ML20207D706

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Insp Repts 50-327/88-27 & 50-328/88-27 on 880501-0530. Violation Noted.Major Areas Inspected:Operational Safety Verification,Review of Previous Insp Findings,Followup of Events & Review of Licensee Identified Items
ML20207D706
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 07/27/1988
From: Jenison K, Mccoy F
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF SPECIAL PROJECTS
To:
Shared Package
ML20207D541 List:
References
50-327-88-27, 50-328-88-27, GL-87-12, IEB-79-14, IEB-80-12, IEB-80-15, IEIN-86-081, IEIN-86-81, IEIN-87-062, IEIN-87-62, NUDOCS 8808150434
Download: ML20207D706 (25)


See also: IR 05000327/1988027

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Report Nos.:

50-327/88-27, 50-328/88-27

Licensee: Tennessee thlley Authority

6N 38A Lookout Place

1101 Market Square

Chattanooga, TN 37402-2801

Docket Nos.:

50-327 and 59-328

License Nos.:

OPR-77 and DPR-79

Facility Name:

Sequoyah Units 1 and 2

Inspection Conducted: May 1, 1988 thru May 30 1988

Team Leaders:

k

Vf' 25,1106

K/M. /ep son, Senior Resident Inspector

Sete Signed

Team Members:

P. Harmon

G. Humphrey-

D. Loveless

W. Poert

Approved by:

W

~/!)788

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F. R. b)t:Co), Chief, Project Section 1

' 'Dat( Signed

  • .

Division of TVA Projects

SUMMARY

Scope:

This routine, announced inspection was conducted on site in the areas

of:

operational safety verification; review of previous inspection

findings; followup of events; review of licensee identified items;

review of NRC Bulletins; and review of Inspector Fellowup Items.

Results: One violation was

identified - Violation 327,

328/88-27-01,

Inadequate Corrective Action for Improper Positive Reactivity Changes

(Paragraph 5).

4

No unresolved items or inspector follow-up items were identified.

BG00150434 800727

PDR

ADOCK 05000327

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REPORT DETAILS

'1

Persons' Contacted

.

Licensee Employees

'H. Abercromb'ie, Site Director

J. Anthony, 0perations Group Supervisor

  • R. Deecken, Maintenance Superintendent

~J'. Eynum, Assistant Manager of Nuclear Power

M. -Cooper,~ Compliance Licensing Supervisor

H. Elkins, Instrument Maintenance Group Manager

R. Fortenberry, Technical Support Supervisor

J. Hamilton, Quality Engineering Manager

.

M. Harding, licensing Group Manager

G. Kirk, Compliance Supervisor

'*J.

La Point, Deputy Site Director

  • L. Martin, Site Quality Manager

R. Olson, Modifications Manager

R. Pierce, Mechanical Maintenance Supervisor

R. Prince, Radiological Control Superintendent

R. Rogers, Plant-Operations Review Staf f

S.. Smith, Plant Manager

,

J. Sullivan, Plant Operations Review Staff Supervisor

B. Willis, Operations and Engineering Superintendant

C. Whittemore., Licensing Engineer

'NRC Employee

F. McCoy

  • Attended exit interview

2.

Exit Interview

'd

- The inspection scope and findings were summarized on June 8, 1988, with

those persons indicated in paragraph 1.

The Startup Manager described the

areas inspected and discussed in detail the inspection findings listed

below.

The licensee acknowledged the inspection findings and 'did not

. identify as proprietary any of the material reviewed by the inspectors

during the inspection.

NOTE:

A list of abbreviations used in this report is contained in

paragraph 9.

Inspection Findings:

One violation was identified, paragraph 5.

No unresolved items or inspector follow-up it'.ms were identified.

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3.

Operational Safety Verification (71707) Units 1 and 2

The inspectors reviewed applicable logs, including the shift logs, night

order book, clearance hold order- book, configuration log, and TACF log;

conducted discussions with control room operators; reviewed LER and PRO

documents; and confirmed the adequacy of corrective actions.

The

inspectors verified compliance with TS LCOs and that WRs had been

submitted as required including follow-up activities and prioritization of

work.

Tours of the auxiliary, control, and turbine buildings were conducted to

observe plant equipment conditions and to inspect corrective actions.

No violations or deviations were identified

4.

Licensee Action on Previous Enforcement Matters (92702)

(Closed) VIO 327, 328/87-61-01, Failure to Accomplish an Activity

Affecting Quality in Accordance with a Prescribed Procedure.

Eight of twenty-five licensed SR0s and R0s failed to complete required

procedure reviews within the prescribed time stipulated in OSLA-1, Nuclear

Generating Plant Operations Training Program Letter. OSLA-1 required

documentation of subsequent action changes to emergency (EI) and abnormal

operating procedures (AOI) within five shift days for snift operators,

however attachment XI-C which records the individual's review of procedure

changes, specified within thirty days. As part of the corrective action

for this violation, and to resolve the inconsistency, the licensee revised

OSLA-1 to agree with Attachment XI-C, requiring review within 30 days.

Changes to immediate action steps of the Els and A01s require review prior

to assuming the shift.

Additionally,- the

licensee has revised

Administrative Instruction (AI)-5, Routine Shift Work, to include a

requirement for all licensed operations personnel and shif t technical

advisors to review and sign for immediate action changes to EIs and A0Is

prior to assuming the shift.

The licensee's actions appear to be

adequate.

This item is closed.

(Closed) VIO 327, 328/87-66-01, Licensee Failed to Adequately Establish,

Implement, and Maintain Procedures for Configuration Control.

This event involved the generation and control of plant system configu-

ration control records. OSLA-58, Maintaining Cognizance of Operational

Status

Configuration

Control,

failed to adequately specify when

configuration control records should start (records only required af ter

check lists completed), and failed to specify an appropriate method for

deviating from SOI checklists (deviations not considered a procedural

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change). The licensee failed to adequately implement OSLA-58 requirements

for the use of working copies of SOI checklists (transfer of all pertinent

data), failed to implement requirements for recording deviations (date not

recorded next to initials), failed to maintain configuration control after

checklists were completed (documented configuration did not agree with

actual configuration) and failed to process revisions to SOI checklists

(log entries cleared without re performing the section of checklist that

had been revised).

The inspector has reviewed AI-58 Maintaining Cognizance of Operational

Status-Configuration Status Control (a conversion of OSLA-58); AI-2,

Authorities and Responsibilities for Safe Operation and Shutdown; AI-4,

Preparation, Revics, Approval and Use of Plant Instructions; AI-30,

,

Nuclear Plant Method of Operation, IR 327,328/87-66, the response to NRC

inspection reports 327,328/87-66 (K. P. Barr/S. A. White) letter dated

December 21, 1987; IR 327,328/88-06; Response to System Alignment

Verification for Unit 2 Heatup Violations, and OSLA-107, System Operating

Instruction Review and Verification Checklist.

