IR 05000382/1987001

Insp Rept 50-382/87-01 on 870101-31.Violations Noted:Failure to Conduct Annual Evaluations of Supplier QA Programs & Failure to Document Designation of Person Responsible for Plant Mgt in Absence of Plant Manager

ML20212A663
Person / Time
Site:	Waterford
Issue date:	02/24/1987
From:	Breslau B, Bundy H, Constable G, Leuhman J, Staker T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20212A501	List: IR 05000382/1987001 ML20212A495 ML20212A566
References
50-382-87-01, 50-382-87-1, NUDOCS 8703030394
Download: ML20212A663 (22)

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APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report:

50-382/87-01 License:

NPF-38 Docket:

50-382 Licensee:

Louisiana Power & Light _ Company (LP&L)

142 Delaronde Street New Orleans, Louisiana 70174 Facility Name: Waterford Steam Electric Station, Unit 3 Inspection At:

Taft, Louisiana Inspection Conducted:

January 1-31, 1987

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Inspectors: "

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J. G. L W ~ Senior Resident Inspector Datef

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' %". F. Bundy, Project Inspector, Reactor Date

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~B. A. /Bresl5u, Project Inspector, Reactor Date

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Project Section C

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. R. 5 taker, Resident Inspector Date Approved:

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G.'T. Constable, Chief, Reactor Project Date Section C

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Inspection Summary Inspection Conducted January 1-31, 1987 (Report 50-382/87-01)

- Areas Inspected:

Routine, unannounced inspection of:

(1) Plant Status, (2) Licensee Event Report (LER) Followup, (3) Followup on Previously Identified Items, (4) Monthly Maintenance, (5) Monthly Surveillance, (6) ESF System Walkdown, (7) Routine Operational Safety Inspection, (8) Containment Integrity Verification, (9) Plant Startup from Refueling /Startup Testing-Refueling,

- (10) Onsite Review Committee, (11) Quality Assurance (QA) Annual Review, (12) License Conditions, and (13) Potential Generic Problems.

Results: Within the areas inspected, two violations were identified (violation of quality assurance procedure for evaluating supplier quality assurance program, paragraph 5; and violation of Technical Specification requirement for designating succession of plant management authority, paragraph 13.a).

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,1 U ~ Persons Contacked

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. J. G. Dewease, Senior Vice President, Nuclear Operations

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1. • R.' P/ Barkhurst, Vice President, Nuclear Operations

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' #N. S.: Carns, Plant Manager, Nuclear

'*T. F. Gerrets," Corporate QA Manager g.~~

S..A. Alleman, Assistant Plant Manager, Plant Technical Staff w

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'J. R. McGaha,: Assistant Plant Manager, Operations and Maintenance J. N. Woods,-QC Manager A. S. Lockhart, Site Quality Manager R. F. Burski,' Engineering and Nuclear Safety Manager

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. #K. L. Brewster, Onsite Licensing Engineer

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G. E. Wuller, Onsite Licensing Coordinator

' ~ T. H. Smith, Maintenance Superintendent, Nuclear M. A. Triggs, Records Manager

• G. J. Brady, Nuclear Personnel Manager W. M. Morgan, Supplier QA Manager J. M. Guillot, System Development QA Supervisor

• N. E. DuBry, Middle South Principal Oversight Engineer

• B. J. Morrison, Licensing Engineer

• Present at exit interview January 9, 1987.

1. Present at exit interview February 2, 1987.

In addition to the above personnel, the NRC inspectors held discussions with various operations, engineering, technical support, maintenance, and administrative members of the licensee's staff.

2.

Unresolved Items One Unresolved item was identified during this inspection and is discussed in paragraph 10 of this report.

Additionally, a previously identified unresolved item is closed in paragraph 5 of this report.

An unresolved item is a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviation.

3.

Plant Status The inspection period began with the reactor in Mode 6 with the reactor coolant loops drained to support steam generator and reactor coolant pump maintenance and inspection.

At 8:28 a.m. on January 7, 1987, the plant experienced a loss of offsite power to the "A" side of the onsite

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electrical; distribution system.

At the time of the loss of offsite power the' reactor was'still in Mode 6, with reactor coolant temperature approximately 100 F.

Upon the loss of power the "A" emergency diesel generator auto-started and sequenced on the required loads.

The "A" shutdown cooling train, which was in operation at the time of the power loss, was' restored to operation at 8:35 a.m.

Because reactor coolant t'emperature remained essentially unchanged the "B" shutdown cooling train, i

which was'available throughout the event, was not started.

Power was restored to the "A" electrical distribution.at 10:39 a.m. and the emergency diesel generator was subsequently secured.

At' 4:40 p.m. on-January 8th, the final reactor vessel head closure bolt was-tensioned and the reactor was placed in Mode 5.

On January 25th at 4:40 p.m. the reactor entered Mode 4 and subsequently. entered Mode 3 at 11:35 p.m. on January 27th.

Because of high temperatures on the upper thrust bearing for the 1A reactor coolant pump a plant cooldown was commenced late in the morning-of January 28th with the reactor reaching Mode 4 at 1:30 p.m. that same day.

The reactor was placed in Mode 5 at 10:31 p.m. on January 28th and the inspection period ended with the plant in that mode and work ongoing on the reactor coolant pump.

No violations or deviations were identified.

4.

LER Followup The following LERs were reviewed and closed.

