ML20082M549

Corrected Ltr Appealing 910710 Technical Decision Denying Request for Amend to License NPF-29,modifying Surveillance Requirements for Flow Biased Simulated Thermal Power Trip Instrumentation.Supporting Info Encl

ML20082M549
Person / Time
Site:	Grand Gulf
Issue date:	08/15/1991
From:	Cottle W ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GNRO:91-00149, GNRO:91-149, NUDOCS 9109050248
Download: ML20082M549 (8)

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PAGE $ of 6' Ent*'sy operatione,ine.

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'C 4 T ( "4 August 15, 1991 U.S. Nuclear Regulatory Commission Mail Station P1-137 Washington, D.C. 20555 Attention: Document Control Desk

SUBJECT:

Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-29 Technical Appeal of the Denial of the Amendment Request For Reactor Protection System Instrumentation GNRO: 91/00149 Gentlemen:

By letter and SER dated July 10, 1991 the NRC denied a proposed aen to the Grand Gulf Technical Specifications which would have modifi*d a surveillance (STP) requirement for the flow biased simulated thermal pos -

trip instrumentation.

enclosed information.Entergy Operations wishes to appeal this technical d discuss this matter further.We would be happy to arrange an appeal m Yours truly, w r e-J0F/be Attachment cc: (See Next Page) hiodbO248910915 A p

PDR ADOCK 03000416 jd PDR G9108151/SNLICFLR pl

Augusc 15.-199L GNRO-91/00149 Page 2 of 3 cc: Mr. D. C. Hintz, w/4 Mr. J. Mathis, w/a Mr. R. B. McGehee, w/a Mr. N. S. Reynolds w/a Mr. H. L. Thomas, w,/a Mr. Stewart D. Ebneter, w/a Regional Administrator U.S. Nuclear Regulatory Commission Region-II 101-Marietta St., N.W., Suite 2900 Atlanta, Georgia 30323 Mr. L. L. Kintner, Project Manager, w/a Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail.Stop 11021 Washington, D.C. 20555 t

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Attachment Page 1 of 6 to GNRO.91/00:49 I Introduction e

" %rence 1, dated April 46, 1991, tc autaitted a request by Grand G u lf to modify V,

~

tally channel mentation in TScheck for the flow biased simulated thermai Table 4.3.1.1'1. rp power (

i The change was requested to reduce itch approved the same change for the Clinton Powj .

t The request was denied by the NkC on July 10, 1991 via Reference 3 .

}

In the SER accompanying the denial of the Grand Gulf request

, the Staff noted:  !

"The STP trip signal is used by the reactor protection syst em to scram the I reactor to limit maximum rector power so as to maintain the minimum critical power ratio loss of foodwater above the appropriate safety limit for events such!

heating." '

i t

On this basis, the NRC concludedt v

"The requested chang 1s to the TS are 4*nied because the del ti e on of the would permit an undetected and unanalyzed incrI

{

As will be discussed in detail below, the STP trio signal is gnused Gulf reactor protection systen to maintaincurrently l

by limitthefor Grand any event. MCPRe safety above th -

In fact, any increase in thk v analyzed for the applicable Grand Gulf event (i.e., scram setpoint is well has been repeatedly reviewed and approved by the NRC..; scram is assumed), and r l i Grand Gulf FSAR  ;,

i briefly discuss the arrangement of the Grand Gulf o FStR.Bef NSSS supplier (GE) for Grand Gulf's first cycle y the A separate volume of opers,j i description current c;rcle of operation. of those Chapter 15 events which For a particular or the were !  ;

i L htroduction current arialysis. includes instructions to the re orjer the to refer  !'

l Chapter 15' also contains appendices which may u, date the Cycle 1 event

-descriptions.  !

j Appendices include r6 analyses for l such initiatives as the Maximus Extended Operating Omain Again, (MEOD).mplicable the evi l Cycle 1 evont description in the appropriate Chapter 15 subsection will (

! the reader to any. applicable appendices for more current info rmation.

refer W

Grand since theGulf has employed-first FSAR update following this Cycle practice 1. (i.e. , CCSA and endices) Chapter 15 app '

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, Attachment Pago 2 of 6to GNRO.71/00143 Loss of FeedW*ar_Heatina (LFWH) Analyses the STP trip surveillance requirement and thengesubseque to historically taken. it is the only safety analysis for whichenial STP because trip credit e

has b The analysis methodology for this event changed significantly during Cyc As originally minimize licensed the calculated for Cycle severity of the 1, the LFVH analysis creditedo the STP transient.

