text

LER-1996-010, on 960829,violations of TS & Design Basis Occurred.Caused by Inadequate Standards for TS Clarification.Deleted Clarifications in Question on 960829 & Revised One Clarification
Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(1); 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
	3731996010R00 - NRC Website
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:on 960829,violations of TS & Design Basis Occurred.Caused by Inadequate Standards for TS Clarification.Deleted Clarifications in Question on 960829 & Revised One Clarification
	ML20148M870
Person / Time
Site:	LaSalle
Issue date:	06/19/1997
From:	Swihart G; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20148M873	List: 05000373/LER-1996-010, :on 960829,violations of TS & Design Basis Occurred.Caused by Inadequate Standards for TS Clarification.Deleted Clarifications in Question on 960829 & Revised One Clarification;
References
	LER-96-010, LER-96-10, NUDOCS 9706250276
	Download: ML20148M870 (19)
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I FACILITY NAME(1):

DOCKET NUMBER (2)

PN35 (3)

LaSalle County Station Unit One 05000373 1 of 19 I TITLE (4)

Inadequate Standards for Technical Specification Clarifications Resulted in Violations of Technical Specifications and Design Basis 1

EVENT DATE (5)

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LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

MOurH DAY YEAR YhAR SWULNHAL Rhyl510N MONTH DAY YLAR FACILITY NAME DOCKET NUMBER LaSale County Station Unit 05000374 Two 08 29 96 96 010 03 06 19 97 FACILITY NAME DOCKET NUMBER ogTipl 1

l THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR i: (Check one or more)(11)

Poun 100 isvst o 20.2201(b) 20.2203(a)(3Xi) 50.73(aX2Xiii) 73.7 Hb) 20.2203(aH I) 20.2003(a)(3 Xii) 50.73(aH2Xiv) 73.71(c) 20.2203(aX2Xi) 20.2003(aX4) 50.73(aX2X v)

OTHER 20.2203(ax2Xii) 50.36(cX l) g3(ax2 Xvii) 20.2203(a)(2XiH) 50.36(cH2) 50.7FaH2XviiiXA)

(Specify in Abstract 50.73(aX2Xi) 50.73(O(2XviiiXB) below andin Text.

20.2203(aX2Xiv) 50.73(aK2Xii) 50.73(a) 2Xx)20.2003(aK2 Xv)

LICENSEE CONTACF FOR THIS LER (12) lNAME TELEPHONE NUMBER (Include Area Code)

Gerald L. Swihart, ReLulatory Assurance (815) 357-6761 Extension 2794 COMPLETE ONP i.INE FOR EACH COMPONENT FAILURE DESCRIHED IN Tills REPORT (13)

CALSE SYSTEM COMPONENT MANUFAGERLR RhPURIABLE

CAUSE

SYSTEM COMPUNENT MANUFACIURER kiPURIABLE TO NPR DS TO NPR DS SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR IO NO SUBMISSION YES (If yes, complete EXPECTED SUBMISSION DATE)

D ATE (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately fifteen smgle-space typeu ritten lmes 16)

On August 29, 1996, LaSalle Station determined that 15 Technical Specification (TS) clarifications and on December 31, 1996, three more TS clarifications (that had been proceduralized) were identified that resulted in past violations of the TS or placed the plant in conditions outside design basis for a short time whenever a situation developed in which the administrative guidance of these clarifications had actually been used.

On April 18, 1997, a review of 39 previously cancelled or deleted clarifications was completed and determined that 11 were in violation of TS or outside the design basis.

The root cause was inadequate standards for the preparation and approval of TS clarificaticas. Compliance with licensing and design basis intent was inappropriately deemed acceptable, rather than adherence to the literal wording and intent. An inadequate Onsite Review program failed to ensure high technical standards.

The corrective actions program failed to detect and/or to prevent these problems.

For the inappropriate TS clarifications identified on August 29, 1996, the immediate corrective actions were deletion of 14 clarifications on August 29, 1996 and revision of 1 clarification on August 30, 1996.

For the inappropriate TS clarifications identified on December 31, 1996, departments associated with affected procedures were notified of the unacceptable clarifications, or the additional restrictions needed to comply with the Technical Specifications.

A temporary Independent Safety Review Overview Group was established to perform an independent review of approved safety evaluations to enforced standards.

Future corrective actions include training, and changes to Corrective Actions Program, and l

Onsite Review process.

Implementation of Improved Technical Specifications is scheduled for 1999. This will improve clarity and flexibility of LaSalle's TS.

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YEAR SEQUENTI AL REVISION NUMRER NUMBER LaSalle County Station Unit One 05000373 96 010 03 2 of 19 t

l (If more space is required, use additional copies of NRC Form 366A)(17)

PLANT ANC SYSTEM IDENTIFICATION General Electric - Boiling Water Reactor j

Energy Industry Identification System (EIIS) codes are identified in the text as [XX).

A.

CONDITION PRIOR TO INITIAL EVENT Unit (s): 1/2 Event Date: 08/29/96 Event Time: 2054 Hours Reactor Mode (s): 1/1 Mode (s) Name: Run/Run Power Level (s): 100%/83%

3.

DESCRIPTION OF EVENT

f j

As a result of a recent NRC Engineering Support Inspection at Zion Station, l

Engineering and Regulatory Assurance performed a review between August 26 and i

August 29, 1996, of the active LaSalle TS Clarifications book for Units 1 and 2.

The results were presented to Station Senior Management at the Plant Operation Review Committee held August 30, 1996.

The team determined the.t most of the TS clarifications were consistent with the Technical Specifications literal wording and licensing / design basis.

However, the team could not confirm through engineering judgment that the interpretations met design and licensing basis for the 17 of the 43 clarifications (written between 1987 - 1995), On August 29, 1996, 2054 hours0.0238 days 0.571 hours 0.0034 weeks 7.81547e-4 months , 16 of those 17 were promptly deleted. One was eppropriately revised and approved August 30, 1996. The Operations Manager ensured that the Operating l

Shift Managers were immediately made aware of the changes and discussed potential for impact on plant operation and compliance with TS.

The 17 inadequate TS clarifications were further reviewed by Regulatory Assurance to determine if any potential violations of TS or if any conditions outside design basis had occurred, and to determine the root causes. Of the l

17 clarificaticns, 15 were issued instead of submitting a TS amendment request

(

to correct information or to implement new requirements, or were less l

conservative than the literal wording of the LaSalle Technical Specifications, or did not fully meet design basis; 2 were no longer needed.

T.nterviews of Operating and Engineering personnel confirmed that these 15 clarifications had been implemented or used in the past.

Regulatory Assurance performed a more extensive review of these 17 TS clarifications, as required in revision 0 of this LER, to determine if there were any more specifics concerning potential violations of TS or i

I conditions outside the design basis and no additional pertinent information was obtained.

