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NUCLEAR nEGULATC 'tY COMMISSION
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J SEP 2 7 1975
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Docket No. 50-348 l
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l MEMORANDUM FOR:
John F. Stolz, Chief, Light Water Reactors Bran ~ch No.1 FROM:
L. L. Kintner, Project Manager, Light Water Reactors j
Branch No. 1 l
SUBJECT:
TECHNICAL SPECIFICATIONS AND ASSOCIATED BASES FOR STEAM'
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GENERATOR TUBE SURVEILLANCE - FARLEY UNIT 1
REFERENCES:
(1) Affidavits of K. Hoge and D. F. Ross in the Matter of Alabama Porar Company (Joseph M. Farley Nuclear Plant Units 1 and 2), llay 3. W76 1
l (2) liemo Ross to Heineman, April 28, 1976 U~~
(3) Decision by Atomic Safety and Licensing Appeal Board in the Matter of Northern States Power Company (Prairie Island Units 1 and 2),
ALAB-343, September-2,1976 BACKGROUND:
In response to the ASLB Order of March 24, 1976, posing a question concerning steam generator tube integrity during long term operation of the Farley plant, the staff prepared affidavits describine among other things, plans for a generic appraisal of the likelihood and consequences of steam generator tube failures concurrent with or resulting froa the transients and accidents customarily considered in safety analyses. This study is expected to provide the technical basis for either maintaining the present NRC position or for altering it if necessary. The present HRC position is that analyses of steam generator tube failures concurrent with steam line break or loss of coolant accident is not a design basis requirement.
Further details of the proposed study are provided in RefJrence (2).
Discussion of this proposed study in R. Heineman's offit.e on May 17, 1976, resulted in a conclusion that additional details of the proposed program were required'before deciding to proceed with it.
Recently, in a related matter, Westinghouse has informed the staff that part of its emergency core cooling system (ECCS) analysis model is not conserva-tive, due to the temperature of water in the reactor vessel being higher
,than previously assumed. Westinghouse had earlier informed the staff that 1
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1, plugged steam generationtubes have an adverse effect on ECCS performance.
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These two adverse effects, (higher water temperature and plugged ste:tm generator tubes) may be offset by reducing the fuel rod heat generator
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t c.s limit in the Technical Specifications.
One of the. major reasons provided by the staff for not considering concurrent tube failures with design basis accidents is that there is ample margin between the limiting fuel rod heat generation ratu
,d the expected operating heat generation rate, even if the maximum tube degradation allowed by Technical Specifications results in a few tube failures as a consequence of accidents (Reference 3). However, the r'
calculations that were the basis for conclusions in Reference (3) to demonstrate ample margin to acconaodate the adverse effect of tube failure's on ECCS performance did not consider the other two adverse
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effects of higher water temperature and plugged tubes.
Implementation in Farley Technical Specifications Representatives of the Division of Systems Safety (F. Schroeder, D. Ross,
. J. Rajan) and the Office of Standards Developement (L. Frank) met on 4
September 21, 1976, with the representatives of the Division of Project Management (D. Vassallo, J. Stolz, L. L. Kintner) to discuss the com-i pletion of the studies described in References (1) and (2) and the impact of these studies and the Appeal Board's decision on Prairie Island (Reference 3) on the licensing of Farley Unit 1.
A significant conclusion of Reference 3 was that" serious tube degra-dation will not invariably be heralded by minor leakage." This conclusion was based on the tube failure (125 gpm flow through the break area) at Point Beach 1 during operation and the " crack area" reported for the tube failure (80 gpm flow through the break area)pported by the recent Obrighe-im Reactor. This conclusion is further su at Surry Unit 1 during reactor operation. The Board further concluded that eddy current testing is the "only now availabl~e method for detecting flaws in steam generator tubes in situ."
The concensus of those present at the September 21, 1976 meeting was that for the long term, the studies described in References 1 and 2 should be j
l completed to determine whether tube degradation greater than that now l
recommended as an acceptance limit in Regulatory Cuide 1.83 should be allowed.
For the short term, the technical specifications for steam generator tube surveillance should require a pronet report to the Connission if "more than 3 of the tubes inspected exceed the plugging limit," in accordance with Regulatory Guide 1.83 Revision 1. July 1975.
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The current Westinghouse Standard Technical Specification 4.4.5 would require nodification, because they now require such a prompt report if p.?29' "" '
"more than 15 of the inspected tubes are defective."
The bases for the T.echnical Specification 4.4.5 should explicitly sta;e the basis-for the selection of the number of tubes as an abnormal degradation limit and thus considered to be a reportable occurrence per Regulatory Guide 1.16.
On September 22, 1976, a representative of the Division of Operating Reactors (D. Brinkman) said in a meeting with DSS (F. Litton) and DPM (D. Vassallo, L. L. Kintner) that such a change in Standard Technical Specification 4.4.5 could be simply made by changing the number in Categories C-2 and C-3 from 1% of the tubes inspected" to "3 of the
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tubes inspected.".However, the paperwork impact on already issued
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standard s be large. pecifications and those planned for operating units would Mr. Brinkman said that directions from DSS would be needed for such changes.
Mr. F. Litton recommended that such changes be considered and agreed to initiate the implementation of the* appropriate directions from DSS.
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L. L. Kintner, Project Manager Light Water Reactors Branch No.1 Division of Project Management t
cc:
D. B. Vassallo F. Schroeder
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- 0. Ross l
D. Eisenhut J. Knight L. Frank J. Rajan l
L. Kintner F. Litton D. Brinkman i
i J. Scinto 1
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i CONNECTICUT YANKEE ATOMIC POWER COMPANY
- 5. Ilbennan Noveaber 24, 1976 "O___
l Steam Generator Inspection Ref.:
- 1) Watinghouse Fie) A Service Report No. NSD-ITiR-1 (1979)
- 2) Westinghouse Field Service Report No. MRS-4.h CW/1 (1973)
- 3) Westinghouse Feild Service Report No. MR.1 h-k CY 6 h) Westinghouac Eddy Current Test (Pzvilminary) 5)
Mimo to R.L. Hale, rrvm n.it. se=re (ste== cowerator service at Connecticut Yankee) dated April 19,1976.
A) Sh.rwrt below are the resdte of Edity Currvest
- resting rrtus 1972 - 19T6 or defected tubes.
h0% OR GREATER 50% OR GRT.ATER Mevd MrTFJU YEAR LOOPS
'Orther A.V.B.
'ORTifER A.V.B.
'IOTAL_ M I
1 2
7 0
1 10 l
2 5
T I
a 13 D
3 0
1 O
O 1
4 2
10 1
3 1%
1 2
12 0
0 lb I
3 0
0 10 19T5 3
0 0
0 0
0 4
2 3
5 0
13 1
2 3
0 0
5 2
2 0
0 0
2 19U 3
0 0
0 0
0 4
1 2
7 2
12 1
www 1972 3
o 0
14 0
14 l
l B) Watinghouse Eddy Current Plu6 sing Cost is as fbilows:
1
$5,000/10 tubes or less l
2
$250/ tube all over ten tubes C)
Radiation Exposure in those areas where Eddy Current Testing is being conducted:
1) 30R/ hrs tube sheet utgion j
9) h r/ hem naam hat a tannat nam i
1 Memo-537 Loomtion of defects that are not relative to anti-vibration bar.
Ri Eddy Current Inspection was taken during 1974.
All Eddy Current Testing vna done on loop three and no deflect less than 50% was recorded.
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