An improved analytic mechodology, first employed during the MEOD analyses during Cycle en 1 and th '

repeated for each reload, assumes no STP trip (or any trip for that matter) .

1.

Initial Cycle 1 LFWH Analysis TheFSAR.

Gulf initial Jycle 1 LFWH analysis is discussed in Section 15.1 1 of .

the G rand With respect to the STP trip (also called the thermal power monitor trip)

Subsection 15 1.1.2.2 of the FSAR notes: ,

l

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"The thermal power monitor sitigating the consequences (of this event.TPM) is the primary protection system Gulf plant design, reactor scram during the loss of transient scrag set would point. occur when the neutron flux exceeds the high APRM flux than the high thermal power scram set point by approximate percent.

Therefore, the loss of feedwater heating transient would be more i severe without the high thermal power trip scram design. This would lead to a higher operation." operating CPR limit and reduce the flexibility of niant In other words, rather than revising the CPR limit which would have unduly restricted In some sense, plant operation, Grand Gulf chose to credit the STP trip for Cycle 1 the need to credit the STP trip was an artificial situation .

necessitatedatby methodology thatthe limitations imposed by tha GE computer codes and analysis time.

Subsection 15.1.1.3.1 of the FSAR and its associated references discuss the mathematical Initial Cycle 1 analysta of (including modeling the LFWH sys.~. a poin' 'inetics core model) employed for the

! point kinetics models yield extremely conservative results for the LFWH event.

2.

Change in Analysis Methodology Subsequent to the initial Cycle 1 analyses for Grand Gulf, GE (and the industry) changed its approach to modeling and analyzing LFWH events.

l i

Rather than a steady state power level and other core conditions before a This approach, which was used on Grared Gulf for the MEOD end all subsequent enalyses discussed below, eliminated the need to credit any plant trip l (including the STP trip) in the course of demonstrating acceptable CPR results .

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Attachment Page 3 of 6to GNRO 91/00149 Some may additional be helpful. background concerning the change in LFWH o o ogy analysis m A

entering loss ofthefoodwater heating event results in a decrease in feedwater reactor vessel. emperature t

the core (i.e., increases subcoolingThis lowers the temperature of the water enterin3 and thereby causes the core average p)o,wer to increase. reduces the average cor causes a void redistribution and a corresponding power redistributionThe While change in su the not change is a total core power increase, not all areas of increase at the same rate.

th.

e core This tends to flatten tne radial power distribution.

Due to the stored energy and mixing of feedwater in the downcomer and lo plenum with the recirculation flow, the temperature reduction at the core i (and, therefore, seconds to minutes.the power increase) occur relatively slowly - onorder the of maintain a quasi-equilibrium in which the water temperature, neutron flux distribution maintain their steady state relationships.

The LFWH analysis presented in FSAR Section 15.1.1 for Cycle 1 evaluated event with represent theacore transient thermal feedback hydraulics code using a point kinetics model to mechanisms.

The point kinetics model is a redistribution that occurs in this event and must assum that bound all the statepoints in the event.

model results in a large overprediction of the LFWH core power increaseConseq .

sudden increases or decreases of important paramete j beginning and and of the event bound those throughout the transiant. Therefore, performing appropriate. the LFWH analysis with a three-dimensional quasi-steady state code state core conditions) are determined based on the positive of the ccider feodwater, and the effect on CPR is calculated.

Since the radial power redistribution effects are accounted for in the methodology, the excessive core power overprediction and associated need to credit a trip is eliminated .

3.

Maximum Extended Operating Domain (MEOD) - Cycle 1 During Cycle operation 1 GE conducted and NRC approved analyses to support Grand Gulf in the MEOD.

l l A bounding analysis was performed using a standard DWR/6 All plant.abnormal

operating transients, including the LFWH event were examined. In analyzing the l LFWH event, GE employed the revised steady state approach discussed above.

1 At the request of the NRC staff, a plant specific LFWH event was performed u the GE three dimensional BWR Core Simulator (Reference 4) docu approved GESTAR Amendment (Reference 5).

by Reference (Reference 7). 6 and approved by the NRC by letter dcted 15, AugustThis 1986 analysis was G9108151/SNLICFLR

Attachment Page 4 of 6 to GNRC-91/00149 was shown to bei bounded by the generic analysis.The core and ana analysis or any other, nortrip.

the Grand Gulf-specific LWH analysis assumed credit for t ,

Gulf FSAR (Revision 2) in December,The 1987.

results of these e Grandanalyses 4.