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$EQUENTIAL REYl$10N NUMBER NUMBER LaSalle County Station Unit One 05000373 96 010 03 3 of 19 (if more space is required, use rJditional copies of NRC Form 366A)(17)

On December 31, 1996, per original "Long Term Corrective Action" number 3, a team of regulatory assurance and engineering personnel with extensive operating experience completed the review of seven TS clarifications that were closed prior to August of 1996 by incorporating the clarifications into station procedures. These clarifications received the review comparable to that given to active TS clarifications in August, 1996. Three of these clarifications were found to be outside of the licensing basis, or not in accordance with the of the Technical Specifications:

1.

TS 3.4.9.1, 3.4.9.2, 3.9.11.1, and 3.9.11.2, Clarification 01-87, page 11, explanation of requirements for a system, subsystem, or I

component to be an Alternate Shutdown Cooling system to satisfy TS action requirements for inoperable shutdown cooling loop (s), which was incorporated into var.ious shutdown cooling and alternate shutdown cooling procedures.

2.

TS Surveillance Requirements (SRs) 4.9.1, 4.9.1.2, 4.9.2, 4.9.3, 4.9.5, 4.9.6, 4.9.8, Clarification 02-89, Refuel Operation Surveillances, which was incorporated into LFS-100-4, " Core Alteration Shiftly Surveillances".

3.

The TS Section 6.0 Administrative Controls, 6.7, Process Control Program, Clarification 05-92, Onsite Review of Radraste Vendor Procedures defining review of minor versus major changes, which was incorporated into LAP-200-6, "LaSalle County Station Process Control Program".

Between December 31, 1996, and January 2, 1997, departments or groups associated with the affected procedures were notified of the identified potential TS violations.

Procedures will be revised to assure the procedures comply with the literal wording of the TS and the design and licensing basis.

This event is reportable per 10 CFR 50.73 (a) (2) (1) (B) due to the violation (s) or potential violation (s) of the Technical Specifications and 10 CFR 50.73 (a) (2) (ii) (B) due to having been in a condition outside the design basis of the plant.

C.

CAUSE OF EVENT

The need for written TS clarifications at LaSalle resulted from inconsistent implementation of a few poorly worded TS sections. The root cause of this event was inadequate standards for the preparation and approval of TS clarif3 cations. Management accepted the TS clarifications as long as licensing and design basis intent was met rather than adherence to the literal wording and intent.

A fundamental cause to this event was inadequate root cause analyses of related problems in the past and/or ineffective corrective actions. There was evidence that the Stetion had previous problems in licensing / design basis work.

When the Station became aware of problems in its ability to perform safety evaluation work, it did not go back and revisit a sampling of the safety evaluation work related to implementation of the TS clarifications to determine if the clarifications were thoroughly justified and met the expected standards. The Station has had several indications over the time period that

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$cQUENTIAL P.EVLSION NUMBER NUMBER LaSalle County Station Unit One 05000373 96 010 03 4 of 19 (if more space is required, use additional copies of NRC Form 366A)(17) these clarifications were written (1987-1995), management did not consistently hold people accountable for strict adherence to the literal wording of controlled documents, such as procedures. There were several events which had as a root cause or contributing cause inadequate or deficient procedures.

This was an indication that the Station needed to improve its OSR/ technical reviews.

Contributing factors were:

1.

The station process to change Technical Specifications is complex, time consuming, and too slow to support Station needs.

a.

The management priority to correct and continuously maintain the licensing and design basis documents current was inadequate.

b, The annual review of TS clarifications for applicability did not include requirements to submit requests for amendment to the TS wording to allow deletion of TS clarifications.

Some clarifications were over 9 years old.

2.

Although implementation of literal compliance with TS had been part of licensed operator training, this same rigor for literal compliance with TS was not applied to development and review of TS clarifications and associated safety evaluation work.

a.

Initial preparation of some of the TS clarifications and evaluation of license and design basis was inadequate. This was due to a combination of inadequate procedures, inadequate training, and not being familiar with increasing performance standards, b.

In some cases, OSR failed to perform adequate independent searches or rigorous reviews to verify compliance with the design and licensing basis.

D.

ASSESSMENT OF SAFETY CONSEQUENCES

Using a more conservative philosophy that TS clarification compliance requires adherence to not only intent but also literal wording, the Regulatory Assurance review determined that 15 of the TS clarifications should be reported because implementation of them could have been considered to be less conservative than the literal wording of TS, or violations of the literal wording TS, or could have placed the plant in a condition outside design basis.

None of these violations resulted in any significant consequences or risk to the health and safety of the public under actual or accident conditions.

See Attachment A for a summary of the individual TS clarifications and the related safety analyses for each.

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NUMRFR NUMBFR LaSalle County Station Unit One 05000373 96 010 03 5 of 19 (If more space is required, use additional copies of NRC Form 366A)(17)

E.

CORRECTIVE ACTIONS

1 The immediate corrective actions upon discovering that some TS clarifications i

may have resulted in potential deviation or violation of licensing or design basis were:

1.

The Station immediately deleted the clarifications in question on August 29, 1996 and revised 1 clarification to ensure proper interpretation and implementation on August 30, 1996.

i i

2.

The results of the TS clarificat. ions review by the team of engineering and Regulatory Assurance personnel were submitted to senior management review at the Plant Operations Review Committee (PORC) August 30, 1996.

4 3.

The Operations Manager immediately communicated item 1 to the Shift q

Managers August 29, 1996 and August 30, 1996, including a detailed review for impact on plant operation. A formal letter was issued Sunday, September 1, 1996 to Operating.

4.

The Site Vice President and Engineering Manager established a temporary Independent Safety Review Overview Group on August 26, 1996 to perform an independent review of approved safety evaluations and engineering evaluations supporting operability determinations prior to implementation of the change being submitted, as a method to enforce standards.

5.'

On September 5, 1996, LAP-1200-17, " Operating License / Technical Specification Clarifications", was revised to require engineering design basio review and Regulatory Assurance licensing basis review of any new or revised clarifications, with the same level of review annually for I'

all active clarifications, 6.

On September 13, 1996, the Station Site Vice President issued Comed's nuclear policy, NOP-OA.38, ' Nuclear Safety Policy," revision 0, to reinforce senior management expectations on conservative decision J

making, emphasize nuclear power plant operational safety culture, and rigorous adherence to procedures.

7.

A team of engineering and regulatory assurance personnel with extensive operating experience performed a review of TS clarifications that have been incorporated in written guidance such as procedures.

Seven TS clarifications have been closed in the past by incorporating the guidance in station procedures. Of these seven, three were found on December 31, 1996, to not meet the design or licensing basis or the

]

literal wording of the TS.

