ML20002D881
| ML20002D881 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/27/1971 |
| From: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Walke G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 8101230263 | |
| Download: ML20002D881 (7) | |
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ATOMIC ENERGY COMMISSION
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WASHINGTON D.C. 20545 u-t; A^7JJ July 27, 1971
'nmie Docket No. 50-155 Consumers Power Company ATTN:
Mr. Gerald J. Walke Nuclear Fuel Management Administrator 212 West Michigan Avenue
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Jackson, Michigan 49201 Change No. 26 Gentlemen:
License No. DPR-6 Your Proposed Change No. 27 dated February 2,1971, and supplements thereto by telegrams dated February 16 and 26, March 2 and 5, and April 8, 1971, requested changes to' the Technical Specifications of Facility License No. DPR-6 to permit operation of the Big Rock Point Nuclear Reactor using a modified post-incident cooling system. We have redesig-nated your request and supplements thereto as Proposed Change No. 26.
Our review of this proposed change is related to our review of the emergency core cooling system for the Big Rock Point f acility in connection with your proposal dated February 9, 1970, and to our comments thereon contained in my letter to you dated June 29, 1970.
On the basis of our review of your request, we have concluded that the proposed modifications to the post-incident cooling system, which include a new backup core spray system and automatic delayed actuation of the alternate containment spray system, should be made as soon as possible to increase the reliability of emergency core cooling. Your current evaluation to determine the need for additional modifications to assure an adequate supply of water to the feedwater pumps which must be relied upon for high pressure core cooling in the event of certain small size breaks (0.4 to 7.2 square inches) should be continued and needed modifi-cations should be implemented. The reanalysis of your ECCS capability, as described in our letter to you of July 20, 1971, should be performed as soon as practical. However, the proposed plant modifications and Technical Specification changes need not be delayed until these further evaluations are completed.
Consequently, we have concluded that the proposed modification of the Big Rock Point Nuclear Reactor post-incident spray system will provide T/O/ASO263
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. Consumers-Power Company July 27, 1971 increased safety and that operation with the modified system does not present significant hazards considerations not described or implicit in your Safety Analysis Report.
There is reasonable assurance that the health and safety of the public will not be endangered by the
. proposed modifications of.the Big Rock Point Nuclear Reactor.
Accordingly, pursuant to Section 50.59 of 10 CFR Part 50, the Technical Specifications of Facility License No. DPR-6 are hereby changed as indicated in Attachment A to this letter.
Sincerely, t
Peter A. Morris, Director Division of Reactor Licensing
Enclosure:
Attachment A - Changes to Technical Specifications cc w/ enclosure:
George F. Trowbridge, Esquire Shaw, Pittman, Potts, Trowbridge & Madden 910.- 17th Street, N. W.
Washington, D. C.
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ATTACHMENT A-CHANGE NO. 26 TO THE TECHNICAL SPECIFICATIONS
- FACILITY LICENSE NO. DPR-6 CONS 11MERS POWER COMPANY DOCKET NO. 50-155 1.
Change Section 3.5.l(d) to read as follows:
"(d). System Actuation' 1 Set Automatic dc Operated, 1 Set Automatic ac Operated" 2.
Change Section 3.5.
a) to read as follows:
"(a) Automatic operation of the containment spray system and backup containment spray system involves a time delay not to exceed five minutes after system actuation to allow the operator:to. override a possible spurious actuation. This time delay feature may be manually overridden to actuate 7
the spray system prior to the expiration of the five-minute delay period."
3.
Change Section 4.1.2(a) to add the following new section under Reactor Emergency Cooling between " Core Spray System" and " Core Spray System Recirculation":
" Backup ~ Core ~ Spray System:
Type Sparger Nozzle Centered Over Core Capacity of Sparger, gpm 470 Nozzle Pressure, psia 115" 4.
Change Section 4.1.2(b), except the tabulation at the end, to read as follows:
"(b) Operating Requirements A minimum of one reactor recirculating loop or its equivalent shall' be used during all reactor power operatio ts when i
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L reactor power level is above -1.0 MRt.
The maximum oper-ating' pressure and temperature shall be the same as 1the reactor vessel.
The controlled rateLof change of_ temper-ature -in the reactor vessel shall be limited to 100*F per hour. All:other components in the' system shall be capable of following this temperature change rate.
The safety.
relief valves shall be set appropriately for all' planned reactor operating pressures so that the allowable pressure of 1870-psia (1700 plus 10%) in the nuclear steam supply
_ system is not exceeded. The emergency condenser, core spray and backup
-t_. apray systems shall be operable and ready for serv.ce at all times during power operation.
The-core sp-
'jstem and shutdown cooling system shall be operable and ready for service during refueling operations and the breakers for M07070 and M07071 shall be tagged
'open'.
