IR 05000382/1998002

Insp Rept 50-382/98-02 on 980202-20.No Violations Noted. Major Areas Inspected:Current Performance & Effectiveness of Corrective Action Program Including Audits & Assessments, Operations & Engineering

ML20248L874
Person / Time
Site:	Waterford
Issue date:	03/20/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20248L869	List: IR 05000382/1998002 ML20248L867
References
50-382-98-02, 50-382-98-2, NUDOCS 9803240324
Download: ML20248L874 (30)

Text

_ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ - _ . ___ _ . _ _ .

.

.

ENOLOSURE U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

l l

l Docket No.: 50-382 l License No.: NPF-38 Report No.: 50-382/98-02 l

Licensee: Entergy Operations, Inc.

.

Facility: Waterford Steam Electric Station, Unit 3 Location: Hwy 18 Killona, Louisiana s l

Dates: February 2-20,1998

'

Inspectors: G. A. Pick, Senior Project Engineer

,

R. L. Bywater, Reactor Engineer i l

Accompanying l Personnel: J. C. Edgerly, Resident inspector Development Program Candidate Approved By: P. H. Harrell, Chief, Project Branch D ATTACHMENT: SupplementalInformation

!

l

'

i

l I

.

l I

9803240324 900320 PDR O ADOCK 05000382 PDR

- _ _ _ _ - _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ __ .-

- _ - _ _ _ _ _ _ - - _ _ - _ _ - _ _

.

I l

l"

! EXECUTIVE SUMMARY Waterford Steam Electric Station, Unit 3 NRC Inspection Report 50-382/98-02 l

The inspectors evaluated the current performance and effectiveness of the corrective action program, which included performance of audits and assessments. The inspectors also reviewed implementation of the operating experience program. Two separate technicalissues reviewed included the licensee evaluation of Potter Brumfield motor-driven relay anomalies and the appropriate containment spray system flow rate value used in the containment pressure accident analysi Ooerations

-

The licensee did not complete the evaluation of potential workarounds in operations surveillance procedures in a timely manner. This corrective action was part of a violation response and took more than 18 months to resolve. It took involvement by the inspectors for the corrective actions to be brought to conclusion (Section 08.1).

Enoineerina

-

Engineering response to Potter Brumfield motor-driven relay failures was good. The reported failures resulted in a 10 CFR Part 21 report being issued by the vendor. The licensee established an appropriate monitoring program to detect malfunctions prior to failure of the relays (Section M2.1).

-

The licensee identified a failure to control design inputs for the containment pressure accident analysis, which resulted in a noncited violation. The licensee credited'a higher containment spray flow than that achievable during originallicensing of the facility. An operability assessment demonstrated that a lower design flow would prevent exceeding peak containment pressure. The licensee implemented corrective actions for the identified condition. Further, the design basis reconstitution program that is established should identify similar design value discrepancies. The licensee continues to experience problems in the fidelity of design basis information (Section E1.1).

-

The corrective action program had improved since the last inspection of the program was completed. This was evidenced by an increased (Jentification of adverse conditions, management involvement throughout the process, and generally thorough resolution of significant condition reports (Section E7).

-

Overall, quality assurance personnel provided critical, effective oversight of the line organizations, as evidenced by the issues identified during the three audits and three assessments reviewed (Sections 08.2, E7.2, and E7.3).

-

The licensee performed very good assessments based on the identified findings and recommendations made in five assessments. The extent of the planned corrective l

l 1  :

i

!

!

- - - _ - - - - - - - - _ - - - 1

__ ____ . _ _ _ _ _ _ _ _ _ _ _ _ _

.

.

Entergy Operations, In l actions to the recommendations reflect a willingness by some managers at the facility to implement corrective actions to address recognized deficiencies (Sections E7.1 l and E7.2).

• The assessment program changes implemented during the last year resulted in greater accountability because of increased formality, which included creating a tracking system monitored at the corporate level and concurrence by the site Vice President of the planned corrective actions (Section E7.1.b.1).

+ The inspectors noted that line organizations did not respond to the recommendations in j quality assurance audits, as reflected in the corporate assessment database. However, '

the responses to assessments had been prepared and/or entered in the corporate assessment database as specified by their process (Section E7.2)

-

Engineering provided effective resolution of 13 condition reports reviewed with one exception. Personnelincorrectly credited the corrective actions for a different condition report when those corrective actions would not have corrected the identified deficiency (Section E7.3).

• Initially, the licensee concluded that an 18-month test of each core protection calculator channel was unnecessary following discussions with the Nuclear Steam Supply System vendor. Following inspector questions, the licensee reevaluated the need to perform the testing and determined that Technical Specification 4.3.1.1 required a test of all four channels each refueling outage (Section E7.3.b.5).

• A noncited violation resulted because engineers failed to perform a commercial grade dedication prior to installation of a nonsafety-related manual valve actuator on a dry cooling tower valve that had an active function to close. The licensee successfully completed a commercial grade dedication of the valve actuator. This deficiency resulted from poor interdepartmental communications (Section E7.3.b.14).

!

!

__ __ _ _ _ _ _ _ - _ _ _ _ _ _ _ - - _ _ _ _ _ _

..

.

Report Details Summarv of Plant Status During this inspection period, the plant operated at essentially 100 percent powe Miscellaneous Operations issues (92901)

08.1 (Closed) Insoection Followuo item 50-382/9724-03: Review of Operator Workaround Evaluations The inspectors initiated this item to follow up on the safety significance related to

,

potential operator workarounds in three surveillance procedures, which required engineering review. The specific items requiring review included: (1) evaluating whether all emergency feedwater flow paths must be verified through both sets of valves for each steam generator, (2) closing Valves CVC-216A test and -216B, pressurizer auxiliary spray isolation, to perform a downstream check valve test, and (3) evaluating the reason for closing Valves CC-125A, -1258, and -125AB, component cooling water discharge isolation, prior to securing the pum The licensee had developed a list of potential operator workarounds in June 1996 in response to the violations documented in NRC Inspection Report 50-382/95-23 and as a corrective action for Licensee Event Report 50-382/96-004. The corrective action also included resolving any items considered operator workarounds through the operations workaround program. The licensee had narrowed the potential workarounds to three items following Refueling Outage 8 in September 1997, but had not added the workarounds to the official list or provided technical resolutio During this inspection, the inspectors determined that the licensee had initiated Condition Report (CR) 96-1509 to document the failure to test both sets of emergency feedwater supply valves to each steam generator. The inspectors verified that Procedure OP-903-014, " Emergency Feedwater Flow Verification," correct!y tested each set of emergency feedwater supply valves and had been changed in November 199 The inspectors concluded that closing Valves CVC-216A and -2168 in order to test the downstream check valves while in Mode 5 did not result in any safety or regulatory concern. Similarly, the inspectors concluded that closing the component cooling water pump discharge isolation valves immediately prior to securing a component cooling water pump following testing did not result in any safety or regulatory concer I

!

On February 12,1998, the Workaround Committee identified that closure of the i component cooling water pump discharge isolation valves did not constitute an operator workaround but rather a good engineering practice to prevent slamming of the discharge check valve. The Workaround Committee concluded that the v ning of the charging system valves, if required, during a surveillance test did not rn .,t in an operator workaround. The inspectors did not identify any safety consequences resulting from these operating configurations; however, the inspectors expressed concem because the licensee failed to address the significance of the workarounds in a timely manner.

!

__ _ - - - - - - _ _ _ _ _ - _ _ _ _

_ _ _ _ _ _ _ _ _ _ -_ . _ - _ _ - - _

.

.

