ML19323H077

Discusses Small Break LOCA ECCS Performance Results. Primarily Effect Function of Sys Characteristics Such as ECCS Flow Rates,Safety Injection Tank Actuation Pressures & Core Power Level.Supporting Info Encl

ML19323H077
Person / Time
Site:	Millstone
Issue date:	06/02/1980
From:	Counsil W NORTHEAST UTILITIES
To:	Clark R Office of Nuclear Reactor Regulation
References
B10005, TAC-11348, TAC-11561, TAC-12505, TAC-42846, NUDOCS 8006110307
Download: ML19323H077 (14)

Text

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June 2, 1980 Docket No. 50-336 B10005 Director of Nuclear Reactor Regulation Attn: Mr. Robert A. Clark, Chief Operating Reactors Branch #3 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

References:

(1) W. G. Counsil letter to R. Reid dated Fbrch 6,1980.

(2) D. C. Switzec letter to G. Lear dated October 12, 1977.

(3) W. G. Counsil letter to R. Reid dated March 30, 1979.

(4) W. G. Counsil letter to R. Reid dated March 22, 1979.

(5) K. L. Ferguson and R. M. Kemper, WCAP-9528, October, 1979.

(6) R. Reid letter to W. G. Counsil dated May 12, 1979.

Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Small Break LOCA ECCS Performance Results In Reference (1), Northeast Nuclear Energy Company (NNECO) provided the results of the non-LOCA Safety Analyses necessary to support Cycle 4 operation. As was noted in Reference (1), this reload will be the first instance in which Westinghouse will supply the fuel and safety analyses for a CE-NSSS. As requested by the NRC Staff on March 18, 1980, this submittal provides both the justification for continued applicability of the Cycle 3 Small Break LOCA analysis for Cycle 4, and the Millstone Unit No. 2 plant-specific Small Break LO A analysis for Cycle 4.

As a result of the incident at Three Mile Island (TMI), various improvements to the Small Dreak LOCA models are being developed by the fuel safety analyses vendors in response to NRC Staff requests. These model changes are currently in progress and will not be available for use in the 1980 reload effort. The Staff has stated that a review of the present Westing-house Small Break LOCA model of a CE-NSSS for Millstone Unit No. 2 without the changes resulting from TMI would not be undertaken because of the scheduled proximity of the review required of the new Small Break LOCA models. As a result of this concern and in accordance with the dis-cussions held between our respective Staffs on March 18, 1980, it is NNECO's purpose to demonstrate that the Cycle 3 Small Break LOCA analysis remains valid for Cycle 4 operation.

In support of this conclusion, an evaluation has been performed related to basic fuel design and operating parameters which could impact small break LOCA transient behavior.

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'l j' . The fuel clad. temperature excursion resulting in peak clad temperatures during small break LOCA's are produced by various phenomena depending on the break size. These eff ects are primarily a function of system characteristics such as ECCS flow rates, safety injection tank acteation pressures, and core power level. These parameters will change the ratio of boil-of f rate to emergency core cooling addition rate during the transient. The ratio will determine how long and to what depth the core will be uncovered. The above parameters are functionally independent 1 of the details of the fuel design.

Fuel design parameters, on the other hand, primarily impact the amount of stored energy in the fuel prior to the event. These effects are important in the determination of the peak clad temperature response to j large break LOCA's, for which the core dries out very quickly and most of the stored energy plus decay heat contributes to the clad temperature excursion. For small break LOCA's, however, the core remains covered for a much longer period of time (prior to uncovery), during which most of the initial stored energy is removed. Therefore, the clad heat up is mainly due to decay heat. Although the fuel clad gas conductance is input to the peak clad temperature calculation, the effect is of second order since the small break is a slower transient.

t j The Cycle 4 reload fuel has been designed to be hydraulically and mechanically identical to the present Millstone Unit No. 2 fuel. It has been compared J to the limiting NSSS vendor fuel, Batch B, in the ECCS analysis areas identified in References (2) and (3). NNECO has concluded that the Cycle 4 reload fuel is sufficiently similar to the present fuel in use at Millstone Unit No. 2 and that the Cycle 3 Small Break LOCA performance i results, Reference (4), will not be effected by the use of Westinghouse fuel in Millstone Unit No. 2.