Cycle 2 Reload and LWH Event Analysis Grand f Gelf contracted with Advanced Nuclear Fuels (ANF) to provide forthe for C cIn fuel 2 and to conduct the relosd analyses necessary cycle. reload to fuelobtain NRC proval The Cycle 2 reload analyses for LWH utilized the ANT 3-D XMBWR stea cors the LWsimulator H event. code which is similar to the GE code used in the MEOD analy s s of Reference 8. The ANF generic analysis of the LWH event is documented in At the request of the Staff, Grand Gulf subeitted by Reference 9 a plant specific analysis of the LFWH event utilizing the XTGBWR code.

This analysis and the resulting delta CPR.provided the steady state conditions of ent the core be No credit was taken for the STP trip.

The Staff's Reference 10. review and approval of the Cycle 2 analysis is documented in

5. Subsequent Relocds steady state approach and NRC approved methodology.The ng a trip (or any trip) credited for achieving acceptable CPR resultsIn .

no instance was the STP employed a relatively new ANF computer code (CASM ame steady state approach described above which did not in trip.

".. . a complete new analysis was run for the LWH.

with the newly approved MICR08 URN-B/ ANT following the ab LWH was analyzed approved for Cycle 4, using an expanded GG1 data base."pproach previously NRC Conclusions in the Denial SER In denying the requested TS change to sodify the STP tri the NRC's SER indicates that the STP trip is necessary "p daily channel check

... to limit maximum reactor power so as to maintain the minimum critical power ratic above the appropriate safety limit for events such as loss of fSedwater heating."

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Attachment-to Page 5 of 6 GNRO-91/00149 This is a correct statoaant only for the initial portion of Cycle rand Gulf, until o 1 at G detail above.perstlom in the MEOD was approved by the Staff.

As discussed in function for the LFVN or other events.. et all subsequent times the STP trip pr In other words, the new LFWH analysis methodology limit. required no limit on reactor power to esintain CPR ey above th The SER concluded that the requested TS change was "... denied e because t deletion of the surveillance unanalyzed increase in the scram setpoint." requirement ... would permit an undetected and 55:  ::17 ::: ;;;rfd:d $7 C:::d C:!" '::

per:: t:  ::::h  :  :::, he;t:: tzed; :t MECT t: d :::S :y:!: ::!::d :

1  ::::!tc.

Th!: :;;::::t i: ; iv:! t t: 1:;;l : d i :x.:::;;; ::;;p!!:;;t!;::: CP"t-:

"The analysis provided by Grand Gulf for  :::19::'

MEOD ; t%:and r ;12t:

each' 112:: c ' 6ha. -

cycle reload allow reactor #

results.

power STP (and any toother) reach trip. a now, higher steady state level and de industry approach to analyzing the nood for the STP complete and extensively reviewed by the NRC. e, trip has bee Previous NRC Approvals for Sfeilar Requests To our knowledge, at least two other BWRs have received NRC approval to the daily channel check for the STP trip - Forei-2 (Referenceand 12)Clinton Power Station (Reference 2).

In preparing requestu for changes to Technical Specifications it is Grand Gulf a practice approved by the NRC. to determine if steilar requests have been pre,viously e and mad

' In doing so, our intent is to ensure that we have NRC in the approval SERs. considered and addressed the safety issues rais conformed our request to include known issume raised as addressing Grand Gulf-specific concerns.

on other dock

, as well l

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Attachmsnt to GNRO-91/00149 Pa9e 6 of 6 4 References 4

1.

Entergy Letter, GNRO-91/00043 dated April 26, 1991; "onReactor Pro System Instrumentation Surveillance Requirements."

2.

NRC Letter, John Hickman to Dale L. Holtzsher, dated January 31

, 1990.

3.

NRC Letter, Theodore R. Quay to William T. Cottle, dated , 1991.July 10 4

J.

1977. A. Wooley, "Threa-Dimensional BWR Core Simulator,", January NEDO-20953 5.

Reactor Fuel" (GESTAR) Ausust 1985 (GE Proprietary or

6. Entergy Letter, AECM-86/0174 Submittal." dated June 9, 1986; " Addendum to ME00 7.

NRC Letter, Lester L. Kintner to Oliver D. Kingsley dated August , 1986. 15 8.

Transient for Boiling Water Reactors", February 1986.XN

9. Entergy Letter, AECM-86/0273 Subalttal Additional Information (LOFWH, LOCA, Fuel Liftoff) " da .

10.

NRC 1986. Letter, Lester L. Kintner to Oliver D. Kingsley, Jr. dated October 24 11.

NRC Letter, Lester L. Kistner to William T. Cottle dated November 15

, 1990.

12.

NRC Letter, Theodore R. Quay to B. Ralph Sylvia, dated June 3

, 1988.

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