maintenance personnel annually. This item is close ; (Closed) Violation 445/9022-02: Procedural inadequacy that resulted i in the inadvertent automatic start of the auxiliary feedwater system i during slave relay testin Completion of corrective actions for this event were documented in '

the closing of LER-90-018-00, " Inadvertent Start of Auxiliary Feedwater Pump Due to Personnel Error," in NRC Inspection Report 50-445/90-26. This violation is considered close . Unit 2 Activities (50071. 50073, 50075. 71302)

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l I During this inspection period, routine tours of the Unit 2 facilit) sere I conducted in order to assess equipment conditions, security, and adherence i to regulatory requirements. In particular, plant areas were examined for evidence of fire hazards and installed instrumentation damage and to ,

l determine the acceptability of system cleanliness controls and general housekeeping. Additionally, the inspector conducted evaluations of existing plant programs for the preservation and maintenance of installed systems and components as well as the utility's preparations for the resumption of construction activities for Unit Unit 2 Diesel Generator Rework The inspectors evaluated the licensee's plans and programs for the rework of the Unit 2 Trains A and B EDGs, This rework resulted from recommendations made by the Owners Group Design Review and Quality Reva11datu Program (DR/QR), which was performed as a result of ,

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generic operational and regulatory issues involving Transamerica Delaval (TDI) diesel generators. These recommendations, which have been implemented for the Unit 1 diesel generators, specify the performance of. detailed inspections and the upgrading / replacement of various components. Additionally, the licensee plans to implement various corrective actions associated with 10 CFR Part 21 issues and lessons learned from Unit I diesel generator rework activitie In general, during their tours and inspections of the Train A EDG rework, the inspectors verified:

Housekeeping / cleanliness was adequat *

Adherence to safe working practices, including rigging operations, using protective gear (safety glasses, etc.), .

checking of confined spaces prior to entry, and controlling solvents used to clean the engin *

Labeling / proper storage of Q-component *

Proper tool / personnel accountabilit Specifically, the inspectors reviewed two major mechanical maintenance procedures being used for the rework:

MSM-CO-3830, Revision 0, " Emergency Diesel Engine Disassembly 1 and Assembly"

MSM-CO-3349, Revision 1, " Emergency Diesel Engine Pistons, Rods, and Rings Maintenance" Both procedures appeared adequate overall. However, MSM-CO-3349 did '

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not adequately. describe the process for blue-checking all of the connecting rods. Therefore, the licensee issued a change to-upgrade this procedure. The revised procedure included those attributes necessary for performing and inspecting the quality of the blue-check proces The inspec' tors also witnessed portions of the disassenibly and r cleaning of the Train A EDG, including the removal of the fuel injection pumps, cylinder heads, pistons, cylinder liners, and connecting rod assemblies. These activities, as well as the cleaning activities, were conducted in accordance.with MSM-CO-3830. No discrepancies were note Also, the inspectors observed selected nondestructive examination (NDE) tests. These included the eddy current testing of the connecting rod threaded joints and engine block, the replication testing of the engine block internals, and the blue-checking of the master and articulating rod interfaces for several connecting rod I

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-19-The inspectors noted that during the disassembly activities, parts were methodically removed, cleaned, packaged, identified, and store The manner of storage was suitable for Q-level components. It was also observed that, during the work activities, access to the work area was being adequately controlled by a security guard and access control personnel. Working together, these personnel ensured proper tool, material, and personnel accountability while maintaining adequate securit The inspectors placed special emphasis on the licensee's blue-checking of the connecting rods. These blue-checks are being performed on the coarse serrations (gear teeth) in the junction between the two mating halves, one of which is a portion of a master

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(main) connecting rod, and the other of which is a portion of a slave (articulating) connecting rod. In particular, the inspectors were nterested in the process by which the licensee determined the amount cf surface area contact between the two halves when they are joined and torqued together. According to vendor guidelines and a licensee commitment (PLN-5600), upon performance of a blue-check, the mating surfaces shall exhibit a minimum of 75 percent contact. The inspectors are closely monitoring the licensee's methodology and interpretation of the individual blue-check More specifically, one of the three connecting rods that have been tested, to date, resulted in an approximate value of 70 percent contact. Accordingly, the licensee documented this on a TV evaluation form. The resolution of this TV evaluation form will be reviewed during a future inspectio This will be tracked as Inspector Followup Item 446/9040-0 In general, the conduct of the overhaul process on the Train A, Unit 2, EDG by the Unit 2 startup and the mechanical maintenance organizations is judged to be a strength, Leveling of Unit 2 Nonsafety Switchgear Base The inspectors also observed construction activities associated with the leveling of the base for Buses 2A2 and 2A3. The inspectors determined that the work was being conducted in accordance with the governing procedures and that the personnel involved were properly trained and qualified to perform the work. Specific activities which were observed included: the switchgear removal, rail and anchor plate placement, welding of the anchor plates to the rails, shimming of the rails, level checking of the rails, and pouring of grou Based on these observations, it was concluded that this activity was accomplished satisfactorily, Storage of Q-Cable Reels During the conduct of routine Unit 2 plant and area walkdowns, the inspector identified several Q-marked cable reels which were staged with other non-Q construction material in the northeast corner of the

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protected are Subsequent to identifying this condition to the licensee, it was determined that adequate controls existed for the storage of Q-cable reels as defined in Construction Procedure ECC 6.08-7, " Control of Materials, Parts, and Components," and that the licensee had taken prompt action to enhance the storage of this material. Accordingly, no significant deficiencies were identified relative to this issu Expansion of Alternate Access Point On Sunday, October 21, 1990, the new, enlarged alternate access point (AAP) was activated. The expansion of the AAP was designed to facilitate the handling of anticipated craft personnel associated with Unit 2 construction activities. The expansion of this facility, coincident with the erection of the new materials staging building, previously identified in NRC Inspection Report 50-445/90-31; 50-446/90-31, are indicative of the enhanced project approach to Unit 2 construction activitie ;

Within the areas examined, no adverse findings were identified and the inspection results generally indicated that the observed activities were adequately performe . Exit Mee, ting (30703)

'An exit meeting was conducted on October 30, 1990, with the persons identified in paragraph 1 of this report. The licensee did not identify as proprietary any of the materials providert to, or reviewed by, the inspectors during this inspection. During this meeting, the NRC inspectors summarized the scope and findings of the inspectio l

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