The NRC inspectors verified that reporting requirements had been met, that causes had been identified, that corrective actions appeared appropriate, that generic applicability had been considered, and that the LER forms were complete.

Additionally, the NRC inspectors confirmed that no unreviewed safety questions were involved and that violations of regulations or Technical Specification (TS) conditions had been identified.

(0 pen) LER 382/86-026 - Inadvertent Fuel Handling Building Ventilation Emergency Filtration System Actuation due to Personnel Error.

The NRC inspector has reviewed this report, verified the licensee's corrective actions and inspected the radiation monitors in the Fuel Handling Building (FHB).

Inexact communications and nearly identical identification tags were contributing factors to the inadvertent actuation of the FHB ventilation described in the LER yet, upon reviewing MI-3-362 and inspecting the radiation monitors, the NRC inspectors found other similar problems.

MI-3-362 Revision 4 Section 7.0 incorrectly identifies the radiation monitors as "the Fuel Handling Building Airborne Isolation Radiation Monitors." while, the labeling on the front of the monitors themselves is nr identical to that given in the LER and MI-3-362.

(Closed) LER 382/86-11, Revision 1 - Walkdown as a Result of Part 21 Report Identified Two Missing Internal Penetration Seals in Addition to Several Deficient Seal Arrangements.

The NRC inspector has reviewed this report and the associated cover letter dated January 19, 1987.

This

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' revised report was -submitted based on the NRC inspector's coments about

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the original report which was discussed and closed in NRC Inspection

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Report 50-382/86-33.

No violations or deviations were identified.

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l 5.

Followup on Previously Identified Items

.(Closed)Open' Item 382/8520-02. QAP 302 required onsite contractors to be audited every 3 years. Beyond that however, the program did not adequately address such areas as; timeframe for audit for a new contractor onsite, placing a contractor on the audit schedule, and tracking closure of audit findings. The licensee has issued NOP-08 which appears to provide programatic controls for the areas mentioned above.

o (Closed) Unresolved Item 382/8520-03. This item noted that Southern Vital Record Company and Siemens-Allis Company had apparently not received the

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required annual evaluations specified in QAP 202. Upon followup of this item the licensee was unable to provide documentation of these

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- evaluations. -In addition,. further review by the NRC inspector found four other annual evaluations not completed within the.12+2 months specified in

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QAP 202. The failure to perform the required evaluations is considered an apparent violation and_is identified as 382/8701-01. The NRC inspector c

reviewed the licensee's 1986 annual evaluation of Seimens-Allis which

contained a reconnendation to. retain the supplier on the qualified supplier list (QSL) but a restriction requiring an audit / surveillance prior to placing an order was added. This restriction was not added to

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the QSL'and subsequently the licensee placed purchase order L-94393 for i

air intake screens without first performing the audit / surveillance. The l

. licensee performed a subsequent review of all suppliers and noted one similar occurrence. The licensee's 1986 annual evaluation of the Foxboro Company also required an audit / surveillance prior to making a purchase but this restriction did not appear on the QSL and the licensee then made four

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L purchases (non-safety) without conducting the audit / surveillance. The failure-to add the restrictions to the QSL are further examples of the failure to follow QAP 202 and as such are considered part of the apparent

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violation identified above.

Licensee management took imediate corrective action as observed by the NRC inspector. A staff meeting was held to review the procedural l

requirements, and a chect on all resource documents was done to identify any similar problems. The NRC inspector was satisfied that the licensee had satisfied the requirements of 10 CFR 2.201 and is now in compliance.

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j No additional response is required.

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- No other violations or deviations were identified.

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6.

Monthly Maintenance Station maintenance activities affecting safety-related systems and l

components were observed / reviewed to ascertain that the activities were

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r conducted.in accordance with approved procedures, regulatory guides and industry codes or standards, and in confonnance with TS.

Portions of the following condition identification work

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authorizations (CIWAs) were observed by the NRC inspectors:

CIWA 018199 - Observed the verification of proper installation of fire seals.

CIWA 027507 - Change out of meter face indications for the meters

discussed in paragraph 7 of this report.

CIWA 030815 - Rerouting of the Emergency Diesel Generator "A" governor cooling water supply from Component Cooling Water to Jacket-Cooling water.

The NRC inspector observed the reinstallation of a RTD in Component

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_ Cooling Water (CCW) Pump "B" per CIWA 020837. The RTD was being

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reinstalled after completion of pump seal replacement per the same CIWA.

The NRC inspector observed that nylon rigging straps were attached to the

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pump suction and discharge piping above the pump. The NRC inspector questioned maintenance personnel on the purpose of these straps. The NRC

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inspector was informed that these straps.had been used to rig lifting equipment during CCW pump seal replacement. Rigging to a pipe snubber was previously identified in NRC Inspection Report 50-382/86-13 when a chain-hoist was rigged to =a snubber in preparation for lifting a containment

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spray pump. The rigging to these components is not specifically addressed

in the maintenance procedures and did not appear to place an undue load on.

these components in these two cases. The licensee has consnitted to the following corrective actions to avoid potential problems in the future.

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An engineering analysis will be performed to determine the effect of

the loads on the CCW piping and supports.

Maintenance personnel will be reinstructed on rigging practices.

• Specific instruction prohibiting rigging from safety-related equipment will be incorporated in general maintenance procedures.

The followup of these actions is considered an open item (382/8701-02).

No violations or deviations were identified.