-2-08.2 Review Quality Assurance Audit of Ooerations Corrective Actions The inspectors reviewed the status of the corrective actions related to two specific findings in Quality Assurance Audit SA-97-034.1," Operations." The findings reviewed included personnel standing watch without the appropriate respirator eyewear, and plant procedures incorrectly allowed plant personnel to credit standing watch outside of a calender quarter to maintain an active operator license. NRC had identified both of these issues as noncited violations in NRC Inspection Report 50-382/97-08, Section O CR 97-0771 documented that a licensed senior reactor operator could not locate his eyewear for the emergency breathing apparatus (respirator). The interim corrective actions included ensuring that all operators and other personnel who could be fire brigade members had the required corrective lenses. As corrective action to prevent recurrence, medical personnel were required to document in the comments section of the physical form that the individual was informed about the need for the corrective lenses, and the employee was required to sign that they were informed. The inspectors considered this corrective action to prevent recurrence weak since neither the procedure nor the form had been changed to document this requiremen The inspectors found that medical personnel had a draft procedure that included the requirement and included a statement on the form to inform the employee of the need for corrective lenses. During the exit meeting, the Director, Plant Support committed to issue a revised procedure by May 31,1998. In addition, the licensee indicated they would evaluate the possibility of requiring that corrective eyewear be in an individual's possession prior to getting a respirator fit tes CR 97-0812 documented that Procedure 01-024-000, * Maintaining Active SRO/RO Status," allowed operators to credit watchstander hours outside of the current calender quarter for maintaining proficiency and an active license status. The inspectors verified that the licensee revised Procedure 01-024-000 to specify that five 12-hour shifts per calender quarter must be stood to maintain the operator license active. As specified in NRC Inspection Report 50-382/97-08, Section 07, the licensee had verified that all personnel with active licenses had stood the appropriate watches within the calender quarte II. Maintenance M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Failure of Train B Suborouo Potter Brumfield Motor-Driven Relavs Scooe (62707. 71707. 40500)

The inspectors evaluated licensee activities related to the failure of normally closed (deenergized state) contacts of a normally energized engineered safety feature subgroup l

,

!

E_____--__--

_-__ _ _ _ - _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _

.

.

-3-relay to go fully closed during emergency diesel generator testing. The licensee had documented this deficiency in CR 97-230 b. Observations and Findinas l On September 22,1997, during the conduct of Procedure OP-903-068, " Emergency l Diesel Generator and Subgroup Relay Train B Operability Test," the Emergency Diesel Generator B output breaker failed to open when Subgroup Relay K-110B deenergized; however, other equipment required to actuate (the low- and high-pressure safety ,

injection pumps) started. Subgroup Relay K-110B actuates these pumps when the i

normally open (deenergized state) contacts of the energized relay open. Subgroup l

Relay K-1108 has seven contacts, with six of those contacts controlling the actuation of safety-related equipmen The inspectors noted that operators entered the appropriate Technical Specifications and performed an operability evaluation for the affected equipment. Electricians replaced

Subgroup Relay K-110B but found that the procedure did not fully test all six safety-related contacts since three of the actuation devices did not change stat Consequently, operators continued to maintain the system in an inoperable statu Electricians successfully bench-tested the replacement relay, locally tested sll j components actuated by the relay, and independently verified all associated subgroup  !

, relay contacts and connections. As a postmaintenance test, the licensee reperformed l Procedure OP-903-068, which demonstrated that the Emergency Diesel Generator B 1 output breaker opened. Subsequently, operators declared the equipment operable. The inspectors found these actions appropriat ,

'

Following review of these immediate actions, the inspectors challenged the licensee with l the following questions: (1) does the test methodology verify operability of Subgroup Relay K-110B, as specified in your Technical Specifications; and (2) are all relay contacts required for testing actually tested?

The inspectors reviewed activities implemented by engineers to determine the root cause of the failure of Subgroup Relay K-110B to actuate and efforts to verify that all required l contacts were being properly tested. The licensee indicated that they had experienced l previous failures of normally closed contacts on energized Potter Brumfield Model 7032  :

motor-driven relays to close when the relay was deenergized. The licensee had sent the relays to an offsite vendor to have a failure analysis performed on the relays and had submitted the test results to the relay supplier (Combustion Engineering).

l Because of other previous failures at Waterford 3, Combustion Engineering issued a 10 CFR Part 21 report detailing this failure mechanism for selected model number Potter i Brumfield motor-driven relays with specific date codes. The 10 CFR Part 21 and '

.

supplement indicated that all Potter Brumfield motor-driven relays with date codes between 93XX and 95XX had the potential to experience failures. Waterford 3 had identified four failed Potter Brumfield motor-driven relays, either Models 170-1 or 7032, with date codes of 94XX. The affected relays were both direct and alternating current

{.

.

( -4-l l

types and all had hardened grease as the failure mechanism. The initial evaluation by l an independent research laboratory identified contaminated grease inserted into the relays during assembly, combined with elevated temperatures, as the most likely root cause. The 10 CFR Part 21 indicates that Combustion Engineering was continuing to evaluate the required corrective action A failure analysis performed by a third-party vendor and provided to Combustion Engineering agreed with the analysis performed by the independent research laboratory for the most part; however, the third-party vendor attributed the root cause of the hardened grease to contamination of the grease with silicon dioxide fibers, which occurred during assembly of the relays without any contribution resulting from high temperature. The third-party vendor tested this theory by using contaminated grease test samples applied to relays and comparing these to control relays without the contaminated grease. The vendor cycled the relays for 162,000 operations and left them energized for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> at 65*C. After completing the test, the relays with uncontaminated grease had fluid grease, while the relays with contaminated grease had grease with the consistency of paste. The silicon dioxide fibers came from contact ring insulator plates, switch rings, or coil epoxy. The silicon dioxide fibers became abrasive, similar to a lapping compound, which caused abnormal wear from the shaft and bearings and resulted in the relay becoming too stiff and difficult to operat The failure analysis bounded the suspected relays by identifying the source of the contamination to startup of a new manufacturing facility in 1993 until the facility altered !

housekeeping procedures in 1995. Additional corrective actions during the j manufacturing process will be implemented to prevent potential contamination of the greas Because of the differing results from the third-party vend ar and the independent research laboratory, the licensee and other facilities with Combustion Engineering engineered safety feature actuation systems contracted with Combustion Engineering to perform a third evaluation of the Potter Brumfield motor-driven relay failure mechanism. The expanded investigation and examination will evaluate the Potter Brumfield motor-driven relays manufactured during different periods from 1988 to the present. The review is intended to identify the population of affected Potter Brumfield motor-driven relays and confirm the failure mechanis The inspectors reviewed the information provided by the licensee on the testing of all contacts actuated by Subgroup Relay K-1108 and concluded that the licensee had received relief from the testing requirements for actuating devices, as specified in Regulatory Guide 1.22," Periodic Testing of Protection System Actuation Functions."

The inspectors based this conclusion on the following: (1) the licensee provided information to the NRC in a letter dated July 21,1983, that described the engineered safety feature actuation devices that could not be tested at power but were tested during shutdown conditions; (2) the licensee indicated in a letter dated May 14,1984, that a i step-by-step review of the subgroup relay testing program had occurred and was

! approved during an April 26,1984, meeting on the contents of the July 21,1983, le ter;

!

-

_ _ _ ____ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _______ - -_

i

!

l

.

'

!

'

-5- l l

and (3) contact plates in a motor-driven relay move along the same shaft, consequently, if a single set of normally closed (open when the relay is energized) contacts open, then ,

the other normally closed contacts will open in the same manne I

- The licensee established an action plan to ensure close monitoring of the potentially affected Potter Brumfield motor-driven relays. The near-term actions included, in part, identifying for evaluation all normally energized Potter Brumfield motor-driven relays with the applicable date codes; performing thermographic scans of all affected relay cabinets; monitoring one normally closed contact for each relay during surveillance; and adding relays that have an 18-month test frequency to the forced outage list. The long-term action was to evaluate possible relay replacement. Review ofimplementation of the Potter Brumfield motor-driven relay action plan and the results of the Combustion Engineering evaluation is an inspection followup item (50-382/9802-01), Conclusions The inspectors considered the planned corrective actions for the Potter Brumfield motor-driven relay failures to be good. The licensee had developed a detailed action plan to ensure that other relays would be identified prior to failure. In addition, the licensee continued to implement actions to resolve the deficiency and identify a root caus Ill. Enaineerina E1 Conduct of Engineering E Containment Sorav Flow Rate Desian Deficiency Insoection Scooe (71707. 37551)

The inspectors assessed licensee performance related to an operability evaluation of containment spray system flow discrepancies. The inspectors reviewed the applicability l of regulatory requirements, design basis for the system, and inservice test requirements and history and discussed the operability evaluation and inservice test history with design engineers. This information affects both the loss-of-coolant accident case and the main steam line break (MSLB) case used in the containment peak pressure accident analysis. Since the containment peak pressures for the main steam line break case provided the highest, most limiting, containment peak pressures, only the main steam line break case is discusse Observations and Findinas t

On November 17,1997, engineers identified in CR 97-2623 that the minimum required l containment spray flow rate used in the containment peak pressure analysis was l 1970 gpm, but recent inservice test results indicated that the containment spray pumps j delivered approximately 1850 gpm, with minimum acceptable design flow rates of  ;

i

)

t - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _

.

