IR 05000382/1989011

Insp Rept 50-382/89-11 on 890327-31.Violations Noted.Major Areas Inspected:Engineering & Technical Support Activities & QA Audits of Activities

ML20246P949
Person / Time
Site:	Waterford
Issue date:	05/08/1989
From:	Barnes I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20246P933	List: IR 05000382/1989011 ML20246P929 ML20246P945
References
50-382-89-11, NUDOCS 8905220317
Download: ML20246P949 (19)

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APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-382/89-11 Operating License: NPF-38 l Docket: 50-382

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Licensee: Louisiana Power & Light Company (LP&L)

l 317 Baronne Street a L New Orleans, Louisiana 70160 ,

Facility Name:- Waterford Steam Electric Station, Unit 3 (W-3)

Inspection At: Taft. Louisiana Inspection Conducted: March 27-31, 1989 Inspectors: . Wagner', Reactor Inspector'

D. Hunter, Senior Reactor Inspector M.-Murphy, Reactor Inspector Accompanying Personnel: S. Klein, Consultant, ERCI -

S. Kobylarz, Consultant, ERCI Approved: I8%

I. Barnes, Chief, Materials and Quality

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.Date Programs Section, Division of ~ Reactor Safety Inspection Summary Inspection Conducted March 27-31,1989(Report 50-382/89-11)

Areas Inspected: Routine, announced inspection consisting of evaluating the engineering and technical support activities and the QA audits of those activities. The engineering organization was evaluated for size, workload, qualification, and training. The quality-of the engineering performed was'

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evaluated by reviewing completed station modification and design change work

/ packages._ The QA audits of the engineering and technical support organization were also' reviewed, and the corrective actions taken with respect to the audit findings were evaluate Results: Within the engineering and technical support activities, two examples of violations related to inadequate design review were identified I (paragraphs 2.c.(1) and 2.c.(2)). Both~of the examples were identified when L the NRC inspectors reviewed completed design changes and discovered a L discrepancy in the original design.- Although the implementing procedures were good, the licensee's engineering evaluation summaries did not, in general, L contain sufficient detail'to ascertain the significance of the project being

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evaluate In addition. the NRC inspectors' identified two instances in which normally accepted levels of conservatism did not appear to have been utilize .

Within'the area of,QA audits,'one unresolved item was identified (paragraph 3.c) pertaining to adequacy of corrective actions taken in response

.to identified discrepancies. The NRC inspectors found that the QA audits and surveillance of-the design. function had been of sufficient scope and depth to detect' the engineering weaknesses in most all cases. The areas of audit checklists, documentation of findings (observations and closed findings),

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!" tracking and trending, and notification of management could be strengthene ,

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DETAILS > Pers'ons Contacted LP&L-s

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.D. . Baker, NuclearL 0perations Support; Manager; ,

t  : T.' Brennan, . Design Engineering Manager. .

A.- Briody, Nuclear Operations Engineering and Construction (N0EC)

.~ Deputy Assistant L N..' Carns, Plant Manager.-

p; , , M. Ferri, Supervisor, Modification Engineering

.S. Fisher,' Principal Oversight Engineer C. Gaines, Events. Analysis, Supervisor

, .T. Gray, Operations-Quality Assurance (QA) Supervisor

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J. Holman, Safety and Engineering Analysis Supervisor J. Howard, Procurement Engineering Manager . . .

D. Klinksiek, Supervisor Mechanical / Civil Design Engineering L. Laughlin, Site Licensing Supervisor L G. Koehler, QA Audit Supervisor ,

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J. McGaha,.N0EC Manager B. Thigpen, Nuclear Operations Construction Manager

< NRC-T. Staker ' Resident Inspector-

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The above personnel attended the exit interview conducted on March 31, 1989. The NRC inspectors also contacted and interviewed other licensee'

personnel during the course o.f the inspectio . Engineering and Technical Support Activities The NRC inspectors evaluated the effectiveness of the LP&L nuclear

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engineering program in the areas of adequacy of staffing levels and experience, training, design changes, and QA audits. The evaluation consisted of documentation reviews and personnel interviews to verify that

' the license requirements included in the Technical Specifications. (TS) and codes and standards were being implemented and that the commitments contained in the Updated Safety Analysis Report (USAR) and other correspondence.were being followe Organization and Staffing (37702 and 40703)

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The NRC inspectors reviewed the Nuclear Operations Engineering Procedure (N0EP) manual, the Nuclear Operations Engineering

