ML20002C819
| ML20002C819 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/15/1977 |
| From: | Bilby C, Bixel D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Desiree Davis Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8101120069 | |
| Download: ML20002C819 (18) | |
Text
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b power CompBuy cene, i on.co.: m2 West Wchigan Avenue. Jackson, Echigan 49201. Area Code S17 788 OSSO September 15, 1977 q
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Director of Nuclear Reactor Regulation 8
Att:
Mr. Don K. Davis, Acting Branch Chief 8;
Operating Reactor Branch No. 2 Y
U.S. Nuclear Regulatory Commission p
Washington, D.C.
20555 W
.k { n%l'. -I DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT -
REQUEST FOR EXEMPTION.
On August 18, 1977 a meeting was held in Bethesda, Maryland between Consumers Power Company and representatives of the NRC staff.
The purpose of this meeting was to discuss certain aspects of the Cycle 15 reload submittal for Sig Rock Point. At this meeting, Consumers Power Company was informed that the adequacy of spray distribution for the ring spray system (primary core spray) was under question by the staff and that Consumers Power Company would be required to justify the primary core spray distribution prior to start-up scheduled for early September 1977.
One method of providing justification for primary core spray distributien in-volved a geometric / trigonometric approach to the problem.
The projected cone angle from cach individual nozzle on the ring spray sparger would be superimposed on a core loading map.
From that projection an assumed spray flow rate could be developed for each fuel bundle. After appropriate correction factors have been applied, this number could then be correlated to the evaporation rate necessary to ensure adequate bundle cooling under LOCA conditions. Although Consumers
, Power Company does not believe that this method reflects actual primary core spray distribution, the results of this analysis have indicated some uncertainty in the primary core spray distribution.
Because of this uncertainty, Consumers Power Company is unable to conclude that the specified design criteria of the primary core spray system (of one gallon per minute per fuel bundle) can be met.
Thus, because of the question of adequacy of the primary ring spray cooling distribution, and because of the lack of time to complete comprehensive test programs evaluating this question, Censumers Power Company requests, under the previsions of 10 CFR 50.12, an exemption until the 1978 Cycle 16 start-up frem the failure criterion requirements imposed by 10 CFR 50.46 and Appendix K, Para-graph 1.D.1 as applied to a Loss of Coolant Accident followed by a concurrent single failure in the redundant core spray system.
The granting.of such an exemption is authorized by law and w!11 not endanger life or property or the H//c2bWof
I 7
consnon defense and security.
In addition, it is otherwise in the public in-terest-for the reasons set forth in the affidavit of C. R. Bilby, which is attached hereto.
Technical justification and proposed course of action for the exemption are provided in Attachment 1 to this letter.
't David A. Bixel Nuclear-Licensing Administrator CC: JGKeppler, USNRC 4
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A'ITACEME:IT 1..
- TECHNICAL JUSTIFICATION FOR EXD!PTION REQUEST a
The purpose.of this attachment is' tc provide. technical justification for.
Consumers Power Company's request for exemption until the 1978 Cycle 16 start-up from the requirements ' imposed by 10 CFR 50.146 and Appendix K, Paragraph I.D.1, as applied to a Loss of Coolant Accident followed by a f
concurrent single failure in the redundant core spray system.
The following topics are discussed in support of the exemptien request:
'I.
The risk assessment of a Loss of Coolant Accident assuming the primary core spray system inoperable.
II.' The operation of the Reactor Feed-Water System to ensure core cooling
~
in the case of a Loss of Coolant Accident caused by a break in the redundant core spray J.ine.
L
-III.
The risk assessment for a Loss of Coolant Accident in the redundant core spray line assuming Reactor Feed-Water System makeup.
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RISK ASSESSMENT: LOCA WITH PRIMARY CORE SPRAY INOPERABLE The r$sk assessment developed in this section deals with the postulated Loss
- of Coolant Accident (LOCA) assu=ing that the. primary ring spray is inoperable.
