ML19281A096

Supports Proposed License Amend to Increase Rated Thermal Power to 2,700 Mwt,Starting W/Cycle 3 Operation.Forwards Addl Info Re Control Rod Ejection,Main Steam Line Rupture & Seized Rotor Event

ML19281A096
Person / Time
Site:	Millstone
Issue date:	02/27/1979
From:	Counsil W NORTHEAST UTILITIES
To:	Reid R Office of Nuclear Reactor Regulation
References
TAC-46174, NUDOCS 7903060279
Download: ML19281A096 (3)

Text

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February 27, 1979 Docket No. 50-336 Director of Nuclear Reactor Regulation Attn:

Mr. R. Reid, Chief Operating Reactors Branch #4 U.

S. Nuclear Regulatory Commission Washington, D. C.

20555

References:

(1)

W. G. Counsil letter to R. Reid, dated December 15, 1978.

(2)

W. G. Counsil letter to R. Reid, dated February 12, 1979.

(3)

W. G. Counsil letter to R. Reid, dated January 17, 1979.

Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Power Uprating In Reference (1), Northeast Nuclear Energy Company (NNECO) proposed a License Amendment to increase rated thermal power to 2700 MWt starting with Cycle 3 operation.

Incorporated into the Reference (1) information were detailed evaluations of the radiological consequentes of five incidents.

These were the Fuel Handling Accident in the Spent Fue.1 Pool, Waste Gas Decay Tank Rupture, Hydrogen Purge, Fuel Handling Accident in Containment, and the LOCA.

In Reference (2), two additional incidents were addressed, namely, the Steam Generator Tube Rupture and the Control Room Dose to Operators Post-LOCA.

Reference (2) also indica

• ed that the remaining three incidents would be addressed in subsequent correspondence.

That additional information is provided below.

(1) Control Rod Ejection In Section 7.3.1 of Reference (2), it is concluded that no fuel pin failures occur as a result of a postulated CEA Ejection.

This result is similar to the FSAR analysis provided for Cycle 1.

Further, in the Millstone Unit No. 2 SER, the Staff calculated offsite doses from a rod ejection accident.

This analysis assumed a core thermal power level of 2700 MW, and concluded that the dose consequences were well within 10CFR100 limits.

Therefore, it is concluded that the radiologi-cal consequences of a postulated CEA ejection are bounded by the SER analysis, and are acceptable, 7003060 279 c

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(2) Main Steam Line Rupture In Section 7.3.2 of Reference (2), it is concluded that for both the two loop full power case, and the two loop no load case, critical heat flux is not exceeded.

This result is similar to that provided in the FSAR.

Based upon the above, more detailed calculations are not required to conclude that the radiological consequences of a postulated MSLB are acceptable.

(3)

Seized Rotor Section 7.3.4 of Reference (2) addresses the Seized Rotor Event. The second page of the description of the event (copy attached) was inadvertently omitted from Reference (2) and should be inserted in front of Table 7.3.4-1.

This page documents the result of a calculated 1.0% failed fuel f raction for a seized rotor event during Cycle 3.

Although this result is greater than the FSAR analysis which calculated a failed fuel fraction of 0.42%, more quantitative radiological evaluations are not warranted because, as stated in the FSAR, the very low proba-bility of this incident allows acceptance of this condition.

10CFR170 considerations regarding the above information are addressed in Reference (3).

We trust the above information is sufficient for you to concur with our conclusion that for all postulated events, the radiological consequences for Cycle 3 operation are acceptable.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY

./fl}L 21 W. G. Counsil Vice President Attachment

I e

O F.

In determination of the percent fuel failure associated with DNB for this transient, the TORC code is used to calculate the DNBR versus radial peaking factor for a given power level. An integral fuel damage calculation is then carried out by combining the results from TORC with Qe number of fuel rods having a given radial peaking f actor. The nueer of fuel rods versus radial peaking f actor is.

taken from a cumulative distribution of the fraction of fuel rods versus nuclear radial peaking f actor. This yields a distribution of the fraction of pins with a particular DNBR as a function of DNBR.

This information is then convoluted with a probability of burnout vs DNBR to obtain the amount of fuel failure. This method is discussed in detail in CENPD-183, "C-E Methods for Loss of Flow Analysis'.'.

This method is totally consistent with the method described in this topical report..and with methods previously used and approved for Calvert Cliffs (Reference 8).

In Table 7.3.4-2, the NSSS and RPS responses are shown for the seized rotor incident initiated fran an Ip value of

.16.

Figures 7.3.4-1 through 7.3.4-4 show core power, core average heat flux, RCS pressure, coolant temperatures during the transient.

A conservatively " flat" pin census distribution ( a histogram of the number of pins with radial peaks in intervals of 0.01 in radial peak normalized to the maximum peak) is used to determine the number of pins that experience DNB. The results show that the number of fuel pins that experience failure based on the criterion stated above is 1.0% in comparison to no predicted fuel failures for Cycle 2.

For the case of the loss of coolant flow resulting from a seizure of a reactor coolant pump shaft, a trip on low coolant flow is initiated to limit the predicted fuel failure to only a small f raction of the total number of pins.

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