The new AI-58 procedure,

as amended by the work ticket process, appears to provide adequate

correctin, of the examples of Violation 87-66-01.

The assignment of

managen.ent personnel to directly supervise the system alignment program in

conjunction with the training given to personnel involved in this program

also appear to be adequate. Finally, QA audit interaction with operations

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on an extended basis has botn increased operator and management attention

in this area and identified a sharp reduction in the number of

configuration related errors. Violations 327,328/88-06-01, 88-06-02, and

88-06-03 also address the subject of configuration status in a more

generic sense. Therefore, because the specifics of this violation have

been corrected, the generic issues aisc appear to be corrected and the

ge ne r.i c issues will be reviewed again on the closure of the other

violations mentioned.

Violation 327,328/88-66-01 is closed.

(0 pen) VIO 327,328/87-71-01, Failure to Implement Adequate Design Control.

This item involved the licensee's failure, due to a misinterpretation of

GDC-56, to install caps on the ends of test connection lines for

instruments30-46B, 30-478, and 30-48B, thereby not providing the required

containment isolation capability.

The licensee's actions regarding this item were as follows:

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DCN X00028-B was issued to add the required caps.

The work was

completed on Unit 2 in accordance with WP 12635.

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The licensee has committed to install the required caps in Unit 1

prior to entry into mode 4.

This item remains open until corrective actions are complete.

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(0 pen) VIO 327,328/87-73-05, Failure to Perform Adequate .10 CFR 50.59

Safety Evaluations for Modifications to the Facility. Which Involved

Compensatory Measures.

As previously reported in Inspection Report 87-76, the inspectr.r has

reviewed the specific USQDs pertaining to this issue and determ',vJ them

to be adequate. TVA's response has been received, however a supplemental

response is due by June 6, 1988. Pending the receipt and review of this

supplemental response, this item remains open.

(Closed)' VIO

327,328/87-76-01,

Failure to Perform Adequate Post

Maintenance Testing

This issue involved an inadequate post maintenance test procedure that did

not test operation of a specific diesel generator voltage regulator under

rapid ioading/ rejection conditions.

The licensee's response to the violation (Gridley/NRC Document Control)-

dated May 4,

1988, (RIMS L44880504811) was reviewed and found to be

adequate.

The licensee has issued LER 327/87070, Revision 3,

that

addresses the generic concerns of this violation.

The licensee has

revised Standard Practice SQN 66, Post Maintenance Testing, to ensure that

testing requirements on a.iy voltage regulator maintenance work are

evaluated for each case until appropriate post maintenance tests are in

place. The licensee's corrective actions appear to be adequate and if

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implemented, will prevent recurrence.

Long term corrective action will be

tracked by LER 327/87070.

This Violation is closed.

(Closed) Unresolved. Item (URI) 327, 328/86-46-06, and IE Bulletin 80-12,

Decay Heat Removal (DHR) System Operability.

This issue involves the potential loss of RHR during Mode 5 operation.

Bulletin 80-12 directed the licensee to perform six reviews and report to

the commission regarding specific actions. The licensee responded on June

19, 1980 by letter from J. C. Ross to J. O'Reilly, indicating the results

of the licensee's reviews and analysis as required.

The licensee

indicated that procedures would be changed to include instructions on loss

of decay heat removal capability during the following modes of operation:

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The RCS partially drained, but before reactor vessel

head

de-tensioning.

The reactor vessel head de-tensioned, but still in place.

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The reactor vessel head removed, but before filling of the refueling

cavity.

A01-14, Loss of RHR S;.utdown Cooling, has been revised to include

alternate cooling for the modes identified above.

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NRC GL 87-12, Loss of RHR while the RCS is partially filled, is a

reiteration of the same basic material asBulletin 80-12, with more

emphasis on detail.

The: inspector reviewed A0I-14 and the conditions described above are

adequately addressed. GL-87-12 was reviewed as well as the licensee's

response (Rims L44 871002 805). GL 87-12 supersedesBulletin 80-12 and

URI 86-46-06, which was written to track some of the above menticned

conditions.

The licensee's actions appear to be adequate.

Therefore, URI 86-46-06, and Bulletin 80-12 are closed.

GL 87-12 will

remain open until the final corrective actions are complete.

(Closed) VIO 327/80-15-01, Piping Support Discrepancies.

NRC Inspection Report 327/80-15 reported drawing discrepancies and pipe

support discrepancies with several safety related supports which had been

previously signed off and inspected by TVA as satisfactory.

The TVA

response to the violation stated that specific discrepancies

had been

corrected and the licensee planned to conduct walkdown inspections and a

review of several hundred supports inside and outside containment.

This

inspection was to be done in conjunction with elements of NRC Bulletin 79-14, Seismic Analysis for As-Built Safety Related Piping Systems.

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TVA transmitted the results of the initial inspection by letter dated June

27, 1980.

The results indicated ". . . discrepancies found were randomly

distributed through several phases of the design, construction, and

operation process

TVA concluded that discrepancies would not

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prevent safe operation at low power; however, additional inspections would

be necessary to provide further assurance that the plant can be operated

at full power.

TVA later identified as part of the Design Baseline Verification Program

(DBVP) design problems with several thousand hangers which led to large

scale re-engineering and plant modification efforts.

The inspector reviewed the inspection report, the TVA acknowledgement and

the initial inspection summary.

The specific discrepancies associated

with Violation 80-15-01 were corrected.

Corrections of programmatic

deficiencies for Unit 2 were addressed under the NRC OBVP verification

inspections and found to be programmatically adequate.

Implementation

activities on Unit I are currently in progress for approximately 1300

hangers and Bulletin 79-14 remains open.

The generic issues associated

with the adequacy of Bulletin 79-14 will be addressed with the closure of

the bulletin.

This violation is closed.

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'(Closed)'URI 327, 328/87-30-06, EDG Circuit Cards.

On April 30,1987, .while trouble shooting erratic v'oltage swings on the-

1 A-A . EDG , the ' licensee determined that 'he voltage swing problem was

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associated with the voltage regular minimum-maximum cards.

The licensee,

with - concurrence and assistance of .the - vendor ~(Basler Electric),

disconnected the minimum-maximum circuit from the EDG voltage regulator

for all EDGs.

Subsequent calculations and operational testing conducted

by the licensee were submitted to the NRC's OSP for review.

OSP issued,

SER NUREG 1232. Volume 2, Part 1 on March 25, 1988, (Revised Preliminary

Report) which accepts the licensee's EDGs as adequate to support restart.

Portions of the SER are excerpted here to detail the boundaries of

acceptance and the licensee commitments:

In January 1988, TVA identified to the NRC data from surveillance

tests which raised significant questions about the operability of the

EDGs at Sequoyah.

