ML19332F184

Application for Amend to License NPF-42,revising Tech Specs 3.4.1.4.2,3.5.4,4.9.8.1 & 4.9.8.2 Re RHR Flow & Safety Injection Pump Availability

ML19332F184
Person / Time
Site:	Wolf Creek
Issue date:	11/30/1989
From:	Withers B WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19332F185	List: ML19332F184 ML19332F241
References
WM-89-0264, WM-89-264, NUDOCS 8912130509
Download: ML19332F184 (11)

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,f 4 9CW T W$LF CREEK NUCLEAR OPERATING CORPORATION -

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. November 30, 1989J

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.VM 89-0264 4, '

U. S. Nuclear Regulatory Coaunission ATTN: , Document Control Desk;

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Washington, D. C. 20555. .

Subjects.--Docket No.150-482: Revision to Technical Specifications.

• 3.4.1~.4.2,.3.5.4, 4.9.8.1, and 4.9.8.2 - Residual Heat

! Removal Flow Rate and Safety Injection Pump Availability: s M Gentlemen:-

Tlie~ purpose: of- this letter is-to' transmit ~an application for amendment:to Facility . Operating 1.icense No. . NPF-42 for, Wolf Creek Generating Stationt

(WCGS),:- _ ' Unit ' ~ No . -- /1.x This-license _ amendment request proposes revisingt Technical Specification 3.4.1.4.'2 to change a note ate the bottom? of page J3/4.4-6 to. add the additional criterion of ensuring the reactor vessel water: 1 ilevel is1above the vessel flange before the-running Residual.- Heat Removal (RHR) ' pump can-be stopped while in. Mode 5 with the, rector coolant loops not '

1 filled.' "In addition,. .this amendment request proposes revising Technical

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' Specifications-4.9.8.1 and:4.9.8.2 to decrease the; required. flow rate of'the W runningL RHRipump. _ Finally, this amendment' request proposes revising the

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Technical- _ Specification ;3.5.4 mode-' applicability. to have 'the' Safety y :Injec.tlon Pumps immediately available to>the' Operators should RHR-cooling be l, ~ lost ~ when' the Reactori Coolant' level: is below the vessel' flange'.

[ Additionally technical ; specification changes related to the removal of the l RHRLautoclosure interlock.are planned for' future submittal by Wolf Creek  ;

'NucleariCperation Corporation. These changes will further enhance the safe l operation of WCGS during' shutdown conditions.'

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, h'h. ALcomplete. Safety . Evaluation and No Significant Hazards Consideration- '!

determination are provided' as Attachments I and II, respectively. The .;

proposed changes to .the Technical' Specifications are provided as  :

H pgg (Attachment'III. Wolf Creek Nuclear Operating Corporation requests that this  !

b8 Technical: Specification amendment request be approved by March 1, 1990 to '

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I facilitate itsuse during the fourth refueling outage at WCGS scheduled to begin on March 15,,1990.

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In accordance with 10 CFR 50.91, a copy of this application, with L, y attachments, is being provided to the designated Kansas State official.

L 'g m Thio proposed revision to the WCGS Tebhnical Specifications will be fully implemented within 30 days of formal Nuclear Regulatory Commission approval.

OD P.O. Box 411/ Burangton, KS 86830 / Phone: (316) 364-8831

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R' ;WH 89-0264 Page,2 of 2'  !

. November 30, 1989-lf- ,

R _If ;you' have ,any' questions concerning this' matter, pleaseicentact me-or' 7

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Mr.y.0'. L. Maynard of my.: staff.

Very truly.yours, y

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'Bart D. Withers President and

• ' Chief Executive officer -l l- BDW/aem .1 i ,,

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Attachments:' I - Safety Evaluation  !

II . Addressing ~The Standarde In 10~CFR'50.92 III - Proposed Technical Specification Changes  :

cca 'G. W. Allen (KDHE), w/a

l. .B. L. Bartlett (NRC), w/a

! . E.-. J. Holler . (NRC) , w/a

'R. D.' Martin (NRC)..w/a 'y

.D. V. Pickett.(NRC), w/a L .

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, STATE OF. KANSAS' ) <

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i Bart D. Wit'hers, of lawful age. - being first duly sworn upon oath says that c- _helis President and Chief Executive Officer of Wolf Creek Nuclear Operating Corporations. that he has read the foregoing document and knows the content thereof that he has executed that same for and on- behalf of said -j

Corporation with full power 'and authority to do so and that the facts  !

therein- stated' are' true and correct- to the best of his knowledge, information'and belief.