For the convenience of the Staff, Table 1 lists various design and operating parameters for both the present Millstone Unit No. 2 fuel

-and the Cycle 4 load fuel.

The above conclusions are further supported by representative calculations g for the Westinghouse fuel reload as discussed in Reference (5). These calculations show similar trend behavior for the various break sizes as compared with the Millstone Unit No. 2 analysis docketed by Reference (4),

and approved by the Staff in Reference ( a). Furthermore, included as i Attachment-1 is a plant-specific Small Break LOCA analysis for the limit'ing break size, as determined in Reference (4). The results of this analysis

! compare favorably to those of Reference _ (4) with respect to peak clad j temperature, local-clad oxidation, and core wide clad oxidation.

. NNECO concludes that, based on. the attached plant-specific analysis, the generic analyses of . Reference- (5) and the continued applicability of the Cycle 3 analysis, that continued conformance to the criteria of 10CFR50.46 and Appendix K for small break 'IDCA's has 'been demonstrated.

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s This conclusion precludes the need for a Staff review of the Westinghouse Small Break LOCA model for Millstone Unit No. 2.

I We trust you find this information sufficient to concur in the above conclusions.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY i

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I W. G.'Counsil Senior Vice President

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i Attachment '

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WESTINGHOUSE PROPRIETARY CLASS 2-k TABLE 1 DESIGN AND OPERATING PARAMETERS i

Millstone 2 ~  !

! ' Fuel Assembly Westinghouse Reference Cycle. Batch B t

Assembly Envelope, inch 8.19 8.19 -i Assembly Pitch, inch 8.18 8.18 j .

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Lower Nozzle Blocked Area, % 64 64 Rod Array 14x14 14x14  !

I Thimble 0.D., inch 1.11 1.115 Rod 0.D., inch .440 .440 i Rod Pitch, inch .580 .580

Assembly

fl/De 3.90 3.90 I  ;

- Number of grids, 9 9 I

• Grid Blocked Area, % 20 22 Upper Nozzle Blocked Area, % 56 57 i *

' Fuel Rod and Pellet 0.440 0.440 Clad 0.D., inch Clad Thickness, inch 0.026- 0.026 1

Pellet diameter, inch: -0.3805 0.3795 Pellat length, inch 0.600 0.450' Pellet Density (% Theoretical) 95.0 94.75-95.0 Active Stack Length,-Cold, inch 136.7 136.7 Reactor Power Level . (MWt) 2,754 2,754 Number of Fuel Rods in Core 38,180 37,796 Technical Specification Peak .

Linear Heat Rate.(Kw/ft) 15.6 15.6 4

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DOCKET NO. 50-336 i

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MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 4

i 1 'SMALL BREAK LOCA ANALYSIS '

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LOSS OF COOLANT ACCIDENTS RESULTING FROM PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE B0UNDARY Intr 6 duction The Acceptance Criteria for LOCA analysis is describeo in 10CFR50.4b as follows:

1. The calculated fuel element peak clad temperature is Delow the requirement of 2200*F.

2. The amount of fuel element cladding that reacts chemically with water 'or steam does not exceed 1 percent of the total snount of Zircaloy in the reactor.

3. The clad temperature transient is terminateo at a time when tne core geometry is still amenable to cooling. The localized cladding oxi-dation limits of 17 percent are not exceeded ouring or after quenching.

4. The core remains amenable to cooling auring and after the break.

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5. The core temperature is reduced ano decay heat is removed for an extended period of time, as required by the long lived raoioactivity remaining in the core.

These criteria were established to provide significant margin in Emer-gency Core Cooling System (ECCS) performance following a LOCA.

1 Mathematical Model

.The requirements of an acceptable ECCS evaluation model are presenteo in Appendix K of 10CFR50.

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Small Break LOCA Evaluation Model The WFLASH program used in the analysis of the small Dreak LOCA is an extension of the FLASH-4 code [3 ] developed at the Westinghouse Bettis Atomic Power Laberatory. The WFLASH program permits a detailed spatial representation of the RCS.

