ML20140G462

Comments on Rev 1 to Task A-39, Determination of Safety/ Relief Valve Pool Dynamic Loads & Temp Limits for BWR Containments. Alternative Approaches for Quencher Application Should Be More Clearly Described

ML20140G462
Person / Time
Issue date:	12/13/1977
From:	Butler W Office of Nuclear Reactor Regulation
To:	Teh-Chiun Su Office of Nuclear Reactor Regulation
Shared Package
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References
FOIA-85-665, REF-GTECI-A-39, REF-GTECI-CO, TASK-A-39, TASK-OR NUDOCS 8604020396
Download: ML20140G462 (2)

Text

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"' ; o UNITED STATES I f - ' ;., NUCLEAR REGULATORY COMMISslON

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WASHINGTON, D. C. 20555 0

'% ,,,,, DEC 131977 MEMORANDUM FOR: T. M. Su, Task Manager (A-39), Containment Systems Branch, DSS FROM: W. R. Butler, Chief, Plant Systems Branch, DOR

SUBJECT:

COPfiENTS ON REVISION 1 TO TASK A-39 We have reviewed your proposed Revision 1 to Category A Task A-39,

" Determination of Safety / Relief Valve (SRV) Pool Dynamic Loads and Temperature Limits for BWR Containments". As a result of our review, we nicommend the following changes to the proposed revision:

(1) The discussion of the application of quencher devices to the Mark II plants on page 4a needs clarification. The Mark II owners' alternative approaches for quencher application should be more clearly described.

(2) The fourth sentence on page 4a is confusing and should be deleted.

(3) Task C.1.c.(3) on page 7 should be clarified as "SRV design load cases". ,

(4) Task C.2.c on page 9 should be modified to reflect the 1/4 scale "T" quencher tests in the Mark I LTP.

(5) Task 2c on page 15 should be modified to include the 1/4 scale "T" quencher tests.

(6) The SRV. program schedule should be modified to reflect the following Mark I milestones, as documented in revision 2 to the Program Action Plan:

Contact:

C. Grimes M8077 t

8604020396 060114 PDR FOIA FIRESTOBS-665 PDR D-21

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T. M. Su i Task 2.c - Mark I quencher evaluatien

a. Small Scale Test Report June 1978

b. Monticello Quencher Test Report July 1978

c. SRV Discharge Steam Mixing Model Jan. 1978

d. SRV Torus Load Definition Sept.1978

e. SRV Pipe Load Definition Oct.1978

f. FSI Test Report Aug. 1978

g. FSI Analytical Model Sep.1978 Task 3.0 - Submerged Structures i

a. LOCA Analytical Model Jan. 1978

b. Simple Geometry Test Report May 1978 i c. Quencher Analytical Model Aug. 1978

d. Data Evaluation Report Dec.1978 4 K' W. R. Butler, Chief Plant Systens Branch Division of Operating Reactors cc: G. Lainas

J. Kudrick B. Buckley J. Guibert L. Shao R. Stuart C. Grimes l

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