'

, 4 l 6-1750 gpm. Further, CR 97-2623 documented that the containment peak pressure analysis currently uses nonconservative (short) delay times for initiation of the containment fan coolers. Preliminary analysis using the GOTHIC 5.0 computer code with the realistic (longer) delay time prior to containment fan cooler initiation and a containment spray flow rate of 1750 gpm indicated that the containment peak pressure did not exceed the design maximum of 44 psig. For the MSLB case, both containment l spray trains yield a combined flow of 3500 gp I Because of the discrepancy in the design flows, the operators had a question regarding system operability but had confidence that the system would be demonstrated operable; therefore, the operators requested that an operability confirmation be performed in accordance with Procedure W4.101, " Operability / Qualification Confirmation Process."

This procedure implemented the guidance contained in Generic Letter 91-18,

"Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions." The inspectors noted that the operability evaluation completed on November 19,1997, for the containment spray system, using the GOTHIC 5.0 computer code, demonstrated that the peak containment pressure would have reached 43.63 psig with a 3500 gpm containment spray flow rate, a 19.5 second delay for initiation of the containment fan coolers, and a 44.62 second delay before full spray into containment is achieved. Also, the licensee indicated that the containment spray flow rates into the containment under design basis accident conditions from the results of the most recent inservice tests exceeded the flow rates used in the operability evaluation. These flow rates for Trains A and B were 1860 and I 1790 gpm, respectively, after accounting for 60 gpm recirculation flo I

!

The inspectors discussed with design engineers the background for the flow rate used in l the safety analysis compared to the flow rate used in the inservice testing program. On September 24,1997, the Inservice Test Program coordinator had questioned an j apparent conflict with the 1970 gpm containment spray flow rate credited in the accident i analysis and the 1750 gpm inservice test flow rate. The design engineer assigned  !

responsibility for resolving this discrepancy concluded that this 1970 gpm flow value i must have been a typographical error, hence the individual did not see a need to resolve the discrepancy immediately. This individual had previously evaluated safety-related pump flows required to demonstrate compliance with the Technical Specifications. This previous evaluation required verification of the inservice testing flow rates, and safety analysis personnel had previously provided the design engineer with information that the required design flow rate was 1750 gp The inspectors reviewed the containment spray system pump curves and system curves and concluded that the highest spray flow into containment that could be expected with a nondegraded containment spray pump under design basis accident conditions was 1940 gp The inspectors found that the licensee used 3790 gpm containment spray flow, 44 seconds delay time until full containment spray flow, and an 8-second delay time for containment fan cooler initiation as the initial design conditions to determine the

- - _ _ _ _ _ _ _ _ _ _ _ -

_ -. _ _ __ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ __ . _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ - _

-

t i

l

!

-7-containment peak pressure achieved during an MSLB accident. The MSLB case had a failure of a main steam isolation valve to close but retaining both trains of containment spray and containment fan coolers as the most limiting condition. The containment peak pressure calculated using the CONTEMPT Mod 26 computer code yielded 43.6 psi Because of nonconservative (short) 8-second delay times for initiation of the containment fan coolers, the architect engineer had performed an evaluation in 1983 using longer delay times for initiation of the containment fan coolers, a longer delay time for full spray initiation, and increased containment spray flow rate. This was documented in Memorandum AP-WK-422, dated April 6,1983. The architect engineer used a 1970 gpm containment spray flow, a 44.62 second delay time for full spray into containment, and a 19.5 second delay until initiation of the containment fan coolers. This MSLB case used the failure of one train of containment spray and containment fan coolers but retained the availability of offsite power, as the most limiting conditions. The containment peak pressure, determined using the CONTEMPT Mod 26 computer code, was 43.74 psi The Root Cause Analysis Report identified the root cause as inadequate update of design basis information in 1983 when the architect engineer recalculated the containment spray flow rate. As corrective actions, the licensee revised the emergency operating procedures to reflect the correct containment spray flow rate, completed a containment peak pressure determination using the CONTEMPT Mod 26 computer code, planned to update the Updated Final Safety Analysis Report with the new design values, and initiated an action to reevaluate the containment spray flow test acceptance criteri The licensee identified that ongoing programs to review the design basis, including mechanical calculation reviews to validate assumptions and design values, should identify other, similar design basis deficiencies. Further, the licensee indicated that communic:tions and process controls have improved significantly since 1983 when this deficiency occurre The inspectors discussed the scope of the design basis reconstitution effort. The licensee indicated that a system function hierarchy is prepared to capture the required system functions to meet the General Design Criteria. This process identified the calculations that support each of the design values in the Updated Final Safety Analysis Report, the inservice testing program, and the accident analysis. In addition, the licensee independently developed a calculation hierarchy that identified the relationship of calculations (design values) to one another for each system. Whenever discrepancies exist, an open item would be generated to resolve the issue. The program coordinator indicated that the goal of the program was to reconcile any differing values among the inservice testing, safety analysis, and calculation design values for parameters required to ensure that the system safety function can be accomplished.

i The inspectors reviewed the Inservice Test Design Basis Document values for the

!

containment spray pumps. This document indicated that the containment spray pumps i are capable of supplying 1,810 gpm flow (includes 60 gpm recirculation flow). Because l the Inservice Test Design Basis Document clearly identified design requirements that

_____ ______

i

'

.

.

i-8-met one Updated Final Safety Analysis Report section but differed from other sections of )

the Updated Final Safety Analysis Report and the supporting calculations, this deficiency would have been an example of an open item that required resolution based on the 3 licensee's description of their design basis review process. Also, this deficiency is an old i design issue that occurred in 198 The inspectors confirmed that the licensee incorporated the containment spray and containment fan cooler delay times but did not incorporate the revised containment spray flow rate into the Updated Final Safety Analysis Report. Because a containment spray pump with no degradation, and ignoring instrument inaccuracies when measuring flow, could only achieve 1940 gpm flow, the inspectors concluded that this was a nonconservative assessment of the capacity of the containment spray pumps during originallicensing. The licensee had failed to properly ensure that the containment peak pressure accident analysis used appropriate values for the containment spray pump flow rate during the MSLB case, as required by 10 CFR Part 50, Appendix B, Criterion Il This licensee-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. Specifically, the violation was identified by the licensee, was not willful, actions taken as a result of a previous violation should not have corrected this problem, and appropriate corrective actions were completed by the licensee (50-382/9802-02).

The inspectors evaluated the basis for the differential pressure value listed in the Technical Specifications. Technical Specification 4.6.2.1.c requires that the licensee demonstf ate operability of the containment spray system quarterly by verifying that, on recirculation flow, each pump develops a total head of greater than or equal to 219 psid when tested pursuant to Technical Specification 4.0.5. Based on reviews of the containment spray pump curves, containment spray system curves, and interviews with the licensee, the 219 psid differential pressure correlated to a minimum spray flow of approximately 1790 gpm under design basis accident conditions. Safety analysis had credited the containment spray system with accident flow rates ranging from 1800 - 1970 gpr The inspectors noted that licensee correspondence indicates that the 219 psid differential pressure listed in the Technical Specifications did not include any allowance for instrument uncertainty, but the 219 psid did allow 10 percent margin from the original capability of the pumps. The inspectors confirmed that Procedure OP-903-035, 1

" Containment Spray Pump Operability Check," Revision 9, had a requirement for instruments to have an accuracy of at least 2 percent as specified by the ASME cod The inspectors also noted that any test that resulted in a value close to 219 psid would be very close to actual flow decreasing below the design value of 1750 gp Specifically, since 219 psid corresponds to a flow of 1790 gpm, a 2 percent loop accuracy allowed by the inservice test program would result in 1790135.8 gpm or a l

lower limit of 1754.2 gpm. The inspectors verified from review of historical test data that

'

the measured differential pressure exceeded 219 psid by a significant margin. The j inspectors determined that the flow corresponding to 219 psid would be acceptable to

! ensure that the design basis flow would be achieved in the event of an acciden _ _ _ _ _ _ _ _ _ _

(_ __ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

l *

.