~
The scenario for the. event is as follows:
A LOCA occurs; Offsite power -(OFFSITE) or the E=ergency Diesel Generator (EDG)
-must operate-to provide power to the h80 V MCC 2B switchboard; with power to
- the switchbcard, the Electric Fire Pump (EFP) is capable of operation or the Diesel Fire Pump (DFP) must operate; assuming that at least ene fire pu=p operates, the breakers for the redundant core. spray valves must remain shut (ERKRS) and the components of the redundant core spray line (RCSV) must op-erate. Given this, adequate core cooling is assured. Graphically, the risk.
functions developed by this scenario are represented as:
P
'P g
g I
P
~
V
~
~
B R
A RISK =
LOCA 480V MCC 93 BP"PS RCSV 0FFSITE DFP P
P O
g Where: RISK = PA (PG0+Py+ PED B+
R
+
Thus, in order to determine the'overall risk assessment, the individual event probabilities must be assigned. It should be stipulated that, unless other-vise noted, all event probabilities were derived from WASH lh00, Appendix III.
The probability assigned to a' LOCA event (P ) e urring is 1 x 10 per hour A
per pipe section. Since this exemption reqcast covers one fuel cycle or approximately one year, the applicable number of hours is-8760. The location of the LOCA is assumed to be anywhere with two exceptions. A break in the primary core spray system is excepted because by memorandum and order dated May 26, 1976 the Con =ission granted Consumers Pov+ r Cc=pany a pla:;t life l
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I exemption for the Big Rock Point Plant against the criteria of 10 CFR 50.h6, Appendix K. Paragraph I.D.1, as applied to the specific case of a LOCA caused by ;a break in either core spray system. The other exception is a LOCA caused by a break in the redundant core spray system since Consumers Power-Company.
vill shov (see Section II) that'this event will not cause core uncoverage when credit is given for the reflood capabilities of the reactor feed syr';eu.
However, the specific number of pipe breaks capable of causing c LOCA has e
not been calculated but is conservatively bound by assuming one hundred (100).
Thus, P =1x 0 x 8760 x 100 or i.76 x 10 A
)
The probability assigned to the inoperability of the Emergency Diesel Gen-I erator (P ) is a combination of individual events. The first event to be G
considered is failure to start, assessed to be 3 x 10-2 per demand. Assu=ing a diesel start, the next event to consider-is failure to run, assessed to be
-3
-3 x 10 per' hour. Since it is. necessary for the diesel to run long enough to open the redundant core spray valves and power the Electric Fire Pump-until the recirculation mode, a time of 2h hours is conservatively applied.
Further, since the 2A-2B tie' breaker must open 'and the Emergency Diesel Generator tie breaker must close for proper operation, these probabilities l
must also be considered. Failure probabilities cf these breakers have each
.been assessed at 1 x 10-3 per demand. Thus, P = 30 x 10-3 + 72 x 10-3 G
or 10h x 10-3, j
1 x 10-3 + 1 x 10-3 4
However, if Offsite power is available, the operation of the Emergency Diesel 3
Generator is not necessary. By letter dated April 19, 1976, the staff as-sessed the probability of failure of Offsite power at Big Rock Point i=medi-ately following a LOCA. This value (P was assessed at 1 x l d per ennt 0
and is accepted.
The next event in the probability chain to be considered is the loss of the LSO V MCC 2B bus (P ).
If this bus were to fail (short to ground), the pcver y
-to open the redundant core-spray valves vould not be available. The proba-U
.i bility of this event is 3 x 10 per hour per wire. Thus, conservatively U
3 allowing'2h hours and assuming 100 vires, F becomes 3 x 10 x 2h x-100 or y
72 x 10-3, I-2 l
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f The inoperability of'the Electric Fire Pump (P is ependent upon row-E 3'
factors. - The>first is failure to start, assigned a probability of 1 x 10 per demand. Then, given a start, the probability of a failure to run, as-
~sessed at 3 x 10
'per hour. Since the pump must ru.: nntil the core spray recirculation system is activated, assume 2h hours, this probability becomes.
