ML20082M549
| 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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CORRECTED COPY TO CORRECT' PARAGRAPH 3 ON ATTACHMENT Entergt
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Operatlons 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 requirement for the flow biased simulated thermal pos -
(STP) 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
PDR ADOCK 03000416 jd p
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.
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Attachment to GNRO.91/00:49 I
Page 1 of 6 Introduction e
" %rence 1, dated April 46, 1991, autaitted a request by Grand G lf tally channel check for the flow biased simulated thermai power tc u
to modify V,
mentation in TS Table 4.3.1.1'1.
The change was requested to reduce rp i
~
itch approved the same change for the Clinton Powt The request was denied by the NkC on July j
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 reactor to limit maximum rector power so as to maintain the minimum em to scram the I
critical power ratio above the appropriate safety limit for events such!
loss of foodwater heating."
i t
On this basis, the NRC concludedt "The requested chang 1s to the TS are 4*nied because the del ti v
would permit an undetected and unanalyzed incr I
e on of the As will be discussed in detail below, the STP trio signal is gn
{
by the Grand Gulf reactor protection systen to maintain MCPR above th currently used l
limit for any event.
In fact, any increase in thk e safety analyzed for the applicable Grand Gulf event (i.e.,
scram setpoint is well v
has been repeatedly reviewed and approved by the NRC..; scram is assumed), and r
l Grand Gulf FSAR i
i briefly discuss the arrangement of the Grand Gulf FStR.Bef o
NSSS supplier (GE) for Grand Gulf's first cycle of opers, i
y the j
description of those Chapter 15 events which were A separate volume current c;rcle of operation.
For a particular or the htroduction includes instructions to the re jer to refer i
L current arialysis.
or the Chapter 15' also contains appendices which may u, date the Cycle 1 event
-descriptions.
l Appendices include r6 analyses for such initiatives as the Maximus Extended Operating Omain (MEOD).mplicable ev l
j Again, the i
Cycle 1 evont description in the appropriate Chapter 15 subsection will l
(
the reader to any. applicable appendices for more current info refer rmation.
W Grand Gulf has employed-this practice (i.e., CCSA and Chapter 15 app since the first FSAR update following Cycle 1.
endices)
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Attachment to GNRO.71/00143 Pago 2 of 6 Loss of FeedW*ar_Heatina (LFWH) Analyses the STP trip surveillance requirement and the subseque nge to historically it is the only safety analysis for which STP trip credit has b enial because taken.
e The analysis methodology for this event changed significantly during Cyc As originally licensed for Cycle 1, the LFVH analysis credited the STP minimize the calculated severity of the transient.
o An improved analytic mechodology, first employed during the MEOD analyses during Cycle 1 and th repeated for each reload, assumes no STP trip (or any trip for that matter) en 1.
Initial Cycle 1 LFWH Analysis The initial Jycle 1 LFWH analysis is discussed in Section 15.1 1 of the G Gulf FSAR.
rand With respect to the STP trip (also called the thermal power monitor trip)
Subsection 15 1.1.2.2 of the FSAR notes:
{
"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 would occur when the neutron flux exceeds the high APRM flux scrag set point.
than the high thermal power scram set point by approximate percent.
Therefore, the loss of feedwater heating transient would be more severe without the high thermal power trip scram design.
i to a higher operating CPR limit and reduce the flexibility of niant This would lead operation."
In other words, rather than revising the CPR limit which would have unduly restricted plant operation, Grand Gulf chose to credit the STP trip for Cycle 1 In some sense, the need to credit the STP trip was an artificial situation necessitated by the limitations imposed by tha GE computer codes and analysis methodology at that time.
Subsection 15.1.1.3.1 of the FSAR and its associated references discuss the mathematical modeling (including a poin' 'inetics core model) employed for the Initial Cycle 1 analysta of the LFWH sys.~.
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.
steady state power level and other core conditions before a Rather than a l
i 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 (including the STP trip) in the course of demonstrating acceptable CPR results l
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Attachment to GNRO 91/00149 Page 3 of 6 Some additional background concerning the change in LFWH analysis m may be helpful.
o o ogy A loss of foodwater heating event results in a decrease in feedwater t entering the reactor vessel.
the core (i.e., increases subcoolingThis lowers the temperature of the water enterin3 emperature and thereby causes the core average p)o,wer to increase. reduces the average co causes a void redistribution and a corresponding power redistributionThe change in su the not change is a total core power increase, not all areas of th While increase at the same rate.