These results were interpreted by TVA as

indicating both a possible aefect in the 2A-A generator's exciter

system and a more general problem in all generators in conforming

with voltage limits during loading as stated in RG'1.9.

A failed

component was replaced in the exciter system of the 1A generator

which corrected the first problem and lef t only' the more general

voltage problem.

A detailed review of the test data by the NRC

identified the following technical issues relevant to the second

problem:

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'The test results were worse than would be predicted by the

calculations methods used to model diesel generator performance.

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The test data, when extrapolated to post accident conditions,

showed that the diesel generators had less margin, in termt of

voltage behavior, than calculations had predicted.

Because of these issues, the NRC required TVA to undertake a major

analytic effort with the following objectives:

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To identify the reasons why calculations did not predict the

severity of the experimental results.

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To improve the calculational methods to provide greater

assurance that the calculational methods would conservatively

predict post accident behavior.

To quantify the margins available between diesel capability and

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specific electrical power system requirements in post accident

operation.

The specific margins in question were the following:

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minimum voltage during the loading sequence.

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diesel generator power rating (KVA and KW).

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motor performance (starting and stalling).

contractor performance _(pick up and drop out).

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motor operated valve performance (torque and timing).

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overcurrent protection misoperation (circuit breakers and

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control fuses).

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sequence timing error (overlap and maximum loading time).

These required analyses and calculations were submitted to the NRC on

March 1, 2, and 11, 1988.

Based on its review of the results of

these calculations the NRC reached the following conclusions:

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The most likely reason that the original calculations did not

bound the test data was that generator voltage did not stabilize

between successive steps in the loading sequence.

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The generator's inability to stabilize voltage was probably

caused by the use of a voltage regulator lacking the speed and

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performance characteristics typical of those used in modern

nuclear applications.

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The most recent analysis methods used by -TVA bounded the

experimental data and predict the EDGs behavior in post accident

loading with adequate margin. However, this margin is less than

was expected when the plant was licensed.

The margin that remains is sufficient to assure safe operation

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of Sequoyah for restart and for the period of time until

corrective action is taken to improve the margin.

In the March 3,

1988, submittal, TVA committed to evaluate the

performance of the EDGs and implement corrective action . prior to

Unit I restart after the naxt Unit 1 refueling outage. This schedule

is acceptable to the NRC staff. The staff SER states that the TVA

calculations on which the staff's findings are based assume the

Sequoyah Unit 1 is in cold shutdown and must be revised to support

Unit i restart.

Further, the staff notes its reliance on TVA's

commitment to undertake,

after restart, a major review and

modification

effort

to

improve

performance

of

the

EDG

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regulator / exciter system."

Based on the above this URI is closed.

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5.

Licensee Event Report (LER) Follow-up (92700)

The following LERs were rev'.ewed and closed. The inspector verified that:

reporting requirements had been met; causes had been identified;

- corrective actions appeared appropriate; generic applicability had been

considered; the LER forms were-complete; the licensee had reviewed the

event; no unreviewed safety questions . were involved; and no further

violations of regulations or Technical Specification conditions had been

identified.

(Closed) LER 327/86038, Two EDG Starts Resulting from Personnel Error and

a Component Failure.

During performance of SI-60, 6.9 KV Unit Board Manual and Automatic

Transfer,

licensee personnel went to the wrong unit board to locate the

components specified in the SI. Operation of the identical components on

another unit board, resulted in inadvertent start of the EDGs.

Partial

cause of this event has been attributed to the close proximity of the Unit

Boards. The licensee's corrective action consisted of clearly identifying

the unit board separation with highly visible stripes and training

personnel to be aware of the unit board layout configuration.

A second

EDG start occurred while transferring Start Bus 28 from alternate to

normal supply.

Approximately two to three seconds after closure of

Breaker 1414, the breaker tripped causing a loss of voltage and starting

of the EDGs.

Investigation revealed that the contacts on the close-trip

switch were dirty.

The manual close-trip switch was replaced, and the

breaker was tested satisfactory and returned to service. While returning

the EDGs to normal following this inadvertent start, EDG 2A-A restarted.

Investigation revealed a loose fuse clip contact.

The fuse clip was

repaired, and the other EDG fuse clips were inspected and found to be

acceptable.

The licensee's correct;,6 ?ctions are adequate.

This LER is closed.

'(Closed) LER 327/87078, Revision 2,

Inadequate Procedure For Reactor

Coolant System Chemical Addition Resulted In Noncompliance With A TS

Action Statement.

This event report describes three occasions on Unit 1 and seven occasions

on Unit 2 when chemical additions to the RCS caused positive reactivity

changes in violation of TS requirements. The chemicals were flushed into

the system using up to 75 gallons of demineralized water which resulted in

a very small dilution of the RCS boron concentration.

The licensee has

calculated these decreases in boron concentration to be approximately four

parts per million (ppm) resulting in negligible positive reactivity

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changes. However, Action Statement (b) to TS 3.7.7 requires that in Modes

5 and 6, when the Control Room Emergency Ventilation System (CREVS) is

inoperable, all operations involving positive reactivity cheqges be

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suspended.

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The NRC inspector reviewed Revision I to the LER in late March .1988 and

identified the following three discrepancies:

The abstract and corrective action-sections of Revision 1 state that

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Technical Instruction (TI)-19, - Chemical Feed _ Controls, and- System

Operating Instruction- (S01)-62.5, Boric Acid Batching, Transferring -

and Storage System, had ~ been revised such that both procedures now

provide instructions for RCS chemical addition that will not cause an

RCS dilution.

In fact, both of these procedures had been revised

with temporary changes that expired on January 30 and February 15,-

1988, respectively-

Thus, when LER Revision 1 was issued on

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February 25, 1988, neither of these . procedure changes remained in

effect.

In addition, two other station procedures that govern

chemical additions which could cause positive reactivity changes were

not addressed; SOI-62.2, Boron Concentration Control, and S0I-62.3,

Reactor Coolant Chemical Addition and Control.

The abstract and corrective action sections of LER Revision 1 state

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that "no planned activities, when either Unit 1 or Unit 2 is in

Mode 5 or 6, .will be performed which could potentially result in a

positive reactivity change if both trains of CREVS are inoperable."

However, there were no formal licensee controls or procedural

limitations in place to assure that such activities could not take

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picce.

The abstract and corrective action sections of the original LER

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contained a commitment to revise the TVA interpretation of the action

statement to TS 3.7.7 so that any planned evolution that could result

in a positive reactivity change is considered prohibited by IS. TVA

also committed to promulgate the details of this interpretation to

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licensed personnel by issuing a management directive by February 29,

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1988.

No management directive or other suitable notification had

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been issued to station personnel. Revisions 1 and 2 to the LER have

deleted references to these commitments without explanation and they

have been closed without action on the licensee's Corporate

Commitment Tracking System.