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Monthly Surveillance The NRC inspectors observed / reviewed TS required testing and verified that testing was performed in accordance with adequate procedures, that test

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instrumentation was calibrated, that limiting conditions for operation (LCO) were met, and that any deficiencies identified were

properly reviewed and resolved.

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During the performance of OP-903-108 "SI Flow Balance" licensee operations personnel observed that reactor coolant system cold leg injection ficw

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indicators were reading flow rates substantially below the minimum test acceptance _ criteria. This condition existed regardless of which high pressure safety injection pump was run and checks of the pumps' flows and pressures indicated that the pumps were performing nonnally. Subsequent review by licensee engineering personnel revealed that the problem was one of improper indication.

A' station modification in progress to update the control room control panels, in part, modified the meter face indications for various control panel meters. The meters indicating cold leg injection flow were improperly changed from meters indicating 0-500 gpm based on a square root output to others indicating a linear output.

In such a condition the indicated flow was well below the actual flow.

Additionally, the non-safety plant computer monitoring points for the injection flow rates were found to be based on a 0-400 gpm monitoring range instead of the present 0-500 gpm range. This second problem has apparently existed since an earlier modification changed the monitoring range. The NRC inspector observed that the licensee had identified this problem and had taken appropriate corrective action including verification that similar problems did not exist.

No violations or deviations were identified.

8.

ESF System Walkdown The Emergency Diesel Generators (EDG) were verified operable by performing a walkdown of the accessible and essential portions of the EDG subsystems on January 7, 8, and 13,1986.

The NRC inspector used the EDG standby system valve lineups specified on attachments 8.1 and 8.2 and the breaker lineups on attachments 8.3 and 8.4 of procedure OP-9-002, Revision 7, in conjunction with the referenced drawings.

On completion of the inspection the NRC inspector had the following comments:

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Section 6.1.3 of OP-9-002, Revision 7, references attachment 8.2

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incorrectly. The correct attachment is 8.3.

I Valves EDA 161A(B) are shown on drawing LOU 1564-G-164 as check valves downstream of the EDG air start valves, and were found to be

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labeled as such. The valve lineups in attachments 8.1 and 8.2 of OP-9-002,Revisien7,incorrectlyliststhesevalves(EDA 161A(B))as air start valves. These air start valves should be listed as EDA 160A(B).

l Valve EGF 1141 is shown on LOU 1564-G-164 as a vent valve installed downstream of the "A" diesel oil transfer pump, this valve is not included in the EDG lineup.

Attachment 8.2 of OP-9-002, Revision 7, incorrectly lists

valves EGA 1318 and EGA 134B as EGA 131A and EGA 134 '

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' There were several errors in the grid location and drawing page

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numbers as referenced in the valve lineup procedures.

Valve EGF-112 is~ inadvertently included'in attachme'nt 8.1 of

OP-9-002, Revision 7.

Attachment 8.1 is the valve lineup for EDG "A" and EGF 112 is'a drain valve located downstream of diesel oil transfer pump "B".

Valves EGF 1211 and EGF 1212 are vent and drain valves located

downstream of the "A" EDG diesel oil transfer pump. These valves are inadvertently included in the lineup of EDG "B".

As documented in NRC-Inspection Report 50-382/86-02, paragraph 8, the licensee has identified a program to upgrade all safety-related checklists, properly tag all plant valves and label all electric breakers.

The upgrading program is: scheduled to be completed at the end of the current refueling outage. The upgrading program completion and correction of the deficiencies identified in NRC inspection reports will be inspected as part of the followup to Unresolved Item 382/8606-01.

The Safety -Injection Tank (SIT) system was verified operable by perfoming a walkdown of the accessible and essential portions of the system on January 13, 1986. The plant was in mode 5 and the SIT tank isolation valves were in the shut position as required.

The NRC inspector used the Safety Injection standby system lineup specified on attachment 8.1 and the breaker lineup of attachment 8.3 of procedure OP-9-008, Revision 6, in conjunction with the referenced

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drawings.

No violations or deviations were identified 9.

Routine Operational Safety Inspection

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By observation during the inspection period, the NRC inspectors verified that the control room manning requirements were being met.

In addition, the NRC inspectors observed shift turnover to verify that continuity of system status was maintained. The NRC inspectors periodically questioned shift personnel relative to their awareness of the plant. conditions.

Through log review and plant tours, the NRC inspectors verified compliance

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with selected TS and limiting conditions for operations.

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During' the course of the inspection, observations relative to protected (

and v_ ital area security were made including access controls, boundary j

integrity, search, escort, and badging.

On a regular basis, radiation work pemits (RWP) were reviewed and the specific work activity was monitored to assure the activities were being conducted per the RWPs. Selected radiation protection instruments were periodically checked and equipment operability and calibration frequency were verified.

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The NRC inspectors kept themselves informed on a daily basis of overall status of plant and of any significant safety matter related to plant operations.

Discussions were held with plant management and various members of the operations staff on a regular basis.

Selected portions of operating logs and data sheets were reviewed daily.

The NRC inspectors conducted various plant tours and made frequent visits of the control room. Observations included: witnessing work activities in progress; verifying the status of operating and standby safety systems and equipment; confirming valve positions, instrument and recorder readings, annunciator alarms; and housekeeping.

On the evening of January 12th, licensee operations personnel were attempting to perform a portion of OP-903-108, Revision 0, "SI Flow Balance".