The RCS is nodalized into volumes interconnected by flowpaths. The transient behavior of the system is determined from the governing conservation equations of mass, energy and momentum applieo through the system. A detailed description of WFLASH is given in Reference [4 ].

The use of WFLASH in the analysis involves, among other things, the representation of the reactor core as a heated control volume witn tne associated bubble rise model to permit a transient mixture neight cal-culation. The mul .aode capability of the program enables an explicit and detailed spatial representation of various system components. In particular it enables a proper calculation of the benavior of tne loop seal during a loss of coolant transient.

Clad thermal analyses are performeo with tne LOCTA-IV code [1] wnich uses the RCS pressure, fuel rod power history, steam flow past tne uncovered part of the core and mixture height history from tne WFLASH hydraulic calculations as input.

The small break analysis was performeo witn the version of the Westing-house ECCS Evaluation Mocel for Westinghouse fuel reloaos of CE plants (refer to References 1, 2, 4, 5, and 6).

i Small Break Results The calculated peak clad temperature resulting from a small break LOCA is less than that calculated for a large break. Based on the results of j Reference r 7] as generally confirmed in Reference [2] the limiting l small break has been found to be a 0.1 ft2 rupture of the RCS cold leg. The results of this analysis for the Westinghouse Fuel Reload of Millstone 2 are summarized in Tables 1 and 2.

Figures 1 through 3 present the principal parameters of interest for the small break ECCS analyses. The following transient parameters are presented:

1. RCS pressure.

2. Core mixture height.

3. Hot spot clad temperature.

The maxim;m calculated peak clad temperature for all small breaks ana-lyzed is 19780F. The analysis results are well below all Acceptance Criteria limits of 10CFR50.46 and are not limiting when compared to the

results presented for large breaks.

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l REFERENCES 1 Bordelon, F. M., et al., "LOCTA-IV Program: Loss of Coolant Tran-sient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary), June 1974.

2. Ferguson, K. L., ano Kenper, R. M., Addendum to ECCS Evaluation Mocel for Westinghouse Fuel Reloads of Comoustion Engineering NSdS",

October 1979.

3. Porsching, T. A., Marphy, J. H., Redfield, J. A. and Davis, V. C.,

" FLASH-4, A Fully Implicit FORTRAN-IV Program for the Digital Sins-lation of Transients in a Reactor Plant," WAPD-TM-84, Bettis Atomic ,

Power Laboratory, March 1969.

4. Esposito, V. J., Kesavan, K. and Maul, B. A., "WFLASH, A FORTRAN-IV

. Computer Program for Simulation of Transients in a Multi-Loop PWR,"

WCAP-8200, Revision 2 (Proprietary) ano WCAP-8261, Revision 1 (Non-Proprietary), July 1974.

5. Skwarek, R. J., Johnson, W. J., and Meyer, P. E., " Westinghouse Emergency Core Cooling System Small Break October 1975 Model,"

WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary), April 1977.

6. Skwarek, R., Meyer, P., Chismar, S., " Westinghouse Emergency Core Cooling System Small Break February 1978 Evaluation Model,"

WCAP-9223 (Non-Proprietary), WCAP-9222 (Proprietary), February 1978.

7. W. G. Counsil to R. Reid, Docket No. 50-336, March 22,1979.

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TABLE 1 SMALL BREAK

[ TIME SEQUEACE OF EVENTS i

0.1 Ft 2 Start 0.0 Reactor Trip Signal (Sec.) 15.2 Top of Core Uncovered (Sec.) 499 I Accumulator Injection Begins- (Sec.) 1313 PCT Occurs (Sec.) 1312 l

i Top of Core Covered (Sec.) 1319 ll h

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TABLE 2 SMALL BREAK 0.1 Ft 2 Results Peak Clad Temp. F 1978 Peak Clad Location, Ft. 11.2 Local Zr/H O 7.4 2 Rxn(max)%

Total Zr/H2 O Location.Ft., 10.6 Total Zr/H O <0.3 2 Rxn, %

Hot Rod Burst Time,sec 991.3 Hot Rod Burst Location,Ft. 10.6 Calculation Assumptions NSSS Power Mwt 102% of 2700 t

Peak Linear Power, kw/ft 15.6 SI Tank Actuation Pressure, psia 225

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