CAQR SQP 871660 which addressed these events was closed on

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January 23, 1988, specifying that a root cause analysis / recurrence

control was not' required.

In addition, the CAQR was erroneously

closed based on a temporary procedure change (expiring January 30,

1988).

The CAQR disposition stated that a procedure was developed

that would ensure no dilution would occur while under LC0 3.7.7

(TI-19 Revised) for addition to RCS.

The plant Operations Review

Staff was apparently not knowledgeable of the corrective action

status or actual actions accomplished at the time Revision 1 to this

LER was prepared and issued.

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The above concerns were discussed in detail with responsible licensee

personnel on April 1,-1988. No indication was made at this meeting that

there was any current intent by TVA to make additional reviews or take

further corrective action.

The licensee subsequently stated (late May

1988) that they had intended to implement additional long term corrective

actions in March, .However, the

licensee could 'not provide any

documentation (meeting notes, memos, log entries, commitments, telecon

notes, etc.) of this intent

that any additional reviews or actions were

assigned or in process. The failure to take corrective action to prevent

recurrence is a violation of 10 CFR 50, Appendix B, Criterion XVI and is

identified as Violation 327, 328/88-27-01.

As a result of the above identified concerns TVA has taken additional-

actions. Revision 2 to the LER was issued which provides more detail on

the positive reactivity change events and committed to permanent revisions

to TI-19, 501-62.2 and S01-62.3 as long-term corrective actions.

These

procedure changes have been reviewed by the NRC inspector and it appears

the procedural corrective action is adequate.

On April 12, 1988, TS

interpretation 111, was issued addressing plant evolutions that could

result in positive reactivity changes. The licensee's corrective actions

appear to be adequate.

This LER is closed.

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(Closed) LER 327/88008, Personnel not Properly Implementing Approved

Administrative Procedures Resulting in Inappropriately Exiting a TS Action

Statement on Radiation Monitoring.

This event involved operations personnel declaring radiation' monitors

(0-RM-90-134 and 141) operable prior to completion of the instrumentation

post modification test PMT-206, Radiation Monitoring Sample Flow Alarm

Function Test and Calibration.

The inspector reviewed the LER and the following corrective actions:

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The immediate corrective actions taken were to enter TS LC0 3.3.3.9

action statements and to resume taking grab samples for analysis.

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Operations personnel on shift were directed to review all PMT

documentation as verification before returning equipment to an

operable status.

This statement was also included as a change to

AI-30, Nuclear Plant Conduct of Operations.

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The personnel involved in the event have been counseled.

The licensee actions appear to be adequate.

This LER is closed.

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(Closed) LER 327/88011, An Inadequate Determination of the Effect of

Routine Diesel Testing Resulted in Both Trains of the Control Room

Emergency Ventilation System Being Inoperable.

This event -involved a routine maintenance' operation on EDG 1A-A that

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administratively made the CREVS train A inoperable while the train B CREVS

was declared inoperable to allow replacement of fire detectors in the

ductwork.

This resulted in both trains of CREVS being inoperable which

required entry into LC0 3.0.5.

This condition had existed from 11:40 p.m.

February 15, 1988 to 12:30 a.m.

February 16, 1988- (approximately 50

minutes) before the shif t supervisor determined that LC0 3.0.3 was in

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effect.

The inspector reviewed LER 327/88011, LC0 3.0.3, LCO 3.0.5 and S0182.1.

The licensee has established a WCG, to help the control room shif t crew

determine the impact of work activities on the plant and has revised the

50I-82 series of instructions to include warnings that certain steps will

cause the EDG to become inoperative.

These factors should reduce the

chances for similar events.

The licensee's actions appear to be

acceptable.

This LER is closed.

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(Closed) LER 327/88012, ESF Actuation.

CVI Caused by Electromagnetic

Interference (EMI).

This event involved an EMI actuated CVI. This problem had been identified

previously and a task force had recommended modifications to eliminate or

at leas' reduce spurious CVIs caused by radiation monitor (RM) circuit

induced EMI.

This event occurred before the modification ~ were installed.

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The inspector has reviewed LER 327/88012, WP 7343-02 and WP 7344-02, ECN

L7343 and ECN L7344. The ECNs and WPs are the result of the task force

recommendations and they have been field completed. This issue was also

addressed as a CAQR which was closed in NRC Inspection Report

327,328/88-19. The licensee's actions appear to be adequate.

This LER is closed.

(Closed)

LER 327/88013,

Radiation Monitor Technical

Specification

Surveillance Requirement Omitted From Surveillance Program Due to an

Oversight.

This issue involved four steam generator blowdown system RMs that had not

been adequately functionally tested.

These RMs are required to be

functionally tested to meet quarterly technical specification SR 4.3.3.9.

The inspector has reviewed LER 327/88013 and discussed the testing of

radiation monitors with the IM department.

Functional testing of

radiation monitors has been completed. SI-1 was reviewed and appeared to

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be adequate. This issue was-also addressed as a CAQR which was closed in

NRC Inspection Report 327,328/88-19. The licensee's. actions appear to be

adequate.

This LER is closed.

(Closed) LER 327/88015, TS SR to. Sample Containment not Properly

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Implemented Before Containment Purge Operations Oue to Incomplete.

Procedure Review.

This event involved the failure ~ to

implement TS SR 4.11.2.1.2 which

requires upper and lower containment be sampled for radioactivity analysis

prior tMall containment purge operations.

The licensee determined that

the cause of this event was an inadequate procedure, SI 410.2, Containment

(Upper, Lower) Purge, that only required sampling of the portion of

containment to be purged.

The inspector has reviewed TS SR 4.11.2.1.2, SI-1, Surveillance Program,

Appendix F, ICF 88-0626 (issued March 18, 1988) and SI-410.2. ICF 88-0626

changed SI-410.2 and resolved the prob em. The licensee's actions appear

to be adequate.

This LER is closed.

(Closed) LER 328/88005, Loosening of Gland Seal Bolts on Speed Increaser

Lube Oil Pumps Causes a Potential Inoperability of Both Unit 2 CCPs.

This event involved the CCP speed increaser lube oil pump failure on CCP

2A-A due to the wrong lube oil pump (a 900 rpm versus the required

1800 rpm)

being

installed during

a

previous

pump

replacement.

Investigation of the Unit 1 and Unit 2 CCPs discovered that three of the

four CCPs had the wrong speed increaser lube oil pumps installed.

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The inspector has reviewed LER 328/88005, WR B257712, WR 8257714, WR

B247090, WR B247453, Maintenance Instruction (MI) 12.3.1 Centrifugal

Charging Pump Speed Increaser Inspection and Maintenance, ICF 88-0729 and

CAQRs SQP 880161 and SQP 880188. The three units with improper speed

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increaser lube oil pumps have had the pumps replaced or rebuilt to proper

requirements. CCP 1A-A had the proper speed increaser lube oil pump, CCP

IB-B had the pump replaced; CCP 2A-A had the pump rebuilt, and proper

gears, bearings and seals installed, CCP 28-B had the pump replaced.