This LER supplement addresses the three additional clarifications.

]

8.

By January 2, 1997, departments and groups were notified of the deficient procedural guidance concerning previously incorporated TS clarifications.

It was verified by cognizant personnel that the LaSalle County Station was in compliance with the Technical Specifications at the time of discovery.

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YLAR SLQUENTIAL kEVlSION NUMDFR NUMBF R LcSalle County Station Unit one 05000373 96 010 03 6 of 19 (If more space is required, use additional copies of NRC Form 366A)(17) 9.

The station has reviewed and concluded that the present procedures for alternate shutdown cooling are adequate for current plant conditions.

Procedures associated with decay heat removal are being evaluated for changes needed to address the requirements for an

Alternate Shutdown Cooling" method to satisfy the design requirement to be capable of removing decay heat within one hour.

The applicable procedures will be revised prior to startup of either unit.

10.

LAP-200-6, *LaSalle County Station Process Control Program", will be revised to clearly distinguish between the review requirements for changes to the Process Control Program versus implementing procedures. Until this procedure is revised any vendor program / implementing procedure changes will be reviewed and approved by Onsite Review.

11.

LFS-100-4, " Core Alteration Shiftly Surveillances" will be revised prior to Core Alterations to assure that Tech Spec refuel mode surveillance requirements for Core Alterations are performed for the required conditions and at the required frequencies.

12.

The initial safety analysis of the 15 clarifications was reviewed and determined to be accurate based on the availability of records regarding the specific uses of the TS clarifications. Attachment A to this LER is updated to evaluate the three additional proceduralized clarifications.

13.

On December 31, 1996, Regulatory Assurance completed a re-evaluation of procedure LAP-1200-17, related to preparation, review, approval and periodic review of TS clarifications and found the procedure to be acceptable to assure compliance with the design and licensing basis.

Corrective actions to prtvent recurrence are:

1.

TS clarifications that were in effect as of August 25, 1996 and any subsequent additions or revisions will be considered during development of Improved Technical Specifications (ITS) wording to ensure the ITS is clearly worded, complete, reasonable, and does not conflict with design bases or other TS sections (can be implemented directly as worded). The clarity and flexibility of ITS will minimize the need for new TS clarifications.

The current active clarifications will be deleted when the ITS are planned to be implemented in by August 31, 1999, 2.

An assessment of the remaining TS clarifications that were closed prior to August, 1996, and not incorporated into station procedures has been made by a licensing engineer. Of the 39 clarifications previously closed, the assessment determined that ten are possibly not in accordance with the design and/or licensing basis at the time the clarifications were in use.

I a.

An in-depth review of the ten questionable clarifications was completed February 28, 1997.

It was determined that nine of the ten clarifications were outside the design and/or licensing basis.

A summary of the safety analysis and description has been added as Attachment B.

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WAR SEQUENTIAL REVISION NUMBFR NUMilER LaSalle County Station Unit One 05000373 96 010 03 7 of 19 (If more space is required, use additional copies of NRC Form 366A)(17) b.

A review of the remaining 29 closed TS clarifications, completed on April 18, 1997, determined that 2 of these closed clarifications were not in accordance with the design and/or licensing basis. A summary of the safety analysis and description has been added as Attachment B.

3.

An Action Plan was issued in December 1996, to implement improvements to the onsite review and investigative function and the technical review processes. Additionally, improvements the safety evaluation process a

were also identified.

Improvements to tne safety evaluation process are complete.

These improvements include the establishment of a temporary Independent Safety Review Overview Group and the issuance of a new six-site common process for performing safety evaluations. Actions contained in the plan to update and maintain the list of qualified participants; improve the control, intent, and expectations of signature authority; and, clarify Onsite Review requirements are complete.

Actions to improve the Technical Review and Onsite Review processes are in progress.

Required improvements have been identified, new procedures have been drafted, and revisions to existing procedures have been identified.

Issuance of these procedures will be completed in October 31, 1997.

These improvements will be completed in parallel with issuing the six-site common Plant Operations Review Committee (PORC) procedure that is being standardized across all Comed Stations.

(NTS 373-180-96-010.14LER).

d.

Training was completed on March 29, 1997, on lessons learned from this event for pecple who prepare and review TS clarifications and safety evaluations.

5.

An improved corrective action program for all Comed nuclear plants has been developed for implementation in 1997.

F.

PREVIOUS OCCURRENCES

LER NUMBER TITLE None.

G.

COMPONENT FAILURE DATA

I Since no component failure occurred, this section is not applicable.

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YEAR SEQUENTIAL >

REVISION NUMBER NUMBER LaSalle County Station Unit One 05000373 96 010 03 8 of 19 l

(If mwe space is required, a.se additional copies of NRC Fonn 366A)(17) l Attachment A 1

l Summary of Safety Analyses i

1.

TS 4.1.3.5.b. Clarification 01-87, page 8 (Control Rod Drive)(CRD) (AA] Accumulator Operability) allowad LaSalle Station to consider a CRD scram accumulator operable when its visual alarm feature failed if, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, a person verified that the indicated pressure (of the accumulator) was I

greater than or %.ial to 940 psig and/or verified there was no water leakage for the affected Control hcd.

The review team determined that the visual alarm is a functioning component needed to meet the operability definition for CRD accumulators. There was minimal risk to the health and safety of the public due to the compensatory actions taken by the operators to assure that the affected CRD accumulator would function as designed, meeting the in'ent of the TS bases. A request for TS amendment should have been submitted.

2.

TS 3.1.3.6 Clarification 01-87, page 9 (Uncoupled Control Rods) was written to be used as a contingency due to having one uncoupled rod.

It allowed the Station to continue operation as if all Control Rods were coupled when an uncoupled Control Rod was declared inoperable, remained fully inserted and electrically disarmed, and as long as the requirements for the number and location of inoperable Control Rods in TS 3.1.3.1. were met.

There was minimal risk to the health and safety of the public due to the affected Control Rod being fully inserted and declared inoperable since the Station followed the action statements in TS, and there was no case found where a condition existed with multiple uncoupled rods.

3.

TS 4.11.2.2.2.b Clarification 01-87, page 18 (Off Gas sampling - Pretreatment noble gas increase > 50 percent).

The clarification allowed the Station to waive sampling if short duration (less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ) increases in release rates as indicated by Off Gas Pretreat noble gas readings were known to be not related to the condition of the fuel or thermal power changes.

Literal compliance with the TS as currently worded would not have provided any additional margin of safety to the health and safety of the public because sampling after such short duration increases would not be likely to have detected any anomaly, since it ic likely that the reading would have returned to normal by the time the sampling was accomplished, and because the increase was due to causes other than failed fuel (such as an electronics spike, a source being in the vicinity of the detector, etc.)

