ML20086D718

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Responds to 740124 Request for Addl Info Re FSAR Analysis of Rod Withdrawal During Reactor Startup Transient & Conditions Existing During 731107 Inadvertent Criticality
ML20086D718
Person / Time
Site: Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date: 02/28/1974
From: Minnick L
VERMONT YANKEE NUCLEAR POWER CORP.
To:
US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML20086D716 List:
References
NUDOCS 8312020469
Download: ML20086D718 (9)
v • d • e


Text

{{#Wiki_filter:f O O VERMONT YANKEE NUCLEAR POWER CORPORATION l SEVCtJTY SCVEN OROVE STRECT RUTLAND, VIMDtONT 05701 REPLY 70: ENGINEERING oft' ICE February 28, 1974 TURNPIKE ROAD WESTDORO M ASSACHUSETTS 01581 TELEPHONE 617 J66-90t1 U.S. Atomic Energy Commission M-Washington, D.C. 20545 .( Attention: Directorate of Licensing gb g y 'di4 5:., [ n .a ,h C

Subject:

Information Relative to Inadvertent Criticality at ss. Vermont Yankee sw.:m R p ytsta .I h'

References:

1) License No. DPR-28 (Docket No. 50-271) f, - g
2) AEC DOL Letter to Vermont Yankee Nuclear Power Corpora January 24, 1974, regarding inadvertent criticality.

Dear Sir:

We have reviewed the Reference 2) letter which requests information relative to the FSAR analysis of the rod withdrawal during a reactor startup transient and to the conditions existing during the inadvertent criticality,on 7 November, 1973. We are supplying the following as our response. Item 1) Curves of rod position and resulting reactivity as a function of time as well as curves of feedback and scram reactivity. Response: Figure 1A is a curve of reactivity insertion versus time for the FSAR transient and Figure IB is a curve of reactivity insertion versus position of control rod 26-23 for the VY transient. The control rod withdrawal rate for both transients is 3 inches per sec. The uncertainty in 1B is < 10%. Figure 2 contains carves of Doppler feedback versus time and scram reactivity versus time for the FSAR transient. Similar curves for the VY transient have not been calculated but the situation is different since the FSAR transient assumes an APRM scram at 120% power level while the actual incident was terminated at a much lower power level by an IRM scram. Thus the dominant shutdown mechanism was the negative reactivity from the rod scram and not Doppler as depicted in Figure 2 for the FSAR transient. Item 2) Curves of reactor power and period and hot pellet power density and energy content as a function of time. 'y 8312020469 74o40s ~[ h PDR ADDCK 05000271 S py COPY SENT REGION re m - _ _ - _v.a __-_-A

O O TO: U. S. Ar,omic Energy Commission Page Two February 28, 1974

Response

Figures 3, 4 and 5 show teactor average power density, peak pellet linear power density and fuel temperatures, respectively, for the FSAR transient. Reactor period can be estimated from Figure 3. Peak pellet linear power density is Curve 1 in Figure 4. As reported in Abnormal Occurrence No. A0-73-31, dated November 14, 1973, the estimated increases in fuel centerline temperature and clad surface temperature for the VY transient were 5000F and 50 F, respectively. These preliminary estimates were based on incomplete information and were conservatively stated. More considered estimates; based upon the reactivity inserted and the power level at which 0 the reactor scrammed, place these temperatures at <50 F and <5OF respectively. The corresponding peak pellet linear power density is estimated to be <5 KW/f t. The estimated uncertainty in these values is <10%. Peak reactor average power density for the transient was <1 KW/L. Item 3) All relevant kinetic parameters such as 6 (delayed neutron fraction), I (average neutron lifetime), and Doppler coeffient. Response:' Parameters used in the FSAR transient analyses were: 8 = 0.0065 i = 38.4 sec Doppler coef = -1.18 x 10-5 joy These values are applicable to the VY transient conditions. Item 4) SRM and IRM alarm and action levels and times of occurrence, with an indication of the relative flux levels at the hot 4 pellet and relevant instruments. i

Response

The analysis of the FSAR transient assumed scram occurred at 120% of full power. 6 The SRM and IRM alarm and action levels are 10 counts and 96% of scale respectively. These correspond to.001% power for both the SRM's and IRM's. ~The times of occurrence from the station computer are as follows: 4 i IRM @ 24-29 tripped @ 21:01:36:46 IRM @ 16-21 tripped 021:01:36:58 SRM @ 32-21 tripped @21:01:37 IRM @ 32-13 tripped @21:01:37:04 j IRM @ 32-37 tripped @21:01:37:14 Two IRM's and three SRM's did not reach trip points as the incident was terminated by the control rod scram. l

TO: U. S. At;omic nergy Commission Pege Three February 28, 1974 j + l Ratios of hot pellet flux to instrument flux during the VY transient are estimated to be: SRM at 32-21 Ratio = 2:1 IRM at 24-29 = 14:1 IRM at.16-21 = 52:1 f We trust that you will find' this inforniation satisfactory; however, should you need additional information, feel free to contact us. t I Very truly yours, 1 l VERMONT YANKEE NUCLEAR POWER CORPORATION Yl, L L e t t Lk< G t f L. E. Minnick Vice President LEM/dm i I f 3 4 i I 1 3 5 i i ) I i i j i I 4 1

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