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Instruction (N0EI) manual,andselectednuclearoperations-administrative procedures, a partial listing of which is included i an Attachment to this repor This review verified that administrative controls had been established which described the

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-4-l responsibilities, authority, and lines of communication for personnel performing support functions. These functions consisted of design, technical support, quality assurance, construction,- and procuremen l The procedures reviewed were in conformance with 10 CFR Part 50 Appendix B, and the licensee's approved QA progra The licensee's organization was structured so that the Nuclear 0A Manager and Vice President-Nuclear reported to the Senior Vice I President Nuclear Operations. Reporting to the Vice President-Nuclear were the Nuclear Plant Operations Manager and the Nuclear Operations Engineering and Construction (N0EC) Manager. Reporting to the 4 Nuclear Plant Operations Manager was the Assistant Plant Manager Technical Services (APMTS). The major engineering and technical support functions reported to the N0EC manager and the APMTS. These functional areas were alltlocated at the W-3 site. The organizational alignment at the time of the inspection, reflected the management reorganization that became effective August 8,198 Plant engineering reported to the APMTS and was composed of two groups; these were systems engineering and reactor engineering and perfonnance. Systems engineers performed all aspect.s of interim modifications, became involved in and tracked regular modifications, and acted as test engineers for postmodification test performanc Reactor' engineers were involved in fuel handling, poststartup physics tests, and surveillance tests on core performanc Procurement / programs engineering, construction, modification control, safety and engineering analysis, and design engineering reported to the N0EC manage Design engineering was composed of mechanical / civil, electrical, and instrumentation and control engineerin The total engineering complement, at the time of.this report, was 144 individuals with specialized experience. Members of the engir,etring staff were required to have, as a minimum, a Bachelors degree in engineering or in an appropriate field. The licensee stated that the average experience of the engineering staff was 12.4 years with 4.39 years average experience at W-3. The engineering turnover rate in 1987 was 2 percent and in 1988 was approximately 8 percen The NRC inspectors determined that the engineering workload was level and manageable in all areas except design engineering. The heaviest backlogs in this area were old modification package closecut and resolving plant engineering information requests. Major modifications for the next outage were scheduled to be handled by  !

consultants. The licensee's goal was to become an independent design engineering group. To this end, the engineering organization had moved from project to discipline type engineers with a significant increase in staff size over the past 2 years. The existing consultant work was being integrated into the LP&L organizations with direct supervision by licensee personnel. Consultant group efforts

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(that function independently were assigned to a senior engineer as coordinator / interfac '

No violations'or deviations were identified in the-area of-organization and' staffin Training' (40703)

The NRC inspectors reviewed the licensee's. training program for the W-3 technical staff. This program had been INPO certified. All new technical-staff members were required to complete " Introductory Training" within the first 6 months. 'Upon completion of the required training, the training department offered courses in applied fundamentals and plant systems. Assignments for further.. training were developed by the staff supervisors dependent on assignment, experience, and nee ' No violations or deviations were identified in the' area ofLtrainin Design Changes and Modifications- (37700 and 37701)

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At W 3, conditions requiring engineering evaluation included station modifications (SMs), (recently instituted) design change packages (DCPs), nonconformances (NCRs), and spare parts equivalency evaluationreports,(SPEERs). The NRC inspectors selected a sampling of modifications to the plant derign for review, including SMS and

DCP In general, the modifications which had been made did not involve major changes- to the plant design and were limited in scop The following modifications were reviewed in detail:

• DC 3056, Revision 2, dated January 17, 1989, " Installation of Pressure Bleed-0ff Valve for Air Accumulator Check Valve Test" SM 896, Revision 5. dateo January 22, 1988, "CCW Surge Tank Vent Isolation"

• - DC 3005, Revision 0, dated February 26,'1988, " Provide Throttle Capability for SI-135A and SI-135B"

SM 1432, dated June 13, 1986, " Add Throttle Settings for SI Throttle Valves"

.SM 83-0035, " Replace bypass Transformers ~for ELGAR Inverters,

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SUPS 3A-S and 3B-S"

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SM 84-0511, " Frequency and Alarm Trip Setpoints for SUPS 3MA-S, 3MB-S, 3MC-S, and 3MD-S" l.