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12.3.1 has been revised to require verification of proper components. The

licensee's actions appear to be adequate.

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This LER is closed.

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(Closed) LER 328/88007, Main Steam Isolation ESF Due to an Inadequate

Calibration Procedure.

This event involved . the lack of communications between operations and

maintenance department personnel. The root cause of this event was

inadequate maintenance department training on performance of procedure,

IMI-99 CC 10.313, Off-line Channel Calibration of Turbine-Impulse Pressure

Channel I during Mode 4 operation.

When operations personnel asked if

bi-stables would be tripped, maintenance technicians thought that tripping

of the bi-stables was not required.

When IMI-99 CC 10.3B Step 5.39.11

was performed the bi-stables were tripped and an ESF actuation resulted.

The inspector reviewed LER 328/88007, IMI-99 CC 10.313 and IMI-99 CC

10.48, and discussed the LER writeup and root cause determination with the

responsible P0RS member.

The licensee corrective action included establishing a work control group,

requiring a field verification of procedures prior to first time usage or

af ter revisions, and revising IMI-99 CC 10.38 and IMI-99 CC 10.4B to

require unit operator authorization prior to' tripping of bi-stables. The

licensee's corrective actions appear to be adequate to prevent recurrence.

This LER is closed.

(Closed) LER 328/88011 Revision 1, A Surveillance Requirement Used to

Verify Baron Concentration in.the Cold Leg Accumulators Was Not Performed

Within the Applicable Time Frames.

This event involved operations personnel not performing SR 4.5.1.1.1.b on

Accumulator 3 as required. The Reactor Coolant System (RCS) inventory had

been leaking into the accumulator, requiring the accumulator to be drained

periodically.

The inspector has reviewed LER 328/88011 Revision 1, 501-63.1, ECCS, SR 4.5.1.1.1.b, and a training letter issued April 27, 1988. The licensee's

actions appear to be adequate.

This LER is closed.

(Closed) LER 328/88012, Two Improper Operability Determinations Relating

to a Level Control Valve Resulted in not Entering the Applicable LCO in a

Timely Manner.

This event involved the failure to enter LC0 3.7.1.2 in a timely manner.

Two separate events were addressed in this LER.

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While performing SI-166.8, Increased Frequency Testing of Category A

and B Valves, level control valve 2-LCV-3-175 could not be controlled

from the control room.

This level control valve was not declared

inoperable at that time.

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Steam Generator Level Indicator 2-LT-3-107 was declared inoperable

due to excessive fluctuation. IM personnel determined that the sense

line for 2-LT-3-107 required backfilling.

Operations and IM

personnel determined that backfilling 2-LT-3-107 per MI

19.1.5,

Backfilling Sensing Lines for System 3 Transmitters Located on Panel

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L-184 ( Accumulator Room 3), would also make 2-LT-3-175 inoperable.

LCO 3.7.1.2 was not entered at the time that the-level transmitters

were made inoperable due to work performed by IM in accordance with

MI-19.1.5.

Contributing factors for both events were:

Failure to properly interpret Technical Specifications (TS).

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to

understand TVA generated Technical. Specification

Interpretations (TSIs).

Failure to make log entries that contain adequate detail.

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Failure to identify the test director properly.

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The inspector reviewed, MI-19.1.5, which has been revised to address the

applicability of LCO 3.7.1.2 and to clarify the effect of backfilling the

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common sense line of LT-3-175 and LT-3-107.

The TSIs have been reviewed

for adequacy and clarity, and returned to the control room.

However,

those TSIs that require revision were not included and are not committed

to be returned to the control room prior to Unit I restart. Training has

been given on AI-6, Log Entries and Review.

The TSIs and MI-19.1.5

revisions are considered to be enhancements.

The licensee's actions

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appear to be adequate.

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This LER is closed.

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The following LERs remain open pending TVA completion of corrective

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actions:

(0 pen) LER 327/88007, Opening of Blast Doors in Secondary Containment

Boundary Outside the Conditions Set for Surveillance Testing of the ABGTS

This report details a situation where the ABSCE configuration did not

coincide with that configuration used during SR testing used to verify TS

operability of the ABGTS.

Specifically, when the TS SR test was performed the blast doors to Units 1

and 2 reactor building were closed and the containment purge system (CPS)

was secured. However, the blast doors could be (and have been) opened and

the CPS operating on a unit in Modes 5 or 6 while the other unit is in

Mode 1-4. At the time of discovery both units were in Mode 5 and the ABGTS

system was not required to be operable. This problem has been identified

on CAQR SQP 880090.

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Short term corrective action was to close the blast doors and to tag the

Unit 1 CPS out of service prior to Unit 2 entry to Mode 4.

To allow

reopening of the Unit 1 blastdoors, an -evaluation of the data from the

previous performance of the ESTS annulus vacuum draw down test was

performed to ensure that the additional annulus inleakage would not'cause

the ABGTS to be declared inoperable.

Technical Instruction- (TI)-77,

Breaching the Shield Building, ABSCE or Control-.. Room Boundaries, was

revised to require an ' evaluation of additional inleakage from primary -

containment and the annulus prior to opening blastdoors (while the

opposite unit is in modes 1-4).

Long term corrective action included a programmatic review of the

administrative controls on the ABSCE boundary and to install logic to

isolate containment purge on an opposite unit auxiliary building isolation

signal.

The. licensee intends to submit a revision to the LER by

September 1, 1988 to provide details of the corrective actions taken.

The inspector reviewed Temporary Alteration Change Form number 1-88-0230

which placed the Unit 1 CPS out of service and the revision to TI-77 and

considers these actions adequate. The inspector verified that resolution

of CAQR SQP 880090 is classed as a Unit i restart item on TROI.

This item remains open pending resolution of CAQR SQP 88090.

(0 pen) LER 327/88010, Revision 1, An Inadequate Review of the Design Basis

of Two Engineered Safety Feature Actuated Valves Resulted in the Potential

for Plant Operation Outside of the Design Basis.

During a review of SI-166, Summary of Valve Tests for ASME Section XI, the

licensee determined the stroke time for valves LCV 62-135 and LCV 62-136

was in error.

Investigation disclosed the correct maximum allowable

stroke time (MAST) to be 15 seconds vice 28 seconds listed in SI-166. The

licensee reviewed previous stroke time records for the subject valves and

determined the stroke times to be within 15 seconds for all previous

tests.

The licensee confirmed the 15 second stroke time to be correct

per .TVA/ Westinghouse evaluations and the TS as approved in May 1987. TVA

retested the valves for both units in November and December 1987 and

verified the stroke times within the maximum allowable.

