B12302, Forwards Responses to 860908 Request for Addl Info on Small Break LOCA Evaluations (NUREG-0737,Item II.K.3.31).Steam Generations Assumed to Be Symmetrically Plugged at 23.4% Tubes Plugged Per Steam Generator

Forwards Responses to 860908 Request for Addl Info on Small Break LOCA Evaluations (NUREG-0737,Item II.K.3.31).Steam Generations Assumed to Be Symmetrically Plugged at 23.4% Tubes Plugged Per Steam Generator

ML20215K675
Person / Time
Site:	Millstone
Issue date:	10/20/1986
From:	Opeka J NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	Thadani A Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.30, TASK-2.K.3.31, TASK-TM B12302, TAC-48179, NUDOCS 8610280254
Download: ML20215K675 (6)

Text

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P.O. BOX 270 HARTFORD. CONNECTICUT 06141-0270 k L J ZI, ((.N,'co*,**,,', (203) 665-5000 October 20,1986 Docket No. 50-336 A06096 B12302 Office of Nuclear Reactor Regulation Attn: 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 Item II.K.3.31, NUREG-0737 Small Break LOCA Evaluations In a letter dated August 29, 1986,(1) Northeast Nuclear Energy Company (NNECO) submitted to the NRC Staff both a "small-break" and a "large-break" LOCA reanalysis for Millstone Unit No. 2. Our August 29, 1986 letter stated that these analyses would be referenced in the near future in a proposed license amendment request regarding Reactor Coolant System flow rate as defined in Technical Specification 3.2.6.

In a letter dated September 8, 1986,(2) the NRC Staff requested that NNECO provide additional information to facilitate the Staff review. Accordingly,

( Attachment I to this letter provides the additional information requested for l Millstone Unit No. 2.

We trust that this information is responsive to your request.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY E b 3* W eka P O l

h0280254861020 p ADOCK 05000336 Senior Vice President PDR A ttachment (1) J.F. Opeka letter to A.C. Thadani, " Item II.K.3.31, NUREG-0737, Small Break LOCA Evaluations," dated August 29,1986. g g (2) D.H. Jaffe letter to J.F. Opeka, " Request for Additional Information on l Millstone 2 LOCA Evaluations," dated September 18,1986.

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Docket No. 50-336 Attachment 1 Millstone Nuclear Power Station, Unit No. 2-Responses to Questions on LOCA Evaluations October,1986 1

Attachment 1 Response to the NRC Request for Additional Information on the Millstone 2 LOCA Evaluations Small Break Analysis

1. " Describe the basis for the core axial power distribution used in the analysis. Justify that this power shape is the worst axial shape allowed by the Technical Specifications."

NNECO Response:

The following are the criteria used in selecting the worst case power shape for Millstone Unit No. 2:

1. 100% power, all control rods out

2. Maximum axial offset with uncertainties

a. maximum power in top half of core

b. minimum power in bottom half of core

3. Maximum local power in top two feet of core

4. Challenges plant Technical Specifications (Fxy, ASl, Linear Heat Rate)

The power shape for Millstone Unit No. 2 meets these requirements.

Specifically, the Millstone Unit No. 2 Technical Specifications state the Axial Shape Index (ASI) can be no more negative than 10% (-0.10) at 100%

power. If a + 6% uncertainty is introduced, the bounding value for the negative ASI is 16% (-0.16). Since a more negative ASI results in more power in the top of the core, the Millstone Unit No. 2 Smali Break LOCA analysis was performed with a power shape based on an ASI of -17% (-0.17) at 100% power. This is more conservative than required by the Technical Specifications and thus, the Small Break LOCA analysis bounds power shapes allowed by the Technical Specifications.

2. "It is stated that the small break spectrum analyses, documented in WCAP-10054, Addendum 1, is based upon the Millstone Unit No. 2 plant. The Millstone Unit No. 2 limiting break analysis resulted in a peak cladding temperature of 21350F or approximately 1600F higher than the results in WCAP-10054, Addendum 1. Describe and justify the differences between the models used in these two analyses and discuss the relative effects of these dif ferences on the temperature increase."

NNECO Response:

The analysis in WCAP-10054, Addendum I is based upon Millstone Unit No.