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The NRC inspectors found that the modification descriptions. Safety Reviews, and Safety Evaluations (performed to meet 10 CFR 50.59), '

were not detailed and did not always clearly describe.the changes (

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-6-being made. Safety evaluations did not always include sufficient rationale to substantiate. the conclusions establisht d. Without improvement in descriptive details and substantiation of the safety evaluations, future. design changes could be made without

< consideration of important' safety elements affected by the modification being evaluated. In addition to.these general observations, the'NRC. inspectors' identified concerns related to the modifications ' discussed below:

(1). DC 305 Valves SI-602A&B are air actuated,' safety-related, containment sump recirculation isolation valves. These valves are required to be closed during the safety injection phase and open on a recirculation actuation signal (RAS) during a design basis accident. The valves are positioned by the instrument air system,-which was not safety-related or seismically cualifie In the event of a loss of instrument air, an air accumulator is supposed to supply. air to operate its associated valve. A check valve was located immediately upstream of each accumulator to maintain' accumulator pressur A modification was performed to install a bleed-off valve upstream of the check valves to permit depressurization of the air line when isolated from the instrument air system to facilitate periodic leak testing of the check valves. The NRC inspectors identified weaknesses in the postmodification testing, the analysis performed to establish leakage acceptance criteria, and a concern related to the design basis for the size of the accumulator tan LP&L Calculation EC-M89-014 (Revision 0, dated March 6, 1989, Allowable Air Leakage Rate - Valves SI 602A&B) was performed to determine the maximum allowable leak rate from the airs accumulator system for= operating Valves SI 602A&B. The input criteria for the calculation indicated that the accumulators were sized to cycle the valves'once (one stroke closed followed by one stroke open) in I hour based'on a referenced Ebasco Specification LOU, 1564.109A. While LP&L was' unable to provide the basis for this requirement, the NRC inspectors concluded that the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> was based on the time required to generate the RAS signal during a design basis accident, which is nominally on the order of 20 minutes. (This is the time necessary to pump down the refueling water storage pool in a postulated large-break loss of coolant accident (LOCA)). In that case, the air accumulators were sized to sustain leakage for up to I hour while maintaining sufficient capacity to operate the valves upon deman However, during a small break LOCA, the time required to drain the refueling water storage pool could be substantially greater

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than I hour. The NRC inspectors were concerned that the accumulators were not adequately sized for these accident '

In their February 21, 1989, response to NRC Generic

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Letter 88-14. " Instrument Air. Supply System Problems.Affecting Safety-Related Equipment,"Lthe licensee indicated that the possibility'that "these valves ~may need to operate beyond.the

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present 60 minute' limit" wasibeing'further; reviewe ,

I L In addition to the above concerns, the NRC 1rispectors determined 1 that.the analysis to establish allowable leakage rates for check valve leak: testing was not conservative. 'The calculation

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(EC-M89-014 above) used an average air system pressure to idetermine allowable leakage, rather than the lowest anticipated operating pressule (95 psig), and no margin was applied to the

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calculated allowable leakage.(pressure-decay rate) for use in the test. Consequently, the full . calculated decay rate of 9 psi / hour was utilized as the acceptance. criteria for leak testing (0P-903-032, Revision 6)._ Using the methodology in the calculation, the NRC inspectors independently determined that 4 the allowable pressure decay rate would be less than 8 psi / hour based on the lowest system air pressure of 95 psig. Testing performed on March 27, 1989, for these check valves, showed t pressure decay rates of 6 and 7' psi / hour; following repairs,

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testing subsequent to the inspection showed decay rates of 0.5 psi / hour on both valve The DCP included a brief Acceptance Test performed for the modification to " verify capacity of accumulators to cycle SI-602A(B) after IA isolated for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />." The test consisted of depressurizing the air system between the _ isolation valve and the check valve by opening the new bleed valve and stroking Valves SI-602A&B after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The NRC inspectors found that:

• There were no: initial or minimum instrument air pressures specified for the test initial conditions,

• There were'no provisions to record pressures observed or time required to stroke the valve, and

• The only documentation required to confirm satisfactory valve performance was a check mark on the data sheet that the results were " SAT" or "UNSAT."