SI-166 was

revised to reflect the proper MAST.

The licensee determined that the cause of this event was an inadequate

determination of the design basis for these valves during the development

of SI-166. Recurrence control actions included a review of MAST for all

safety related motor operated valves against the requirements of the FSAR,

1

Design Criteria and the TS. This review is being conducted as part of the

corrective action for CAQR SQP 871446.

TVA has committed to make any

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necessary design document changes by July 29, 1988 and revise the FSAR if

necessary at the next scheduled FSAR update.

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The NRC inspector reviewed the change to SI-166 and other supporting

' documentation for this LER. The actions taken regarding LCVs62-135 and-

'62-136 appear to be adequate.

The review of design criteria against the

. MASTS identified 32' additional valves that had MASTS greater /less than

-specified in Design Criteria.

Pending completion of the. corrective actions for CAQR SQP 871446 this item

remains open.

(open) LER 327/87012, Loss of Shutdown Decay Heat Removal Resulting from

False Indications of RCS Level in Sightglass.

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This LER reported that shutdown decay heat removal was lost in Mode 5 for

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approximately 1 1/2 hours because of an inadvertent water level loss in

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the reactor coolant system.

The water level had been lowered to allow

for planned maintenance in one of the SGs and the event occurred when a

buildup of debris blocked the inlet to the sightglass and caused a faulty

level indication. The inspector reviewed the LER dated February 27, 1987,

revised LER dated April 30, 1987, and associated documentation which

stated the cause of the event and proposed corrective action.

The inspector made the following determinations:

SI-673, Reactor Coolant System Level Verification Using Sightglass or

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Tygon Hose, was revised to include the requirement for weekly

,

flushing of the sightglass and radiochemistry lab analysis of debris

collected from the flush water.

SI-673 was also revised to include the physical elevations of

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pertinent features in the open reactor coolant system such as RHR

nozzles, loop centerlines, SG manways, vessel flange, etc.

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The licensee evaluated the need to modify the design of the existing

1/2 inch sensing lines to the sightglass and determined that

increasing line size would have little or no effect on the problem.

The lines were designed with continuous upward slope, but were

observed to be discontinuous as the result of heavy objects being

placed on the tubing and loose hanger clips allowing tubing to slip.

WR B232170 was written to correct these deficiencies.

The task force which iavestigated the cause of the event recommended

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various actions which included the installation of a diverse level

indicating system for the RCS. CCTS item NC0-37-0077-002 was written

to trati this commitment which was scheduled for completion by the

,

next refueling outage.

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The inspector agreed that the corrective actions as stated in the revised

response should preclude recurrence of the event and that the proposed

schedule that the licensee has committed to is adequate.

However,

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installation of the diverse level indicating system remains to be

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completed.

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This LER remains open.

(open) LER 328/88013, Rod Control System Deficiencies Caused Inaccuracies

in the Rod Group Demand Position Indication Resulting in Three Manual Reactor Trips.

These events involved three manual reactor trips initiated because of

problems associated with the rod group demand position indication.

Two

events were caused by faulty individual rod position indications and one

was caused by an open switch in the rod lift coil circuitry.

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The inspector has reviewed LER 328/88013, WR-B267610, replacement of group

1 demand step counter for control bank B, WR-7B290346, determine . problem

on open disconnect switch, and WR-8290792, troubleshooting to determine rod

control problem and to replace bad components if any.

The licensee's

actions appear to be adequate for Unit 2 restart.

However, long term

commitments associated with Unit 1 restart are being tracked by the CCTS

and are listed as follows:

NC0 88 0110 001, Continued Troubleshooting of Circuit Boards.

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NCO 88 0110 002, Procedure to Cycle Rod Control Logic. Circuitry.

NCO 88 0110 003, Prepare a Supplement to this LER if Troubleshooting

.

Requires it.

NCO 88 0110 004, Clean and Test Demand Step Counters.

NCO 88 0110 005, Review of Technical Specification (TS) 3.1.3.3.

Necessary actions to close this LER are verification of completion of the

above corrective actions.

This LER will remain open.

(0 pen)

LER 327/88014,

Nonconformance with

Configuration

Control

Requirements Following a Post Modification Test of a Radiation Monitor

Resulted in a Containment Ventilation Isolation.

This event involved an EMI induced signal in the radiation monitor of

sufficient strength to cause a CVI.

This EMI occurred when instrument

maintenance personnel returned the local sample pump switch to the normal

(run) position, without notifying operations. Thus the cause of this CVI,

was an actuation of a control without proper authorization. The licensee

has issued a memorandum to all IM personnel stating that operations

personnel shall be contacted, and the RM trip signal blocked before

performing any work on RMs capable of actuating ESFs.

CVIs have been

generated several times by EMIs resulting from contact chatter. A team

was established to determine causes of EMIs and to suggest a fix to

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eliminate or reduce the occurrences.

This team ' was established as a

result of a CAQR which was closed in NRC IR 327,328/88-19. Modifications

have-been completed to the involved RMs. The licensee's actions appear to.

be adequate. However, the licensee intends to issue a supplement to this

LER to detail specific actions taken to prevent recurrence.

This LER remains open pending issuance and review of the revised LER.

(0 pen) LER 327/88017 Inadequate Tagging of a Radiation Monitor Pump Switch

Results in a Containment Ventilation Isolation.

.This event involved an EMI induced signal of sufficient magnitude to cause

a RM actuated CVI. This CVI was generated in the same canner as the one

described in LER 327/88014.

The inspector has reviewed LER 327/88017 and concurred with the root cause

determination. The licensee has placed a hold order (H0) on the RM local

sample pump switch as an interim corrective action. Long term corrective

action will consist of a permanent sign on the RMs, that would indicate

that improper operation could cause ESF actuation and instruct that RMs be

blocked prior to any work performance.

A contributor to this event was

contact chatter (see LER 327/88014 in this report).

The licensee's actions

appear to be adequate.

,

This LER remains open pending review of completed long term corrective

actions.

(0 pen) LER 327/87030, Revision 1, Blown Fuses in Emergency Start Circuits

Results in Spurious Emergency Diesel Generator Starts on Two Occasions.

This event involves the spurious start of the standby diesel generators

due to mechanical failure of the FLAS-5 fuses supplied by Littlefuse, Inc.

The licensee contacted Littlefuse, Inc. and was informed that the

production process was changed after lots 2 & 3 were delivered to TVA.

This change should make the fuses more reliable. This LER also fulfills

the 10 CFR 21 report commitments.

The inspector has reviewed the LER, the licensee's response to the problem

and Information Notice (IN) 87-62, Mechanical Failure of Indicating-type

Fuses.

The problem appears to be in the manufacturing process of fuses

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prior to. lot number 4.

No fcilures have been attributed to fuses supplied

after lot 3.

