A06096, Forwards Addl Justification for 861020 Application to Amend License DPR-65 to Facilitate NRC Review of Small & Large Break Locas.Amend Incorporates Changes Relative to Revised pressure-temp Limit Curves

Forwards Addl Justification for 861020 Application to Amend License DPR-65 to Facilitate NRC Review of Small & Large Break Locas.Amend Incorporates Changes Relative to Revised pressure-temp Limit Curves

ML20214V293
Person / Time
Site:	Millstone
Issue date:	12/01/1986
From:	Mroczka E NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	Thadani A Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.31, TASK-TM A06096, A6096, B12348, TAC-63133, TAC-63198, NUDOCS 8612090610
Download: ML20214V293 (8)

Text

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UMO General Offices

• Selden Street, Berlin. Connecticut

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I:5 Sc co*~~ P.O. BOX 270 HARTFORD. CONNECTICUT 06141-0270 L t a !.l'f. 0"l'.",% 'c"o ,," (203) 665-5000 December 1,1986 Docket No. 50-336 A06096 B12348 Office of Nuclear Reactor Regulation Attn: Mr. Ashok C. Thadani, Director PWR Project Directorate #8 Division of PWR Licensing - B U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Additional Information {

Small Break LOCA Evaluations Pressure-Temperature Limit Curves In a letter dated October 20, 1986,(1) Northeast Nuclear Energy Company (NNECO) submitted to the NRC Staff additional information to facilitate the Staff review of both "small break" and "large break" LOCA reanalysjs for Millstone Unit No. 2 that were provided by letter dated August 29,1986.t2) In response to a subsequent telephone conversation with the Staff, NNECO is hereby providing the attached (Attachment 1) additional information as a supplement to that provided in Reference (1).

In a letter dated October 20, 1986,(3) NNECO proposed to amend its Operating License DPR-65 for Millstone Unit No. 2, pursuant to 10CFR50.90, by incorporating the changes therein relative to revised heat-up and cool-down curves. As discussed in a telephone conversation with the Staff, NNECO is hereby submitting additional justification relative to 10CFR50.92 for one of the changes proposed in Reference (3). This additional justification is included as Attachment 2 to this submittal. The proposed change described in Attachment 2 was implicitly addressed in Reference (3), but more explicit justification is hereby provided in response to the Staff request.

(1) 3. F. Opeka letter to A. C. Thadani, " Item II.K.3.31, NUREG-0737, Small Break LOCA Evaluaticns," dated October 20,1986.

(2) 3. F. Opeka letter to A. C. Thadani, " Item II.K.3.31, NUREG-0737, Small Break LOCA Evaluations," dated August 29,1986.

(3)3.F. Opeka letter to A.C. Thadani, " Proposed Revision to Technical Specifications, Pressure-Temperature Limit Curves," dated October 20,1986. o hhk P

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x 2-We trust that the information is responsive to your needs.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY A

E. Y M(o6zka Senigi/Vice President

.g Attachment STATE OF CONNECTICUT )

) ss. Berlin COUNTY OF HARTFORD )

Then personally appeared before me E. J. Mroczka, who being duly sworn, did state that he is Senior Vice President of Northeast Nuclear Energy Company, a Licensee herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Licensees herein and that the statements contained in said information are true and correct to the best of his knowledge and belief.

'7hd6xi Wjsh Notary Pup My Comnussion Dpires March 31, 1988

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Docket No. 50-336 A06096 B12348 Attachment l~

Millstone Nuclear Power Station, Unit No. 2 Additional Information item II.K.3.31, NUREG-0737 Small Break LOCA Evaluations December,1986

o Attachment 1

3. "The Staff is not convinced that the 4 inch cold leg pump discharge break is the worst case small break. It is noted that, prior to accumulator actuation, cladding temperature was continuously increasing. The brief accumulator actuation resulted in an approximate 2 foot level increase in the core mixture level which terminated the cladding temperature increase.