During the course of the testing a high pressure safety injection (HPSI) pump was started and approximately 50-100 gallons of reactor coolant was inadvertently discharged, through reactor vessel head vent valve RC-101, onto the reactor head and into the surrounding areas of the refueling cavity.

RC-101 is a normally shut manual valve which was found to be partially open.

In following up this event the NRC inspector reviewed both OP-903-108 and MM-8-035, Revision 2, " Reactor Vessel Head Purge and Vent Line Removal and Installation." The latest change to MM-8-035 deleted the requirement to reinstall the blind flange downstream of RC-101.

It is normally installed when the purge and vent line is removed during shutdown for refueling.

This change was apparently made because following shutdown the reactor coolant system is normally vented through this line and the blind flange would only have to be removed a second time to perform the vent if it was reinstalled following the use of RC-101 for the draining done prior to the removal of the vent line.

The revision to MM-8-035 made changes to the initial conditions in OP-903-108 appropriate.

The NRC discussed the lack of detailed initial conditions in OP-903-108 with licensee management and the operations superintendent agreed that a check of the head vent lineup should be made if the vessel head has been removed and a subsequent lineup of the head vents had not been made prior to the test.

He went on to state that this change would be made to OP-903-108.

On January 22nd, the licensee informed the NRC inspectors of an event that resulted in the inadvertent transfer of radiological contamination to areas outside the owner controlled area.

On January 21, 1987, an instrument and control (I&C) technician apparently became radiologically contaminated while working in the Reactor Auxiliary Building (RAB).

Subsequently, the individual was able to leave the plant and proceed to and return from his residence before the contamination was discovered.

The licensee performed an indepth review of the above event and the findings are summarized below:

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• The individual was estimated to have received an exposure of approximately 150 mrem to the skin as a result of the contamination.

The contaminated individual' performed two whole body frisks incorrectly (one upon leaving the RAS job site and a second at the exit of the Radiologically Controlled Area).

The I&C technician and another individual were allowed to work in an area of the RAB which was under control of a Radiation Work Permit (RWP) without following the RWP requirements, based solely on

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the verbal instructions of a Health Physics (HP) technician.j ih.is.is.

contrary to the-approved procedure which only allows approved written changes to downgrade the requirements of an RWP.

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The individual who was contaminated was not logged in or ou', of the RCA on the HP computer as required by approved procedures.

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The contaminated I&C technician was able to exit the site due in

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part, to a lack of interdepartmental communications.

Minutes before he left the site, plant operations persennel performed an electrical switching evolution which deenergized the Protected Area (PA) exit radiation monitors.

Neither the HP or security departments were-notified of the evolution in order to implement compensatory measures.

The day following the initial contamination event the individual attempted to exit the PA during his lunch break.

He was properly

stopped by a security officer when the exit radiation-monitor alarmed.

After. going through the monitor a second time and again alarming it, he was detained until an HP technician could respond.

After only his coat was surveyed and found to be contaminated the individual was allowed to exit the PA.

Followup surveys of the coat showed radiological contamination on the inside of the garment indic'ating other articles of clothing which had not been surveyed were also possibly contaminated.

' After relocating--the individual and questioning him it was determined

.that_the he had actually left the site contaminated the day before

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and returned.

Based on this information HP personnel went to his

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, residence where they did find a shirt that was contaminated; however,

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This event and the licensee's critique of it have been discussed with the NRC Region IV Facilities Radiation Protection Section.

Because of the licensee's indepth review and extensive corrective actions no further actions were deemed appropriate.

Included in the licensee's corrective actions were:

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Retraining of the individuals involve _

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ms Formal briefings for all'. plant' personnel concerning this event.

Revision of. operating procedures.to include notifications for deenergizing the PA exit radiation monitors. Additionally, these monitors have been equipped with lights to indicate t'neir operational status.and instructions to plant personnel on what *c do if the

. monitors are inoperable.

.No violations or deviations were identified.

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Containment Integrity Verification Containment integrity was verified by observing mechanical barriers and isolation valve positions' for several containment penetrations, witnessing

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portions of two, unsuccessful air lock local leak rate test, and walking

_ down the Containment Atmosphere Release (CAR) system prict to commencement

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'of reactor coolant system heatup in order to enter mode 4.

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ThelNRCinspectorreviewedsurveillanceprocedures.PE-5-023," Containment Air > Lock DooreSeal Leakage Test,". Revision 3 and PE-5-024, " Containment

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s ' ; Airc Lock Overall Leakage. Test' and Door Interlock Chock," Revision 2.

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Section 8.3.oftPE-5-024 requires the updating of the total leakage rate

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-and~the bypass' leakage rate based on the results of the test and the

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A section also specifies the maximum acceptable leakage value for-each

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' parameter. :These" maximum values are TS limits; however, the test

[ acceptance criteria are written solely for the specific tests the

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procedure -is designed to perform, which are the airlock leakage rate and the door interlock ~ test.

As mentioned above, the results of the specific

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w tests could change the values of the other two TS parameters; therefore, they too.should beLincluded in the. acceptance criteria. ' Unlike most local leakage rate tests!which are performed with the plant shutdown, the test of the airlocks is required to be performed at design pressure every 6 months.

In such a situation-it is not acceptable to wait and evaluate

.the effects on overall and bypass leakage until the data is transferred to i

the licensee's local leak rate log book because the plant could easily be

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.in'a mcde where containment integrity is required.

