ML17250A504
| ML17250A504 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 08/07/1980 |
| From: | White L ROCHESTER GAS & ELECTRIC CORP. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-05.B, TASK-3-5.B, TASK-RR NUDOCS 8008120419 | |
| Download: ML17250A504 (8) | |
Text
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:8008120419 DOC ~ DATE: 80/08/07 NOTARIZED:
NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Planti Unit ii Rochester G
05000244 AUTH'AME AUTHOR AFFILIATION NHITErL.D.
Rochester Gas L Electric Cor p.
RECIP ~ NAI'lE RECIPIENT AFFILIATION CRUTCHFIELD<D.
Ooerating Reactors Branch 5
SUBJECT:
Responds to NRC 800624 request for review of SEP Topic
~
III-5.B< "Pioe Break Outsjde Containment" L for schedule of implementation of NRC positions. Alternative diesel generator coolinz method test to be conducted by June 1981.
DISTRIBUTION CODE:
A035S COPIES RECEIVED:LTR g'NCL jg SIZE:
TITLE: SEP Tooics NOTES: 1 copy:SEP Sect.
Ldr.
05000244 REC Ip IENT ID CODE/NAME ACTION:
CRUTCHF IELD INTERNAL: A/D MATLLQUAL16 D/DIRiHUM FAC S
HYD/GEO BR 17 NRC 02 G FIL 01 COPIES LTTR ENCL REC IP IENT ID CODE/NAME CONT SYS A
13 DIRg HUM FAC SFY ILE 08 OR ASSESS BR 15 SEP BR 04 COPIES LTTR ENCL 1
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1 3
EXTERNAL: ACRS NSIC 23 07 16 1
1 LPDR 03 1
A J6 La )so0 3f 0
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!//////////II ROCHESTER GAS AND ELECTRIC CORPORATION o
89 EAST AVENUE, ROCHESTER, N.Y. l4649 LEON D. WHITE, JR.
VICE PRESIDENT TELEPHONE AREA CODE Tla 546-2700 August 7, 1980 Director of Nuclear Reactor Regulation Attention:
Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch t5 U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
SEP Topic III-5.B, "Pipe Break Outside; Containment" R.E.
Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Crutchfield:
This letter is provided in response to your June 24, 1980 request for a review of the NRC's SEP assessment of this topic, and for an RGRE implementation schedule for the five staff positions resulting from that-assessment.
Comments:
1)
Reference is made on p.
3 to an NRC evaluation of SEP Topic III-4.C, "Internally Generated Missiles".
We would appreciate receiving this evaluation.
2)
It is stated, on p. 7, that the Intermediate Building basement elevation could be flooded to 30 inches.
No reference is given for this value.
The RGEE evaluation of this scenario states that the maximum water level above the basement floor due to a complete feedwater line break is 7.2 inches (Nuclear Services Corporation report, "Thermal Hydraulic Analysis of Pipe Rupture Outside Containment for Robert E. Ginna Plant", Section 5.2 dated June 13, 1973).
3)
In the second paragraph on p. 9, third sentence, "Turbine Building" should be "Intermediate Building".
4)
On the bottom of p. 11, it should be noted that a pressure diaphragm wall has also been installed between the turbine building and the mechanical equipment room.
8008X2O ti 5)
At the bottom on p. 13, Auxiliar... of the IB, it is stated that the AFW lines are high energy lines which could result in pipe whip and jet impingement on cable trays and electrical penetrations.
It was pointed out during the Ginna site visit on this subject that there are check valves located in this line, immediately adjacent to the feedwater line.
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This piping system is therefore not pressurized during normal operation, and should not be categorized as a high energy line.
6)
The final sentence on p.
16 is not clear.
Nhy is the NRC continuing this SEP reevaluation of pipe breaks outside containmentP Nas this not the full review? If additional review is planned in this area, we believe that the review should be conducted before any commitments are required of RG&E to perform modifications.
7)
In Table 1, reference is made to two SEP Topic evaluations for Ginna, "Station Service and Cooling Water Systems" and "Safe Shutdown Systems".
Ne would appreciate receiving these reviews.
Implementation Schedule:
Staff Position 1 Xt is planned to conduct the alternative diesel generator cooling method test by June 1981.
0 Staff Position 2
Based on RG&E's review of this scenario, we find the proposed solution to be unnecessary.
Present routine walk-through inspections of the Intermediate Building would detect a pipe leak long before there were any danger of flooding safety-related equipment.
Xf the postulated leak occurred at a level above the sub-basement, leakage into the sub-basement via the floor drains would be obvious during the routine once-per-shift walk-throughs.
And even a large secondary side break would result in only a 2-foot depth in the sub-basement.
If the lea'k were in the Service Water piping located in the sub-basement of the Xntermediate Building, there would be a significant time interval between the initiation of the crack and the flooding of safety-related equipment.
The Intermediate Building sub-basement has a volume of approximately 50,000 ft.
With a service water leak rate of about 585 gpm (as calculated on p.l3 of the NRC assessment),
it would take over 10 1/2 hours to begin flooding the basement level. It does not seem conceivable that a sizeable leak rate such as this would not be detected, visibly or audibly, by personnel during the walk-throughs, or by personnel monitoring the control board (the 585 gpm leak would be a significant fraction
>108 of the Service Nater pump flow).
Staff Position 3
Protection from the effects of the Turbine Building Intermediate Building cinder block wall failure on the atmospheric dump valves and main steam safety valves will be integrated into the modification program resulting from RG&E's review of I&E Bulletin 80-11, "Masonry Wall Design"
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Our initial response to this bulletin is contained in a July 7, 1980 letter from L.D. White, Jr.
(RG&E) to Mr. Boyce H.
Grier (NRC Region I Director).
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Staff Position 4
It is presently planned to separate the battery rooms from the mechanical equipment room, where the source of Service Water leakage exists, by replacing the doorway with a watertight wall.
This modification should be completed by June 1981.
Staff Position 5
RGEE is performing an evaluation to determine the effects of a CVCS Letdown or steam heating line break in the Auxiliary Building in the vicinity of safety-related equipment.
The results of this study, and proposed modifications, will be submitted to the NRC for review in January 1981.
Pending the resolution of any noted concerns, present once-per-shift inspections, together with the procedures available for isolation of the steam heating line, should provide adequate protection against the effects of significant adverse environment damaging safety-related equipment.
Very truly yours,
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