ML15238A572
| ML15238A572 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/21/1982 |
| From: | Wagner P Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 8205060554 | |
| Download: ML15238A572 (9) | |
Text
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4 APR 21 1982 DISTRIBUTION:
ORB#4 Rdg Gray File DocketFile DEisenhut" OELD NRC PDR RIngram AEOD.
L PDR PWagner IE Dockets Nos. 50-269, 50-270 ACRS-10 and 50-287 HOrnstei rr EBlackwood.
LRubenstein AP R 27198220 Mr. William 0. Parker, Jr.
6
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a*as Vice President - Steam Production
=9 Duke Power Company P. 0. Box 33189 422 South Church Street Charlotte, North Carolina 28242
Dear Mr. Parker:
We have completed our review of the information you submitted by letter dated March 19, 1981 in response to our January 14, 1981 letter con cerning Induced Neutron Flux Errors.
The findings of our review are contained in the enclosed Evaluation; and we have concluded that no changes &re requi red.for present-operating parameters.
If you have any questions on this matter, please contact me.
Sincerely, Original signed b Philip C. Wagner, Project MPanager Operating Reactors Branch #4 Division of Licensing
Enclosure:
Evaluation cc w/enclosure:
See next page 8205060554 620421 PDR ADOCK 05000269 ORB#4:DL ORB#4:DL O#4:DL O FFIC Ek..
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82 4fi I..........
/.DATE PY USGPO: 1981-335-960 NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD C
Duke Power Company cc w/enclosure(s):
Mr. William L. Porter Duke Power Company P. 0. Box 33189 422 South Church Street Office of Intergovernmental Relations Charlotte, North Carolina 28242 116 West Jones Street Raleigh, North Carolina 27603 Oconee County Library
.501 West Southbroad Street Walhalla, South Carolina 29691 Honorable James M. Phinney County Supervisor of Oconee County Walhalla, South Carolina 29621 1Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission, Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Regional Radiation Representative EPA Region IV 345 Courtland Street, N.E.
Atlanta, Georgia 30308 William T. Orders
.Senior Resident Inspector U.S. Nuclear Regulatory Commission Route 2, Box 610 Seneca, South Carolina 29678 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodmont Avenue Bethesda, Maryland 20814 Manager, LIS NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 J. Michael McGarry, III, Esq.
DeBevoise & Liberman 1200 17th Street, N.W.
Washington, D. C. 20036
Eyaluation of the Induced Neutron Flux Error for Babcock & W11co' Reactors Introduction In October 1980 Babcock & Wilcox (B&W) indicated (Ref. 1) that studies recently performed had concluded that event induced errors in the neutron flux detector readings and thus effective flux trip levels could be larger for some events than those normally assumed in analyses. The staff responded, following conversations with B&W, by requiring information from utilities (Ref. 2).
The utilities with operating B&W reactors have responded (Ref. 3) and the response has been reviewed. The response and review are summarized here.
In brief the problems are (1) for some cooldown events the colder water in the downcomer region increases neutron flux attenuation thus potentially increasing the transient flux error on the excore nuclear instrumentation (NI) beyond the 2% normally used in analysis, and (2) for control rod ejection events the neutron flux distrfbution change resulting from the abnormal control rod pattern causes effective levels in the excore detectors to change (for a given core average level).
Both effects affect trip levels and potentially in an amount beyond that normally assumed.
All of the responding utilities, except Duke (Oconee reactors) presented a similar response, based on B&W calculations which were in turn primarily based on the calculations for the WPPSS-WNP 1/4 reactors which had initiated the problem concern.
Duke carried out their own calculations and presented therefore, a somewhat different viewpoint. All concluded that the result of potential flux error increases were suitably bounded within the existing
-2 operating parameters of-their respective reactors. In the following discussion the two presentations will be referred to as the B&W and Duke analyses.
Both analyses were used in forming a judgment for the review.
Evaluation Based on the WPPSS study B&W. concluded that a limiting maximum overcooling event, among the small steamline break, feedwater and turbine bypass events, was a turbine bypass with peak inlet (and downcomer) temperature reduced by 16*F. They concluded that larger steamline breaks would be terminated by a building pressure or variable low pressure trip. Duke studied (analyzed) several overcooling events, including turbine bypass failure with ICS failure, and also studied the larger steamline breaks assumming a high flux trip was required.
Normally B&W has used a 2% transient flux error. This, along with other assumed errors and a trip setpoint of 105.5% of full power gives a trip in analyses of 112%. Based on ANISN calculations (from the WPPSS study) to translate downco;ner temperature changeso UANI the maximum transient inlet temperature reduction of about 16*F corresponds to 13% ANI, giving an effective trip point of 123%. Duke examined data from a number of test programs relating temperature and flux readings. Based on these te'sts they developed a relationship (linear with temperature) between inlet temperature and ANI (at a 95% confidence level).
It would provide a 12% ANI at 16*F. For~much of their analysis, however, they used a 1%6NI/1*F factor (16%ANI at 16*F).
-3.-.
Using the calculated ANI vs inlet temperature relationship, B&W developed, for each reactor, at its minimum pressure (trip setpoint) a (graphical) relationship between reactor power, outlet temperature, trip lines (high flux With error and variable low pressure - outlet temperature) and thus regions protected by the reactor protection system.
(This is best described in the Davis-Besse submittal).
They superimposed on this DHBR values calculated using design power distributions.