In addition, the station has multiple indications of failed fuel and changes in release rates, such as main steam high rad monitors, wide range gas monitor, etc. to confirm that offsite releases were within limits. A request for TS amerdment should have been submitted.

4.

'1S 4.5.1.d.1 Clarification 04-87 (Automatic Depressurization System)(ADS) [SB)

Bottle Bank Pressure Switch) allowed when ADS bottle bank low pressure alarm switches were out of service for surveillance or for maintenance, that ADS was not l

considered to be inoperable if special logs were maintained to record the pressure verified locally by operator with alarm function inoperable. The review team l

determined that the alarm is a functioning component required to meet the operability definition for ADS accumulators. There was minimal risk to the health and safety of the public due to the compensatory actions taken by the operators to assure AIS would function, meeting the intent of the TS bases.

A request for TS amendmee should have been submitted.

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$EQUEN HAL RtVISION j

NUMBFR Nt%fMFR j

LaSalle County Station Unit One 05000373 96 010 03 9 of 19 j

(If more space is required, use additional copies of NRC Fonn 366A)(17)

I 5.

'TS 3.6.4 Clarification 07-88 (Drywell Vacuum Breakers position indication) allowed i

the Station to call the Drywell Vacuum Breakers and position indication operable when one of two power supplies to a single Drywell Vacuum Breaker position j

indication was not available.

The guidance for use of 3.0.5 was nonconservative.

The review team determined that both power supplies to both position indications should be operable functioning components to meet the operability definition for i

Drywell Vacuum Breakers and position Indication. There was minimal risk to the health and safety of the public due to at least one power supply and therefore one l

position indication being available and the (mechanical) vacuum breaker' remained j

capable of opening and closing.

6.

TS 3.3.7.1 Clarification 13-89 (Control Room Ventilation) (VC) (VI] (Radiation j

Monitors) informed the operator that TS 3.7.2 was the governing specification if the train was inoperable (e.g. OOS) then the rad monitor instrumentation was not i

required to be operable.

Thus the operator would not start the other train in pressurization mode as required by TS 3.3.7.1 There was minimal risk to the health f

and safety of the public due to not operating an Emergency Makeup (EMU) train with j

inoperable rad instrumentation.

If minimum rad instrumentation was not operable, i

the action was followed for EMU train being inoperable.

(Note affected Control Room / Auxiliary Electric Room Ventilation (VC/VE) system is not operated when EMU train is inoperable.)

No cases were found when there was any valid radioactivity i

concern at the EMU air intake and there was no risk to the habitability of the Control Room, l

7.

TS violation of literal wording TS 3.7.1.1 Clarification 01-90 (Resida.1 Heat j

Removal (RHR) (BO) Service Water power sources) allowed the Station to call RHR j

operable in the shutdown cooling mode if there was 22 feet of water above the

]

reactor vessel flange and the associated RHRSW (RHR Service Water) subsystem was l

operable with a single power supply, which is less conservative than the literal j

3 wording.

The review team determined that both power supplies to RHRSW subsystems l

must be operable to meet the literal TS 3.7.1.1. wording for operability of RHRSW in conditions 4 and 5.

There was minimal risk to the health and safety of the public due to a system having one power source available for operability (same as j

j associated RHR pump in operating condition 5).

The risk of having only one RHR j

system is allowed in T/S 3.9.11.1.

No additional risk was created by allowing an i

. inoperable RHR system's service water system to be inoperable. Although the

{

flexibility is technically acceptable and other TS sections have power source

" notes' similar to that in this clarification, the confusing wording and apparent 1

conflicts between TS 3.7.1.1 and 3.9.11. along with the absence of a typical note j

concerning normal and emergency power supplies in condition 5 should have been 4

resolved by a request for TS amendment.

i 8.

TS 4.7.2 Clarification 07-90 (VC - control room boundary penetrations) allowed the

}

Station to call Control Room Ventilation operable during work activities that caused an opening in the VC pressurization boundary (such as a hole in the floor or breaking fire seals) if a means was established to return the boundary to a closed

]

status in a reasonable amount of time upon VC emergency train operating, thus avoiding entry into 3.0.3.

This interpretation did not have corresponding rigorous engineering testing or calculations nor design bases documentation to ensure that I,

the Control Room and Auxiliary Electric Rooms could be maintained at the required pressure and flow rates with holes in the boundary. There was minimal risk to the health and safety of the public due to preplanned compensatory actions to ensure a hole could be sealed upon VC Emergency Makeup train actuation.

4 i

l l

..u.

%RC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-01N (S 92)

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LaSalle County Station Unit One 05000373 96 010 03 10 of 19

_ If more space is required, use additional copies of NRC Form 366A)(17)

(

9.

TS 3.1.5 Clarification 09-90 (Standby Liquid Control (SBLC) [BR] power supply) allowed the Station to call the Standby Liquid Control system operable when the 1

second power supply was not avt11r.ble. The review team determined that both power l

supplies should be operable to meet the most conservative interpretation of l

operability definition for SBLC.

There was minimal risk to the health and safety of the public due to at least one power supply being available to each SBLC train.

l This did not meet literal wording of TS although the flexibility is technically i

acceptable and other TS sections have power source notes" similar to that in this a

clarification. A request for TS amendment should have been submitted.

l 10.

TS 3.8.1.1, 3.8.1.2, 4.8.1.1.1 Clarification 10-90 (Offsite Lines) allowed the Station to use the 138 KV lines as a third offsite power source without adequate engineering testing or analyses to confirm they could adequately supply shutdown loads as required. The UFSAR does not include the 138 KV line in the discussion of offsite power sources. There was minimal risk to the health and safety of the public since the Station did not experience the need to actually use the 138 KV lines power supply during a situation when full S/D load demand was present. The UFSAR should have been changed after appropriate analyser.

11.

TS 3.3.7.5, 3.5.3, 3.6.2.1 Clarification 13-90 (Suppression Pool Level Alarm actions) (CG) allowed the Station to consider suppression pool wide range water level instrumentation channels to be operable and that there was no need to enter the action statements for TS 3.5.3 or 3.6.2.1 when o*.? or more of the high and/or low suppression pool water level alarms became inoparable. A policy using a special log was initiated and suppression pool level was to be verified to be within TS limits at least once per shift by remote or local indication in these cases. The review team determined that the alarm is a functioning component needed to meet the operability definition. Upon deletion of this clarification, Unit 2 entered an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months shutdown timeclock per TS 3.6.2.1 Action d. on loss of narrow range alarm functions due to inoperable excess flow check valve associated with both suppression pool high level alarms. The repair was completed, tested and declared operable within the allowed time. There was minimal risk to the health and safety of the public due to the compensatory actions taken by the operators to assure that the i

suppression pool level met TS level limits, meeting the intent of the TS bases and due to the short time in which the situation existed. A request for TS amendment should have been submitted.

j i

12.