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specifying the maximum allowable values in Section 8.3 was meant to I.

accomplish the immediate evaluation; however, these values should be moved

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to the acceptance criteria or they should be clearly referred to as TS limits.

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The NRC inspector performed a walkdown of the accessible and essential

_ portions of the CAR system prior to entry into mode 4.

The NRC inspector used the standby system valve lineup of attachment 1 of OP-8-002, the

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breaker lineup of attachment 2 of OP-8-002, Revision 3, and the referenced L

drawing.

The NRC inspector found the CAR system exhaust fans downstream l

isolation valves (CAR-206A and CAR-2068) and the exhaust system crossconnect valve (CAR-205)'in the open position.

The standby system

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lineup of attachment 8.1 of OP-8-002, Revision 6 requires these valves to be in the shut position when the CAR system is not in operation. Because the CAR system is connected directly to the containment atmosphere, the

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systemcontainmentis$lationvalves;werecheckedandfoundtobe. shut',as

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. required,'thus/ isolating the system from'the containment atmosphere.

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Valves CAR-206A and CAR-206B are required to be open for CAR system.

operation. 'Although,thiere is no direct safety function of these valves, s

as discussed above when the system is not'in ' operation, the system

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operating procedure requires them to be shut.

The NRC inspector was informed byfplant' operations' personnel that these valves were normally

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left in the open p6sition because of frequent use of the CAR system for

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-containment pressure-control. -The NRC inspector was concerned with this

. practice for two reasons.

First, the CAR system interfaces directly with; the. containment atmosphere when'in use and therefore proper system alignment needs to be controlled by procedures.

Second, even with.'such

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< controls in place. system alignment was found to be improper because

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established procedures were not adhered to because of frequent. system

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operations. The NRC inspector was also concerned because the CAR system

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' is. not included <in the system valve lineup checklist of attachment 8.14

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of; !' General. Plant 0perations," OP-010-001, Revision 9.

These concerns

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were discussed with-the plant Operations Superintendent who-stated that a mreview would:be performed on OP-8-002, that the CAR system would be added

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to attachment 8.14 of OP-010-001, and that a review would be performed to

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veriff that' attachment 8.14 of OP-010-001-contains all of :the required

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• While inspecting containment penetration isolations, the NRC inspector s'

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, observed several drain, valves installed between containment penetrations

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7,sandithe~ system penetration isolation valve (outside of containment) in the Safety. Injection (SI) system.

These valves were found to be connected to

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drain piping which ended in open funnels.

The valves were not locked and in the shut position. Although SI system containment penetration

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isolation valves are not subject to containment leak rate -testing, Section 6.2.4.1.2;(a) of the Waterford-3-Final Safety Analysis Report (FSAR) requires that the SI system containment isolation valves

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hsve remote manual operation capability to isolate malfunctioning portions of,the system from the containment.

Section 6.2.4.1.2 of the FSAR

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y requires all lines penetrating the cm.tainment (excluding fluid instrument

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lines) to have a combination of isMaiion valves including on the outside of.the containment either autm@ cd' i operated, remotely operated or a

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F locked closed valve.

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containment isolation penetrations.

Nis section states that test, vent, and drain valves will have a cap or blind flange installed during operation for a double isolation barrier.

The locking or installation of caps on drain valves in safety injection system is considered'an unreso Ned item pending evaluation by the licensee and is identified as

'

Unresolved Item 382/8701-03,

.

t No violations or deviations were identified.

'

11.

Plant Startup From Refueling and Startup Testing - Refueling

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During this inspection period the NRC inspectors reviewed the licensee's preparations for plant startup following refueling, verified that selected f

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starturi including heatup and testing of the reactor coolant pumps.

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.' '(As part _of"the review of the licensee's preparations for plant startup the NRC insp~ ecto & examined the recently approved startup test procedures m. _* i_

=li.sted below:

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-4.NE-2-060 Revision 0, " Isothermal Temperature Coefficient Measurement" NE-2-110' Revision 0, " Fuel Symmetry Verification"

NE-2-120 Revision 0, " Nuclear and Thermal Power Calibration"

NE-2-130 Revision 0, "CPC Verification" NE-2-140 Revision 0, " Core Power Distribution Measurement"

NE-2-150 Revision 0, " Radial Peaking Factor and CEA Shadowing Factor Measurement" Overall, these procedures appeared adequate for conducting the startup test program following the refueling.

The NRC inspector has discussed the following comments about the above listed procedures with a member of the licensee's reactor engineering group.

Steps 5.1 and 8.0 of NE-2-040 reference NE-2-001 instead of NE-2-003.

Step 3.8 of NE-2-150 specifies that reactor power must be

+/-10 percent instead of +/-1.0 percent.

• The numbered substeps in Step 3.8 of NE-2-150 are incorrectly labelled 3.11.1, 3.11.2, and 3.11.3.

The initial conditions of NE-2-130 specify that the test is to be conducted at the specified power +/.5 percent but the specified power is not given in NE-2-130. Additionally, NE-2-130 is not

sequenced or referenced in NE-2-003. Through discussions with

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licensee reactor engineering personnel as well as review of the Final Safety Analysis Report the NRC inspector determined that performance of NE-2-130 is not required during the startup test program.

The NRC inspector performed walkdowns of the EDG and the SIT systems as described =in section 9 of this report.

