ML20095K693

Application for Amends to Licenses NPF-11 & NPF-18,deleting TS LCO 3.4.2,Action Statement B

ML20095K693
Person / Time
Site:	LaSalle
Issue date:	12/21/1995
From:	Benes G COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20095K694	List: ML20095K693 ML20095K697
References
NUDOCS 9512290184
Download: ML20095K693 (7)

Text

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] 3 Commonweahh Mson Osmpany i100 Opm Place Downers Grove. it. (d K i 5 December 21,1995 l

I U.S. Nuclear Regulatory Commission l Attn: Document Control Desk

! Washington, D.C. 20555  ;

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SUBJECT:

LaSalle County Nuclear Power Station Units 1 and 2  :

Request for Technical Specification Amendment l Facility Operating License NPF-11 and NPF-18 Deletion of LCO 3.4.2 (Safety / Relief Valves) Action Statement b

NRC Docket Nos. 50-373 and 50-374 i

j Pursuant to 10 CFR 50.90, Commonwealth Edison (Comed) proposes to

! amend Appendix A, Technical Specifications, of Facility Operating License NPF-11

and NPF-18. The proposed change deletes Technical Specification Limiting Condition for Operation (LCO) 3.4.2. (Safety / Relief Valves), Action Statement b. The Action i Statement requires placing the reactor mode switch in the Shutdown position, thus

manually scramming the reactor, if unable to close a stuck open safety / relief valve

(SRV) within two minutes, or if suppression pool average water temperature is 110 l degrees F or greater. The operator would still be required to manually scram the i

reactor if suppression pool average water temperature is 110 degrees F or greater in accordance with LCO 3.6.2.1 (Depressurization Systems), Action Statement b.1.

Also enclosed with this package are marked up Index pages XII for LaSalle

Units 1 and 2. These pages were originally included in the May 23,1995 G. Benes j letter to USNRC. This letter described Technical Specification Bases changes that l were performed by Comed pursuant to 10 CFR 50.59. From recent discussions with the LaSalle NRC Project Manager it has been determined that the Index pages for j these bases changes can not be changed by the 10 CFR 50.59 process, but instead j need to be issued as an amendment pursuant to 10 CFR 50.90. The Index page

, changes are editorial, as only page number references are changed. Therefore, j Comed proposes that the enclosed Index pages Xll for LaSalle Units 1 and 2 be

! changed pursuant to 10 CFR 50.90.

i i This proposed amendment request is subdivided as follows:

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1. Attachment A gives a description and safety analysis of the proposed changes in this amendment.

2. Attachment B includes a summary of the proposed changes and the marked up Technical Specifications pages for LaSalle Units 1 and 2, with the requested changes indicated.

3. Attachment C describes Comed's evaluation performed in accordance l E with 10CFR50.92(c), which confirms that no significant hazard j consideration is involved.

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4. Attachment D provides an Environmental Assessment Applicability Review per 10 CFR 51.21.

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USNRC (2) Dncomber 21,1995 This request for a Technical Specification Amendment has been reviewed and approved by Comed Senior Management, as well as On-Site and Off-Site Review in accordance with Commonwealth Edison procedures.

Comed believes this amendment is needed to support continued safe operation of the plant and should be classified a Priority 2 per the NRC Prioritization Process. If the two minute requirement to manually scram after a SRV becomes stuck open is not removed, the operator has to scram the reactor, thus challenging the Reactor Protection System, the reactor vessel, and other associated components and systems.

Therefore, Comed requests that this amendment be approved by the NRC within about six months, i.e., NRC approval by approximately June of 1996, with an implementation time of 60 days.

To the best of my knowledge and belief, the statements contained above are true and correct. In some respect these statements are not based on my personal knowledge, but obtained information furnished by other Commonwealth Edison employees, contractor employees, and consultants. Such information has been reviewed in accordance with company practice, and I believe it to be reliable.

Commonwealth Edison is notifying the State of Illinois of this application for amendment by transmitting a copy of this letter and its attachments to the designated state official.