2 with the following assumptions:

1. 9.4% Steam generator tube plugging

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2. Thermal Shield in place

3. Safety Injection flow of 75% HPSI,75% LPSI, and 75% charging

4. Reactor vessel minimum guaranteed flow of 375,000 gpm This analysis resulted in a peak clad temperature of 1975oF. The Small Break LOCA reanalysis in question was performed with the following assumptions:

1. 23.4% Steam generator tube plugging

2. Thermal Shield removal

3. Safety Injection flow of 75% HPSI,50% LPSI, and 50% charging

4. Reactor vessel minimum guaranteed flow of 335,000 gpm This analysis resulted in a peak clad temperature of 21350F. These changes account for a 1600F increase in peak clad temperature between the two analyses. The two biggest contributors to the increase in the peak clad temperature are the increase in steam generator tube plugging and the reduction in minimum guaranteed flow. As stated in the 3.F. Opeka letter to A.C. Thadani, dated August 29, 1986, the purpose for assuming such a large number of steam generator tubes plugged was to fully bound _the effects of any reductions in RCS flow rate or reductions in steam generator heat transfer area on peak clad temperature. ,

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 in-crease. 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."

NNECO Response:

Regulation 10 CFR 50.46 requires that 'ECCS cooling performance shall be calculated in accordance with an acceptable evaluation model, and shall be calculated for a number of postulated loss-of-coolant accidents of dif-ferent sizes, locations, and other properties sufficient to provide assurance that the entire spectrum of postulated loss of coolant accidents is covered.'

Currently, for small break loss of coolant accidents, the spectrum of breaks to be analyzed per Westinghouse methodology consists of standard break sizes of 2,3,4,6 and 3 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 reference CE-plant. From these results it was shown that the peak clad

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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 does appear that the limiting break size would be the largest break in which the clad temperature excursion would be terminated  !

by safety injection flow alone (i.e., no accumulator injection). However, recent Westinghouse smatt break LOCA analyses have shown that, even with accumulator injection, a larger break resulting in increased break flow and deeper core uncovery at higher decay heat levels may be more limiting than slightly smaller break sizes without accumulator injection. On the basis of the current methodology used by Westinghouse, a spectrum of break sizes was analyzed for Millstone Unit No. 2 to assure the worst break size has been identified.

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9 Large Break Analysis

1. "On June 2,1986, Westinghouse notified the staff of errors in its 1981 ECCS evaluation model with respect to modeling of the control rod thimbles. Determine whether this model error is present in your analysis.

If this error is present, assess its impact on your plant to demonstrate compliance with 10 CFR 50.46."

NNECO Response:

The LOCA analysis for Millstone Unit No. 2 was performed in January of 1986 before the guide tube modelling methodology was changed. There-fore, the old method of modelling control rod thimbles exists in the Millstone Unit No. 2 LOCA analysis. The cross sectional area for the guide tubes for Millstone Unit No.2 is 6.34 sf. For typical Westinghouse plants, the guide tube cross sectional area ranges between 3 to 6 sf. The Millstone Unit No. 2 LOCA analysis was performed with the 1981 evaluation model consisting of SATAN /WREFLOOD/ COCO /LOCTA without BART Westing-house evaluated the impact of the modelling change for the 1981 evalua-tion model without BART and found the impact on calculated peak clad temperature to be less than 200F, Currently the calculated peak clad temperature for Millstone Unit No. 2 is 21420F. On this basis, the Millstone Unit No. 2 LOCA ' analysis is still valid and complies with 10 CFR 50.46.

2. " Describe whether the steam generator tube plugging was modeled sym-metrically or asymmetrically and justify the approach used."

NNECO Response:

For Millstone Unit No. 2, the steam generators are assumed to be symmetrically plugged at 23.4% tubes plugged per steam generator. This approach is conservative as long as the steam generator with the highest plugging level remains below 23.4% tube plugged. The assumption that both steam generators are at the same plugging level of 23.4% results in a higher pressure drop across the steam generator for the LOCA analysis than actually would be the case in the plant. The higher pressure drops conservatively restrict flow through the loop during a LOCA, so it is justified to model the steam generators symmetrically.

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