The NRC inspectors reviewed the data from the DCP required test and noted that some data not specifically required by the test procedure had been recorded. The data indicated that the valves could fully stroke at 95 psig. However, the initial pressures recorded were higher than the lowest anticipated air system operating pressures. In addition, a handwritten note on the

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! Surveillance Procedure OP-903-032, Revision '6L was) performed 'on" . ,

Valves SI-602A&B to confirm that;the valves" stroke times were ~

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within acceptable-limits. 'While~the data indicated that stroke-times for both valves 1were within limits, there was'no basis for; '

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15 second stroke time accepta'nce criteria used.in the. test.' '

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LP&L- personnel . informed the NRC inspectors that vendor-.

recommended stroke = times were used, if no other requirement was

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identified. ' The NRC inspectors foun'd that the vendor..l(Fisher ~,

. oControls data' sheet.for the valves). indicated aLvalve) stroke time of 5 seconds, maximum., Using.this criterion,:

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Valve SI-602B, which tested at an actual stroke' time of- '

6 seconds, would.have, failed the. tes As a' result of: these observations, the _NRC inspectors were -

y concerned.that:

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": The accumulators may not'have adequate 1 capacity to operate these critical safety-related valves on demand.

I' * Weaknesses may exist in the definition of postmodification testing requirements and acceptance: criteria _at W- '

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n.* > - .* :This modification to the plant design appeared to have been

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made without'an adequate review of the original-design basis for the system to confirm that a sound basis existe for the; changes being mad , , , . .

1. - In response to the.NRC inspectors! concerns. LP&L personnel indicated that corrective actions had been initiated relative;to

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(a) 'A Non Conformance Report. Condition Identification (NCRCI)

1. - re t an 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

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+> (b) Administrative procedures were initiated to maintain instrument air pressures above 107 psig header pressur '

The 107 psig header pressure was based on'an LP&L evaluation, performed after the' inspection, of small-break

' LOCAs, which could require more than.1. hour to drain the refueling water storage pool. In addition', these

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Administrative' procedures would include a set of curves depicting allowable system pressure decay rate versus required system pressure to assure two full strokes of the recirculation valves.for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> subsequent to a loss of instrument air. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> requirement was based on the LP&L evaluation which determined that the refueling

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- water stora'ge pool' would be drained in less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for break sizes exceeding .01 square foot. Two full

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strokes'of the valves'were; required to-account for.the

^ possibility that the1 valves could be open prior to the-accident. The curves would also. form the basis for establishing monthly; surveillance '

testing allowable leak  ;

rate acceptance criteri , T O licensee' discussed the.above commitments.with NRC .

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Region IV and NRR personnel subsequent to the inspection.-

In addition to. refining those consnitments, LP&L personnel-stated that an administrative requirement had been established for the instrument' air system; i.e., if system

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pressure were to reduce to less than 90 psig and could .not

[ - be restored;to greater than 90 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the; plant would_be shutdow , . .

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Violation:-LThe. failure to control the design of.the

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accumulators: adequately in order to provide assurance'of containment sump: recirculation valves (sis 602A and 602B) .

operation is an apparent violation (382/8911-01).

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(2) SM 83-0035'

SM 83-0035 upgraded the bypass transformers for the Class IE f static' uninterruptible power supplies (SUPS) 3A-S and 38-S.

!- Although no problems were identified with this modification, the

< NRC inspectors identified a potential problem with the availability of power from the inverter during conditions when a fault occurred on a branch feeder circuit fed from a downstream distribution pane The NRC inspectors reviewed the instruction manual for SUPS 3A-S and-3B-S, which were manufactured by ELGAR, and determined that

, the model UPS 103-1-151 inverter would shut itself down if the inverter output was subjected to a fast overload condition. A

" fast overload condition". shutdown could occur at an inverter load current greater than 165 percent of rated, as noted in paragraph;4-4.8.1 of the manual, which could result from a fault

condition'on the output of the inverter. An inverter shutdownc on fast overload would be automatic and would last for approximately ore-half second before the inverter would restart.

l- EUpon restart after a fast overload, the o6tput voltage of the '

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inverter would be slowly " ramped-up," and this could take on the

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order of seconds to ' complete. '

Since the design of the SUPS were such that the inverter bypass power source was utilized only during maintenance of the inverter.and since the bypass source could only be connected to '

the distribution bus by manual operator action, the Class IE  ;

distribution bus powered by the inverter would be temporarily _

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-10-de-energized during the fast overload or fault condition described abov FSAR Section 8.3.1.1.2.11. " Electric Circuit Protection Systems," stated that the electrical protection had been designed for selective tripping, so that only the affected circuit, close to the point of fault, would be isolated. During a fast overload, the inverter shutdown mode of operation would effectively be racing the operation of the downstream circuit