7 2 x 10~h There is a relief valve and check valve at the pu=p discharge; therefore,- failures of these components must be considered. Premature cpen-
~5 ing of a relief valve is assumed to be 1 x 10 per hour; with a conservative assumption of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> until isolation, the probability of this event becomes 1'x 10-5 Failure of a check valve is assigned the probability of 1 x 10 per'de=and. 'Thus Pg becomes 1 x 10-3 +.72 x 10-3 +.01 x 10-3
,1 x ig-3
~3 i
or 1.63 x 10 The operation of'the Diesel Fire Pu=p is also dependent upon the sa=e four i
factors, except that there are two relief valves. The probabilities assigned to the relief valves pre =aturely opening and the failure of the check valve
~5 aa.2 x 10 and 1 x 10, respectively..The probability of the pu=p failing.
t
-2 to start is 3 x 10 per demand and the probability of the pu=p failing to
-3 run once started is 3 x 10 per hour (again assu=e 2h hours of operation).
Thus, the overall probability of Diesel Fire Pump failure (P ) is 72 x 10-3 D
or 102 x 10-3,
- 30 x 10-3 +.02 x 10-3
,1 x 1g-3 The next event to be considered in the probability chain is the failure of the breakers for M0-7070 and M0-7071 to open. This value is assessed at 1 x 10~
per hour per valve. Since the breakers must stay shut only for the motor-operated valves to operate, a conservative time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is assigned.
~
Thus, P becomes-l'x 10 x 2 or 2 x 10 g
The final event to be considered is the failure of components in the re-dundant core spray line (P ).
The redundant core spray line consists of R
two motor-operated valves and one check valve. The probability assigned for
~ a failure of the motor-operated valves.to open is 1 x 10-3 per demand and the probability assigned ta a failure of the check valve to cpen is 1 x 10~ per
-3
-3 or 2.1 x 10-3' demand. Thus, P becomes 1 x 10 x 2 +.1 x 10 g
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---e-p t-g.e, y
-e--,y=
7
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Thus, the overall failure risk for one cycle of operation for a Loss of Coolant Accident assuming an inoperable primary core spray system is evalu-ated to be:
R
+
= 8.76 x 10-(104 x 10 x 1 x 1g-3 +.72 x 10-3 + 1.83 x 10-3
-3
-3 x 102 x 10
.002 x 10-3 + 2.1 x 10-3)
= 8.76 x lo- ( 3.1 x 10-3)
.. RISK = 2 7 x 10 / cycle Consumers Power Company concludes that the assessed risk for this event is acceptable and supports our execption request for an additional cycle of operation.
IL
LII. Core Cooling Folleving a LOCA Caused by a Break in the Redundant Core Spray Line
- Appendix A of Consumers Power Company's letter.to the NRC (dated March 26,
- 1976) presented a blevdown analysis ~ for' a break in the redundant core spray
'line (nozzle core spray line). This analysis considered a h" line break, about 12 feet above the top of the fuel and took no credit for feed-water makeup. The results show that the minimum level reached is one foot above
. the bottom of.the core (core never completely uncovers) and that this level is reached in 595 seconds.
Taking credit for feed-water system availability at 10 minutes after the Loss of Ccolant Accident (LOCA) and that the minimum level is reached in 595 see-onds, it can be assumed that total feed-water flow would be used for reflood-ing of the core. Primary system volume considerations indicate that 1,815 6allons of water are required -to raise the water level from the minimu= level reached during the blos icvn to the top of the fuel. Using the capacity of one feed pu=p (1600 gpm. reflood vill be completed in 68 seconds. The time of core reflood is calculated to be 678 seconds. This assumes a 10-minute wait for feed pu=p start, a 10-second starting time and a 68-second reflood time.
To determine the fuel temperature rise, the following method was used. From the blevdown analysis; it was determined that the hot node of the fuel would be uncovered at 520 seconds. Usin6 ANS standard plus 20% for calculating decay heat (at 520 seconds) an adiabetic heatup enalysis was perfor=ed. This calculation determined'the temperature rise between the time when the hot node is' uncovered and the t'me when the top of the fuel is reflooded. The calculated te=perature rise was h.56 F per second or 720 F for the time period between hot node uncovery and. core reflood.
.To develop a maximum fuel clad temperature, we referred to Figures A-9, A-10 and A-16 of our July 25, 1975 submittal, " Big Rock Point Plant Loss of Coolant Accident Analysis for General Electric Fuel in Conformance with 10 CFR 50, Appendix K (Nonjet Pump 3 oiling Water Reactor), July 11, 1975."- These figures j
. indicate that with an uncovery time of between 2h0 and 600 seconds, the het II-l r
+, i node clad temperature at. time of uncovery is typically 700 F.