-9-l l Conclusions A noncited violation occurred because the licensee did not use the proper design basis

'

information to perform the accident analysis for exceeding the peak containment pressure in 1983. The licensee failed to update the Final Safety Analysis Report after the architect engineer revised the containment spray and containment fan cooier design basis values used in the containment peak pressure accident analysi E7 Quality Assurance in Engineering Activities E Assessment Activities i Scoce (40500)

The inspectors reviewed the actions implemented to formalize the assessment proces The Systematic Assessment of Licensee Performance report for Cycle 16 (50-382/96-99)

criticized Entergy Operations, Inc for their ineffective assessments. In response to this Systematic Assessment of Licensee Performance report comment, each site was required to develop a method to track, review, and respond to assessment recommendations. Waterford 3 established Performance improvement Plan Goal 3. that required personnel to develop guidelines and implement a site-wide assessment oversight proces The inspectors reviewed selected corporate assessments and site-based assessments to determine the effectiveness of the corrective actions and assess the implementation of the assessment databas Observations and Findinas Assessment program The inspectors determined that Procedure W1.108, " Assessment Oversight (Waterford 3 Self Assessment Program)," described the formalized assessment process, which included requirements for identifying, tracking, and responding to recommendations contained in audits, inspection reports, onsite assessments, and corporate assessments, et cetera. For corporate assessments the corrective actions must be concurred with by the Vice President Operations. The inspectors reviewed the following plant assessments:

- Waterford 3 Plant Engineering Assessment Corporate

. Waterford 3 Operating Experience Assessment Corporate

- Assessment of the W310 CFR 50.59 Program Site-based

- W3 Corrective Action Program Assessment Site-based (Section E7.2.b.3)

The inspectors confirmed that the licensee had appropriately included assessment report recommendations in their assessment database. The inspectors concluded that none of

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ - - _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _

.

e-10-the assessment recommendations should have been initiated as CRs. The inspectors noted that, although the required assessment recommendations were listed in the assessment database, the responses to the operating experience and plant engineering corporate assessment recommendations had not been accepted by the Vice President Operations at the time of this inspectio b.2 Corporate assessment evaluations The inspectors discussed the planned actions with the Manager, Operating Experience to implement the recommendations from the corporate assessment. The issues identified by the corporate assessment included the lack of easy access to operating experience by facility personnel, management expectations for the use of operating experience information had not been clearly communicated, and the lack of evidence of the use of industry operating experience into day-to-day plant processes such as work package planning and corrective action process. Items in the corrective action plan developed to respond to the assessment included: (1) identifying a point of contact in each facility work group for receiving industry information and training the point of contact to perform operating experience searches; (2) proceduralizing operating experience input into the planning and scheduling process; (3) retraining of root cause analysis evaluators in the use of operating experience for root cause analyses; and (4) developing site-wide searchable databases. The inspectors could not verify implementation or could not assess the effectiveness of the changes since they had not been implemented or in place for a long enough period of time, as of the date of this inspectio The inspectors discussed the planned actions in response to the corporate assessment of plant engineering with a technical assistant to the Plant Engineering Manager. The issues identified by the corporate assessment were a lack of clear, written expectations, including roles and responsibihties for performance of the plant engineers, a lack of balance between emergent and base engineering tasks, and the work load not having been clearly defined or prioritized. The inspectors determined that the licensee had developed an action plan to address each of the recommendations; however, at the time of this inspection, the licensee had not implemented each of the corrective action Consequently, the inspectors could not assess the effectiveness of the corrective action CFR 50.59 Assessment The licensee had licensing and quality assurance personnel perform this assessment to ensure compliance with regulations and evaluate program effectiveness in June 199 The assessment evaluated 28 CRs documenting 10 CFR 50.59 discrepancies since  ;

'

initiation of CR 96-0471, which documented that discrepancies in 10 CFR 50.59 t evaluations constituted an adverse trend. The inspectors noted that CR 96-0471 remained open at the time of this inspection. The assessment resulted in initiation of several CRs that documented inadequate reviews of licensing basis documents, failure to meet industry standards for 10 CFR 50.59 screenings, and failure to properly address i the evaluation questions asked. The assessment indicated that personnel did not use ,

I i  !

l l

l L- - -

_ __ __________ ______ __ _________-_ _ _ _ _ _ _ _ - _ - _ _ _ _ - - _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ - _ _ - _ _ _ _ - _ _

a

.

-1 1 -

the published Entergy "10 CFR 50.59 Safety Evaluation Guidelines." The recommendations included modifying the safety evaluation forms to remind personnel to review the Technical Specifications, Technical Requirements Manual, and the Updated Final Safety Analysis Report (including the most recent changes).

The inspectors confirmed that Procedure W2.302, "10 CFR 50.59 Safety and Environmental Impact Evaluations," Revision 4, had included appropriate changes to address all of the recommendations identified in the assessment of the 10 CFR 50.59 program. The inspectors verified that the recommendations and responses were ,

contained in the assessment database. The inspectors could not evaluate the I effectiveness of the process change corrective actions since the process had not been implemented for a sufficient period of tim Conclusions The licensee performed very good assessments based on the identified findings and recommendations made. The extent of the planned corrective actions to the recommendations reflect a willingness by some managers at the facility to implement corrective actions to address recognized deficiencie E7.2 Corrective Action Program Review Scoce M0500)

.

The inspectors evaluated licensee activities utilized to assess and evaluate the effectiveness of their corrective action program and evaluated management review of identified conditions. The inspectors reviewed quality assurance corrective action program implementation audits, observed Corrective Review Group and Corrective l Action Review Board meetings, and reviewed the corrective action program assessment I committed to in the Waterford 3 Performance Improvement Program, item 3. Observations and Findings 4 Review of Corrective Action Audits The inspectors evaluated the last two documented 6-month corrective action program audits, SA 96-004.2 and SA 97-004.1, that had been completed at the time of this inspection. Quality assurance auditors evaluated the effectiveness of the corrective actions implemented in response to selected CRs, NRC violations, licensee event reports, discrepancy notices, and operating experience group reviews of information notices and 10 CFR Part 21 reports. Other audit criteria included evaluation of the implementation of the corrective action program as described by procedures, which included evaluation of backlogs, root cause analyses, and generic issue The inspectors concluded that auditors initiated CRs for all conditions adverse to quality and made appropriate recommendations for apparent performance weaknesses. For

_ _____-_____ - -

[

l

.