TS Table 1.2 Footnote Clarification 01-91 (Mode Switch to Run - rod insert verification) allowed administrative control using an out-of-service in place of a second licensed operator or other technically qualified member of the unit technical j

staff to ensure that control rods remained fully inserted when the reactor mode l

switch was placed in Run or Startup/ Hot Standby to test the switch interlock functions. There was minimal risk to the health and safety of the public since the

)

Station used an alternative method to ensure rods were maintained fully inserted via OOS on the withdrawal button. A request for a TS amendment should have been l

(

submitted.

i

i iNRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92)

EXPIRES 05/31/95 j

ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECrlON REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER)

MMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714). U.S. NUCLEAR TEXT CONTINUATION REGULATORY COMMISSION, WASHINGTON. DC 20555-0001, ANDTO l

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YkAR SEQUENTIAL RtWISION NL'MBFR Nt'MBER LaSalle County Station Unit One 05000373 96 010 03 11 of 19 (If more space is required, ise additional copies of NRC Form 366A)(17) l l

13.

TS 3.4.3.1, 3.4.3.2. Clarification 02-92 (Reactor Coolant System leakage) rev 0 stated that when a portion of the leakage detection instrumentation was inoperable, TS 3.4.3.1 and the Action statement of TS 3.4.3.1 should apply. As long as TS 3.4.3.1 did not require the unit shutdown due to inoperable leakage detection instrumentation, there was sufficient instrumentation to monitor reactor coolant system leakage per TS 3.4.3.2.

The instruments used to perform TS 4.4.3.2.1 surveillance was dependent on the operable instrumentation of TS 3.4.3.1.

This clarification was revised to ensure that when declaring instruments inoperable, applying 4.0.2 and 4.0.3 is the appropriate action. A request for TS amendment should have been submitted.

14.

TS 3.6.5.1 Clarification 2-95 (Secondary Containment Integrity holes) allowed waiving the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months timeclock for Secondary Containment integrity as long as the total hole area size or combination of hole area sizes did not exceed a given value based on leak rate testing and some design calculations.

The controls and documentation to support the license basis were weak.

There was minimal risk to the health and safety of the public due to limiting total hole size to ensure operability of Secondary Containment.

Since there was no guidance on holes in Secondary Containment for maintenance or testing purposes, a UFSAR amendment should have been submitted.

15.

TS 3.7.2, 3.7.7 Clarification 06-95 (VC/VE Emergency Filtration System vs.

Refrigeration) interpreted the lack of refrigeration for the Control Room as not required for habitability and thus did not affect operability of the Control Room and Auxiliary Electric Equipment Room Emergency Filtration Ventilation (VC/VE) systems. However, the UFSAR does include the refrigeration equipment in the design basis.

There was minimal risk to the health and safety of the public because the control room habitability was ensured by compliance with T/S 3.7.7 actions. Changes should have been submitted to resolve the apparent conflict between the TS and design basis or UFSAR.

16.

TS 3.4.9.1, 3.4.9.2, 3.9.11.1, and 3.9.11.2, Clarification 01-87, page 11 (explanation of requirements for a system, subsystem, or component to be an Alternate Shutdown Cooling system to satisfy TS action requirements for inoperable shutdown cooling loop (s)).

The design requirement for the alternate shutdown cooling system to be capable of starting to remove decay heat within one hour was not incorporated into the clarification for alternate shutdown cooling originally, nor the procedures for shutdown cooling and alternate shutdown cooling. At the time of the discovery, an alternate shutdown cooling method was available and capable of or removing the decay heat from the reactor coolant system on Unit 1, which was in Cold Shutdown.

Unit 2 was defueled and thus did not require shutdown cooling to be Operable.

One of the methods previously used as an Alternate Shutdown cooling loop may not have been capable of removing decay heat in one hour, depending on availability of certain flow paths. Therefore, there may have been times when a backup (second decay heat removal system (alternate or normal) may not have been able to begin removing decay heat within one hour.

However, there was minimal risk to the health and safety of the public because the decay heat was controlled and a decay heat removal system (either alternate or normal) always available or in operation as required by the TS.

l l

_.m._

.._...m._m.__..

~~_.____m._____

TNRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92)

EXPIRES 05/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLEC110N REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER)

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YEAR SEQUENTIAL REV!SION NtIMRER NUMBER LaSalle County Station Unit One 05000373 96 010 03

'12 of 19 (If more space is required, use additional copies of NRC Form 366A)(17) 17.

TS Surveillance Requirements (SRs) 4.9.1, 4.9.1.2, 4.9.2, 4.9.3, 4.9.5, 4.9.6, 4.9.8, Clarification 02-89 (Refuel Operation Surveillances) interpreted the definition of CORE ALTERATIONS versus the special operation condition of Core Alterations to be distinct from each other in the Refuel Operational Condition. The clarification stated that routine breaks between shifts performing Core Alterations of less than or equal to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months were not considered to be ' stopping" core alterations, provided surveillances were otherwise current.

Because the all capitals version of Core Alterations is used in the specifications, there is no clear distinction between a Core Alteration versus ongoing Core Alterations as a mode. There was minimal risk to the health and safety of the public because the periodic and other conditional SRs were performed at the required intervals during the time periods in which Core Alterations were being performed, i

18.

The TS Section 6.0 Administrative Controls, 6.7, Process Control Program, Clarification 05-92 (Onsite Review of Radwaste Vendor Procedures defining J

review of minor versus major changes) interpreted TS section 6 7 as allowing j

minor' changes to the Process Control Program (PCP) as requiring a Technical j

Review versus an Onsite Review..TS section 6.7 requires changes to the PCP to be reviewed and approved in accordance with the Onsite Review and Investigative Function.

TS section 6.2 allows the PCP implementing procedures to be reviewed and approved by the Technical Review and Control process described in 6.2.C and 6.2.D, which allow minor changes. The clarification does not distinguish between what is the PCP and what procedures implement the PCP, and thus could cause a violation of TS section 6.7.

There is minimal risk to the health and safety of the public because the minor change process that was used assured that changes did not change the final waste form.

Examples of minor changes included typographical errors, name or title changes, formatting changes.

v.

' NRC FORM 366 U.S. NUCLEAR REGULAT ORY COMMISSION APPROVED BY OMB NO. 3150-0104 i

(9-92)

EXPIRES 05/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATlON COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER)

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P 4GE (3)

YEAR SEQUENTIAL REYlSION NUMBER NUMBFR LaSalle County Station Unit One 05000373 96 010 03 13 of 19 (If more space is required, use additional copies of NRC Form 366A)(17)

Attachment B Summary of Description and Safety Analysis of Previously Deleted Technical Specification Clarification 1.