Using this Lyalue' for initial clad temperature, the peak clad temperature for the case of a no :le spray line' break is approximately 1h20 F.
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We have concluded'that the above cohparative analysis provides a reasonably
. conservative determination of PCT and demonstrates that the PCT will be ac-ceptablE for a LOCA caused by a redundant core spray line break with feed-water reflood.
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III. -RISK ASSESSMENT: LOCA IN REDUNDAI:T CORE SPRAY LINE WITH FEED-WATER MAKEUP The risk assessnent for this section deals with the postulated LOCA occurring in the redundant core spray line assuming that the primary ring spray is inop-erable and that the core is kept covered by the operation of the reactor feed system.
In general, the risk assess =ent can be deter =ined by the following:
RISK = Pg (PF*P)
M Where: P represents the probability of a LOCA in the redundant core spray sys-3 tem, P is the probability that the reactor feed-water syste= is inoperable and F
P is the probability that fire water cannot be made up to the main condenser g
hot-vell.
l The value for P has been assessed by the NRC staff in their letter of April 19, 3
-5 1976 as being 9 x 10 per year and is accepted.
To derive P, it is necessary to develop a logic diagram as follows:
F I
P III C1 CV Pl CV P
P h80 V Colm P #1 V
FEED P #1 U2
~
U O
S LC3 1 V
C' *{ V d
P o OFFSITE 2h00 y
{
V BUS 9
h80 V CCLD P #2 GCK FEED P E2 GCK LG 2 V
V P
P P
P P
g C2 CV P2 CV The operation of the-feed syste= to provide core coverage is assured if Where:
l Offsite power (OFFSITE) is available, the 2h00 V bus (2h00 V BUS) does not short, either of the condensate pu=ps (COND P
- 2) and their associated power supplies 1
(k80 V LCE1 or ~2) and check valves (CHCK V) operste, either of the reactor feed-water pumps (FEED P
- 2) and their associated check valves'(CHCK V) operate, 1
the feed-water flev control valve (FCV) -operates and that the feed header check or step check (CHCK V) open as designed.
9 III-l l
4
1
- ( __
- 1 Thus:
~
(P CV P
- CV CV FC CV
+
+
+
' PF"PO S
L
- C ' + CV L
+ CV 1
1 1-1 2
2 u 1 2
Note: All probabilities were taken from WASH 1400 Appendix III unless -otherwise noted.
3 P = Probability of -loss of Offsite power which is 1 x 10 as derived by the NRC O
--staff's letter of April 19, 1976.
. P = Pr bability that the 2h00 V bus faf1s by shorting to ground. Assu=e'100 S
vires and a 1-hour run time'until core is covered.
(See Section II.)
I
~
= 3 x 10 /h-vire x 100 wires x 1 h 5
~
= 3 x 10 Probability that h80 V LCB 1 is lost due to short to ground, transformer P
=
~ shorting.or open circuiting, or breaker inadvertently opening. Again assume 100 vires and a 1-hour need.
v.
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-5 For short: P = 3 x 10 /h-wire x 100 wires x 1 h = 3 x 10 l
For breaker: ~ P = 1 x 10
'x 1 h = 1 x 10-For transformer: P = 1 x 10" /h x 1 h + 1 x 10~ /h x 1 h = 2 x 10'.
3'x 10-5 +.1 x 10-5 +. 2 x 10-5,
=
3.3-x 10 '.
=
Probability that #1 condensate pump fails to operate. The factors involved P - =
C are the probability that the pump fails to start (1 x 10" ) and the probability
~5 that it' fails to run for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (3 x 10 x 1).
E 1 x 10
+ 01 x 10-3,
~
=
1.01 x 10".