-12-example, any instance of procedure noncompliance with Procedure W2.501, " Corrective Action," would result in a CR being initiated. Recommendations included items such as identifying that: (1) root cause analyses did not explicitly identify an exact root cause but instead listed several contributing causes, or (2) the need to reinforce management expectations that self-identification of adverse conditions is desire The inspectors noted that recommendations contained in Audits SA-97-034.1, SA-96-004.2, and SA-97-004.1 were entered into the database; however, no responses to the recommendations had been made at the time of this inspectio b.2 Condition Review Group and Corrective Action Review Board meetings The inspectors attended several Condition Review Group meetings. The Condition j Review Group members included the plant manager, several site directors, and line

{

managers from all disciplines. During this meeting, the managers discussed in detail the  !

event described on a CR, assigned a category (significance level) for review and assignment of resources, and assigned a responsible individual. The assignment of the category results in determining whether the condition was considered a significant condition adverse to quality versus a condition adverse to quality and results in a timeliness requirement for review of the issue. The assignment also determined the level of management oversight required to address any required corrective action l The inspectors attended the Corrective Action Review Board meeting on February 5, i 1998, that discussed draft Root Cause Analysis Report, " Danger Tagged Components." l The root cause analysis evaluated two events related to improper implementation and control of danger tags. The first event related to the removal of a pipe section by welders in accordance with their work authorization and clearance approval that had a valve attached that had a danger tag from another clearance order. The second instance involved a valve handwheel disconnected from the stem even though a danger tag was attached to the handwheel. There were good discussions among the board members regarding the subject events, including the root cause determination, safety significance, and corrective action determinatio However, the inspectors expressed concern that the members had not adequately addressed the impact on personnel safety posed by the events or corrective actions related to planning and scheduling of emergent work. The inspectors discussed these issues at the meeting with the licensee. The revised Root Cause Analysis Report addressed the personnel safety issue more clearly and implemented an additional proposed corrective action to address the planning and scheduling of emergent wor b.3 Corrective Action Program Assessment The licensee had performed the Corrective Action Program Assessment in August 1997 as specified in the Waterford Performance Improvement Plan, item 3.B.1, in order to

'

identify weak areas that could be corrected to improve the effectiveness of the CR process. The inspectors reviewed the recommendations and their responses with l

l l

__.________________U

_ __ _ _ _ . _ - _

,

.

? .

I I -13- )

I personnelin the in-house Events Analysis group. The inspectors found the assessment to be critical of the corrective action process. The assessment identified the following concerns: (1) ability to clearly identify, segregate, prioritize, and focus appropriate resources on significant conditions; (2) timely documentation of adverse conditions; (3) timely completion of root cause analyses; (4) operability reviews performed that failed to fully evaluate the deficiency, did not address operability, and did not meet the guidance contained in Generic Letter 91-18, ' Degraded / Nonconforming Conditions";

(5) weaknesses evaluating generic implications and utilizing operating experience information; (6) collectively, poor documentation; and (7) timely completion of corrective action The inspectors verified that the Superintendent, in-House Events Analysis provided training to site personnel related to the revision to Procedure W2.501, " Corrective Action," Revision 7. During training of plant personnel, the Superintendent, in-House Events Analysis discussed the need for increased use of industry operating experience, timeliness of CR generation and completeness of CR documentation and emphasized use of the Root Cause Analysis desk guide. The licensee modified Procedure W2.501 to address several of the recommendations in the Corrective Action Program Assessmen The revised procedure modified the categories for classifying CRs according to significance. More significant CRs would receive more detailed root cause analyses,

'

whereas the procedure allowed for less labor intensive root cause determinations for l

'

CRs that were less significant. The procedure increased the flexibility of the analysis tools allowed for determining the root cause. The licensee modified the form for performing operability assessments by the shift technical advisor and/or the shift !

superintendent and added guidance, in the form of examples, for performing operability determination The inspectors concluded that the licensee made appropriate changes to their corrective action program to address the recommendations contained in the Corrective Action Program Assessment. However, the licensee had not been implementing the changes long enough to allow the inspectors to make conclusions relative to the effectiveness of the changes. The inspectors noted that plant personnel continued to identify adverse conditions at a low level. Further, the inspectors agreed with the licensee that the new category definitions for CRs will, by their nature, result in application of resources to those adverse conditions that had the greatest potential to affect plant safet c. Conclusions The inspectors found the corrective action audits to be thorough, with critical recommendations. The audits were compliance based but followed the corrective action process for identification of conditions adverse to quality. The inspectors identified no instances of failure to properly classify items as recommendations or conditions adverse I to quality. The inspectors concluded that the line organizations did not respond to recommendations contained in all the quality assurance audits reviewed as promptly as i

the recommendations contained in site-based and corporate assessments.

l

_ _ _ _ _ _ _

l*

..

! -14-E7.3 Condition Reoort Reviews Scoce (40500)

The inspectors evaluated significant CRs to identify the ability of the licensee to resolve conditions adverse to quality. The inspectors assessed the operability determinations, the quality of the engineering response, the extent of the corrective actions, and the adequacy of the root cause determinatio Observations and Findinas CR 96-1405 Personnel initiated this CR after identifying that components were listed as nonsafety-related in the Station Information Management System database for the safety-related dry cooling towers. Personnel had identified this discrepancy while ordering replacement parts. Consequently, the licensee conducted a review of the safety classification of a sample of components listed in the Station Information Management System database and identified that several components were misclassified as nonsafety-relate The inspectors reviewed the CR, the root cause analysis, and the corrective action The licensee initiated the Q-List Verification Project as a corrective action to prevent recurrence. In accordance with the Q-List Verification Project, engineers reviewed all components contained in safety-related/ quality-related systems identified in Station Information Management System database as non-nuclear safety to ensure that they were correctly classified by comparison with primary source documents (the design documents that specified provide information regarding component safety classification).

The engineers initiated CRs to document review and disposition of incorrectly classified components (including operability determinations) and initiated engineering requests for components that lacked a primary source document or were potentially misclassifie Engineers corrected the Station Information Management System database entries for any misclassified components. The inspectors concluded that the Q-List Verification Project provided effective corrective action for component classification deficiencie CR 96-1585 This CR documented that the component cooling water surge tank had a one-inch nonsafety-related check valve installed to provide vacuum relief protection. NRC Inspection Report 50-382/97-25 documented review of the 10 CFR 50.59 for a calculation related to this modification, but indicated that the CR corrective actions had not been reviewed. Also, NRC Inspection Report 50-382/96-202 documented that this check valve was installed, that the licensee initiated the CR to document the deficiency, and that review of draft calculations identified no concerns. The inspectors reviewed the approved calculations and identified no problems.

[

!

l

__ __. _ _ _ _ _ _ . . _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _

_ _ __ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _

.  !

-15-

i I

b.3 CR 96-1680 Quality assurance personnel initiated this CR after identifying from work authorization package reviews that seven parts of the wrong quality level had been installed in safety-related systems. Two parts remained installed in the plant at the time of the quality assurance audit. The licensee confirmed operability of the affected systems by performing commercial grade dedication of the parts. The inspectors reviewed the associated root cause analysis and corrective action The licensee implemented additional administrative controls to ensure that, in the work planning and postwork review processes, personnel obtained and installed parts with the correct quality level. Initially,1200 commercial grade items had been procured for use in safety-related applications but had not been dedicated. The licensee implemented a project to dedicate this backlog of components and at the time of this inspection approximately 500 unmedicated parts remained. The licensee indicated that the remaining part dedications would not be completed until 1999. The inspectors evaluated the administrative controls and concluded that the administrative controls should preclude installation of commercial grade nonsafety-related parts. If a part was required prior to the program review, the licensee would perform expedited part dedications. The inspectors considered these corrective actions satisfactor b.4 CR 96-1807 This CR identified the potential inaccuracy in Calculation MN(Q)-6-45 (a calculation of water seal in selected safety system) that justified exemption from 10 CFR Part 50, Appendix J, leakage testing for certain containment penetrations. Specifically, CR 96-1807 documented that the required 30-day water seal may not have been capable of being maintained because of leakage through Valve CVC-209, the charging header stop valve. Engineers determined that the calculation had demonstrated that a sufficient water seal existed in the charging line to withstand containment pressure for 30 days with maximum leakage from the valve packing outside the containment boundar The inspectors reviewed the sections of Calculation MN(Q)-6-45 applicable to the charging line and identified no concerns with the assumptions or the methodology use The inspectors verified that leakage testing through the valves and outside of the system demonstrated compliance with the design basis. The inspectors confirmed that the licensing basis for the charging line did not require Appendix J leak testing of the charging line valves. The licensee had established a compensatory action to close manual Valve CVC-208 in the event Valve CVC-209 can not be closed. This interim action will remain in effect until the end of the refueling outage that begins in 199 Closure of Valve CVC-209 is needed to ensure that the water seal will be maintained for 30 days postaccident if a charging line shear were to occur. This corrective action had

been reviewed by the NRR technical staff prior to the licensee starting up from the last refueling outage that ended in July 199 _ _ _ .

