ML18220A293
| ML18220A293 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/28/1978 |
| From: | Jurgensen R American Electric Power Service Corp |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18220A293 (12) | |
Text
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS >
" DISTRIBUTION FOR INCOMING MATERIAL 50-316 ~15
, REC:
KNIGHTON G W
NRC-ORG:
JURGENSEhl R
W AMER ELEC PWR SVC DOC DATE: 03/28/78 DATE RCVD: 04/05/78 DOCTYPE:
LETTER NOTARIZED:
NO
SUBJECT:
REQUEST REMOVAL OF ROBERT S.
HUNTER FROM MAILLING JURGENSEN IN HIS PLACE.
I COPIES RECEIVED LTR 1
ENCL 0 LIST AND ADDING ROBERT W.
PLANT NAME: COOK UNIT 2
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REVIEWER INITIAL:
XRS COOK UNIT 1
DISTR IBUTER Ih!ITIAL:~
DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS NOTES:
SEND 3 COPIES OF ALL MATERIAL TO IOE CHANGES OF PERSONNEL/ADDRESS
< DISTRIBUTION CODE 8007)
FOR ACTION:
FOR INFO:
FOR'NFO:
INTERNAL:
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BroadIIIaII, 'i nO YOrk, X Y. 10004 AEP (2l2) 422.4800 AMERICAN ELECTRIC POWER Serv',Ire'.,
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March 28, 1978 Mr. George W. Knighton, Chief Environmental Projects Branch No.
1 Division of Site Safety and Environmental Analysis U. S. Nuclear Regulatory Commission Washington, D.
C.
20555
Dear Mr. Knighton:
Would you please remove Robert S. Hunter from your mailing list and add Robert W. Jurgensen in his place.
Mr. Jurgensen is Chief Nuclear Engineer, Nuclear Engineering Division.
Very truly yours, RWJ:clb R.
W. Jurgensen Chief Nuclear Engineer V'
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NRC FORM 195 U.S. NUCLEAR REGULATORY
'VIISSION
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" NRC DISTRIBUTIONFoR'ART 50 DOCKET MATERIAL DOCKET NUMBER FILE NUMBER Mr. Edson G. Case FROM:
Westinghouse Elect Corp Pittsburgh, Pai 15230 Mo H, Judkis DATE OF DOCUMENT DATE RECEIVED 05/04/78 PC.ETTE R 0ORIGINAL SCOP Y ONOTORIZED
+CANC LASS IF I E D PROP INPUT FORM NUMBE R OF COPIES RECEIVE D tc~.
DESCRIPTION 4~i~ 3 S)(@f7'NCLOSURE Forwarding Amend 81 to the Final Safety Analysis Rept consisting of revisions to the FSAR to reflect the steam line break protection changes
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DONALD C COOK UNITS 1 8c 2 jcm 05/05/78 i'R FOR ACTION/INFORMATION ENVIRONMENTAL
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HARLESS INTERNALDISTRIBUTION NT SYSTEMS SITE SAFETY G
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EXTERNALDISTRIBUTION XO ACRS CONTROL NUMBER
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Westinghouse Electric Corporation Power Systems e
AEW-7095 PWR Systems Divtshn Box 355 Pittsburgh PernsyNania t5230 Mr. Edson G.
- Case, Acting Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission 7920 Norfolk Avenue
- Bethesda, Maryland 20014 March S. 0')
6, M78 AMP/60 tvS
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'CD MD e tie cn, CU r
AMERICAN ELECTRIC POWER PROJECT Donald C.
Cook FSAR Amendment'81
Dear Mr. Case:
Enclosed please find 70 copies of FSAR Amendment 81 applicable.to the Donald C.
Cook Docket 50-315 and 50-316.
The legal papers for this amendment are being forwarded by American Electric Power Service Corporation by a separate
~,cover letter.
This amendment revises the FSAR to reflect the steam line break protection changes.
M. Oper/je Attachment Very truly yours,
=M.
. dudk Mana er American E
ctric ower Projects CC:
R.
W. Jurgensen, 1L, 250A R.
S.
Hunter 1L R.
F. Hering 1L S.
H. Horowitz 1L S. J. Milioti 1L J.
G. Feinstein 1L
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DONALD C.. COOK NUCLEAR PLANT 4
AMENDMENT 81 INSTRUCTION SHEET REMOVE
-INSERT FRONT/BACK 14;2.'5-1/14.2.'5-2+
14.2.5-3/14.2.5-4 FRONT/BACK 14.2.5-1/14.2.5-2 14.2.5-3/14.2.5-3 ~
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RUPTURE OF A STEAM PIPE DISCUSSION OF ACCIDENT A ru ture of a steam pipe results in an uncontrolled steam release from a ste generator.
The steam release results in an initial increase in steam fl w which decreases during the accident as the steam pressure falls.
The energy removal from the Reactor Co'olant System causes a reduction of coolant temp ature and pressure.
