ML20054H948
| ML20054H948 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 06/14/1982 |
| From: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| REF-SSINS-6820 FRY-82-64, FYR-82-64, IEB-80-04, IEB-80-4, NUDOCS 8206250242 | |
| Download: ML20054H948 (2) | |
Text
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YANKEE ATOMIC ELECTRIC COOPANY e
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2.C.2.1 1671 Worcester Road, Framingham, Massachunits 01701 yyg g2_sg
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June 14,1982 United States Nuclear Regulatory Commission Washington, D. C. 20555 Attention:
Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch #5 Division of Licensing
References:
(a) License No. DPR-3 (Docket No. 50-29)
(b) YAEC Letter to USNRC, dated May 8, 1980 (WYR 80-50)
(c) YAEC Letter to USNRC, No. 173, dated July 22, 1981, (FYR 81-52)
(d) USNRC Letter to YAEC, dated July 22, 1981 (e) YAEC Letter to USNRC, dated June 30, 1981, (FYR 81-95)
(f) USNRC Letter to YAEC, dated January 20, 1982
Subject:
Response to Request for Additional Information on PWR Main Steam Break with Continued Feedwater Addition
Dear Sir:
The subject of core reactivity response during the cooldown imposed after a major steam line break was discussed in the Reference (b) response to IE Bulletin 80-04.
A review was performed and reported in Reference (c) of the potential for recurrence of criticality following scram for Core XV which could result if core shutdown margin was reduced to zero during cooldown. The Core XV analysis was accepted by the NRC via Reference (d).
In Reference (e),
two design modifications were discussed that reduce the amount of cooldown during a steam line break.
These modifications include automatic tripping of the main condensate pumps on coincident signals of high containment pressure and low steam line pressure, and ensured main feed pump auto-trip at power levels greater than 15 MWe.
These design changes were implemented in addition to the automation of nonreturn valves, which provides for main steam system isolation upon receipt of either a high containment pressure or low steam line pressure trip signal.
Reference (e) states that potential for a return to power exists when the primary coolant system is rapidly cooled by increased energy removal due to the blowdown of secondary coolant. The combined effect of all various reactivity contributions determines whether recriticality can occur following shutdown. Ultimately, the amount of cooldown depends upon the available secondary coolant inventory and feedwater system performance following reactor t ri p.
Following automation of NRV closure during the Core XIV-XV refueling outage, blowdown of more than one steam generator either inside or outside containment will be prevented when credit is taken for either (1) automatic closure of the NRV's in fulfilling their function as isolation valves, or (2) prevention of reverse flow through the NRV's of the remaining unbroken steam lines in fulfilling their function as reverse flow check valves. The i
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United States Nuclear Regulatory Commission June 14, 1982 Attsntion:
Mr. Dennie M. Crutchfield PJgm 2 discussion in Item 2 of Reference (f) does not reflect these substantial and preventive design modifications that will prevent recriticality by limiting cooldown to final temperatures in excess of 700F.
Analysis is performed for each reload core to determine the shutdown margin required to prevent recurrent criticality due to a cooldown associated with complete blowdown of all secondary coolant from each steam generator.
Most recently, the Core XV analysis indicated that recriticality was prevented without crediting main steam system isolation on reactor trip on low main steam line pressure. The Core XV analysis is similar to the Core XIV analysis. The Core XIV analysis demonstrated for both full power and zero power cases that, even if cooldown of the primary loop to 700F following shutdown was considered, the core still retained sufficient shutdown margin to ensure subcriticality.
This calculation included moderator temperature defect, fuel temperature Doppler defect, provision for the most reactive control rod stuck out, boron injection, and conservative nuclear design uncertainties. Rod insertion limits are imposed for all operating modes to provide adequate shutdown margin to preclude recriticality, even when main steam isolation is not credited and blowdown of all four steam generators is assumed.
Both the Core XIV and Core XV reactivity calculations for cooldown following steam line breaks also demonstrated that, if conservative nuclear design uncertainties were not assumed, cooldown to 700F will not result in recriticality following scram even in the absence of boron injection.
The post-Core XIV design modifications described above, however, will limit cooldown to final temperatures of 2120F or greater.
Thus,_ boron injection is unnecessary to preclude recriticality at this higher temperature, even when conservative nuclear design uncertainties are assumed. The 700F temperature is, therefore, unrealistically low when credit is taken for automatic isolation of the main steam system. Shutdown margin requirements presently preclude recriticality following steam line breaks, without crediting boron addition from safety injection.
Since the Reference (b) response to IE Eulletin 80-04 was prepared prior to these design modifications, the questions in Item 2 of Reference (f) may be answered by noting that neither automatic safety injection system operation nor manual operator action is required to prevent recriticality following steam line breaks.
Therefore, the questions regarding dynamics of introducing boron to prevent recriticality are moot for Yankee Ibclear Pover Station.
This conclusion is assured by imposing stringent Technical Specification requirements for shutdown margin in each mode of operation.
Very truly yours, EE ATO I ELECTRIC COMPANY 4.
J. A. Kay Senior Engineer - Licensing JAK: dad 1