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protective devices (both molded case circuit breakers and fuses) i to interrupt the fault. The licensee's engineers contacted the inverter manufacturer, during the inspection, to confirm whether the inverter would or would not go into a shutdown mode for a fast overload or fault condition. The inverter manufacturer could not confirm that the inverter would not shutdown during fast overloads such as fault conditions. Since the electrical overload protective devices for the inverter and the distribution system branch circuits were apparently not  !

coordinated for all fault conditions, the design did not appear to satisfy the FSAR commitment for selective electrical protectio The NRC inspectors were concerned that the vital distribution system powered from SUPS 3A-S and 3B-S could be subject to a common loss of power event, as a result of fault conditions, if the Class IE power distribution panels powered Non-Class IE circuits. Non-Class 1E circuits may fail during design basis accidents, such as a seismic event, causing a loss of power in both of the redundant safety / shutdown control trains if each redundant train powered Non-Class IE control circuits. The licensee's design engineers reviewed the loads on the distribution panels powered that Power Distribution Panel by(the SUPS PDP) N A-Ssupplied 390-SA, and 3B-S by and found SUPS 3A-S, powered one Non-Class 1E centrol circuit. Drawing No. LOU 1564, B-289, Sheet 147A, Revision 8. for PDP 390-SA showed a Non-Class 1E Telephone Cabinet (PEC) was powered from circuit No. 65. This was the only Non-Class 1E circuit found on the SUPS 3A-S and 3B-S distribution system The NRC inspectors also reviewed the Associated Circuit Analysis Calculation EE5-32-02, Revision 0, which was prepared as part of the licensee's fire protection program. Part 4B of the analysis demonstrated the selectivity and the coordination of the electrical protective devices on the 120V AC SUPS panel Part 4B.3.1 stated that the SUPS 3A-S and 3B-S operated in a similar manner as SUPS 3MA-S, 3MB-S, 3MC-S, and 3MD-S, which i were manufactured by Solidstate Controls, Inc. The NRC inspectors found this analysis to be incorrect in that the fast overload shutdown mode of the ELGAR SUPS (3A-S and 3B-S) resulted in a loss of power; licensee personnel stated that the inverters

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• s suppl'iedbySolidstateControls,Inc.didnotexhibit'the automatic fast overload shutdown feature of the ELGAR inverter During the course of the inspection, the licensee initiate NCRCI 262261 to address the potential loss of power event for SUPS 3A-S during design basis accident conditions. At:the exit'

meeting, the licensee indicated that the immediate corrective-

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action would be to relocate the feeder for the:PEC:from SUPS 3A-S to the Non-Class 1E computer SUP ' Violation: The failure to adequately control the design of the SUP5 3A-5 loading is an additional example of the apparent violationdiscussedinparagraph2.3.(1)above(382/8911-01).

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(3) SM 896 i

This modification installed a check' valve in the component- I cooling water (CCW) surge tank vent line to prevent waste gas activity from entering the tank. A va;uum breaker was also added to the line to preclude excessive vacuum pressures which might result from reductions in tank' level.-

The NRC inspectors.found that the modification package contained-no documented analysis to substantiate the size of the vacuum breaker. In" addition, there_was no analysis'to demonstrate that the size of the vent line was adequate. The licensee initiated calculations to determine the maximum vacuum' pressure the tank could. safely accommodate and stated, during the exit meeting, that the results indicated the tank could sustain exter_nal pressures near 125 psig. The NRC inspectors had no further

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concerns with the capability of the tank to sustain vacuum pressure levels. However, this was an example which contributed to the NRC inspectors' concern that modifications to the plant design could be made without an adequate review of the design basis for the systems being~ change .

- Audits of the Support Functions' (37702)

The~ NRC inspector reviewed applicable QA program procedures, audits,

_ surveillance, and corrective actions associated with plant modification activitie a-. QA Audits-Modifications The licensee scheduled annual audits of selected design control

- activities. :Two audits were scheduled and performed; one in January - February 1987, and the other during September -

November 1988. The NRC inspector reviewed these two audits for '

scope, content, and auditor complement. The QA audits appeared to be acceptable and were performed by lead auditors with additional personnel assigned to the audit team. The certifications of the