The primary coolant shall be sampled and analyzed
- daily during periods of power operation.
The following are absolute limits which if exceeded shall necessitate reactor shutdown.
Corrective action will necessarily be taken at more stringent limits to minimize the possibility of'these absolute limits ever being reached."
5.
In Section_4.2.l(a), add-the following items to the list of equip-ment which receive their-power supply from the 125-volt de battery system:
" Post-Incident Enclosure Spray Valves Core Spray Valves"
- 6.
In the first paragraph of Section 4.2.6 - Fire Protection System, change " Alternate method of ' core cooling by flooding the reactor pressure vessel" to read " Backup Core Spray Cooling System,"
7.
In item 6 [High ' enclosure pressure (4 pressure switches)) of the tabulation in Section 6.1.2, change the two columns indicated below to read as follows:
" Scram Setting and Tolerance Special Features 4
1.5 + 0.2 psi Closes containment sphere above atmospheric isolation valves."
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8.
In Section 6.1.4 - Related Systems:
Renumber present Section "6.1.4(b) Emergency Condenser Control" a.
to "6.1.4(d)...".
b.
Insert new paragraphs 6.1.4(b) and (c) as follows:
"(b) Backup Core Spray System Control The backup core spray system shall be automatically actuated by simultaneous tripping of the low reactor water level sensor along with the ' low reactor pressure device'.
The ' low reactor pressure device' consists of a pressure switch interlock which prevents backup core spray system admission valves opening while reactor pressure is above 200 psig."
"(c) Core Spray, Backup Core Spray, Containment Spray and Backup Containment Spray System Set Points The following tabulation gives the actuation set points for der?ces associated with the core spray, backup core spray, containment spray and backup containment spray systems:
(See next page)
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k Actuation Setting and
-Sensor Contacts Tolerance Function
' Low reactor water-2 switches / valve Elevation 610' Actuates M07051 and level (4 level
. 1 out of 2 6" + 1 inch 7061 in conjunction switches).
coincidence-with reactor pres-sure.
Low reactor water 2 switches / valve Elevation 610' Actuates M07070 and level (4 level 1 out of 2-6" i.1' inch 7071 in conjunction switches) coincidence with reactor pres-sure.
. Low reactor pres-2 switches / valve 200 psig i 20 psi Actuates M07051 and sure switches (4 1 out of 2 7061 in conjunction pressure switches) coincidence with low reactor water level.
Low reactor pressure
- 2 switches / valve 200 psig i 20 psi Actuates M07070 and switches 1 out of 2 7071 in conjunction coincidence with low reactor water level.
High enclosure pres-2 - 1 of 2 1.5 1 0.2 psig Actuates a time-sure' switches (2 required delay mechanism pressure switches) that initiates M07064 (containment spray system) within 3
5 minutes unless the control is manually overridden.
a High enclosure pres-2 - 1 of 2 2.0 1 0.2 psig Actuates a time-l sure switches required delay mechanism that initiates i
M07068.(backup containment spray system) within five minutes unless t
M07064 is open or the control is manually overridden.
Valve position 1
75-100% of full Blocks automatic switch.M07064-open actuation of M07068 when M07064 is full open."
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Change Section 6'1.5(b) to read as follows:
"(b) The core spray system, backup core spray system and emer-gency condenser system control initiation sensors shall be functionally tested not less frequently than once every-12 months. This testing of th* control initiation sensors of the core spray system and s
- ..p spray system shall include actuation of valves M07051, M07061, M07070 and M07071. The check valves between M07051 and M07061 and M07070 and M07071 will be tested once every 12 months to assure that they are operable."
10.~
Change Section 6.1.6 to read as'follows:
' "The automatic actuation of the containment spray system and the backup containment spray system shall occur within five minutes of the receipt of a high containment pressure signal.if more than one-half of the fuel bundles in the core are zirconium-clad."
11.
Delete paragraph (d), Supplemental Core Cooling, of Section 8.2.1-in its entirety.
6
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Docket No. 50-155 Consumers Power Company ATIN: lir. Gerald J. Walke Nuclear Fuel Fm agement Administrator t
212 West Michigan Avenue Jackson, Michigan 49201 Change No. 26 License No. DPR 6 Gentlemen:
Your Proposed Change No. 27 dated February 2, 1971, and supplements thereto by telegrams dated -February 16 and 26, March 2 and 5, and April 8, 1971, requested changes to the Technical Specifications of Facility License No. DPR-6 to permit operation of the Big Rock Point Nuclear We have redesig-Ecactor using a modified post-incident cooling system.
nated your request and supplements thereto as Proposed Change No. 26.
Our review of this proposed change is related to our review of the emergency core cooling system for the Big Rock Point facility in connection with your proposal dated February 9, 1970, and to our comments thereon contained in my letter to you dated June 29, 1970.