=
f 9
a III-2 t
ex-r-
w T-
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N P
'4Y T
4
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- W ly ir--p yy-y m-rvy g1@y Wyy g--,9yr>-i4-fi
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A P
= Probability that h80 V LCB 1 is lost due to short to ground transformer 2
shorting or open circuiting, or breaker failure. This is assessed at the same value as P, "1
= 3 3 x 10-5, P
= Probability that #2 condensate pu=p fails to operate. This is arsessed at C2 the same probability as PC*
1
= 1.01 x 10-3, P
= Probability that #1 reactor feed-water pump fails to operate. This is p1 assessed at the same probability as PC*
1
= 1.01 x 10-3, P
= Probability that #2 reactor feed-water pump fails to operate. This is p2 also assessed at the same probability as PC*
1
= 1.01 x 10-3, P
= Pr bability that check valves in feed and condensate system fail to open.
CV
= 1 x 10-(Note: This assumes one valve for each pu=p.)
l P
= Pr bability that the feed-water flow control valve fails to operate.
FC
= 1 x 10-3, bis is felt to be overly conservative since the bypass valve vill still be available to allow some flow if the FCV fails shut.
P Pr bability that check valve and stop check valve in feed header fails CV1 to open:
= 2 x 10'.
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Thus:
P
=P
(
- C + CV L
C + CV (P + CV}( P
- CV
+ CV FC F
O S
_L 1
1 2
1 1
2 2
-3
"(.033 x 10-3 + 1.01 x 10-3 +.1 x 10-3)(.033 x 10-3
= 1 x 10" +.03 x 10
-3 +.1 x 10-3)(1.01 x 10-3 +.1 x 10-3)~
+ 1.01 x 10-3.,1.'x 10-3)
+ (1.01 x 10
~
-3
~
~
+.2 x 10
+ 1 x 10
= 1 x'10-3 +.03 x 10-3+1.30x10k+1.21x10-6 +.2 x 10-3 +-1 x 10~3
-3 P
2.23 x 10 F
-Although this is a rather simplistic approach to the probability that the reactor feed-water system will not operate to provide flow for one hour, it does show that the most limiting-factors in the probability chain is the availability of Offsite power and the operation of the feed-water flow control valve.
c To derive F it is also necessary to develop a logic diagram:
M 1
FH.'
V V
P P
4 1
2 MU P,, =
FIRE h80 V h80 V MU VALVES EdR MCC 1A DIST FN1P I
Where: Emergency makeup water flow to the main condenser hot well is assured if the. fire header operates (FIRT. HDR), power is available to Motor Control Center 1A (h80 V MCC 1A) and to districution Panel 1P (h80 V Dist FN1P), and that the Motor-Operated Valves MO-7073 and M0-70Th (MU VALVES) operate.
Thus:
- I
+I
- I M"ITH V
V MU*
3 g
FH, the probability that the fire header (including at least one pu=p) does not P
l eperate, can be derived by utilizing the factors developed in Secticn I.
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t
PTh. = PG O V + P.,PD P
+P r.
= (104 x 10- )(1 x 10-3) +.72 x 10-3 + (1.83 x 10-3)(102 x 10-3)
= (.10h x 10-3) + 72 x 10-3 +.167 x 10-3
= 1.01 x 10 The probabilia that the h80 V Motor Control Center does not have power (ie, breaker inadeeltently opens) or shorts out is derived by (a) breaker failure:
1 x' 10 /h times 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or 1 x 10- and (b) bus short (assume 100 vires):
3 x 10 /h-vire x 1 h x 100 vires or 3 x 10 Thus: P becomes.1 x 10-5 +
-I
-5 y
-5
-5 1
+ 3 x 10 or 3.1 x 10 The probability what distribution Panel 1P does not have power, shorts, or does not supply power to the makeup feed valves (Py ) is derived by assur. lng that two circuit brea'.ers fail and that any one of 100 hires shorts to ground. Thus, P
becomes 2 x 1 x 10 plus 3 x 10 or 3 2 x 10-5,
-b
-3 y
The probability that the emergency makeup water valves do not operate is given by 2 x 1 x 10 or 2 x 10-3,
-3 Thus: The probability that the emergency makeup system does not operate is:
P = 1.01 x 10-3 +.031 x 10-3 +.032 x 10-3 + 2 x 10-3 g
1 3 07 x 10 Again, althoug'n this is a simplistic approach, it is evident that the overriding event in this chain is the failure probability of the emergency makeup water valves.