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-12-auditors were reviewed by the NRC inspector and found to be acceptable. The audit schedule was provided to the Safety Review Committee (SRC) subcommittee for concurrence in order to ensure SRC cognizance of the QA audit program. The completed audits were i distributed to the audited organization and other licensee management i personne l The review of the detailed audit checklists by the NRC inspector revealed that the design activities were, in most instances, checked only against the procedure requirements and not against both the program and procedure requirements. Interviews revealed that the procedures were considered to be acceptable, based on previous.QA i group independent reviews of the procedures and procedure change The audit checklists could have been more comprehensive (e.g.,

included a comparison of selected modification procedures / procedure changes with the QA program and regulatory requirements). This mattar was. discussed with the licensee repre:,entative for consideration. Interviews revealed that the licensee was in the process of including members of the SRC on the audit team to strengthen the overall audit program and to increase SRC involvement on the performance of the QA audit The NRC inspector reviewed the audits to verify that the audit results and effectiveness of the OA program associated with the area audited were addressed. The specific documentation of the sumary of the audit results was not apparent; however, interviews with licensee representatives revealed that the results of the audit were considered to be indicated by the audit findings (or lack of negative findings) and represented the effectiveness of the QA program elements which were audited. This observation was discussed with the licensee representatives, in that an annual or biennial sumary of audit results provided to management based on the QA audits, QA surveillance, and the associated findings would likely provide a significant enhancement to the effective implementation of the overall QA progra Discussions with licensee personnel and records review revealed that the QA group performed additional, special audits as requested. Also unscheduled audits of an onsite contractor's (Paul-Monroe) activities were routinely audited. Unscheduled and special audits were addressed in the Nuclear Operations Management Manual (Section VII, '

Chapter 9. and Section V, Chapter 18); however, the quality program lacked specific definition of unscheduled and special audits. This item was discussed with the licensee for consideration. The NRC inspector had no further questions regarding this matte b. QA Surveillance - Modifications

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Document review and personnel interviews conducted by the NRC inspector revealed that QA surveillance of specific activities were i performed to supplement the QA audit program. The NRC inspector l

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L -13-selected eight surveillance concerning modification activities for review..'The surveillance were performed by QA group personnel when deemed appropriate or as a result of special requests. The NRC inspectors had no questions regarding this matte A Audit and Surveillance Findings The NRC -inspector reviewed selected QA audit and surveillance findings to assess the reports and to determine if the dispositions of the findings were thorough and the corrective actions implemented in a timely fashion. In most instances, the findings were acceptable. The following comments were note The review of QA scheduled Audit SA-88-005.1 (Organization and Quality Assurance Program) revealed that a finding regarding the apparent " misclassification of. findings" had been identified by the licensee. The finding was documented on a quality notice (QN)

(QA-88-036) and corrective action was nearing completion. The finding identified the practice of the QA group of identifying a

" condition adverse to quality" and closing the finding without issuing a QN. Interviews and document reviews revealed that the licensee had identified this item previously and planned to change the procedures to require that a QN be issued to identify all

" conditions adverse to quality" to ensure tracking and trending. The NRC inspector had no further questions regarding this ite

The review of QA scheduled Audit SA-87-006.1 (Station Modification / Design, January 28 - February 11,1987) revealed that a

finding regarding the established controls for safety-related structural welding inspection activities was identified. Similar deficiencies regarding structural and ASME welding activities had alsc been identified previously and a QN (QA-86-133, January 7,1987, welding - testing and inspection) had been issued. Since the condition had been previously identified and the QN was still "open,"

an additional QN based on the specific audit findings was not initiate Document review and interviews revealed that the earlier QN (QA-86-133), which addressed deficiencies concerning AWS-D1.1 and ASME requirements, was not revised, or modified to address the specificaudit(SA-87-006.1) finding In addition to the specific welding deficiencies, it was noted in QN QA-86-133 by QA that the findings had potential generic implications to other safety-related welding activities. The QA audit (SA-87-006.1) also identified a deficiency regarding the established controls for safety-related grouting activities and a finding was issued (QN QA-87-065 grouting-procedural controls).