By Bart b. Withers President and Chief Executive Officer I

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. SUBSCRIBED and sworn to before me this 3 o day of M M 1989.

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ATTAcalENT I SAFETY EVALUATION-o l ?

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- Attachment I to WH 89-0264 i Page 'l of 4 q SAFE N EVALUATION 1

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Description of Channes ,

n Technical Specifications 4.9.8.1 and 4.9.8.2 currently requires ' a flow of greater than 2800 gpm from at least one Residual Heat' Removal (RHR) pump.

Technical Specification 3.5.4 requires.both safety injection pumps to be inoperable, when the plant is in H0DE 5 or in AODE 6 with the reactor vessel ;l head on. The note at the bottom'of the page for LCO 3.4.1.4.2 allows 'the )

running RHR pump to be-stopped for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, while-in MODE 5 with the-reactor coolant loops not filled, provided nothing is done that would allow

-a dilution and core outlet . temperature remains at- least 100F- below saturation temperature.  ;

A The proposed changes to 4.9.8.1 and 4.9.8.2, decreases the required flow rate of the running RHR pump from t 2800 gpm to t_1000'gpm and: sufficient to y maintain the RCS temperature at 5 1400F. The proposed change to '3.5.4- H changes th,e APPLICABILITY from MODE 5 and MODE 6 with'the reactor vessel j head on, to MODE 5 and MODE 6 (with the reactor vessel head on)_with water 1 .'

level 'above the reactor vessel flange. The proposed change to the note at the_ bottom of the page-to LCO 3.4.1.4.2 adds the additional criteria of j ensuring the reactor' vessel water level is above the vess'el flange before- 1

-the running RHR pump can be stopped while in HODE 5 with'the reactor coolant loops not filled. .The' Bases to 3/4.5.4 and 3/4.9.8 are alsor being changed

. to-reflect these proposed changes.

,Baskaround NRC Generic Letter 88-17 reco:mmended several program enhancements that each nuclear unit should consider. Among these recommendations, _ was the following

" Technical Specifications that restrict or limit the safety benefit of the actions ' identified- in this letter should be identified and appropriate changes should be submitted."

Wolf Creek Nuclear Operating Corporation (WCNOC) has identified these changes as enhancements to the present Wolf Creek Generating Station (VCGS)

Technical Specifications.

The flow requirements for 3/4.9.8.1 and 3/4.9.8.2 ensures that; (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the Reactor Vessel below 1400 F as required during the refueling mode,- and (2) sufficient coolant circulation is maintained through the core ,

to minimize the effect.of a boron dilution incident and prevent boron I stratification.

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Attachment I to WM 89-0264 Page 2~of 4 The requirement'to have no operable Safety Injection Pumps in MODE S or 6 with. the reactor vessel head on, provides assurance that a mass addition.

pressure transient can be relieved by the operation of a single Power Operated Relief ~' Valve (PORV) or RHR suction relief valve. The proposed-change to 3.5.4 will allow the safety injection pumps to be' immediately available to the operators, should RHR cooling be lost. 'when the reactor coolant level is below the vessel flange. As technical specifications are currently written, both safety injection pumps must be safety tagged out of servi'ce with their breakers racked down in MODES -5 and 6 whenever the reactor vessel head is in' place. For the operators to use a safety injection pump after a loss of.RHR they must remove the safety tags and rack up the breaker prior to running the pump. It is estimated that it will take approximately 15 to 20 minutes to get a pump running. With the proposed change, the operators could have a safety injection pump running in approximately one minute.

The note at the bottom of the page for Technical Specification Limiting condition for Operation (LCO) 3.4.1.4.2, presently allows the running RHR pump to be stopped for up to I hour with the reactor coolant loops not filled in MODE 5. The two criteria that must be met when the RHR pump is i stopped prevent stratification of the coolant boron concentration and j prevent . core boiling. The proposed change to reduce the RHR minimum flow j rate in MODE 6, while maintaining decay heat removal requirements, reduces I potential problems due to vortexing and subsequent loss-of the RHR system. l

~The'added restriction'to the Technical Specification 3.4.1.4.2 footnote is '

proposed so that the operating RHR pump cannot be intentionally deenergized for one hour when reactor vessel water level is below the vessel flange.-

Thiv- will ensure that -the operating pump will not be intentionally deenergized during operation with the RCS loops partially filled.