On the besis of our review of your request, we have concluded that the proposed modifications to the post-incident cooling system, which include a new backup core spray system and automatic delayed actuation of the alternate containment spray system, should be made as soon as possible Your current to increase the relisbility of emergency core cooling.
evalutaion to determine the need for additional modificatinus to assure an adequate supply of water i.o the feedvater pumps which muct be relied upon for high pressure core cooling in the event of certain sea 11 size breaks (0.4 to 7.2 square inches) should be continued and needed modifi-cations should be implemented. However, the proposed plant modifications and Technical Specification changes are not affected and need not be delayed until these current evaluations are completed.
Consequently, we have concluded that the proposed modification of the Big Rock Point Nuclear Reactor post-incident spray system will provide glop 30M3 1
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Consumers Power Company
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increased safety and that operation with the modified system does not present significant hazards considerations not described or implicit in your Safety Analysis Report. There is reasonable assurance that the health and safety of the public will not be endangered by the proposed modifications of the Big Rock Voint Nuclear Reactor.
Accordingly, pursuant to Section 50.59 of 10 CFR Part 50, the Technical Specifications of Facility Licence No. DPR-6 are her2by changed as indicated in Attachment A to this letter.
Sincerely, Peter A. Morris, Director Division of Reactor Licensing
Enclosure:
Attachment A - Changes to Technical Specifications cc: George F. Trowbridge, Esquire Shaw, Pittman, Potts, Trowbridge & Madden 910 - 17th Street, N. W.
Washington, D. C.
20006 Distribution W. Dooly, DR R. H. Vollmer, DRL R. Engelken, CO (2)
D. L. Ziemann, DRL H. Shapar, OGC J. J. SHea, DRL N. Dube, DRL (5)
R. M. Diggs, DRL J. R. Buchanan, ORNL R. DeYoung, DRL T. W. Laughlin, DTIE R. S. Boyd, DRL Docket File PDR DR Reading DRL Reading ACRS (3)
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l EL27197; Docket No. 50 -155 Consumers Power Company ATTN; Mr. Gerald J. Walke Nuclear Fuel Management Administrator 212 West Michigan Avenue Jackson, Michigan 49201 Change No. 26 Gentlemen.
License No. DPR-6 Your Proposed Change No. 27 dated February 2, 1971, and supplements thereto by telegrams dated February 16 and 26, March 2 and 5, and April 8, 1971, requested changes to the Technical Specifications of Facility Licens; No. DPR-6 to permit operation of the Big Rock Point Nuclear Reactor using a modified post-incident cooling system. We have redesig-nated your request and supplements thereto as Proposed Change No. 26.
Our review of this proposed change is related to our review of the coergency core cooling system for the Big Rock Point f acility in connection with your proposal dated February 9,1970, and to our comments thereon contained in my letter to you dated June 29, 1970.
On the basis of our review of your request, we have concluded that the proposed modifications to the post-incident cooling system, which include a new backup core spray systen and automatic delayed actuation of the alternate containment spray system, should be made as soon as possible to increase the reliability of emergency core cooling. Your current evaluation to determine the need for additional modifications to assure an adequate supply of water to the feedwater pumps which must be relied upon for high pressure core cooling in the event of certain small size breaks (0.4 to 7.2 square inches) should be continued and needed modifi-cations should be implemented. The reanalysis of your ECCS capability, as described in our letter to you of July 20, 1971, should be performed as soon as practical. Bovever, the proposed plant modifications and Technical Specification changes need not be delayed until these further evaluations are completed.
l Consequently, we have concluded f. hat the proposed modification of the Big Rock Point Nuclear Reactor post-incident spray system will provide fleL23ep(o3 V
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P Consumers Power Company increased safety and that operation with the modified system does not present significant hazards considerations cot described or implicit in your Safety Analysis Report. Taere is reasonable assurance that the baalth and safety of the public will not be endangered by the proposed modifications of tue Big Rock Point Nuclear Reactor.
Accordingly, pursuant to Section 50.59 of 10 CFR Pu t 50, the Technical Specifications of Facility License No. DPR-6 are hereby changed as indicated in Attachment A to this letter.
Sincerely, pri in:1 Signed by Leter A.Mxris Peter A. Horris, Director Division of Reactor Licensing
Enclosure:
Attachment A - Changes to Distribstion Technical Specifications W. Dooly, DR R. Engelken, CO (2) s.,
cc w/ enclosure:
H. Shapar, OGC George F. Trowbridge, Esquire N. Dube, DRL (5)
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Sha.t, Pittman, Potts, Trowbridge si Hadden J. R. Buchanan, ORNL.'
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T. W. Laughlin, DTIE 910 - 17th Street, N. W.
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