The NRC inspector reviewed the documentation associated with the condition adverse to quality documented in ON QA-86-133. The NRC review revealed a protracted closure of the issues - March 1987 through final closeout of all identified matters in January 198 The NRC inspector's review of the corrective actions associated with i

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l QN QA-86-133 revealed that the finding placed certain station modification packages (SMPs), performed after 1985 through identification i of the deficiency in January 1987, in question. The documentation . ci review revealed that the' engineering department performed a selected sample of SMPs (12) which had been completed during the identified 'l time period; however, it was not apparent to the NRC-inspector that I the sample was representative of all completed work activities or the .l specific findings (observations) identified in QA audit SA-87-00 i Further, the closeout of QN QA-86-133 by the QA group'did not address the specific QA audit findings (SMP-84, SMP-1297, and SMP-195) nor, the generic implications (review of all applicable work activities completed during the period of concern to ensure that all the requirements were addressed) associated with the SA 87-006.1 audit '

findings. This apparent deficiency was discussed with the licensee for consideratio Subsequent to the completion of the required corrective actions concerning the program and procedures in early 1988, a QA surveillance (QS-88-062) was conducted in May and June 1988, to check a specific modification (SMP-138, Reactor Vessel Water Level Indication System). This surveillance identified significant deficiencies concerning safety-related grouting activities (ON QA-88-084, grouting - attention to details and QN QA-88-085, grouting - procedural requirements and clarification) and safety-related welding andinspections). activitieswelding The specific (QN QA-88-082, welding i.e.,

activity findings -(testing failure to perform required nondestructive examination of welds)

resulted in a formal report to the NRC (LER 88-022, Revision 1).

The licensee performed the required nondestructive examinations during a forced outage in September 1988. Training of selected personnel was also provide The noted recurring conditions raise a concern regarding the adequacy of actions taken to preclude recurrence. This subject is considered an unresolved item pending further NRC review during a future inspection. (382/8911-02)

The review of QA scheduled Audit SA-88-005.1 (Organization and Quality Assurance Program, November 21, 1988 - February 10,1989)

revealed that a number of findings were identified, including the failure to identify recurring findings (QN QA-89-031) and the misclassification of findings (QN QA-89-036). Both issues were bein o actively pursued by the licensee to develop a mechanism to identify -

recurring ONs and to ensure that all conditions adverse to quality were identified and entered in a tracking and trending system. The practice of utilizing closed findings and observations in the tracking program was also noted by the NRC inspector during the review of selected audit and surveillance repor The review of QA unscheduled Audit 0A-88-003 (Modification and Test of Hydraulic Snubbers - April 8 - May 6,1988) revealed that the audit results (findings and observations) were transmitted to the contractor (Paul-Monroe), licensee contract management, and licensee

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-15-managemen However, the licensee audit results were not sent directly to the vendor's audit group to provide information to be considered in the next vendor audi The review of QA scheduled Audit SA-88-006 (Design / Station Modifications - September 12 - November 29, 1988), revealed that the audit of the design activities identified a. number of programmatic-and procedural deficiencies as a result of the design program revisions which occurred in 1988. The deficiencies had been addressed by the licensee, including the finding regarding " interim modi fications.

The review of QA surveillance QS-88-072 ~(Post Audit Sampling, June 25, 1988) revealed that an observation was documented regarding gas sample disagreernent. The samples were required to agree within a factor of two. The samples were perforraed again on September 1, 1988, and the three samples were acceptable. The QA the QS open until the observation (deficient samples) wasgroup maintained corrected and verified. This was an example where the condition adverse to quality could have been documented on a QN, to ensure specific and generic corrective action and tracking and trending of the findin The licensee had identified the misclassification of findings on QN QA-89-036 and was pursuing the corrective action The NRC inspector had no further questions regarding the above item Temporary Modifications The NRC inspector reviewed selected controls established regarding temporary modifications to ensure that the licensee requirements were properly implemented. The review of Procedure UNT-5-004, Temporary Alteration Control, Revision 6, revealed the licensee provided controls of temporary jumpers, lifted leads, flanges, hoses, relays, and setpoints. The procedure addressed the review of proposed quality-related temporary alterations (modification / change) by the plant operations review committee (PORC) and approval by the plant inanager (PM). For nonsafety-related alterations, immediate implementation could be accomplished, provided the PORC review, and PM approval was performed within 14 days after implementation. The procedure Attachment 6.3,Section II, Item B.11, addressed the update of necessary drawings; however, the procedure did not address the update of necessary procedures. This item was discussed with licensee personnel for consideration and the NRC inspector had no further questions regarding this matte . Unresolved Item An unresolved item is one about which more information is requested in order to determine whether or not it is a violation, a deviation, or acceptable. One unresolved item concerning corrective action adequacy is delineated in paragraph 3.c of this report.

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p-O 4. . Exit Interview (32703) .