Evaluation i A reduction of RHR flow in the shutdown cooling mode could potentially l affect the decay heat removal rate, boron stratification, and the boron 7 dilution accident analysis. The present Bases for Technical Specification 3/4.9.8 specifies that at least one RHR train be in operation to ensure

, thats (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140 0F as required during the refueling mode, and (2) sufficient coolant circulation is maintained through l the core to minimize the effect of a boron dilution incident and prevent l boron stratification.

The WCGS Technical Specifications do not contain provisions to minimize the potential for vortering and air entrainment in the RHR system which may occur during Reactor Coolant System (RCS) operation under certain conditions. Although the current requirement for a minimum 2800 gpm flowrate is conservative for decay heat removal and uniform boron mixing, it is clear that minimizing the potential for vortexing was not considered when i the minimum flow rate requirement was established. I l

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Attachment I to WM 89-0264

= Page= 3 of 4-Operation with the RCS partially drained in MODES 5 and 6 ise.necessary for required inspection and maintenance on RCS components such as reactor coolant pumps and steam 1,enerators.- A reduced RHR flow rate would provide a greater margin- to vortexing and preclude an inadvertent loss of decay heat removal capability due f.o air entrainment and cavitation of the RHR. pumps.

The likelihood of vortering at the auction of the RHR' pumps could be lowered by' reducing the RHR firsw rate.

As the time.after plant shutdown increases decay heat removal requirements:

for the RHR system flow are reduced since decay heat decreases'as a function of time after initial reactor shutdown. However, continuous de ay heat removal- capability is still required. Since the consequences'of a loss of decay heat removal during the RCS partial drain conditions can be severe, the RHR system flow rate should not only meet decay heat removal.and uniform coolant boron mixing requirements, but should also consider minimizing, the potential' for loss of RER flow by vortexing and air'entrainment at the RHR pump suction.

I Revisions to Technical Specifications 4.9.8.1 and 4.9.8.2 are proposed so  !!

that RER system flow rate may be reduced to the greater of the, flow required j to maintain the water in the' reactor vessel below. 1400F or' 1000 . gpn . R Although a reduced flow rate is not required for: prevention of-vortexing and air.entrainment for water levels above the . reactor vessel flange greater than 23 feet (Technicel Specification 4.9.8.1)..-this change is being made j for consistency to minimize operator confusion and human factors-problems. -1 i

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'The added restriction to the Technical Specification 3.4.1.4.2 footnote is

• proposed.so that the operating RER pump cannot be intentionally deenergized for one, hour when teactor vessel water level is below the vessel flange. i This will ensure that the operating pump will not be intentionally deenergized-during operation with the RCS loops partially filled.

An evaluation of boron dilution and boron stratification at RHR flow rates of 1000 gpm has been performed. The'results of this evaluation indicate  ;

that the minimum required RHR flow rate is established at 1000 gpm for these I two concerns. .

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Attachment I to WH 89-0264

-Page: 4 of 4 Allowing the Safety Injection Pumps to be operable affects the~ cold over-pressure. analysis. Whenever the reactor coolant level is below the vessel flange, there is a minimum empty volume of approximately ,17,000 gallons. These pumps would only be used in an emergency situation, while attempting to return an RHR pump to. service. This empty space will. allow

.the- operators- sufficient. room to provide cooling to the core and still not completely fill the reactor coolant system.

Based on the above discussions and the considerations- presented in-Attachment II, the proposed revision to the WCGS Technical Specifications does not increase the probability of occurrence or the consequences of an accident .or malfunction of equipment important to safety previously evaluated in the safety analysis reports or create a possibility for- an accident or malfunction of a'different. type than any previously' evaluated in the safety analysis report lor reduce the margin of safety as defined in the. basis for any technical specification. Therefore, the proposed revision.

doesinot adversely affect or endanger the health or safety of the general public'or involve a significant safety hazard.