Prior to the relocation of Radiological Effluent Technical Specifications (RETS) by License Amendments 85 (Unit 1) and 69 (Unit 2), Technical Specification (TS) gaseous effluent sampling of the Main Vent Stack and Containment Vent and Purge per Table 4.11.2-1, Note b required the following:

" Analyses shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period."

l Clarification 01-87, page 17 stated that if there are several power transients that exceed 15 percent, samples may be delayed until after the last transient provided it is within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the first transient.

The allowance for one sample following more than one power change exceeding 15 percent within one hour was not in the TS and therefore did not literally comply with the requirements of TS 3/4.11.2.1.

However, in circumstances where transients cause multiple power changes greater than 15 percent of rated thermal power, the required samples could not be done between each change.

Also, taking a sample for each power change after the last power change would have provided unnecessary data and would provide no new information. Normal startups and shutdowns do not cause multiple power changes of greater than 15 percent in one hour, so this clarification is not of routine use.

Therefore, this clarification is of little safety significance. This clarification was incorporated into the Offsite Dose Calculation Manual (ODCM) by a revision in accordance with section 6.8.2, and can be used to comply with sampling requirements.

2.

TS clarifications 01-87, page 29, 08-87, and 04-89 involved TS 3.4.1.1 action a.1, with only one reactor coolant system recirculation loop in operation. The original problem was that action a.1.a required placing the recirculation (recirc) flow control system in the Master Manual mode, which was a more automatic flow control mode than

lower" modes. The primary. concern involved the ability for offsite load control with the recire flow control system in Master Auto mode per TS 3/4.1.6, which was not allowed in single loop operation. TS 3.4.1.1 action a.1 was changed by license amendments 60 (Unit 1) and 40 (Unit 2), dated September 7, 1988, to change it to read Master Manual mode or lower. This resolved the one concern with specifying a particular mode.

When this part of the clarification (prior to September 7, 1988) was in use, the safety significance was low, because the concern was to prevent offsite load control to change reactor thermal power with only one recirc loop in operation.

m-

_.-._m..

. _ ___ _ _~ _,_ -,. _ _. _. _ - ~ _. _ _ _.. _ _ _. _ _ _. _. _ _

,a'o o

NNRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92)

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YEAR SEQUENILAL REV LSION NUMBER NUMBER LaSalle County Station Unit One 05000373 96 010 03 14 of 19 (If more space is required, use additional copies of NRC Form 366A)(17)

The second part of these clarifications stated that the TS 3.4.1.1 required actions per action a.1, except action a.1.a.

do not apply below 25 percent power, because the referenced LCOs do not apply below 25 percent power. Only the section 3.2 TS LCOs are not applicable at less than 25 percent Thermal Power.

Also, Rod Block Monitor is not applicable below 30 percent power.

TS 2.2.1 and 3.3.6 for APRM flow biased simulated thermal power-upscale scram and rod blocks are required to be applicable in Operctional Condition 1.

TS Safety Limit 2.1.2 is applicable in Operational Conditions 1 and 2.

Therefore, TS 3.4.1 would have been violated if the applicable actions had not been done at less than 25 percent thermal power as allowed by this clarification.

There is significant margin to the reactor fuel thermal limits when power is less than 25 percent, so the consequences are minimal if this allowance had been used below 25 percent power with only one recirculation loop in operation. Also, there have been only a few cases where unit operation with only one recire loop in

- operation has occurred, and typically operation has been at greater than 25 percent of rated thermal power. Based on the above, the safety significance of the use of this clarification was minimal.

3.

Prior to license amendments 55 (Unit 1) and 36 (Unit 2), TS Surveillance Requirement 4.4.1.2,1, which verifies Reactor Coolant System Jet pump operability, l

required the surveillance to be performed "when both recirculation loops are operating at the same flow control valve position".

During a LaSalle Unit 1 planned mid-cycle outage, the 1B Recirculation Pump was repaired due to high vibration. As part of the repair, a new pump impeller was installed which was not a like for like replacement resulting in a rated flow reduction of about 2.5 percent. This resulted in about a 6 percent position difference between Recirc Flow Control Valve position l

between the two recirc loops in order to balance flows. With pumps of equal pumping capacity, equal flow control valve positions should result in the same drive flows between loops. The detection of inoperable jet pumps depends on equal loop flows and so with the reduced pump flow, a TS change was required.

The TS change was submitted August 29, 1987, and Clarification 07-87 was approved September 4, 1987, and Unit 1 was resumed operation on September 15, 1987. Clarification 07-87 was written and approved after the request for a TS change was submitted on August 24, 1987, to cover the time until the license amendments were issued based on an agreement with the LaSalle NRC resident and the LaSalle NRR Project Manager.

The license amendments were issued on March 4, 1988, so Unit 1 operated less than six months outside of TS 3.4.1.2.

The safety consequences were minimal, because jet pump' operability was maintained by conducting the jet pump surveillances at equal recirculatien loop drive flows.

4.

TS clarification 03-88, approved April 25, 1988, to provide guidance to the Operating Shift personnel on how to comply with TS 3.8.1.1, A.C. Sources -

Operating, for the circumstance when an additional diesel generator (s) become l

inoperable while one diesel generator is already inoperable. The clarification j

allowed entry into separate action statements for various combinations of diesel l

generators.

In particular, the following combinations were given as in compliance with the TS:

i 4

a.

If, while diesel generator 0 is inoperable, the unit diesel generator B becomes inoperable, then enter the actions required for each one being for each of inoperable. This allowed separate entry when the LCO was not met.

these diesel generators to be on its own 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed outage time.

i I

I

.,s.*s l

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)

)LAR SEQUENTIAL kEVISION NUMBER NUMBER LaSalle County Station Unit One 05000373

~

96 010 03 15 of 19 (if more space is required, use additional copies of NRC Form 366A)(17) b.

If, while diesel generator 0 is inoperable, the opposite unit diesel generator A becomes inoperable, then enter the actions required for each one being inoperable. This allowed separate entry when the LCO was not met for each of these diesel generators to be on its own allowed outage time, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months i

for diesel generator 0 and 7 days for the opposite unit diesel generator A.

j j

On October 27, 1989, a change to the TS was submitted to incorporate these examples and other combinations as separate action statements into TS 3.8.1.1.

License

]

r Amendments 72 (Unit 1) and 56 (Unit 2) approved these changes on March 15, 1990.

(corrected April 26, 1990).

For the time from April 25, 1988, to March 15, 1990, LaSalle would have been in violation of TS 3.8.1.1 if this clarification had been used.

Even during that time diesel generator maintenance on line is conducted on i

only one diesel generator.at a time.