The licensee has replaced all fuses from lots 2 and 3 and

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has completed testing of FLAS-5 fuses from lots 4, 6, 10, 11 and 12. The

test results indicate an average expected life of 80 months and a minimum

expected life of 23 months. The licensee has committed to evaluating the

application of FLAS-5 fuses by March 31, 1989. This commitment is tracked

by CCTS Number NCO 870 227 003, Evaluate Application of FLAS-5 Fuses.

This LER, remains open pending review of the TVA evaluation.

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(0 pen) LER 327/87043, Control Room Isolation Caused by Electromagnetic

Interference on a Radiation Monitor During Flow Switch Testing.

During surveillance testing of sample flow switches for the main control

room intake process radiation monitor 0-RM-90-206, electrical disturbances

from- the switch contacts -chattering caused an inadvertent control room

ventilation isolation due to a high radiation signal on radiation monitor

0-RM-90-126.

The plant staff verified that no high radiation levels

existed.

The licensee has revised Design Change Request (DCR)-2276 to

modify monitor 0-RM-90-206.

Modification of 0-RM-90-126 is complete.

Modification of 0-RM-90-206 is not complete.

This item remains open pending completion of corrective action.

(0 pen) LER 327/87070, Revision 1, Diesel Generator Voltage Low When Output

Breaker Closes Because of a Component Defect Found During Surveillance

Testing and Design Deficiency Outside of Plants Design Basis.

During the Conduct of Special Test Instruction (STI)-77, Loss of Offsite

Power with Safety Injection on October 21, 1987, EOG 2A-A failed to

achieve the required output voltage at the time of the output breaker

closure. The cause of the low voltage condition was determined to be due

to a defective voltage regulator component (installed November 8,1986)

and output breaker close permissive relays tolerance that allowed breaker

closure at less than minimum design voltage.

The defective voltage

regulator component was repaired and 10 CFR Part 21 notifications were

appropriately accomplished. The remaining EDGs and a spare component were

verified to be correct.

Following repairs to the voltage regulator, EDG

. voltage rise still exhibited slow development with respect to breaker

closure. Further investigation revealed that the EDG output breaker close

permissive relays CX and AX tolerance was excessive, allowing the breaker

to close prior to achieving the required voltage.

Adjustments were made to the EDG governor actuators to achieve the

required voltage output at time of breaker closure. STI-77 was completed.

The performance data was analyzed by NRC staff in March 1988. Results of

the analysis are presented in NUREG 1232, Volume 2, Part 1 (SER) dated

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March 25,

1988, Revised Preliminary Safety Evaluation Report which

accepted the DGs for Unit 2 restart. However, the SER noted that the

conclusions for Unit 2 restart assumed Unit 1 in a cold shutdown condition

and that additional TVA calculation is required to support Unit I restart.

This LER also identified the PMT following maintenance as inadequtte. The

guidelines issued May 28, 1988, were reviewed, discussed with the licensee

and appear adequate as interim guidelines. TVA has requested an extension

to the commitment in the LER (30 days after Unit 2 Mode 2) to provide PMT

procedures for EDG testing following maintenance.

The following items require resolution prior to this LER closure.

Acceptable TVA calculations supporting two unit operation.

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Acceptable PMT procedures.

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This LER remains open.

6.

Inspector Follow-up Items

Inspector Follow-up Items (IFIs) are matters of concern to the inspector

which are documented and tracked in inspection reports to allow further

review and evaluation' by the inspector.

The following IFIs have been

reviewed and evaluated and the inspector has either resolved the concern

identified, determined that the licensee has performed adequately in the

area, and/or determined that actions taken by the licensee have resolved

the concern.

(0 pen) IFI

27/85-17-06, 327/85-17-05, Review Licensee QA Exceptions for

Feed Water Isolation Valve Stem Replacement.

This IFI was initiated to evaluate TVAs process for controlling the

installation of safety related material that has been identified as

nonconforming or has inadequate documentation. This question arose when a

valve stem with outstanding licensee identified QA exceptions to vendor

documentatien was installed on valve 1-FCV-3-47.

Installation of items prior to final disposition of nonconformances is

addressed in Section 7.6 of AI-11, Receipt Inspection, Nonconforming

Items. QA Level / Description Changes and Substitutions.

This procedure-

requires

plant manager approval

for the installation,

and the

Nonconforming Item (NCI) tag to remain attached to the item and Site

Quality Managers approval. However, the responsibility for ensuring that

the equipment or associated system is not declared operable until the NCI

is cleared is given to the requesting organization. No formal controls or

documentation requirements are currently in place to assure that. equipment

or systems are not declared operable with outstanding non-conforming

conditions. CAQR SQN 880350 has been issued to address this concern.

It

is expected that AI-46, Supplemental Requirements for Release of Material

From Power Stores-Unit 0,

will be revised to provide the required

controls.

-The inspector reviewed Maintenance Request A-300025 which removed and

installed the replacement stem in 1-FCV-3-47 and supporting documentation.

The nonconformance was related to the fact that the vendor's Certificate

of Conformance (C0C) had not been provided.

Documentation shows the C0C was received and the nonconformance cleared in

May 1985.

The inspector reviewed the C0C. There is no technical concern

related to this individual material.

This item remains open pending completion of programmatic corrective

actions.

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(Closed) IFI 327,328/86-69-05, Possible Loss of ENS With Loss of Offsite

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Power.

This item involves the requirement, identified in IE Bulletin 80-15, that

the ENS have a reliable power source. Sequoyah has a station package that

requires on-site power. The ENS hcs an uninterruptible power supply that

is backed up by a battery operated power supply (thirty minute minimum).

ECN 7088 and WP 7088-1, installed an inverter powered from a reliable

backup source-that completes the necessary ENS modifications.

The inspector has reviewed Bulletin 80-15, NRC IR 327.328/86-69, the

licensee's response (Gridley/ Grace) dated November 5, 1986, ECN 7088, WP

7088-01 and the installation.

The licensee's actions appear to be

adequate.

This item is closed.

(Closed) IFI 327,328/87-19-01, Control of Shelf Life On In-Plant Stored

Material.

This item involved a question raised regarding the shelf life of material

requisitioned from power stores and subsequently stored for extended

periods- of time in the plant.

Specifically, the inspector raised

,

questions regarding the shelf life of the reactor cavity seal, which is

stored in the plant between uses.

In addition, the inspector requested

information regarding procedural controls for in plant storage of other

shelf life material.

Although the manufacturer specifies a warehouse shelf life of twelve years

for the reactor cavity seal, once the seal is used it may continue to be

reused as long as it is not damaged or worn or subjected to sufficient

radiation to cause material embattlement.

MI-1.2, Removal And Replacement

Of RPV Head And Attachments, Revision 24, Section 9.2.1, requires that a

visual inspection for damage or wear be performed and that a durometer

check be performed to detect material degradation prior to each use.