_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ - _ - _ _ -

.

I

.

-16-b.5 CR 96-1913 Personnel initiated this CR when they identified that surveillance testing for the core protection calculators (CPC) and control element assembly calculators may not have been performed as required by Technical Specification 4.3.1.1. This Technical Specification surveillance requirement specified that each channel shall be demonstrated operable by the performance of the channel functional test every refueling outag Further, the test shall include the injection of simulated process signals into the channel as close to the sensors as practicable to verify operability, including alarm and/or trip functions. The licensee considered it acceptable to use reactor trip system response time testing for the CPCs, as specified in Technical Specification 4.3.1.3, as an acceptable method to test all four channels as required by Technical Specification 4.3.1.1. However, the licensee performs the response time test on each channel once every 72 months (i.e., one out of four channels every 18 months) instead of all four channels every 18 month Upon recognition of this deficiency, the licensee declared the three channels that had not been tested during the previous refueling outage inoperable and entered Technical Specifications 3.0.3 and 4.0.3 until they completed testing and demonstrated operability of the untested channels. The NRC discussed this issue in detailin NRC Inspection Report 50-382/96-14 and identified this as a noncited violation of Technical Specification 4.3.1.1. The initial deportability determination completed on February 12, 1997, concluded past deportability was indeterminate because the licensee had requested that Combustion Engineering review the technical requirements for the tes The inspectors verified that the licensee performed Technical Specification 4.3. surveillance tests for all CPC channels during the 1997 refueling outag In the final deportability determination, the licensee concluded that the issue was not reportable and that the testing they had performed met the requirements of Technical Specification 4.3.1.1. The licensee based their conclusion on a draft report, " Review of the CPC Functional Testing for the CE Owner's Group " performed by Combustion Engineering that the testing met the requirements of General Design Criterion 21 and IEEE Standard 338-1987, " Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems." The inspectors noted that the deportability evaluation did not explicitly describe how the requirements of Technical Specification 4.3.1.1 were me )

The licensee indicated that Combustion Engineering had not completed its review of this i issue. During discussions, licensee personnel were unaware that the NRC had identified that a violation of Technical Specification 4.3.1.1 had occurred. In NRC Inspection Report 50-382/96-14, the inspectors concluded that the corrective actions (performing l I

additional testing) were appropriate. After the inspectors identified that NRC concluded that a violation of Technical Specifications had previously been identified, licensing l

personnel indicated that the deportability determination would be evaluated and that i personnel would reconsider the testing requirement l During discussions with licensee personnel on February 26,1998, the inspectors learned that Combustion Engineering continued to review the testing requirements after licensee l I

l l

I . . _ _ _ _ . _ _ . _ _ _ _ _

_ _ _ _ _ _ _

.

I (

?' l-17-personnel indicated that the initial response was an industry survey instead of an answer to the concern of whether testing of each channelis required. On March 5,1998, the inspectors learned that the licensee had received a second response from Combustion Engineering that indicated Technical Specification 4.3.1.1 required testing of all four channels every 18 months. The licensee indicated that a licensee event report would be initiated to document past violations of Technical Specification surveillance requirement The inspectors will review corrective actions during followup of the licensee event repor b.6 CR 97-0288 This CR documented that Essential Chiller AB tripped on low refrigerant pressure while swapping air handling units and replacing Essential Chiller A with Essential Chiller A During the beginning of this inspection, the inspectors determined that the licensee had declared essential chiller control system problems as an adverse trend, as documented on CR 97-2698. The licensee initiated this CR because the compressor inlet vanes for Essential Chiller A would not change position, which allowed the outlet essential chilled water temperature to exceed 42*F and rendered Essential Chiller A inoperable. At the same time, Essential Chiller B was out of service because of Freon leaks; consequently, the operators entered Technical Specification 3.0.3 until they replaced Essential Chiller B with Essential Chiller A At the request of the Plant Engineering Manager, the system engineer had researched the work history database and identified 19 failures related to the essential chillers with 15 failures related to the control systems. Previously, system engineers had initiated repetitive tasks for several of the failures. The licensee had developed a modification for this operating cycle to replace the relay module for each essential chiller. The essential chiller control system has 37 components (includes 25 relays), consequently, the system engineer initiated a cost analysis study to determine the feasibility of replacing the control systems in their entiret Because the licensee had identified the essential chiller system problems as an adverse trend, the inspectors did not perform a detailed evaluation of any individual CRs. The inspectors considered the actions of the licensee to be satisfactory and were encouraged by this self-initiative to resolve the essential chiller system trouble b.7 CR 97-0352 The licensee initiated this CR when a weld crack and leak was identified on an auxiliary component cooling water (ACCW) heat exchanger drain pipe. A similar failure had occurred on the same drain pipe on a nearby weld, as documented on CR 97-0040. The licensee performed a metallurgical examination of the welds and a root cause analysi The licensee determined that fatigue cracking caused by bending stresses and vibration i resulted in the weld failures.

!

t The inspectors found that excessive vibration resulted from low-flow induced vibration during decreased ACCW system flows. The licensee had been operating the ACCW f

_

. _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - . _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _

_

O

'

.

I

!

-18-pumps continuously at low flows during cold weather months to maintain the system filled with water to prevent water hammer and to prevent overcooling the component cooling water system. Also, the heat exchanger drain pipe had additional, abandoned corrosion monitoring piping attached, which created a moment arm that induced additional stresses. As corrective action following the second occurrence, the licensee removed the corrosion monitoring system piping from both ACCW trains. The inspectors determined that the licensee performed magnetic particle examinations of the drain pipe welds on the other train. The inspectors agree from review of the examination report that there were no indications. The onsite metallurgical engineer indicated that the welds examined were the ones most susceptible for cracking, given the low-flow induced vibration that had occurred. The engineer had also performed system walkdowns and concluded by observations that the heat exchanger drain lines were most susceptible to crackin The licensee installed Design Change 3470, "ACCW Water Hammer," during the 1997 refueling outage. This modification installed a keep-fill system that allowed the ACCW pumps to be secured and eliminated the low-flow induced vibration. The inspectors concluded that installation of the keep fill system and removal of the corrosion monitoring system piping were effective corrective actions. However, at the time of the inspection, the licensee had not completed a corrective action identified in the CR to perform vibration monitoring of the Train A heat exchanger drain pip The inspectors reviewed other vibration-induced problems in the ACCW system. During low flow operation, cracks developed in the 0.5 inch threaded pipe nipple connections on the ACCW pump suction and discharge flanges for the ACCW pump bearing oil cooling system. The licensee had initiated CR 97-0547 and concluded that these cracks were also caused by low-flow induced vibration. The licensee replaced the Schedule 40 ASME Section Ill, Class 3, pipe nipples with Schedule 160 ASME Section Ill, Class 3, threaded pipe nipples, which were less susceptible to vibration and had greater strengt Also, the inspectors reviewed a weld crack and leak deficiency that resulted on the Charging Pump A discharge relief vent and documented on CR 97-2277, Again, they attributed the cause of the weld failure to vibration-induced fatigue. After the licensee

'

repaired the weld joint using the same configuration as the original, the weld joint failed again (refer to NRC Inspection Report 50-382/97-22). Subsequently, the licensee modified the socket weld and added a support to the discharge relief vent pipe to reduce vibration Around this same period as the charging pump failures, the licensee incorporated the information contained in Electric Power Research Institute Report TR-107455, " Vibration Fatigue of Small Bore Socket-Welded Pipe Joints," into their design process. The report provided information on this problem and provided guidance for the evaluation of piping systems. The licensee determined that previous guidance for evaluating vibration

,

concerns was too general. The licensee also had actions in progress to evaluate other i

systems to identify piping configurations susceptible to vibration-induced weld failure and to determine whether piping changes should be proposed or whether additional vibration

_ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ - -

_ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ - - _ - _ - _ _ _ _ _ _ - _ ___

O l'

.

