ML20141A885

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Informs That Proposed Change 7 to License DR-15,revising TS Re Qualification Requirements for Key Personnel,Acceptable
ML20141A885
Person / Time
Site: 05000231
Issue date: 12/02/1971
From: Skovholt D
US ATOMIC ENERGY COMMISSION (AEC)
To: Wolfe B
GENERAL ELECTRIC CO.
Shared Package
ML20140G249 List: ... further results
References
FOIA-97-34 NUDOCS 9705150039
Download: ML20141A885 (3)
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visininui10A W. Dooly, DR R. Engelken, CO (2)

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,. Docket File j

DRL Reading s

Docket No. 50-231 Branch Reading i

ACRS (3) i R. Boyd i

R. DeYoung f

General Electric Company D. Skovholt ATTN:

Dr. Bertram Wolfe R. Vollmer General Manager R. Schemel i

Breeder Teactor Development Operation P. Erickson 310 DeGuigne Drive R. Woodruf f Sunnyvale, California 94086 Chakteggs7 T

Gentlemen:

License No. DR-15 1

l

'Your Proposed Change No. 2 dated June 12, 1970, subsequently amended on July 30, 1970, requested a change to the Technical Specifications of Provisional Operating j

License No. DR-15 that would:

(a) revise the qualification requirements for key j

personnel, (b) clarify limits for the release of radioactive affluents from plant j

stack, and (c) increase the allowable flow in the pump-around loop.

Items (a) j and (b) were designated as Change No. 3 and authorised, as modified, by letter l

dated April 12, 1971. We have now reviewed item (c) that would increase the allowable flow in the pump-around loop. We have designated this action as Change No. 7.

1 During the course of our review, we found that certain modifications were neces-j sary to meet our licensing requirements. These modifications, as discussed with j

you, have been made. We conclude that the change, as modified, does not present significant hasards considerations not described or implicit in the Safety Analysis Report and that there is reasonable assurance that the health and safety l

of the public will not be endangered. Accordingly, pursuant to Section 50.59 of

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10 CFR Part 50, the Technical Specifications of Provisional Operating License l

No. DR-15 are hereby changed as follows:

j (1) Substitute replacement page 3.4-2 attached.

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~(2) Change 3.4.B to read "the reactor vessel pump-around loop shall be oper-i sting with the flow maintained at a nominal value of 2.0 spa or less".

I sincerely, IS/

l Donald J. Skovholt Assistant Director for Esector Operations l'

Division of Rosetor Licensino

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Docket No. 50-231 General Electric Company ATTN:

Dr. Bertram Wolfe General Manager Breeder Reactor Development Operation 310 DeGuigne Drive Sunnyvale, California 94086 Change No. 7 Gentlemen:

License No. DR-15 Your Proposed Change No. 2 dated June 12, 1970,. subsequently amended on July 30, 1970, requested a change to the Technical Specifications of Provisional Operating License No. DR-15 that would:

(a) revise the qualification requirements for key i

personnel, (b) clarify limits for the release of radioactive effluents from plant stack, and (c) increase the allowable flow in the pump-around loop.

Items (a) and (b) were designated as Change No. 3 and authorized, as modified, by letter dated April 12, 1971. We have now reviewed item (c) that would increase the allowable flow in the pump-around loop. We have designated this action as Change No. 7.

During the course of our review, we found that certain modifications were neces-sary to meet our licensing requirements. These modifications, as discussed with you, have been made. We conclude that the change, as modified, does not present significant hazards considerations not described or implicit in the Safety Analysis Report and that there is reasonable assurance that the health and safety of the public will not be endangered. Accordingly, pursuant to Section 50.59 of 10 CFR Part 50, the Technical Specifications of Provisional Operating License No. DR-15 are hereby changed as follows:

(1)

Substitute replacement page 3.4-2 attached.

(2)

Change 3.4.B to read "the reactor vessel pump-around loop shall be oper-ating with the flow maintained at a nominal value of 2.0 gpm or less".

Sincerely, lp //ll /.C Donald Jc kovholt Assistant Director for Reactor Operations Division of Reactor Licensing

Enclosure:

l Replacement page 3.4-2 cc:

Paul B. Van Buren, Attorney General Electric Company

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.ases

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' ach of the four sodium coolant loops, including pumps, heat exchangers, and 5

1 associated contro's and coolant equipment, must be operable during reactor opera-a

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to assure adequate Iore cooling capability for normal and emergency conditions.

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s 300~F minimum temperature in the sodium loops assures that the sodium temperat'.:ri i

vill be maintained above the plugging temperature to avoid potential oxide plugging o

i p rob l ems.

The 300 / value provides a reasonable margin above 275 F, which is o

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expecteu to be tne towest plugging temperature that can be clearly determined.

1 Plunginn temperatures below 275 F are difficult to determine, because the character.

istic drop in finw wich decreasing temperature is not clearly distinguishable at

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lower temperatures.

Ir addition, the 300 P minimum temperature in the primary loops O

askures adequate shutdown margin for the core as specified in 3.3.A.

4-The pu.rp-arcund loop circulates sodium continuously between the reactor vessel and I

'e crimarr drain tank.

This loop must be operable during reactor operation to i

mcycain the reactor sodium level within prescribed limits and to provide assurance t.at tr e loop is availabic for accident situations.(1) the pump around flow-rate will be set high enough to avoid numerous low flow alarms

ue te r.ormal flow rcce variations.

The average flow rate over a reasonable period i

1 ci tim.* will not exceed 2 gpm during reactor operation.

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i The arvon ecver 21s system is required to be operable to maintain the conditions l

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j nrihed in specification 3.4.D.

The vent vacuum pump is required to reprime the i.

auxiliary primary coolant system in the event sodium is lost from that system during seme abnormal (accident) condition.

1he cover gas pressure in the secondary system is set equal to the cover gas pressure 2

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I-t!e reactor (which is at a lower elevation) so that the secondary sodium pressure

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fr the IHX will be greater than the primary sodium pressure in the IHX. This will

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assure that leakage of radioactive sodium from the primary coolant system to the secondary coolant system will not occur.

Under normal operating conditions, the i

secondary sodium pressure will exceed the primary sodium pressure by about 30 psi

.in the main IHX and about 43 psi in the auxiliary IHX. Small leaks may occur in the

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IHX, but the dif ferential pressure will prevent the radioactive primary sodium from entering the 4

3.4-2 Change No. 7 1971 DEC 2 -

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