The results, which of course take advantage of the improved DNBR value at the lower inlet temperature conditions, demonstrate that DNBR limits (both 1.30 and 1.43 which includes a 10.2% rod bowing penalty) fall within the protected region for overcooling conditions out to, and beyond, 16*F overcooling.
Power distribution calculation for 125% full power conditions were also done to check perturbations in distributions at these limiting conditions.
These were also used to demonstrate margin to DB and center fuelmelt (CFM) limits.
Duke performed plant specific analyses for each overcooling transient, including the turbine bypass event (also giving the maxinum overcool,ing as above) and the larger steamline breaks accidents (assuming a high flux trip i srequired).
They used 1% NI/F to identify maximum (non trip) power levels.(giving about 11% ANI for the turbine bypass) and assumed ICS failures to maximize overcooling and analyzed for DINB'using design peaking factors.
They found that DNB and CFM limits were not exceeded,--even without the reduction which would have been provided by a lower trip level which would occur using the derived ANI -
temperature error rather than 1% ANI/*F.
-4 The review of all of the submittals has lead to the conclusion that the magnitude and extent of the effect and its consequences during events of interest have been suitably examined. - The B&W calculations -and the Duke measurements complement each other on the-magnitude of ANI vs temperature as do the complementary calculations.for the magnitude of temperature decrease to -be considered during -maximum events.- Using this information the protection system will be able to provide protection before exceeding liimiits on DNB and CFM. However, all future submittals which require analysis-of.
overcooling events by B&W reactors should include in the analysis and presenta tion an equivalent of the information involved in the present submittals and the use of the penalties resulting from inlet cooling similar to these unless new values are justifie'd.
The other event involving a potential indication error for the flux. signal, which-in turn is involved-in terminating the-event. by a trip signal, is the rod-ejection accident.
In this case the error-arises from the change in power distribution-caused by the-ejected rod.making the effective power.
level as seen by the flux detector. different from the average used in (point kinetics) analyses.
The problem, as related to trip, would only exist for small worth rods (neighborhood of. 0.2% Ak or less) since the rise in flux level is too large-to significantly affect trip occurrence and timing for larger reactivity insertions.
Since the B&W "zero power" event -analyses normally involve high pressure trips rather than high flux trip for smaller rod worths, the problem is only relevant to the full power analyses which are normally analyzed as tripping on high flux.
-5 The B&W submittals argued on the basis of engineering judgment that, if heattransfer out of the fuel pin during the transient were included in the ejection analysis (as.has not been the ca-se in past submittals), the power and peaking increases for the range of reactivity insertion that might not cause flux trips would not result in peak enthalpies exceeding limits (280'
-cal/gm).i Duke presented results -of calculations of flux errors resulting from a number-of rod configurations, providing a basis for a correlation of error with rod worth, and also presented typical power histories as a function of rod worth. From these-it can'be concluded that there would be a high flux trip for a rod worth above about 0.1%2Ak at a trip level of about 120%
(rather than the usually assuned 112%).
For rods under this level there.
might-not be a fluxstrip, however, power-levels.and peaking factors associated with'these rod worths-are sufficientlyllow that the limit for the event (280 cal/gm) is not approached; The initial transient is minor and the quasi steady state is similar to that of the single rod withdrawal event. The latter is described in the Midland SAR where it is indicated, in an analysis with heat transfer, that 280 cal/gm is not approached (nor is DNB reached) for even larger rod worths than are involved here (e.g., greater than 0.3% Ak).
The review of the submittals.has lead to the conclusion that the flux error associated with the changed power distribution for rod ejection does not significantly affect the trip function for the larger rod worth-events and that the consequences for the smaller worth events are not of a magnitude to approach limits when considering the heat transfer that occurs.
Sum-mary and Conclusions The effective neutron flux trip level in B&W reactors may be raised above that normally used in analyses because of increased flux attenuation in the downcomer in cooldown events and because of power (flux) distribution changes in the rod ejection event. However, analyses of extreme cooldown events requiring high flux trip indicate that sufficient margin exists in the trip levels, as augmented by the improvement in DNBR provided by the cooldown, that limits on DNB and CFM are not exceeded in operating reactors. The review of this analysis has resulted in agreement with this conclusion for.
operating reactors. However, all future analyses of these events. for B&W reactors should include in the effective trip level for cooldown events a suitable flux error term of a magnitude as discussed in this review, e.g.
13% ANI for a 16*F cooldown,,)r as specifically derived for the reactor as has been done by Duke. For the rod ejection event the analysis of the increased error indicates that the only events which may be significantly affected are those with smaller rod worths for which the consequences are below limits even without a.
high flux trip.
The review-has concluded that no changes are needed in operating parameters for currently operating reactors because of this error.
m9 7
References
- 1. Letter from James Taylor (8&W) to Victor Stello (NRC), October 29, 1980:
"Results of Recent Induced Flux Error Investigations."
- 2. NRC letter to Duke Power Company dated January 14, 1981.
.3. Letters from the following utilities on the indicated dates to the NRC, Operating Reactor Branch 4.
Toledo Edison, March 18, 1981 Duke Power Co., March 19, 1981 Sacramento Municipal Utility District, March 20, 1981.
Metropolitan Edison Co., September 29, 1981.
Florida Power Corp., March 20, 1981 Arkansas Power & Light Co., January 30, 1981.