This tends to flatten tne radial power distribution.
e core 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, the power increase) occur relatively slowly - on the seconds to minutes.
maintain a quasi-equilibrium in which the water temperature, order of neutron flux distribution maintain their steady state relationships.
The LFWH analysis presented in FSAR Section 15.1.1 for Cycle 1 evaluated event with a transient thermal hydraulics code using a point kinetics model to represent the core feedback 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 beginning and and of the event bound those throughout the transiant.
j performing the LFWH analysis with a three-dimensional quasi-steady state code Therefore, appropriate.
state core conditions) are determined based on the positive of the ccider feodwater, and the effect on CPR is calculated.
power redistribution effects are accounted for in the methodology, the excessive Since the radial core power overprediction and associated need to credit a trip is eliminated 3.
Maximum Extended Operating Domain (MEOD) - Cycle 1 During Cycle 1 GE conducted and NRC approved analyses to support Grand Gulf operation in the MEOD.
l A bounding analysis was performed using a standard DWR/6 plant.
l operating transients, including the LFWH event were examined.
All abnormal LFWH event, GE employed the revised steady state approach discussed above.
In analyzing the l
1 At the request of the NRC staff, a plant specific LFWH event was performed the GE three dimensional BWR Core Simulator (Reference 4) docu approved GESTAR Amendment (Reference 5).
by Reference 6 and approved by the NRC by letter dcted AugustThis analysis was (Reference 7).
15, 1986 G9108151/SNLICFLR
Attachment to GNRC-91/00149 Page 4 of 6 was shown to bei bounded by the generic analysis.The ana core and analysis nor the Grand Gulf-specific LWH analysis assumed credit for t or any other, trip.
Gulf FSAR (Revision 2) in December,The results of these analyses 1987.
e Grand 4.
Cycle 2 Reload and LWH Event Analysis Grand Gelf contracted with Advanced Nuclear Fuels (ANF) to provide for C cIn 2 and to conduct the relosd analyses necessary to obtain NRC f
reload fuel for the fuel cycle.
proval The Cycle 2 reload analyses for LWH utilized the ANT 3-D XMBWR stea cors simulator code which is similar to the GE code used in the MEOD analy the LW H event.
The ANF generic analysis of the LWH event is documented in s s of Reference 8.
At the request of the Staff, Grand Gulf subeitted by Reference 9 a plant specific analysis of the LFWH event utilizing the XTGBWR code.
and the resulting delta CPR.provided the steady state conditions of the core be This analysis No credit was taken for the STP trip.
ent The Staff's review and approval of the Cycle 2 analysis is documented in Reference 10.
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 steady state approach described above which did not i ame 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 reactor power so as to maintain the minimum critical power ratic above the
... to limit maximum appropriate safety limit for events such as loss of fSedwater heating."
G9108151/SNLICFLR
Attachment-to GNRO-91/00149 Page 5 of 6 This is a correct statoaant only for the initial portion of Cycle 1 at G Gulf, until o detail above.perstlom in the MEOD was approved by the Staff.
rand 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 required no limit on reactor power to esintain CPR above th limit.
ey The SER concluded that the requested TS change was "... denied because t deletion of the surveillance requirement... would permit an undetected and e
unanalyzed increase in the scram setpoint."
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"The analysis provided by Grand Gulf for MEOD and each
- 19::' ; t%: r ;12t: ' 112:: c ' 6ha.
cycle reload allow reactor #
power to reach a now, higher steady state level and de results.
STP (and any other) trip.
industry approach to analyzing the nood for the STP trip has bee complete and extensively reviewed by the NRC.
e, 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 (Reference 12)
Power Station (Reference 2).
and Clinton In preparing requestu for changes to Technical Specifications Gulf a practice to determine if steilar requests have been pre,viously mad it is Grand approved by the NRC.
In doing so, our intent is to ensure that we have e and NRC in the approval SERs. considered and addressed the safety issues rais conformed our request to include known issume raised on other dock as addressing Grand Gulf-specific concerns.
, as well l
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Attachmsnt to GNRO-91/00149 Pa9e 6 of 6 References 4
1.
4 Entergy Letter, GNRO-91/00043 dated April 26, 1991; " Reactor Pro System Instrumentation Surveillance Requirements."
on 2.
NRC Letter, John Hickman to Dale L. Holtzsher, dated January 31
, 1990.
3.
NRC Letter, Theodore R. Quay to William T. Cottle, dated July 10
, 1991.
4 J. A. Wooley, "Threa-Dimensional BWR Core Simulator," NEDO-20953 1977.
, January 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 15
, 1986.
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 Letter, Lester L. Kintner to Oliver D. Kingsley, Jr. dated October 24 1986.
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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