In the presence of a negative coolant
,temperature co ficient, the cooldown results in a reduction of core shutdown margin. If the st reactive RCCA is assumed stuck in its fully withdrawn position, there is n increased possibility that the core will become critical and return to power.
A return to power following, a steam pipe rupture is a potential problem mai ly because of the high hot channel factors which exist when the most react ve assembly is assumed stuck in its fully withdrawn position.
Assuming the mos pessimistic combination of circumstances which could lead to power generation following a steam line break, the core is ultimately shut down by boric ac d delivered by the Emergency Core Cooling System.
t
,The analysis of a steam pipe rupture performed to demonstrate that:
1)
Assuming a stuck assembly, with or thout offsite power, and assuming a single failure in the engineered sa ty features there is no consequential damage to the primary system and the co remains in place and intact.
2)
Energy release to the containment from the rst steam pipe break does not cause failure of the containment stru ture.
3)
There will be no return to criticality after reacto trip, for a break equivalent to the spurious opening, with failu to close, of the largest of any single steam dump, relief or saf ty valve.
14.2.5-1
The following systems provide the necessary protection against a steam pipe rupture:
1)
Safety Infection System actuation from any of the following; a)
One out of three coincident low pressurizer pressure and low pressurizer level signals.
b)-
Two out of three differential pressure signals between a.steam line and the remaining steam lines.
c)
High steam line flow in two out of four main steam lines (one out of 'two per line), in coincidence with either low Reactor Coolant 'System average temperature (two out of four loops) or low main steam line pressure (two out of four lines).
d)
Two out of three high containment pressure signals.
2)
The overpower reactor trips (neutron flux and hT) and the reactor trip occurring in conjunction with receipt of the Safety Injection Signal.
3)
Redundant isolation of the main feedwater lines:
Sustained high feedwater
'low would cause additional cooldown.
Therefore, in addition to the normal control action which will,close the main feedwater valves, a safety infection,.signal. will rapidly close all feedwater control valves, trip the main feedwater
- pumps, and close the feedwater pump discharge valves.
4)
Trip of the fast acting'steam line stop valves (designed to close in less than 5 seconds) on:
14.2.5-2
a ~
High steam flow,in any two steam lines in coincidence with either low Reactor Coolant System'verage temperature or low steam line pressure".
b.
High containment pressure.
Each steam line has a fast-closing stop valve capable of stopping flow in either direction.
These four valves prevent blowdown of more than one steam generator for any break location even if one valve fails to close.
For example, in the case of a break upstream of the stop valve in one line, closure of any:three stop valves will prevent blowdown of the other steam generators.
In particular, the arrangement precludes blowdown of more than one steam generator inside the containment and thus prevents structural damage to the containment.
In addition each main steam line incorporates a 16 inch diameter venturi type flow restrictor which is located inside the containment.
These components serve to limit the rate of release of steam for an outside break.
2.
METHOD OF ANALYSIS The analysis of the steam pipe rupture has been performed to determine:
'I 1)
The core heat flux and Reactor Coolant System temperature and pressure resulting from the cooldown following the steam line break.
A full plant digital computer simulation has been used.
2)
The thermal and hydraulic behavior of the core following a steam line break.
A detailed thermal and hydraulic digital-computer calculation has been used to determine if DNB occurs for the core conditions computed in (1) above.
3)
The containment pressure response to a large area break.
14.2.5-3
0 The following conditions were assumed to exist at the time of a steam break r
accident:
1)
A 1.6% end of life shut down margin at no-load, equilibrium Xenon conditions, and with the most reactive assembly stuck in its fully withdrawn position:
Operation of the control rod banks during core burnup is restricted in such a way that addition of positive reactivity in a steam break accident will not lead to a more adverse condition than the case analyzed.
2)
The negative moderator coefficient of reactivity corresponding to the end of life rodded core with-the most reactive assembly in the fully withdrawn position.
The variation of the coefficient with temperature and pressure has been included.
The k ff versus coolant temperature eff corresponding to the negative moderator temperature coefficient used is shown in Figure 14.2.5-1.
In computing the power generation following-a steam line break, the local reactivity feedback from the high neutron flux in the region of the core near the stuck RCC assembly has been included in the overall reactivity balance.
The local reactivity feedback is composed of the doppler reactivity from the high fuel temperatures near the stuck RCC assembly and moderator feedback from the high water enthalpy near the stuck assembly.
The effect of power generation in the core on the total core reactivity is shown in Figure 14.2.5-2.
3)
Minimum capability for injection of high concentration boric acid solution corresponding to the most restrictive single, failure in the Safety Injection System.
The injection curve used is shown on Figure 14.2.5-3.
This corresponds to the flow delivered by one high head centrifugal charging pump delivering its full contents to the cold.leg header.
Low concentration boric acid (2000 ppm) must be swept from'the injection lines downstream of the Boron Injection Tank isolation valves, prior to the delivery of'igh concentration boric acid (20,000 ppm) to the main coolant loops.
This effect has been allowed for in.the analysis.
14.2.5-4