It appears that the worst case break would be a slightly smaller break which does not rely upon accumulator injection to terminate the transient.

Provide additional spectrum analyses to demonstrate that the worst case break has been identified."

Response

The choice of the break spectrum for Combustion Engineering (CE) designed plants has evolved from the Westinghouse experience with small break LOCA thermal-hydraulic transient phenomena and the detailed analysis calculations of the plant response. The transient response determines the limiting break size.

The physical phenomena governing the transient response of a CE plant are fundamentally the same as in a Westinghouse designed plant. The competing effects of the rate of RCS depressurization, as governed by the mass inventory depletion through the break with the decay heat removal processes, and the emergency core cooling system response, will determine the limiting break. The limiting small break size choice is determined from the evaluation of the Millstone Unit No. 2 transient analysis results.

10CFR50.46 requires the ECCS cooling performance shall be calculated in accordance with an acceptable evaluation model, and shall be calculated for a number of postulated LOCAs of different sizes, locations and other properties sufficient to provide assurance that the entire spectrum of postulated LOCAs is covered. Currently, for small break LOCAs, the spectrum of breaks to be analyzed per Westinghouse methodology consists of standard break sizes of 2,3, 4,6, and 8 inch diameter breaks.

Specifically in WCAP-10054, Addendum 1, a 3, 4 and 6 inch diameter pump discharge break was analyzed using Millstone Unit No. 2 as the CE plant. From these results it was shown that the peak clad temperature was bounded between a diameter of 3 and 6 inches, and a discrete break diameter of 4 inches produced the limiting peak clad temperature. It appears that the limiting break size would be the largest break in which the clad temperature excursion would be terminated by safety injection flow alone. However, recent Westinghouse small break LOCA analyses have shown that, even with accumulator injection, a larger break resulting in increased break flow and deeper core uncovery at high decay heat levels may be more limiting that slightly smaller break sizes without accumulator injection. This is particularly true for the Millstone Unit No. 2 analysis with the nonequilibrium NOTRUMP computer code. While an equilibrium model of accumulator injection during a small break LOCA may result in massive amounts of accumulator injection due to local depressurization, the NOTRUMP nonequilibrium model will realistically calculate the amount of accumulator injection appropriate to the local conditions. On the basis of the current methodology used by Westinghouse, a spectrum of break sizes was analyzed for Millstone 2 to assure the worst break size had been identified.

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The dominating factors in determining whether or not accumulator actuation will occur in a small break LOCA are the accumulator back-pressure and the pressure transient of the associated small break. For Millstone Unit No. 2, the accumulator back-pressure is 215 psia, so the determination of accumulator

' injection is important in terminating the clad heat-up. For instance, in WCAP-10054, Addendum 1, the 4 inch diameter pump discharge break does have

. accumulator actuation at 1,670 seconds with the peak clad temperature occurring at 1,810 seconds. On the other hand, in WCAP-10054, Addendum I, the 3 inch diameter pump discharge break does not have accumulator actuation with the peak clad temperature occurring at 2,275 seconds. From these two small break cases performed in WCAP-10054, Addendum 1, it is conceivable that there could be a break size between the 3 and 4 inch break sizes that will depressurize to a point just above the accumulator injection cover gas pressure.

System pressure, core power, core mixture level, and core steam flow are the important parameters that go into the calculation of the clad heat-up. Other key parameters effecting the transient that will be looked at are. break flow, vessel liquid mass and total system mass. A very important observation for the Millstone Unit No. 2 small break LOCA analysis is that, in the 4-inch break, the vessel liquid mass and the total system mass displayed an increase approximately 300 seconds before accumulator injection while, at. the same time, the core mixture level stabilized at the 6.0 foot elevation in the core. The rate of

' depressurization of the 4 inch break is 30 psia per 100 seconds just prior to accumulator injection, which is similar to the break cases in WCAP-10054, Addendum 1.

Break sizes slightly smaller than 4 inches would still have accumulator injection.