Between uses the seal is stored in an air-tight stainless steel box in the

Auxiliary Building, as it is not desirable to store it in the power stores

warehouse due to the presence of radioactive contamination.

On February 10, 1987, a Directive (R00-870210-910) was issued to all TVA

sites, requiring that no spara material be issued from power stores

without a specifically designated end use.

This action should prevent

storage of materials within the plant and preclude exceeding shelf life

intervals of material that is outside warehouse control.

Corrective

action for this item appeared to be adequate.

This item is closed.

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7.

10 CFR Part 21 Report Follow-up (90712)

(0 pen) Part 21 (P21)-86-01, Atwood and Morrill Main Steam Isolation Valves

Spring Failure Caused by Quench Cracks.

This issue was addressed in NRC Information Notice (IN) 86-81. .The

inspector reviewed the licensee's response to IN 86-81.

The inspector

verified that corrective actions had been completed on Unit 2 and that the

Unit 1 actions were being tracked on the licensee's TROI system under the

title, Isolation Valves and is scheduled for completion prior to entry

into Mode 3.

This item remains open.

(0 pen) P21-87-01, Failure in Silicone-Rubber Insulated Cables Manufactured

by Rockbestos, Anaconda-Erickson, and American, During High Potential

Test. (Possibly Damaged During Transport).

NRC Inspection Report 327,328/87-76, previous discussions between the NRC

and TVA management and TVA commitments specify additional testing of

selected cable samples during the next refueling outage.

The requirement

for performance of additional cable testing is committed to in the

licensee's CCTS (NC0 87 0322 002).

Failed silicone rubber cables have

been replaced.

This item remains open pending completion of additional testing.

(0 pen) P21-87-02, Defective MIS-5 Indicating Fuses.

This report was documented in LER 327/87030 (paragraph 6) and describes a

possible generic problem with Littlefuse, Inc., FLAS-5 fuses manufactured

prior to lot number 4.

Subsequent to lot 3,

Littlefuse changed the

production process and the solder material used in FLAS-5 fuses.

This

change appears to have corrected the observed problem. TVA has conducted

testing of FLAS-5 fuses from lots 4,

6,

10, 11 and 12.

The results

indicate an average expected life of 80 months with a minimum expected life

of 23 months.

TVA is cor.tinuing to evaluate the application of FLAS-5

fuses.

This evaluation is expected to be completed by March 31, 1989.

This issue remains open.

(Closed) P21-87-03, Miswired Diode Causes Slowed Voltage Response of EDG

Voltage Regulator.

Regulator Manufactured by Basler Electric.

On November 8, 1986, during troubleshooting of voltage fluctu6tions of the

2A-A EDG the licensee replaced the voltage regulator, supplied by Basler

Electric Company. Post maintenance testing failed to detect the defective

replacement part. However, during performance of STI-77, Loss of Offsite

Power With Safety Injection-2A-A Containment Isolation Test, the DG

(i

-

.

.

.

.

.

.

.

.

.

24

t'

voltage . failed to respond as specified.

Subsequent troubleshooting

revealed that a diode in the voltage regulator was installed incorrectly.

The licensee's corrective actions properly wired the diode in the voltage

,

regulator for 2A-A EDG, verified the diode was installed correctly in all

~

other EDG regulators, inspected a spare voltage regulator as properly

'

configured, and confirmed that this type voltage regulator is not used at

other TVA nuclear facilities.

A 10 CFR 21 notification was provided to

NRC Region II on November 18. 1987, followed by written LER 327/87070.

Failure of the post maintenance testing program to detect the defective

voltage regulator at the time of installation is being tracked by

Violation 327,328/87-76-01. Based on the review conducted, the licensee's

actions pertaining to this particular 10 CFR 21 item are adequate.

This item is closed.

8.

NRC Bulletins (92701)

(Closed)Bulletin 80-12, Oecay Heat Removal (DHR) System Operability

The closure of this oulletin is discussed in paragraph 4 under URI

327,328/86-46-06, a follow-up item to the bulletin.

This bulletin is closed.

.

9.

List of Abbreviations

ABGTS

-

Auxiliary Building Gas Treatment System

ABSCE

-

Auxiliary Building Secondary Containment Enclosure

AI

-

Administrative Instruction

Abnormal Operating Instruction

A0I

-

BIT

-

Boron Injection Tank

C&A

-

Control and Auxiliary Buildings

CAQR

-

Conditions Adverse to Quality Report

Centrifugal Charging Pump

CCP

-

Corporate Commitment Tracking System

CCTS

-

COPS

-

Cold Overpressure Protection System

CREVS

-

Control Room Emergency Ventillation System

Containment Ventilation Isolation

CVI

-

DC

-

Direct Current

DCN

-

Design Change Notice

ECCS

-

Emergency Core Cooling System

EDG

-

Emergency Diesel Generator

Emergency Instructions

EI

-

ENS

-

Emergency Nctification System

Engineered Safety Feature

ESF

-

Flow Control Valve

FCV

-

GDC

General Design Criteria

-

GL

-

Generic Letter

,

.- - - .. - ..

-

.

.

.

.

.

.

.

.

-

25

HIC -

Hand-operated Indicating Controller

H0

-

Hold Order

IE

-

Inspection and Enforcement

IEB -

Inspection and Enforcement Bulletin

IM

Instrument Maintenance

IMI

-

Instrument Maintenance Instruction

IR

-

Inspection Report

KVA -

Kilovolt-Amp

KW

-

Kilowatt

Kilovolt

KV

-

LER -

Licensee Event Report

LCO -

Limiting Condition for Operation

MI

-

Maintenance Instruction

NOV .-

Notice of Violation

Nuclear Regulatory Commission

NRC

-

OS LA -

Operations Section Letter - Administrative

OSLT -

Operations Section Letter - Training

OSP -

Office of Special Projects

PMT -

Post Modification Test

i

PORS -

Plant Operation Review Staff

PRO -

Potentially Reportable Occurrence

QA

-

Quality Assurance

Reactor Coolant System

RCS

-

,

RG

-

Regulatory Guide

RM

Radiation Monitor

-

RO

-

Reactor Operator

Residual Heat Removal

RHR

-

RWST -

Reactor Water Storage Tank

SER -

Safety Evaluation Report

SG

-

Steam Generator

SI

-

Surveillance Instructien

System Operating Instructions

SOI

-

SR

-

Surveillance Requirements

Senior Reactor Operator

SR0

-

STI -

Special Test Instruction

TACF -

Temporary Alteration Control Room

TROI -

Tracking Open Items

TS

-

Technical Specifications

TVA -

Tennessee Valley Authority

Violation

VIO

-

Work Control Grouc

WCG

-

WP

-

Work Plan

WR

-

Work Request