These walkdowns were performed to verify the lineups for systems that either had maintenance performed during the outage or are inaccessible during normal plant operation.

No violations or deviations were identified.

12.

Onsite Review Committee On January 16, 1987, the NRC inspector attended the regularly scheduled weekly meeting of the Plant Operations Review Committee (PORC). The NRC inspector attended this meeting to ascertain the workload imposed on the PORC under outage conditions as well as to verify conformance with the requirements of TS 6.5.1.

Overall, the meeting progressed smoothly largely because the committee quickly tabled any reviews that were improperly prepared / written rather than using the committee members' time -

to correct errors or omissions that should have been corrected prior to submission for approval.

The NRC inspector found the PORC workload to be of an acceptable volume and the meeting to be in conformance with TS 6.5.1.

No violations or deviations were identified.

13. QA Annual Review The NRC inspectors reviewed various areas of the licensee's QA program to ensure that is was being implemented in conformance with regulatory requirements, licensee's connitments, and applicable industry guides and standards.

The areas discussed below were covered during this inspection included:

a.

QA Program Changes and Administrative Controls The NRC inspectors reviewed QA program-and implementing procedures made during 1986 to ensure continued compliance with commitments.

Also, certain requirements in TS Section 6 were reviewed with regard to current compliance.

A significant staff reorganization was initiated via a memorandum dated October 6,1986, from the Senior Vice President-Nuclear Operations, to various senior managers.

Implementation of the reorganization is ongoing.

Completion of Nuclear Operations policies and interdepartmental procedures is tentatively scheduled for April 1987.

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The NRC inspector reviewed the following memoranda and Quality Assurance Procedure (QAP) which collectively appear to provide satisfactory bases for initiating implementation of the reorganization:

R. P. Barkhurst to all site supervisors dated November 25, 1986, titled, " Site Reorganization-Purpose and Implementation."

QAP-309, dated November 18, 1986, titled, " Quality Department Instructions Transition."

F. J. Drummond to J. G. Dewease with copies to various nuclear operations managers, dated December 5,1986, titled,

" Responsibility Reassignment as a Result of the Implementation of the New Organization."

N. S. Carns to nuclear operations personnel, dated January 5, 1987, titled, " Cross-Reference Lists for Plant Operations Manual / Departmental Changes."

The NRC inspector commented to various licensee senior managers that although the above documents appear adequate to initiate reorganization, they do not appear to be timely.

The response to this comment was that the purpose and effects of the reorganization were discussed in meetings with all employees within a few days after announcement and a good understanding of the reorganization has been displayed by employees during the current outage.

Changes to QA program documents resulting from the reorganization are being overseen by the Nuclear Operations Management Manual (NOMM)

Steering Committee which is chaired by the nuclear services manager and includes the plant manager and nuclear QA manager.

Planning, l

scheduling and work directing for changes is performed by the NOMM j-working committee which includes representatives of the NOMM steering

'

committee members.

,The NRC inspector reviewed a draft schedule of QA document changes dated January 8, 1987, and commented to senior management that while i

the schedule appears adequate, he would have expected it to have been

'

developed much closer to the beginning of the reorganization.

The. inspector interviewed various management personnel concerning the licensee's commitment management program.

Overall responsibility for commitment tracking is in nuclear services and implemented per

.

NSP-106, Revision B, " Commitments Management System." This is a

!

nuclear services internal procedure and does not have the necessary distribution to assure adequate instructions are available to all personnel.

The licensee had previously identified this roblem and provided'a draft copy of Nuclear Operations Procedure (N.-) N0P-011, Commitments Management System, for review by the NRC inspector.

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. must be followed by.all nuclear operations personnel and appeared

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adequate to satisfy commitment management objectives. Minor comments where discussed with the licensee.

The licensee ut!1izes a tiered manual and precedures system to implement the QA program. The NOPE is the central policy document.

Policies contained therein reflect external commitments contained in

documents such as the license, TS, FSAR, industry codes and standards, and commitments.

The internal commitments are contained in documents such as Nuclear Operations Executive Directives (EDs)

and other corporation policy documents.

The NOMM is implemented procedurally by the NOP manual.

It contains' procedures applicable to more than one department.

The next tier down involves the departmental procedures and instructions.

They cannot cover activities performed _by individuals outside the issuing departments.

The NRC inspector ~ observed that this system appears functional and made the-following comments on implementation:

(1) It appears the NOMM contains some information which would be more appropriate for the N0P manual.

'

'(2)' For some activities, essentially the saine information was found in three or more documents.

This complicates change control.

The following are'ainong documents reviewed by tha NRC inspector to

"asceitain the licensee's compliance with QA program commitments:

,.

.(1). ED-001, Revision 3, Nuclear Operations Executive Directive System.

(2)' ED-013, Revision 2, Quality Reporting I

'

(3) ED-025, August 20, 1984, Interface With the Licensing Section (4) ED-041, December 11, 1985, Personnel Qualifications / Verification Program Plan (5) NOMM, Revision 0, Forward and Policy Statement (6) NOMM,Section I, Revision 0, :atroduction (7) NOMM,Section III, Revision 1, Managerrcut Documentation I

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-(8) NOMM,Section IV, Revision 1, Quality Requirements Matrix l

(9) N0P-002, Revision 1, NOMM Steering Committee l-The licensee produced detailed functional organization charts

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reflecting the new organization.