'

'

monitoring should be conducted. The inspectors concluded that, after the second charging pump discharge vent failure and evaluation of the Electric Power Research

, Instituted information, the licensee had initiated comprehensive corrective actions.

l i CR 97-0726 Quality assurance auditors initiated this CR to document inconsistencies related to the use of measuring and test equipment between the Station Information Management System database and the " blue card" used to document removal of measuring and test equipment from inventory for quantitative / qualitative uses instead of diagnostic use !

Specifically, on occasion maintenance personnel listed diagnostic uses and did not list all quantitative / qualitative uses on both the Station Information Management System database and the blue card. The licensee attributed the confusion to a lack of understanding of the definition of quantitative / qualitative use of measuring and test equipment. The licensee concluded that the definition as written in the procedure was clear and provided training to maintenance personnel to ensure familiarity with the definition. The Superintendent, Instrumentation and Control established an extended corrective action due date to allow for evaluation of technicians' understanding of the definition of qualitative / quantitative use at a later dat The inspectors did not identify any concems with the evaluation of the root cause or the planned corrective actions. The inspectors concluded that the licensee could capture all safety-related uses of measuring and test equipment since, upon identification of uncalibrated equipment, the licensee reviews both the Station Information Management l

'

System database and the blue cards for uses of the equipmen CR 97-0869 j l

Operators initiated this CR to document the inability to comply with the requirements of an operability evaluation (performed for CR 97-0174) that documented the conditions under which piping systems penetrating containment became inoperable. The licensee performed the CR 97-0174 operability evaluation in response to Generic Letter 96-06,

" Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions." The operability evaluation did not specify the conditions when system operability could be affected. During the power decrease in April 1997, the reactor coolant system temperature decreased below the specified values in the j operability evaluation; therefore, operators correctly entered the operability confirmation i procedure and declared the charging system inoperabl The corrective actions involved revising the operability evaluation for CR 97-0174 to i clarify that the restrictions applied following a large break loss-of-coolant accident when containment temperatures could cause pipe temperatures to exceed 260'F. The i

'

inspectors concluded that the operators implemented appropriate actions to comply with the operability evaluation. The engineers performed a weak evaluation in that the exact l l conditions for applicability and the impact on plant operations had not been fully l

considere l i

J l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . - - - - - - - - - - - - - - - - _ - - - _ - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _

.

.

-20-b.10 CR 97-1605 and CR 96-1857 On November 24,1996, the licensee initiated CR 96-1857, after personnel determined that Containment isolation Valves MS-116A and -1168 did not have valve position indication available that conformed with Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," recommendations for postaccident monitoring instrumentation. The licensee had requested deviations from the Regulatory Guide 1.97 recommendations and the NRC had approved these deviations in a Safety Evaluation Report dated June 5,1997. As a corrective action in the CR, the licensee stated that it s

would conduct an independent review of its compliance with Regulatory Guide 1.97. The inspectors noted that an extension request had been approved that extended the due date to April 30,199 The inspectors found that an independent assessment would be performed by a contractor rather than by Entergy personnel. The scope of the review included containment isolation equipment identified in Updated Final Safety Analysis Report Table 7.5-3 (Regulatory Guide 1.97 information) and Updated Final Safety Analysis Report Table 6.2-32 (containment isolation valve information). Specifically, the review would detect any cases where Regulatory Guide 1.97 recommendations were not met and determine if instrumentation provided for verification of containment isolation function provided direct indication and if the equipment is qualified as required. The inspectors considered these appropriate corrective actions for CR 96-185 The inspectors also reviewed CR 97-1605, which documented discrepancies between l Updated Final Safety Analysis Report Table 7.5-3 and control room ventilation I components installed in the plant. The corrective actions included implementing appropriate corrections to the Updated Final Safety Analysis Report. However, engineers closed CR 97-1605 with the justification that corrective actions for CR 96-1857 would result in the performance of a Regulatory Guide 1.97 submittal review and the required Updated Final Safety Analysis Report updates. The inspectors determined that l l

the Updated Final Safety Analysis Report had not been corrected. Because the licensee limited the scope of the corrective actions to containment isolation, the corrective actions for CR 96-1857 would not have addressed the subject components and could have resulted in failure to correct the failure identified in CR 97-1605. The inspectors discussed this deficiency with licensee personnel who agreed that CR 97-1605 had been inappropriately closed. The licensee initiated an action to ensure the identified Updated ,

Final Safety Analysis Report discrepancies would be correcte The licensee indicated that another opportunity would have existed to recognize that CR 97-1605 corrective actions were incomplete during the CR closure review. The inspectors agreed that the potential existed for the process to have captured the error.

l f

I

• I

\

l

.

-21-b.11 CR 97-1804 Engineers initiated this CR to document that personnel had installed Hilti Kwik bolt anchors on ACCW system seismic restraints using Hilti Kwik Bolt 11 installation procedures. The Hilti Kwik Bolt 11 installations differed from those of the Hilti Kwik Bolt installations in terms of embedment depth, torque, and hole diameter for anchor bolts of comparable size. The inspectors reviewed the engineering evaluation that demonstrated acceptability of the as-built configuration. The inspectors agreed that the configurations were acceptable for the expected stress levels during design basis accident The inspectors identified one minor concern with the engineering evaluation. A key test document used to support the operability of the Hilti Kwik bolts installed in accordance with the Hilti Kwik Bolt 11 procedures was not adequately referenced in the CR. The inspectors relied upon this test document to convince them that the configurations were acceptable since they had been tested. Specifically, this test document contained test data supporting adequate Hilti Kwik bolt tension capacity at reduced Hilti Kwik bolt embedment depths. The test report (File H2189-S1, Report 8783 R) was referenced in j the CR as "Hilti Fastening Systems Catalog " The licensee indicated that corrections to the evaluation references would be made. The engineering evaluction adequately supported the operability of the as-built Hilti Kwik bolt b.12 CR 97-2691 This CR documented that a crack was found in the valve manual actuator housing for Valves CC-161 A, Dry Cooling Tower A Tube Bundie 1 outlet, and CC-167A, Dry Cooling Tower A Tube Bundle 4 outlet. Engineers visually inspected other, similar dry cooling tower outlet valve manual actuator housings to ensure that similar deficiencies did not exist. Operators considered the system operable because the valves stroked smoothly l without the cracks getting larger and because the valve actuator housing functions to maintain the grease and protect the gears from the elements. In addition, these valves had a passive open safety function since they are protected by the dry cooling tower tornado missile shield and no active closed safety function. The licensee evaluated the valves upon removal and found no evidence of deformation of the screw stop that would i

'

provide indication that excessive force was used to manipulate the valve actuator. The licensee had an offsite laboratory evaluate the material composition and identify a probable root cause for the cracked actuator housin i An offsite laboratory confirmed that the valve actuator housing was cast grey iron, ASTM A126, Class A material and identified that a warped adjoining surface at the base of the actuator caused the cracking. The laboratory recommended that the licensee replace the valve actuator housings and measure the flatness of the adjoining surfaces using a 0.005" feeler gauge with the mounting bolts detensioned. The licensee also l initiated plans to change the vendor technical manual to reflect the step to check the

[ flatness of the adjoining surfaces during installation / repair. The inspectors found the I corrective actions appropriate to the circumstance _ - _ _ - - - _