The shutdown AC sources TS 3.8.1.2 is less restrictive in regard to diesel generator operability and would allow more than one diesel generator to be inoperable at a time.

During the time this clarification was in use, approximately twenty-three months, this clarification may have been used, but the number of times would have been limited, since diesel generator maintenance

)

l on line is scheduled one at a time and the LaSalle diesel generators have been highly reliable (greater than a 97.5 percent reliability per Station Blackout requirements). Also, the maximum total time (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ) that the more than one diesel generator could be inoperable during unit operation per occasion is small considering the total time (23 months) that TS 3.8.1.1 could have been violated.

Based on the above, the safety significance is minimal.

5.

Clarification 08-89 allowed diesel generator B (High Pressure Core Spray (HPCS) diesel generator) to be cons;-N. ed Operable per TS 3.8.1.1 under the following l

restrictions in order to avoid running the other diesel generators per the TS 3.8.1.1 actions for an inoperable diesel generator:

a.

The HPCS diesel is fully Operable prior to losing the AC bus fed by the diesel and/or the DC bus that provides control power to that electrical division.

l b.

No work, repairs, modifications or surveillances are to be performed on the i

HPCS diesel generator.

c.

No other condition exists to otherwise make the diesel generator inoperable 1

(except due to loss of AC or DC power supply).

d.

The purpose of placing the HPCS diesel generator in Maintenance (a key-locked control switch which prevents start signals (auto or manual) from starting the diese ) is to prevent energization of the AC bus, e.

Applicatle TS are followed:

3.8.2.1, AC distribution systems - operating and 3.8.2.3, DC distribution systems - operating, or 3.8.2.2, AC distribution systems - shutdown and 3.8.2.4, DC distribution systems - shutdown.

1

~ -. ~ - - - - - - -. - - - - - -. _. - - - - - - -

.. (*,

a

$kC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92)

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AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR TEXT CONTINUATION REGULATORYCOMMISSION. WASHINGTON DC 20555-0001. ANDTO THE PAPERWORK REDUCTION PROJECT (3150-OlN). OFFICE OF MANAGEMENT AND BUDGET. WASHINGTON. DC 20503.

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YEAR SEQUENTIAL REYlSION NUMRER NUMBER LaSalle County Station Unit One 05000373 96 010 03 16 of 19 (If more space is requ' red, use additional copies of NRC Form 366A)(17) i i

l The TS definition of Operable includes the requirement for a normal and emergency i

electrical power source in order for a system, subsystem, train, component, or device to be Operable.

Therefore, the HPCS diesel generator was inoperable and the TS 3.8.1.1 or 3.8.1.2 actions as applicable should have been taken. The use of this i

clarification was in violation of the TS at the time. TS 3.8.1.1 was changed by l '

diesel generators when a diesel generator is removed from service for preplanned license amendments 72 (Unit 1) and 56 (Unit 2) to not require the starting of other maintenance or testing on March 15, 1990. This clarification was available for use from April 3, 1989, until January 28, 1992, when it was deleted.

The clarification was primarily intended for use with preplanned maintenance, which required deenergization of the AC and/or DC buses. The TS change approved March 15, 1990, i

did not require starting other Operable diesel generators for preplanned maintenance on a diesel generator. Although not allowed by TS 3.8.1.1 at that time, the consequences were minimal because:

a.

Electrical division 3, which includes the HPCS system and the HPCS diesel generator, is dedicated to only the HPCS system operability and HPCS was inoperable.

i l

i b.

The HPCS diesel generator was inoperable due to planned maintenance or components not associated with common mode failure of the other diesel l

generators.

6.

Clarification 10-89, approved on August 8, 1989, defined operability of the Automatic Depressurization System (ADS) as only requiring 138 psig in the accumulator associated with each ADS valve. TS SR 4.5.1.d.1 and 4.5.1.d.2.c) require the ADS accumulator backup compressed gas system low pressure alarm to have a channel functional test once per 31 days and a channel calibration once per 18 months as part of the surveillance requirements for assuring the operability of ADS.

The clarification allowed ADS to be Operable without the backup compressed gas system, which is in violation of TS 3.5.1.

The safety consequences were minimal, since the design function of ADS is met by the accumulator associated with each ADS valve. The backup compressed gas system was i

i designed primarily for long term cooling requirements as a backup to Shutdown Cooling, due to a common SDC suction line. The clarification was deleted on November 27, 1990, so the time it was in use was also limited and may not have been used, since the clarification was written as a result of a LaSalle Unit 2 shutdown l'

due to inoperable ADS accumulator low pressure alarms.

7.

Clarification 11-89, approved September 25, 1989, allowed changing modes with secondary containment isolation dampers inoperable provided the TS 3.6.5.2 actions for inoperable dampers were met and secondary containment was Operable per TS 3.6.5.1 and the definition of Secondary Containment Integrity was met.

The modes in question were special operating conditions, defined in footnote

of TS 3.6.5.2 as handling irradiated fuel in the secondary containment, Core Alterations, and

(

operations with a potential to drain the reactor vessel, which applied when a unit

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was not in Operational Conditions 1, 2, or 3.

This was in violation of TS 3.0.4 prior to license amendments 94 (Unit 1) and 78 (Unit 2), issued February 24, 1994, based on Generic Letter 87-09.

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'qNRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 l

(542)

EXPIRES 05/31/95 l

i ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMEN'S REGARDING BURDEN ESTIMATE TO 1 HE INFORMATION LICENSEE EVENT REPORT (LER)

AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR TEXT CONTINUATION REGULATORY COMMISSION. WASHINGTON. DC 20555-0001. ANDTO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF l

MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

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YEAR St@UENilAL KEVISION NUMBER NUMBER LaSalle County Station Unit One 05000373 96 010 03 17 of 19 I

(If Inore space is required, use additional copies of NRC Fonn 366A)(17)

The safety consequences for the mode changes are minimal, because based on the current TS 3.0.4, the mode changes related to special operational condition

would be allowed, based on compliance with TS 3.6.5.2 action statements, operation could l

continue in the special operational condition entered.

i l

Also, the clarification stated that the dampers were operable in condition

when either the emergency or normal power source was inoperable, based on exception regarding shutdown power sources given for TS 3.6.5.3 for Standby Gas Treatment System (SBGT).

During Operational Conditions 1, Run mode; 2, Startup; and 3; Hot Shutdown, TS 3.0.5 allows either the normal or emergency power source to be inoperable, however this does not apply in condition *. TS 3.6.5.2 did not have the exception that allowed either the normal or emergency power source to be inoperable in condition *, so this was also in violation of the TS 3.6.5.2.

The clarification was deleted on January 25, 1995, as a result of the above license amendments. During the time this clarification was approved for use, it may have been used one or a few times.