The NRC inspectors met with'the personnel _ identified in paragraph 1 on March 31, 1989, to discuss the findings and conclusion reached during the inspection.- The licensee. personnel acknowledged the findings. No information was presented to the NRC inspectors that was: identified by the licensee as proprietar ,

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ATTACHMENT LIST OF DOCUMENTS REVIEWED i

Drawings N Title LOU 1564, G-285 Main One Line Diagram LOU 1564, G-286 Key Auxiliary.One Line Diagram LOU 1564, G-287 125V DC and 120V AC One Line Diagram LOU 1564, B-289, Power Distribution ano Motor Data 120V Distribution Panel No. 390-SA

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Sheets 147 & 147A LOU 1564, B-239, Power Distribution and Motor Data Sheets 148 & 148A 120V Distribution Panel No. 391-SB Specifications N Title LOU 1564.282A, Static Uninterruptible A-C Power Supply for Class 1E Revision 9 Control Systems LOU 1564.109A, Butterfly Valve Data Sheet, Revision 7. Sheet 16 Revision 4

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Instruction Manual Title 457000387 Volume 1 ELGAR Model UPS 103-1-151, Uninterruptible Power Supply 457000387, Volume 2 Instruction Manual for ELGAR AC Power Line Conditioner, Model PLC 253-1-04 Vendor Drawings Title 1564-1897, Revision 5 Schematic, 20 kVA Inverter Solidstate Controls, In (Drawing No. 014D10915. Sheet 1 of 2)

1564-1898, Revisior. 7 Schematic, 20 kVA Inverter Solidstate Controls, In (Drawing No. 014D10915 Sheet 2 of 2)

Calculations N Title EE5-32-02, Revision 0 Associated Circuit Analysis EC-M89-014, Revision 0 Allowable Air Leakage Rate - Valves SI602A&B l

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Procedures ,

NOP-14, Revision 1. Design Changes

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'NOP.-15.4 Revision 1, Justification for Continued Operation LQAP-000,' Revision 4, Quality Assurance Charter

'QAP-301, Revision-2, Quality. Assurance Review of~ Station Modification Packages x QAP-302, Revision 6 Conduct of Operations' Quality Assurance Audits-QAP-304, Revision 1, Quality Assurance Group Revision of Programs, n' Procedures, and Instructions e

QAP-305.. Revision' 1, Planning' and Scheduling of Operations Quality Assurance Audits

. QAP-306, Revision'1,' Conduct' of Operations, Quality Assurance : Surveillance

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QAP-350,' Revision 1. Review of Work Authorization Packages QAP-366, Revision 0,- Operations Inspections - General PE-TEM-012 Revision 1, Plant Engineering Station. Modification ,

UNT-5-004, Revision 6 Temporary Alteration Control UNT-5-015,, Revision 8. Work Authorization OA Audits - Scheduled (SA) and Unscheduled (UA)

SA-87.006.1, Station Modification / Design, January 28 - February 11, 1987 SA-88-005.1, Organization an'dQuality Assurance Program, November 21, 1908 -

, February 10,1989 SA-88-006, Design / Station Modifications, September 12 - November 19, 1988'

UA-88-033, Modification and Testing of Snubbers, April. 8 - May 6,1988 i SA-88-016.1, Fire Protection and Loss Prevention Program, December 2-9, 1988-QA Surve111ances QS-87-076, Yellow / Orange Jumper Wires, May 20-27, 1987 05-88-050, Main Steam Isolation Valves, April -18 .May 22,1988 05-88-053, Station Modification (Paul Monroe), April 27 - May 6,1988

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b QS-88-062. Station Modification Package 138, May 31 -' June 20, 1988 QS-88-072, Postaccident Sampling' System. June 25,.1987 QS88-087,Acceptanceof-StationModification'818) August 8-9, 1988

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'QS88-089, Evaluations.on SMP-1332, May 5-8,1988 ,-

b QS-89-010, Status of Safety-Related SMs and'DCs, March 2-3, 1989

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Other Documents'

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Condition Identification #262265, 3/30/89, Accumulators.for SI Sump Outiet

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Isolation Valves Are Sized For One Hour Supply Of Air to Valve Operators (N0P-19).

OP-903-032 Revision '6, Surveillance Procedure, Quarterly ISI Valve Tests,'

Section 8.24 Instrument Air. Check Valves'(performed 3/27/89), Section 8.3, )

Safety Injection (1/24/89)

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LP&L Letter to U.S. Nuclear Regulatory Commission, W3P89-0028, dated 2/21/89, Generic Letter 88-14

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