Please direct any questions you may have concerning this submittal to this office.

Sincerely, f 'hl Gary . Benes Nuclear Licensing Administrator

,r======v Subscribed on this - and 7/Swor7to

1. beforedaymeof  :! OFFICIAL SEAL lh h ie( 1995.

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?? l Nojdry Fublic Attachments:

A. Description and Safety Analysis of the Proposed Changes B. Marked-Up Technical Specification Pages C. Evaluation of Significant Hazards Considerations D. Environmental Assessment Applicability Review cc: H. J. Miller - Regional Administrator, Region lli P. G. Brochman - Senior Resident inspector, LaSalle County Station M. D. Lynch - Project Manager, NRR Office of Nuclear Facility Safety - lDNS

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1 ATTACHMENT A l

DESCRIPTION AND SAFETY ANALYSIS OF THE PROPOSED CHANGES Descriotion of the Prooosed Change A proposed license amendment to the Technical Specifications for LaSalle County l Station Units 1 and 2. This amendment proposes to delete LCO 3.4.2. (Safety / Relief Valves), Action Statement b. The Action Statement requires placing the reactor mode switch in the Shutdown position, thus manually scramming the reactor, if unable to close a stuck open safety / relief valve (SRV) within two minutes, or if suppression pool average water temperature is 110 degrees F or greater. The operator would still be required to manually scram the reactor if suppression pool average water temperature is 110 degrees F or greater in accordance with LCO 3.6.2.1 (Depressurization Systems), Action Statement b.1.

Descriotion of the Current Ooerating License / Technical Soecification Reauirement LCO 3.4.2 (Safety Relief Valves), Action Statement b. currently requires the following:

With one or more safety / relief valves stuck open, provided that suppression pool average water temperature is less than 110 F, close the stuck-open relief valve (s); if unable to close the open valve (s) within 2 minutes or if suppression pool average water temperature is 110 F or greater, place the reactor mode switch in the Shutdown position.

LCO 3.6.2.1 (Depressurization Systems), Action Statement b.1 currently requires the following:

With the suppression chamber average water temperature greater than 110 F, place the reactor mode switch in the Shutdown position and operate at least one residual heat removalloop in the suppression pool cooling mode.

Bases for the Current Reauirement The LaSalle Technical Specification Bases for Section 3/4.4.2, Safety / Relief Valves includes the following:

The safety valve function of the safety / relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in A1 e

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accordance with the ASME Code. Analysis has shown that with the safety function of one of the eighteen safety / relief valves inoperable, the reactor pressure is limited to within ASME Ill allowable values for the worst case upset transient. Therefore, operation with any 17 SRVs capable of opening is allowable, although all installed SRVs must be closed and have position Indication available to ensure that the integrity of the primary coolant boundary is known to exist at all times.

Technical Specification Bases Section 3/4.6.2, Depressurization Systems, also includes the following in regards to SRVs:

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event of safety / relief valve inadvertently opens or sticks open. As a minimum this action shallinclude:

(1) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck open safety relief valve to assure mixing and uniformity of energy insertion to the pool.

Descriotion of the Need for Amendina the Technical Soecification i

in compliance with current LaSalle Technical Specifications, the operator must manually scram the reactor within two minutes of a SRV becoming stuck open. Two I minutes may not be long enough for an operator to take all the necessary mitigating actions for a stuck open SRV prior to manually scramming the reactor. However, as discussed in the upcoming section, " Bases for Amended Technical Specification Request", the reactor scram is only appropriate if the suppression pool average water temperature approaches its Technical Specifications limit of 110 degrees F.

It is estimated that a stuck open SRV at LaSalle results in a suppression pool average water temperature rise of about two degrees F for every minute the SRV is open.

Therefore, the maximum allowed time for the operator to take the necessary mitigating actions for a stuck open SRV event depends on the initial suppression pool average water temperature at the time the event occurs and the number of stuck open SRVs.

