ML17318A568
| ML17318A568 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/17/1980 |
| From: | Maloney G INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0630, RTR-NUREG-630 AEP:NRC:00322, AEP:NRC:322, NUDOCS 8001220451 | |
| Download: ML17318A568 (12) | |
Text
ACCESSION NBR FACIC';50-3i5 50-3i6 AUTH BYNAME MALONEYrGiPe ABC'If NAME OEATUA,H;H.
REGULATORY INFORMATION DISTRIBUTION SYSTEM (BIOS)
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~ 8001220 151 OOC ~ DATE 80/01/17 NOTARIZED ~
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'Donald C.'ook Nuclear Power Plant~ 'Onit lr Indiana 8
0500031i D'onald C.
Cook Nuc)eer Power Plantr Unit 2r Indians L
050003i6 AUTHOA AFFIL'LATION indiana 4 Michigen Electric Co, RECIPIENT AI FILIATION Office of NucT'e'ar" Reactor Regulation 1 A ~ I
SUBJECT:
Responds to 791109 1 tr re impact of clad behavior data on cur rent ECCS'enelys>s
~ Current limits remain valid, DISTH IBUTION CODE ~
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INDIANA Er MICHIGAN ELECTRIC COMPANY P. O.
BOX 18 BOWLING GREEN STATION NEW YORK, N. Y. 10004 January 17, 1980 AEP:NRC:00322 Donald C.
Cook Nuclear Plant Unit Nos.
1 and 2
Docket Nos.
50-315 and 50-316 License Nos.
Subject:
ECCS Analysis (Draft NUREG-0630)
Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Denton:
This letter responds to Mr. D. G.Eisenhut's letter of November 9, 1979 concerning the impact of the clad behavior data published in draft NUREG-0630 on the current ECCS analysis for operating plants.
In our letter of January 8,
1980 (AEP:NRC:00322A),
we stated that our pre-liminary review of the impact of the new NRC data on the Unit 1
ECCS analysis showed that our current limits remain valid.
A more complete review of this situation substantiates our preliminary conclusion.
Our evaluation which is the basis of this finding is included as Attachment A
to this letter.
The impact of the NRC data on Unit 2 was discussed in our letters of January 8,
(AEP:NRC:00322A), January 9,
(AEP:NRC:00322B),
and January ll, 1980 (AEP:NRC:00322C).
Very truly yours, GPM:em cc:(Attached)
G.
P.
1 o ey Vice P
si ent 80Ox 2 2o+ 5 I 5i/i
STATE OF NEW YORK ss.
COUNTY OF NEW YORK )
G.
P.
Maloney being duly sworn, deposes and says that he is the Vice President of licensee Indiana 8 Michigan Electric Company that he has read the foregoing statement con-cerning the 10 CFR 50.46 limit of the Donald C.
Cook Nuclear Plant and knows the contents thereof; and that said contents are true to the best of hi's knowledge and belief.
~Subscribed and sworn to before me this day of Notary Public 1%ATPLEEN DARKY HOTAt!Y t'U8r.iC, Stnto of How Yor>>
N0 ill ov6s93 Qunfjffad in Qucuns County Corljficnto ljjod in How Yor'r. County Corrrnu.s on cxtsrrus IJ>rcpt 30, 1'
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fir. Harold R. Denton, Director AEP:NRC:00322 cc:
R.
C. Ca'lien G. Charnoff R.
S. Hunter R.
W. Jurgensen D. V. Shaller
-Bridgman
ATTACHMENT A TO AEP:NRC:00322
EFFECT OF NEW CLAD SWELLING QND RUPTURE DATA ON 10 CFR 50.46 LIMITS FOR DONALD C.
COOK UNIT 1
Exxon Nuclear Company (ENC) has performed plant specific analysis for Donald C.
Cook Unit 1 to determi'ne the effects on the Peak Clad Temperature (PCT) of the NRC-proposed Clad Swelling and Rupture Model(1).
This analysis shows that the differences between the PCT's calculated with the NRC models are sufficiently small such that the current total peaking factor limits for Donald C.
Cook Unit 1 protect the 2200cF PCT limit.
Therefore, it is concluded that Donald CD Cook Unit 1 continues to be in conformance with the criteria of 10 CFR 50.46.
The changes in calculated PCT for ENC fuel at Donald C.
Cook Unit 1
when the NRC model for clad swelli'ng and rupture is used in place of the ENC model are summarized i'n Table 1.
The calculations wpre performed for the 1.0 DECLS limi'ting break at Donald C.
Cook Unit 1 (2).
The PCT impact is seen to be relati'vely small ranging from 22oF to 35oF depending upon fuel exposure.
Previ'ous ENC calcufagions for Donald C.
Cook using the ENC WREM-IIA ECCS Evaluati'on Model (3) showed the available margin to the 2200of limi't to be im excess of 40oF.
.Therefore, the PCT differences due to the use of the NRC Clad Swelling and Flow Blockage Model all fall within the av~i'lable margin to the 2200oF limit.
These sensitivity calculations were made in accordance with ENC's approved WREM-11A PWR ECCS Evaluation Model (3,4, 5).
The fuel rod in-ternal pressure for each exposure corresponds to the nominal conditions identified in the Donald C.
Cook exposure sensi'tivity analysis (6).
IA all cases, the. pC7 wi.'th. the NRC clad swelling and rupture model is less than 2200oP'.
4,
TABLE' EFFECT OF NRC RUPTURE; AND FLOW BLOCKAGE MODEI, ON D.
C.
COOK 1
ECCS ANALYSlS.
Peak Pellet Exposure (MWD/MTM)
Total Peaking F~
Heatup Rate at Rupture
( C/S
)
1.95 2.0 20,000 1.95 1.5 1.774 1.65 1.0
.5 30,000 40,000 PCT Impact of NRC Model vs; ENC Model
( F)
+22oF 280F
+35oF
+29oF Current Technical Speci ficati ons
REFERENCES 1.
D. A. Powers and R. 0. Meyer, "Cladding Swelling and Rupture Models for LOCA Analysis, Draft NUREG-0630, November 8, 1979.
2.
"Donald C.
Cook Unit 1
LOCA Analyses Using the ENC WREN-Based PWR ECCS Evaluation Model (ENC WREM-II'), " XN-76-51, October 1976.
3.
"Exxon Nuclear Company NREM-Based Generi'c PWR ECCS Evaluation Model Update ENC WREM-llA:
Responses to NRC Request for Additional Infor-mation,"
XN-NF-78-30 (A) 5 XN-NF-78-30, Amendment (A), May 1979.
4.
" Exxon Nuclear Company WREN-Based Generic PWR ECCS Evaluation Model, "
XN-75-41:
a.
Volume I, July 1975 b.
Volume I'I, August 1975 c.
Volume II'I, Revision 2, August 1975 d.
Supplement 1, August 1975 e.
Supplement 2, August 1975 f.
Supplement 3, August 1975 g.
Supplement 4, August 1975 h.
Supplement 5,
Revi.si.'on 5, October 1975 Supplement 6, October 1975 j.
Supplement 7, November 1975.
5.
"Exxon Nuclear Company WREN-Based Generi'c PWg
. ECCS Evaluation Model Update ENC WREM-I'P.," XN-76-27, July 1976; Supplement 1,
September 1976; Supplement 2, November 1976.
6, "Flow Blockage and Exposure Sensi'ti'vity Study for ENC D..C, Cook Unit 1 Reload Fuel Ustng ENC WREM-ll Model, "XN-76-51; Supplement 1,
January 1977; Supplement 2, February 1978; Supplement 3, April 1978.