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ATTACBIENT II-d > j' q i

ADDRESSING THE STaunantut IN 10 'CFR'50.92 -- ;

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n- 5 Attachment II to WM 89-0264 page~ 1 of 2 l

anmen8ING THE STAlmanns IN 10 CFR 50.92 The' proposed changes revise Technical Specification 3.4.1.4.2 to change a note at1 the bottom of page 3/4.4-6 to add the additional' criterion of ensuring the reactor vessel water level is above the vessel flange before .i the running Residual Heat Removal (RHR) pump can be stopped in MODE 5 with l the reactor coolant loops not filled. In. addition, the proposed changes'  !

would revise Technical Specifications 4.9.8.1 and 4.9.8.2 to decrease the j required flow rate of the running RHR pump. Finally,- the proposed: change would revise Technical Specification.3.5.4, changing the mode applicability so-that the safety injection pumps would be immediately available to the operators should RHR cooling be lost when the reactor coolant level is below the vessel flange. The following sections discuss the proposed _ changes i

- under the three 10 CFR 50.92 standards:

Standard 1 - Involve a Significant Increase in the Probability' or Consequences of an Accident Previously Evaluated These changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The change to the RHR

-flow rate reduces the probability of a loss of decay heat removal .due to. ,

- vortexing-and~ cavitation, while the probability of accidents analyzed in the

' Updated Safety Analysis Report (USAR) are unaffected. A reduction of RHR flow could potentially affect decay heat-removal rate, boron stratification. '

and the boron dilution accident analysis. An evaluation concluded that the consequences of these accidents are not increased. Allowing the-safety injection pumps to be operable affects the cold over-pressure analysis.

Whenever _the rector coolant level is below the vessel flange, there is'a minimum; empty volume of approximately 17,000 gallons. .These pumps would ,

only be used in an emergency. situation, while attempting to return an RHR

' pump to service. This empty space-will allow the operators sufficient room to provide cooling to the core and still not completely fill the reactor coolant system.

Standard 2 - Create the Possibility of a New or Different Kind of Accident Frosa any Accident Previously Evaluated These changes do not create the possibility of a new or different kind of g

accident from any accident previously evt.luated. The WCGS Technical Specifications do not contain provisions to minimize the potential for vortexing and air entrainment in the RHR system which may occur during RCS operation under certain conditions. Although the current requirements for a minimwn 2800 gpm flow rate is conservative for decay heat removal and uniform boron mixing, it is clear that minimizing the potential for vortexing was not considered when the minimwn flow rate requirement was established.

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L Attachment II to WM 89-0264

', Page 2 of 2-1 0 -Revisions to Technical Specifications 4.9.8.1 and 4.9.8.2 are proposed so

.that RHR system. flow rate may be reduced to the greater of the flow required to maintain the water in the reactor vessel. below 140 0 F or 1000 gpm.

Although a reduced flow rate is not required for prevention of vo'!texing and-

- air'entrainment for water level above the reactor flange greater than 23 feet (Technical' Specification 4.9.8.1), this change is being made for -l consistency to minimize operator confusion _and human factors problems. -l Allowing the Safety Injection Pumps to be operable gives the operators added flexibility in providing cooling to the core whenever- RHR cis lost. The requirement that the reactor coolant 1cvel be below the vessel flange prior to making the pumps operable.- provides a sufficient empty volume to not 1 completely fill the reactor coolant system, i Standard 3 - Involve a Significant Reduction in a Margin of Safety 1

These changes do not involve a significant reduction in a margin or safety.

Operation with the RCS partially drained in MODES 5 and 6 is necessary for I required inspection and maintenance on RCS components such as reactor coolant pumps and steam generators. A reduced RHR flow rate would provide a greater margin sto vortexing and preclude an-inadvertent loss of decay heat removal capability due to air entrainment and cavitation _of_the RHR. pumps.

- The. likelihood of vortexing at the suction of the RHR pumps could be lowered by reducing.the RHR flow rate.

The proposed minimum flow rate of 1000 gpm and ensures a sufficient. flow rate as assumed in the boron dilution and boron stratification evaluations.

Having reactor coolant level below the vessel flange prior to making tha safety injection pumps operable provides sufficient empty volume in the ,

reactor coolant system, to where cold over-pressure concerns are negligible.

The' change that adds the restriction to the Technical Specification

. 3.4.1.4.2 footnote will ensure that the operating pump will not be

-intentionally deenergized during operation with the RCS loops partially filled.

, . Based on the above discussions and those presented in Attachment I, it has been detennined that the requested Technical Specification revisions do not involve a significant increase in the probability or consequences of an accident or other adverse condition over previous evaluations or create the possibility of a new or different kind of accident over previous evaluations; or involve a significant reduction in a margin of safety.

Therefore, the requested changes do not involve a significant hazards consideration.

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