They would have less inventory loss and shorter core uncovery times than the 4-inch break. The shorter cover uncovery time would offset the slight delay in accumulator injection for the smaller break sizes.

As stated before, it is conceivable that a break size between 3 and 4 inches will depressurize to a point just above the accumulator back-pressure. However, since the system liquid masses are increasing before accumulator injection and the core mixture elevation has stabilized, the conclusions made in Reference (1) apply equally for the plant-specific analysis for Millstone Unit No. 2, which has

, shown that a break between 3 and 4 inches without accumulator injection is bounded by the 3 and 4 inch break analyses.

Reference (1): Letter NS-NRC-86-3172, E. P. Rahe, 3r. to D. M. Crutchfield, dated October 23,1986.

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4 Docket No. 50-336 A06096 2

B12348 1

4 Attachment 2 i

Millstone Nuclear Power Station, Unit No. 2

, Additional Justification in Accordance with 10CFR50.92 Proposed Revision to Technical Specifications

Pressure-Temperature Limit Curves r

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December,1986 4

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E Justification in Accordance with 10 CFR 50.92 The proposed changes to the Technical Specifications for Millstone Unit No. 2, submitted to the NRC Staff on October 20, 1986,(1) revise the pressure-temperature limits and the maximum heat-up and cool-down rates for the reactor coolant system. The proposed changes to page 3/4 4-17 included a change to the ACTION Statement which replaces the current wording of " fracture toughness properties" with the recommended wording of " structural integrity." This wording identifies the particular aspect of the reactor coolant system that is to be evaluated in the event any of the heat-up or cool-down limits specified in Specification 3.4.9.1 are exceeded. These wording changes represent a much more conservative course of action and reflect the actual scope of evaluation that NNECO would be conducting if any of the prescribed limits were exceeded.

" Fracture toughness properties" (current specification) is only one subset of

" structural integrity" (proposed changes). This proposed change would specify consideration of all the structural implications of an out-of-limit condition such as plasticity, fatigue, and others, as well as fracture toughness.

NNECO has reviewed the attached proposed changes pursuant to 10 CFR 50.59 and has determined that they do not constitute an unreviewed safety question. The probability of occurrence or the consequences of a previously analyzed accident have not been increased and the possibility for a new type of accident has not been created, nor has the margin of safety as defined in the basis of any Technical Specification been reduced. The proposed specification change represents a more conservative requirement for those situations when a limit is exceeded.

NNECO has reviewed the proposed change, in accordance with 10CFR50.92, and has concluded that they do not invol/e a significant hazards consideration in that the change do not:

1. Involve a significant increase in the probability or consequences of an accident previously analyzed. This proposed change reflects a more restrictive requirement for a broader-based evaluation in the event an out-of-limit condition occurs.

2. Create the possibility of a new or different kind of accident from any previously analyzed. This proposed change does not alter current requirements but simply makes them more comprehensive and conserva-tive.

3. Involve a significant reduction in a margin of safety. This proposed change represents increased requirements. Safety margins as defined in the bases of the Technical Specifications are enhanced by the inclusion of the more comprehensive wording.

NNECO's conclusion that the three criteria of 10 CFR 50.92(c) are not compromised is supported by our determination made pursuant to 10 CFR 50.59.

Our review of the given examples in 51 FR 7750, March 6,1986, of amendments that are considered not likely to involve a significant hazards consideration, has (1) 3.F. Opeka letter to A.C. Thadani, " Proposed Revision to Technical Specifications, Pressure-Temperature Limit Curves," dated October 20,1986.

o, .,

determined that example (ii), a change that constitutes an additional limitation, restriction or control not presently included in the technical specifications, is the one most applicable to this proposed change. The proposed revised wording represents a more conservative technical specification requirement but reflects what NNECO would actually do in the case of an out-of-limit condition.

i The Millstone Unit No. 2 Nuclear Review Board has reviewed and approved the attached proposed revision and has concurred with the above determinations.

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