A TS Change Request submitted to

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the NRC on October 15, 1986, proposed changes to TS, Section 6 l

(Administrative Controls) reflecting the effects of the the l

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reorganization.

A revision to this request, which provided functional organization charts was submitted on November 19, 1986.

The.NRC inspector reviewed the detailed organization charts and position functional descriptions to determine thb realignments did not degrade the position functions involved,- degree of management control, and available resources to fulfill organizational responsibilities.

No degradation was identified.

For individuals having new positions or revised duties, their resumes were. reviewed to deterniine that they satisfy the education and experience requirements established by the licensee for that-position.

No exceptions to current position requirements were observed.

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With regard to the reorganization, the NRC inspector made the

.

following-positive observations:

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(a)~ The organizational structure appears more functional for the 9 L_

facility in its present status I

(b) ~With the reporting line changed from Plant Operations to QA, QC

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(Inspe'ction). independence has been increased.

Tnis could result

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!in'better performance.

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larification~ of--the duties of the Safety Review Committee (SRC)

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Chairman and assignment nf additional resourcas to the SRC

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w.thould result -in a positive effect on nuclear safety.

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s-(d)'Havingtbe=IndependentSafetyEvaluationGroup(ISEG) reporting

. to a higher management level should result in increased visibility which will prove beneficial.

(e) < Creation of an Event Analysis and Reporting section reporcing

'directly to the Assistant Plant Manager-Technical Services o

should have a positive impact on nuclear safety.

.

With regard to compliance with the TS, the licensee was unablo to

l identify at the exit meeting on January 9,1987, documentation satisfying the intent of TS 6.1.1. which states, "The Plant

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l Hanager-Nuclear shall be responsible for overall unit operation and i

shall delegate in writing the succession to this responsibility

during his absence." The licensee produced ::;emoranda issued by the Plant Manager to delegate his responsibility during planned absences from the site.

However, no docunentation delegating his l

responsibility in the event of unforeseen absences was identified.

l This is an apparent violation. (382/8701-04)

No other violations or deviations were identifiea.

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q.dascertain conformance with TS, regulatory requirments, commitments

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Distribution and Use of Plant Operating Manual (POM) Procedures

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virtually all c.ontrolled. documents.

Most plant documents are-

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F controlled by a centralized document control section'in the Nuclear Services Department; however, the POM is controlled by a records

.j-management section in. Plant Operations.

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The NRC inspector interviewed the records manager to obtain a better-understanding:of system operation and management goals.

The Document-Control section has a management goal of issuing drawing revisions the next working day after the approved revision.is received.

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delay cf I week is considered unacceptable by LP&L.

Periodic

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internal assessments of distribution and cuntrol accuracy are performed.

A discrepancy between the as-constructed facility and.an

as-built drawing can be resolved through issuance of a Plant En0ineering Information Request in accordance with Nuclear Operations

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Administrative Procedure 18. This proce.ss would probably result in issuance of a staticn modification package (SMP).

Outdated drawings

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are stored an microfilm.

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The NRC inspector sampled document control system performance, by

selecting certain documents from the master indices and verifying their existence at the following libraries:

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The.following documents were selected:

FSAR, Chapters 13 and 17; NOMM, Sections 2 and 4; N0Ps-002 and 003; TS Pages 3/4 3-6 and 3-12; POPS UNT-6-001, OP-2-018, OP-10-001, OP-901-013 and OP-903-001; Technical

- Manuals (TMs) 166 (reactor coolant pumps) and 1225 (emergency diesel generators); and Drawings G00167 through G00172 including all sheets.

The document status indicated in the master index was noted; for example revision level, temporary change notices (TCNs), amendment numbers, outstanding SMP numbers, and interior drawing revisions.

The copies surveyed reflected the status in the master indexes with the following exceptions:

.

Drawings in the engineering library were not up-to-date.

Five SMPs (67,R2; 257,R8; 338,R4; 1004,R0; 1047,R3) had been received in the library, but not posted on the appropriate drawings.

One SMP (107,R5) had been received by Document Control, but not transmitted to the library.

The 6 SMPs not posted had been in Document Control from 6 to 20 days prior to transmittal, which does not comply with the licensee's unofficial timeliness criterion. One SMP (1004,R0) not posted had been received by the library 38 days prior to the inspection date.

SMPs 1700, 1779, and 1795 were posted to TM 1225 but were not entered in the master computer index.

The licensee believes all document entries to the computer file on January 7, 1987, had been dumped when they experienced problems with the computer; however, this was not suspected until the above discrepancies were identified.

The SMP posting sheet for TM 1225, Volume 1 was not found in the volume located in the maintenance library.

The records management

-

supervisor believed that a borrower of the manual had inadvertently removed the posting sheet, and its absence went undetected the next time it was loaned.

SMP 1427R0 was not posted to TM 166, Volume 2 in the maintenance library (Copy 106).

Document Control produced an acknowledgment sheet that it was received by the maintenance library on November 7, 1986.

The NRC inspector commented to senior management that although the

.

above discrepancies do not involve the safety significance necessary to constitute violation of commitments, they may be indicative of i

weaknesses wnich deserve management attention and support.

For

example, it.should be determined if the overall timeliness of s

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document issuance is satisfactory.

In view of the data loss, it should also be determined if there is sufficient QA for the computer file.

In general, there appeared to be the basic framework for an.

excellent document control system.

No violations or deviations were identified.

c.