The safety consequences for operability in special operational condition

with either the normal or emergency power source inoperable is minimal, because the l

secondary containment isolation dampers fail closed on loss of power, which is the isolated position.

8.

Clarification 05-91 was approved April 20, 1991, due to a design change during the fourth refuel outage, which changed the logic for the Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT).

The logic change required a change to the TSs, but was not recognized as a TS change until near the end of the outage. The request for a license amendment to change TS 3/4.3.4.1, ATWS Recirculation Pump Trip System Instrumentation, was submitted to the NRC on June 4, 1991.

License amendments 79 (Unit 1) and 63 (Unit 2) were issued by the NRC on October, 8,

1991.

The clarification was conservative with respect to the new logic and the TS prior to the license amendments, and therefore the safety consequences were minimal. The TS was changed prior to the design change being completed on Unit 2.

i 9.

Clarification 07-91 was approved January 3, 1992, due to a problem identified with the surveillance requirements for the Intermediate Range Monitor (IRI!) scram and rod block functions and the Source Range Monitor (SRM) rod block functions per TS Tables 4.3.1.1-1 and 4.3.6-1.

These trip functions are reactor mode switch dependent and/or power dependent. Therefore, when shutting down a unit, prior to changing modes from Operational Condition 1 to Operational Condition 2 when the IRMs were required to be Operable, TS 4.0.4 requires that surveillances have been performed within the stated surveillance interval, or as otherwise specified. Trips are bypassed in Run Mode and/or by power within the Startup mode and thus the Channel Functional Tests could not be completed until after the mode change or the trips were automatically unbypassed. The clarification allowed completion of the l

surveillances as much as possible with the trips still bypassed, then change modes and complete the surveillances as soon as practicable once the mode or condition in which the surveillance is required is entered. The stabilization of the plant was l

to be considered first and the surveillances completed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of entering j

the required condition, i

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%NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150 0104 (5-92)

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A TS change was requested to allow mode or special condition change without current Channel Functional Tests as part of the IRMs and SRMs as part of another change submitted to the NRC on December 14, 1994, and approved by the NRC on August 2, 1995.

Prior to the license amendments, the mode change was in violation of TS 3.0.4.

I Beginning in mid 1993, unit shutdown was changed to shutdown with a manual scram from about 10 percent of rated thermal power, just after taking the main turbine offline. This bypassed the startup mode during the shutdowns since then, so that the IRM and SRM surveillances were not required to be performed. The number of normal unit shutdowns that manually inserted the control rods that involved the unit being in the Startup mode during shutdown was less than 12.

Other shutdowns were by scram.

The safety consequences were minimal, because other scram and rod blocks were operable, with current surveillances in Run and Startup modes, and the portions of the channels ~not functionally tested for the IRMs and SRMs were limited.

Also, when the surveillances were performed, they were completed satisfactorily, so the instruments would have functioned properly if trip setpoints had been exceeded.

10.

Clarification 01-87, pages 32a through 32c, was approved March 23, 1987, to identify what instrumentation satisfied the requirements for the gaseous effluent monitoring instrumentation for both normal range and accident range by the instrument equipment piece number (EPN).

For the Main Condenser Off Gas Treatment System Effluent Monitoring System, there are two monitors or subsystems, A and B.

The clarification indicated that either A or B would satisfy the requirements of TS 3.3.7.11, Radioactive Gaseous Effluent Monitoring Instrumentation, which only required one channel to be Operable. However, this instrumentation is required to cause automatic isolation of the pathway and required a trip of both channels to cause an isolation. Therefore, if one channel was inoperable, the inoperable channel would have to be tripped to retain trip capability. The TS did not require both channels to be Operable, but did require automatic trip capability.

l License amendments 85 (Unit 1) and 69 (Unit 2) relocated the Rad Effluent TS to the Offsite Dose Calculation Manual (ODCM) were issued September 1, 1992. This relocated the associated TS 3.3.7.11 to the ODrM.

The ODCM was revised to incorporate this clarification and this part was corrected to indicate that both channels were required to trip to isolate the pathway. The clarification was deleted February 6, 1996, after the changes to the ODCM were completed.

The consequences were minimal, because the instrumentation was reliable and surveillances were conducted in accordance with the TS and later, the ODCM.

11.

Clarification 12-92 interpreted the term "other specified condition" in TS 3.0.4 regarding mode changes for Shutdown Cooling TS 3.9.11.1 and 3.9.11.2.

TS 3.9.11.1 is applicable in Operational Condition 5 when irradiated fuel is in the reactor vessel and the water level is greater than or equal to 22 feet above the top of the reactor pressure vessel flange. TS 3.9.11.2 is applicable in Operational Condition 5 when irradiated fuel is in the reactor vessel and the water level is less than 22 feet above the top of the reactor pressure vessel flange. The only difference between the specifications is the reactor water level.

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The clarification interpreted 'other specified condition" as only applying to conditions not defined in Table 1.2 of the TS.

As a result., an applicability statement which is limited to a portion of an operational Condition defined in Table 1.2 was interpreted not to be an 'other specified condition".

Therefore, with a shutdown cooling loop inoperable and the action statements met for the LCO being I

entered, reactor water level could be changed. TS 3.0.4 did not literally support this interpretation prior to the approval of the license amendment changing the 3.0 and 4.0 TS per Generic Letter 87-09, " Sections 3.0 and 4.0 of the Standard Technical Specifications (STS) on the Applicability of Limiting Conditions for Operation and Surveillance Requirements". License amendments 94 (Unit 1) and 78 (Unit 2) were issued by the NRC February 24, 1994, making the clarification unnecessary. The clarification was deleted based on the approval these amendments on January 25, 1995.

The safety consequences were minimal, because although not allowed by the TS at the time, the clarification enforced compliance with the action statements for the LCO being entered and " operation" could continue indefinitely in the LCO.

I

05000373/LER-1996-010, :on 960829,violations of TS & Design Basis Occurred.Caused by Inadequate Standards for TS Clarification.Deleted Clarifications in Question on 960829 & Revised One Clarification

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CAUSE

DESCRIPTION OF EVENT

CAUSE OF EVENT

ASSESSMENT OF SAFETY CONSEQUENCES

CORRECTIVE ACTIONS

PREVIOUS OCCURRENCES

COMPONENT FAILURE DATA

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05000373/LER-1996-010, :on 960829,violations of TS & Design Basis Occurred.Caused by Inadequate Standards for TS Clarification.Deleted Clarifications in Question on 960829 & Revised One Clarification

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CAUSE

DESCRIPTION OF EVENT

CAUSE OF EVENT

ASSESSMENT OF SAFETY CONSEQUENCES

CORRECTIVE ACTIONS

PREVIOUS OCCURRENCES

COMPONENT FAILURE DATA

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