_

la 1 a-22-b.13 CR 98-0056 Unlike the valves described in Section E7.3.b.12, which had a passive safety function, the valves for the issue identified on this CR had an active safety function to close during a tornado under certain conditions. Procurement engineers initiated this CR to reflect that a commercial grade Kenneth Elliot 2M valve actuator had been installed without having been dedicated for safety-related use. As documented in NRC Inspection Report 50-382/97-28, Section M2.1, the licensee had installed the Kenneth Elliot 2M i valve actuator on January 6,1998, on the Valve CC-1778, Dry Cooling Tower B Tube l Bundle 9 outlet, when the original valve actuator shaft was found corroded to the actuator housing. Valve CC-1778 is located outside the dry cooling tower missile shield and, therefore, has an active safety function to be manually closed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following a tornado if the associated dry cooling tower bundle becomes damage This CR documented that the valve safety classification did not take into account the active safety function of the valve, as a result, the licensee procured the actuator as commercial grade. Although the CR indicated that no other nonsafety-related replacement actuators had been installed, the inspectors determined from discussions that the originally installed valves were also commercial grade. As documented in the operability evaluation, the licensee inspected other actuators and found no others visibly damaged or incapable of being stroke The root cause determination identified that the safety function of the valve was not identified because the procurement engineer failed to review the most current Updated Final Safety Analysis Report information (Licensing Document Change Request 97-0135), and information related to the license basis change was not communicated to all required departments. As corrective actions, in part, the licensee evaluated the condition of the other valves, confirmed that all valves had been procured commercial grade, and initiated a commercial grade evaluation in order to qualify the actuators for safety-related us As actions to prevent recurrence, the licensee will: (1) for any manual valves with an active safety function, perform commercial grade evaluations, as necessary, to determine whether the valves can perform their safety function; (2) evaluate existing commercial grade evaluations for manual active valves to ensure the evaluations are appropriate; and (3) route the CR to other engineers in an attempt to sensitize the engineers to communication issues. Other long-term projects ongoing that could result in issues similar to this and will aid in preventing the problem in the future include the O-List verification project, discovery phase of the design basis review program, and other ongoing actions to make review of pending Updated Final Safety Analysis Report changes more user friendly.

l For the Kenneth Elliot 2M actuators, the inspectors reviewed the seismic qualification i packages, the design drawings, and a recently drafted commercial grade dedication and

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

,,

o-23-discussed the actuator operation with engineers. The inspectors concluded that the actuator housing was designed as a nonload bearing component and agreed with the commercial grade dedication of the valve actuator Procedure NOECP-153, " Commercial Grade Item Dedication Evaluation," Revision 2, requires the engineer preparing the commercial grade evaluation to gather documents as necessary to understand the function and design of the item being evaluated. The commercial grade item evaluation, prepared in October 1997, failed to consider the design functions for several safety-related actuators as identified in Licensing Document Change Request 97-0135, approved in January 1997. This licensee-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll. of the NRC Enforcement Policy. Specifically, the violation was identified by the licensee, was not willful, actions taken as a result of a previous violation should not have corrected this problem, and appropriate corrective actions were completed by the licensee (50-382/9802-03). Conclusions Based on the CRs reviewed, the inspectors concluded that the corrective action program provided appropriate resolution to significant conditions adverse to quality. The inspectors determined that operability assessments were appropriate, root causes were identified and, generally, effective corrective actions were implemented. Engineering provided good support to resolve adverse condition $

E7.4 Ooeratina Exoerience Program Insoection Scoce (40500)

The inspectors evaluated the operating experience group evaluations and engineering response for five NRC Information Notices and six Independent Technical Reviews of l

'

10 CFR Part 21 report Observations and Findings The specific information notices reviewed are listed belo Subject 96-031 Cross-Tied Safety injection Accumulators96-045 Potential Common-Mode Post Accident Failure of Containment Coolers

!

l 97-009 Inadequate Main Steam Safety Valve Setpoints and issues I Associated With Long Main Steam Safety Valve inlet

_ - _ _ _ _ _ _ - _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ ___ _ .

S

.

l .

i -24-

!

,97-013 Deficient Conditions Associated with Protective Coatings at

!

Nuclear Power Plants l

l 97-025 Dynamic Range Uncertainties in the Reactor Vessel Level Instrumentation

!

l The specific areas of 10 CFR Part 21 reviewed are listed below.

l Subiect 97-001 ITT Barton Model 763 Gage Pressure Electronic Transmitters97-004 ABB/CE Main Steam Safety Valve Piping Pressure Line Loss Considerations

<

97-007 Counterfeit Limitorque Parts97-009 Excessive Failure of Borg Warner Pressure Switches l 97-043 Cooper-Bassemer KSV Emergency Diesel Generator Piston

,

Crown Thickness Conclusions I The inspectors found the evaluations appropriate for the information notices reviewe The inspectors found the 10 CFR Part 21 evaluations appropriate. In addition to the evaluations reviewed above, the inspecbrs evaluated the licensee response to a 10 CFR Part 21 report related to Potter Brumfield motor-driven relays (refer to Section M2.1).

l-E8 Miscellaneous Engineering issues (92903)

E (Closed) Insoection Followuo item 50-382/9725-011: Evaluation of Configuration Control Issues

- The inspectors initiated this open item to ensure that the licensee had a means to identify and evaluate configuration controlissues. The inspectors had identified one engineering  ;

request that identified a configuration control deficiency and two CRs that identified i l

configuration control deficiencie i f

During this inspection, the inspectors determined that the licensee had recently closed l an adverse trend CR related to configuration control. From discussions with the individual who closed the CR, two factors contributed to this dacision. The first reflected a general decline in the subcategories that constituted configuration control (e.g., wiring discrepancies, Station Information Management System errors, fastener discrepancies, installation discrepancies) and the realization that configuration control discrepancies l

_ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ ___ _____ __

o I

-

l

-

l

-25-i would always occur at some frequency with such a broad characterization. The individual indicated that drawing discrepancies weie being monitored closely since the

.ecent number of errors reflected a declining trend and this category was on the trend watch list. Further, the individuals who trend CRs perform " key word' searches on CRs daily as they are processed.

!

V. Management Meetings l

X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management on February 20,1998. The licensee acknowledged the findings presented, and the Directnr, Plaret Support committed to revise a procedure to require that the personnel who need special eyewear for respirators would be informed and sign that they understan The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie l

'

l l

!

l l

- - _ - _ _ - _ _ _

_ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ . _ _ _ _ _ . _

9 e

'

ATTACHMENT SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED f Licensee R. Allen, Manager, Operating Experience Evaluation R. Burski, Director, Plant Modifications and Construction F. Drummond, Director Site Support C. Dugger, Vice President, Operations i

T. Gaudet, Manager, Licensing E. Ewing, Director, Nuclear Safety & Regulatory Affairs G. Fey, Supervisor, in-House Events Analysis j J. Hoffpauir, Operations Manager J. Howard, Procurement / Programs Engineering Manager T. Leonard, General Manager, Plant Operations D. Matheny, Manager, Outage G. Pierce, Director of Quality D. Vinci, Superintendent, System Engineering A. Wrape, Director, Design Engineering NflC T. Andrews, Emergency Preparedness Analyst J. Edgerly, Reactor Engineer J. Keeton, Resident inspector INSPECTION PROCEDURES USED 37551 Onsite Engineering 40500 Effectiveness of Licensee Controls in identifying, Resolving, and Preventing ;

Problems l 62707 Maintenance Observation 71707 Plant Operations 92901 Followup - Operations 92903 Followup - Engineering  !

I

_ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - - _ - - - - - _ - - . - - - - . . - - -

, ,

e I

.

m o

l-

-2-l l

t l ITEMS OPENED AND CLOSED J

Ooened L 50-382/9802-01 IFl Evaluate resolution of Potter Brumfield MDR relay failures (Section M2.1) {

50-382/9802-02 NCV Failed to control design basis containment spray flow rate (Section E1,1)

50-382/9802-03 NCV installed nonsafety component in safety-related application

.

(Section E7.3.b.13)

t l

l Closed

!

50-382/9802-02 NCV Failed to control design basis containment spray flow rate (Section E1.1)

50-382/9802-03 NCV installed nonsafety component in safety-related application (Section E7.3.b.13)

50-382/9724-03 IFl Review workaround evaluations (Section 08.1)

50-382/9725-12 IFl Evaluation of configuration control issues (Section E8.1)

LIST OF ACRONYMS USED

!

ACCW auxiliary component cooling water l

CFR Code of Federal Regulations CPC core protection calculator CR condition report l gpm gallons per minute MSLB main steam line break )

NRC U.S. Nuclear Regulatory Commission PDR Public Document Room

,

psid pounds per square inch differential l

_ _ _ _ _ _ _ _ _ _ -