~ Audit Program During this inspection period, the NRC inspector conducted a review to determine whether the licensee has developed a program relating to audits of activities that is in conformance with regulatory requirement commitments in the FSAR and industry guides and star.dards.

Additionally, he reviewed the implementation of this

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program by reviewing the licensee's NOMM, QAPs, NOPs, lead auditor training and certification records, audit reports, and associated checklists utilized in the conduct of performing audits.

The NONM, QAPs, N0Ps, and other related procedures / instructions reviewed did not reflect organizational changes.

The licensee is currently reviewing necessary changes to update these and other related documents.

The licensee received an independent management audit in October 1986 by the Middle South Services (MSS) QA organization. The NRC's review of these results indicated the nine functional areas audited were assessed by MSS as being implemented effectively in accordance_with

,

the policies and procedures provided for those activities.

The MSS report identified 8 audit findings for which the licensee is preparing a response to MSS and is committed to have complete by February 5, 1987.

A review of licensee's Radiological Environment Monitoring Program

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Audit Plan (SA-87-022.1) and checklist which was scheduled to l

commence January 19, 1987, appeared adequate except for the

following:

Technical Specification 6.5.2.8.p requires auditing the l

performance of activities required by the QA program to meet the provisions of Regulatory Guide (RG) 1.21, Revision June 1974 and l

l RG 4.1, Revision 1, April 1975 at least once per 12 months.

Neither the audit plan nor checklist specifically address these RGs, the

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checklist does address procedures NSP-244 and N6P-242 which have RG 1.21 and RG 4.1 listed as references.

The licensee stated a review would be performed to ensure requirements of the RGs were l

adequately being implemented via these procedures or to include the RG specifically within the audit plan / checklist.

This is considered an open item (382/8701-05) pending completion of this review by LP&L.

Additionally, the NRC inspector reviewed the previous audit conducted on the Radiological Environmental Monitoring Program.

He noted two quality notices were issued, one was adequately closed, the other l

(QA-86-045) was issued because the licensee's auditor noted on

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March 20, 1986, that the N0fm didn't_ address the Effluent Monitoring.

program.'

TS section-6.8.1.k requires written procedures to be

established for Effluent and. Environmental monitoring and section 6.8.2 requires this program to'be reviewed and approved by

-

the plant manager-nuclear prior to implementation.

The licensee has-a program policy which has been approved by all senior managers on October 3,1986, but had not been issued as of January 9,1987.

The l licensee stated this. policy would be issue immediately to satisfy full compliance with TS.

This.is consider'ed an open item (382/8701-06).

The NRC inspector reviewed the following completed audits to verify.

audit plans / checklists.are in conformance with TS, RGs, and industry

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standards.

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(1) Q.3_-A35.39-124-85.1, Audit report for Southern Vital Records

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2) Q.3-A35.39-149-85.1, Audit of Supplier-Siemens-Allis Company.

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(4)" SA-W3-QA-86-17, Procurement of Materials and Services.

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,Within-theaYeasinspected,theNRCinspectordeterminedtheseaudits

~=were adequately performed within the licensee's audit program which,-

if followed, will provide adequate conformance to commitments.

' - No other violations or deviations were identified..As a. result of

'

the-QA~ Program Annual Review, one apparent violation was observed as s

discussed section 13.a above.

14.'

License Conditions

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Electrically independent neutron flux indication at LCP-43~to be installed prior to startup from the:first refueling.

In order to meet the requirements of this license condition the licensee

. installed a panel, in the cable spread area of.the +35 elevation of the

+

Reactor Auxiliary Building, where in the event of a control room fire an operator would have to perform a number of electrical plug-disconnection /reconnections.

The NRC inspector reviewed the installation of this modification, walked thru the control room evacuation procedure, and then discussed'the concerns listed below with licensee management:

Emergency lighting should be provided at the panel to give the operator enough light to make the necessary disconnections /

reconnection,

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Even with additional emergency lightin'g the-labelling of.the Edifferent cables should be evaluated to'see if some more easily g'

recognizable labelling could be provided.

LThough access'to this' panel is.important, establishing neutron.

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indication at LCP-43 is not an immediate action upon evacuation of~

the control room.

There are other switches, breakers, and valves,

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required-to be operated during the immediate actions which the

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. f analysis prior to startup following the.first refueling outage.

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%,upon the're'sults i f.the spurious signal analysis study a number of changes

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_ ] were~ndde ~to'both~ equipment. and procedures required for plant control

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<rs ctor had concerns with'the labelling of switches and breakers

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being readily This concern was also discussed with plant X 7, J anagement.. ;jidentifiable.T v,.

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Potential Generic Problems i=

i-The NRC inspector provided the Licensee with copies of the'following

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10 CFR3Part 21'. reports:

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ITT Grinnell air-operated diaphragm valves found to have natural frequencies less than 33Hz (86-15)

.

.

Advisory on handling Foxboro N-Ell and N-E13 transmitters (86-13)

T Spring failures in Valcor solenoid values using reactor chemistry water at temperatures above 440*F.

(86-09)

Defective steel manufactured by Inland Steel Company with heat number 17802 (87-02)

No violations or deviations were identified.

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16.

Exit Interview

'

The inspection scope and findings were summarized on January 9, 1987, and February 2, 1987, with those persons indicated in paragraph 1 above.

The licensee acknowledged the NRC inspectors findings.