Thus the current two minute requirement to scram with a stuck open SRV would be overly conservative if a single SRV becomes stuck open when the suppression pool temperature is initially at 70 degrees F. The operator would have almost 20 minutes before the suppression pool average water temperature reaches 110 degrees F, but would shut down the reactor unnecessarily because of the requirement to manually scram the reactor within two minutes if unable to close the stuck open SRV. Initiating a manual scram after the SRV has been stuck open for two minutes should be avoided because it adds another unnecessary challenge to the reactor protection system (RPS), the reactor vessel, and the associated components.

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V This proposed change is consistent with the improved Technical Specifications (NUREG-1433/1434 for BWR-4/BWR-6, respectively).

Therefore, Comed requests a Technical Specification amendment to delete LCO

. 3.4.2. (Safety Relief Valves), Action Statement b.

Descriotion of the Amended Technical Soecification Reouirement Comed proposes to delete LCO 3.4.2 Action Statement b., and rename LCO 3.4.2 Action Statement c. to LCO Action Statement b., so that the revised Action Statement

b. states:

With one or more of the above safety / relief valve stem position indicators

. Inoperable, restore the inoperable stem position indicators to OPERABLE status l within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in l COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Comed also proposes that the last paragraph of Technical Specification Bases Section 3/4.6.2, be modified accordingly to state:

in addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety-relief valve inadvertently opens or sticks open. As a minimum this action shall include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown when suppression pool average water temperature is 110 F or greater, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-

. open safety relief valve to assure mixing and uniformity of energy insertion to the

. pool.

4 flases for the Amended Technical Soecification Reouest i

The design basis for SRVs is primarily to protect the reactor vessel from the overpressure condition, and a stuck open SRV or an inadvertent open SRV does not violate this design basis requirement. The opening of a SRV allows steam to be discharged into the suppression pool. The sudden increase in the rate of steam flow leaving the reactor vessel causes the reactor vessel coolant mass inventory to decrease. The pressure .*egulator senses the nuclear system pressure decrease and 4 closes the turbine control valve far enough to stabilize reactor vessel pressure at a

, slightly lower value, and reactor power settles at nearly initial power level. Minimum critical power ratio (MCPR) is essentially unchanged, safety margin unaffected and A3 I

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fuel barrier unchallenged. Thus there is no radiological consequence and this event is indeed a mild depressurization transient. The acceptable results from this analysis require no operator action to protect fuel or maintain radiological limits. ,

However, a stuck open or inadvertently open SRV during power operation heats up ,

the suppression pool. The operator must try to close the SRV in order to cease inserting reactor heat energy into the suppression pool. The design basis for the suppression pool requires that it should accommodate a total reactor blowdown event at all conditions. The upper limit of the suppression pool average water temperature to meet this requirement is 110 degrees F (Technical Specification 3.6.2.1 Action b.1),

at which time the reactor must be scrammed to limit the reactor blowdown energy to the suppression pool.

This proposed change is consistent with the improved Technical Specifications  !

(NUREG-1433/1434 for BWR-4/BWR-6, respectively).

The design bases for both the SRV and the suppression pool during a stuck open SRV event are satisfied by the requirement to manually scram if the suppression pool average water temperature is 110 degrees or greater. The requirement to manually scram the reactor within two minutes if unable to close the stuck open SRV(s) is not needed, and can therefore be removed with no safety impact.

Schedule j There are no specific schedule requirements associated wH this amendment proposal. Therefore, Comed requests that this amendment le approved by the NRC within about six months, i.e., NRC approval by approximately June of 1996, with an implomentation time of 60 days.  !

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ATTACHMENT B PROPOSED CHANGES TO THE LICENSE / TECHNICAL SPECIFICATIONS NPF-11 NPF-18 XII Xll 3/4 4-5 3/4 4-6 3/4 6-16* 3/4 6-19*

B 3/4 4-2* B 3/4 4-1a*

B 3/4 6-3* B 3/4 6-3*

B 3/4 6-4 B 3/4 6-4 l

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4 There are no changes to these pages, they are provided for information only 4

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