The risk assessment for the postulated LOCA occurring in the redundant core spray line, with a concurrent inability to provide feed flow is assessed at:
III-5
l RISK = PN (P
+P F
M I 9 x 10-3 (2.23 x 10-3 + 3 07 x 10-3) 2 9 x 10-5 (5 3 x 10-3)
I 4.77 x 10-I The overriding factors in this evaluation are the necessity to maintain Offsite power available, the operation of Nbtor-operated Valves M0-7073 and MO-707h, and the operation of the feed-water con rol valve.
Consumers Power Company concludes that the assessed risk for this event is acceptable for an additional cycle of operation.
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i 1
l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION.
In the Matter of
)
)
CONSUMFhs POWER COMPANY.
)
Docket 50-155
)
1(Big Rock 1'oint Plant)
).
License DPR-6
-)
AFFIDAVIT OF C. R. BILBY STATE OF MICHICAN )
)
SS.
COUNTY OF JACKSON )
I am Vice President of Production and Transmission for Consumers Power Company.
as such I am in charge of operation and maintenance of all electric production facilities of the Applicant.
It would be in the public interest for the Com-mission to grant the requested exemption from 10 CFR 50.46 and Appendi, K.
Paragraph 1.D.1, since it would result in a significant saving of fossil fuel
- and its cost to Consumers Power Company and its customers.
If the requested exempt.an is not granted, Consumers Power Company's Big Rock Point Plant will need to remain shut down until appropriate tests and/or modifications are completed..This will require testing the primary core spray system in a steam environment and, if necessary, modif"ing or replacing the existing system with one that conforms to the specified criteria.
If.the core spray testing were required and modification found necessary, we estimate that the plant st. art-up would be delayed to at least February 1978.
-To assess the economic impact of delaying the Big Rock Point Plant start-up on Consumers Power Company and its customers, the cost of replacement power 1
..=
- 2 j
was evaluated. This cost is est.imated to be $3.6 million for the period.
The replacement fuel used to generate this power (if produced by Consumers Power Company's other generating stations) would consist of about 185,000 barrels of oil and 37,000 tons of coal.
CONSUMERS POWER COMPANY
/
Q, By d-
-/
'C.
R. Bilby, Vice P dent Sworn and subscribed to before me this 15th day of September 1977.
@ !!YJ2)
O/_X Linda R. Thayer, Notarf PubtE (S E A L)
Jackson County, Michigan My Commission Expires July 9, 1979 O'
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y
u.s. NUCLEAa RE00LAToiiY CIN'.tISSION DOCKET NUMBE A [
4Cp;nu 195 q
/f t
top NRC Dl'2TRIBUTION rem PART 50 DOCKET MATERI AL h
Mr. Don K. Davis Consumers Power Co.
09/5fff Jackson, Michigan 49201 David A. Bixel DYf/NW'"
[zrTen CNoToni:Eo PROP INPUT PoRM NUMS:;R OF COPtES RECEIVED Yms:4NAL
([yNCLAS3lplg D CCoPY 3.5/6A/60 EN C:.0$U M E lsCEXPTICNRequest for exemption until the 1978 Consists of Technical justification
'ycle 16 start-up from the failure criterion and proposed course of action for the exemption..
C requirements imposed by 100FR 50.46 and Appen-Affidavit of C. R. Bilby Notorized 09/15/77...
dix K, Paragraph 1.D.1 as applied to a Loss of Coolant Accident followed by a concurrent single frilure innthe redundan, core spray system...
15p Trans The Following:
2p
'LANT NAME: BIG ROCK POINT jcm 09/16/77 l
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//6 EA>CL -
SAFETY FOR ACTION /INFORMATION I BRANCH CHIEF: W -A6 34 t/15 Moe4D 3 cu, s./ 4 ADVN.
l I
I AN INTERNAL DISTRIBUTION fREG FILE _3 TNRC PDR i I C F (2) iOELD I
!HANAUER ICHECK
\\STELLO l
lEISENHUT SHis0 BAER s
BUTLER GRIMES j J. COLLLNS I
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EXTERNAL DISTRIBUTION CONTR CL NUMBE R LPDR: :ifD28tM M/GH-TIC
/?Iy hqp$9f)b0?
NSIC 16 CYS ACRS SENT CATEU0ItY.8 I
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