ML20151B360: Difference between revisions
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| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS & PERMITS | | document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS & PERMITS | ||
| page count = 2 | | page count = 2 | ||
| project = | | project = TAC:43333, TAC:43335, TAC:48832, TAC:51387, TAC:52156, TAC:54259 | ||
| stage = Request | | stage = Request | ||
}} | }} | ||
| Line 43: | Line 43: | ||
: 1. Change to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications 3.3.3.8, 4.3.3.8.2, Table 3.3-14, Technical Specifications 3.7.9.1, 3.7.9.2, 3.7.9.3, 3.7.9.4, Table 3.7-4 and Table 3.7.5 concerning Fire Detection and Fire Suppression Systems. | : 1. Change to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications 3.3.3.8, 4.3.3.8.2, Table 3.3-14, Technical Specifications 3.7.9.1, 3.7.9.2, 3.7.9.3, 3.7.9.4, Table 3.7-4 and Table 3.7.5 concerning Fire Detection and Fire Suppression Systems. | ||
A. Time Required to Implement This change is to be effective upon NRC approval B. Reason for Change (Facility Change Request 80-263) | A. Time Required to Implement This change is to be effective upon NRC approval B. Reason for Change (Facility Change Request 80-263) | ||
This is to comply with R.' Reid's letter dated September 23, 1980, Log No. 609 C. Safety Evaluation The proposed revision to Appendix A, Technical Specifications 3.3.3.8, 4.3.3.8.2, 3.7.9.1, 3.7.9.2, 3.7.9.3, 3.7.9.4, Tables 3.3-14, 3.7-4 and 3.7-5 incorporate changes made in conjunction with the Fire Protection System Upgrade and to comply with the Nuclear Regulatory Commission's Technical Specification format. | This is to comply with R.' Reid's {{letter dated|date=September 23, 1980|text=letter dated September 23, 1980}}, Log No. 609 C. Safety Evaluation The proposed revision to Appendix A, Technical Specifications 3.3.3.8, 4.3.3.8.2, 3.7.9.1, 3.7.9.2, 3.7.9.3, 3.7.9.4, Tables 3.3-14, 3.7-4 and 3.7-5 incorporate changes made in conjunction with the Fire Protection System Upgrade and to comply with the Nuclear Regulatory Commission's Technical Specification format. | ||
These chan'ges"have been evaluated as an integrated part of the Davis-Besse Nuclear Power Station Unit No. 1, Fire Hazards Analysis Evaluation and Report. The report was submitted to the NRC for their review and was endorsed in Amendment 18 to the Davis-Besse Nuclear Power Station Unit No. 1, License No. NPF-3. | These chan'ges"have been evaluated as an integrated part of the Davis-Besse Nuclear Power Station Unit No. 1, Fire Hazards Analysis Evaluation and Report. The report was submitted to the NRC for their review and was endorsed in Amendment 18 to the Davis-Besse Nuclear Power Station Unit No. 1, License No. NPF-3. | ||
mj d/8 | mj d/8 | ||
| Line 508: | Line 508: | ||
i | i | ||
.. -3 2.4 Loss of DHRS Cooling This transient is mitigated by the DHRS relief valve and should be accounted for in this safety evaluation. A separate license amendment request is being made to the NRC in response to their letter dated June 11, 1980. This change requires operability of at least two of the following ecolant loops in modes 3, 4 and 5. | .. -3 2.4 Loss of DHRS Cooling This transient is mitigated by the DHRS relief valve and should be accounted for in this safety evaluation. A separate license amendment request is being made to the NRC in response to their {{letter dated|date=June 11, 1980|text=letter dated June 11, 1980}}. This change requires operability of at least two of the following ecolant loops in modes 3, 4 and 5. | ||
: 1. Reactor Coolant Loop 1 | : 1. Reactor Coolant Loop 1 | ||
: 2. Reactor Coolant Loop 2 | : 2. Reactor Coolant Loop 2 | ||
Latest revision as of 07:04, 11 December 2021
| ML20151B360 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/26/1980 |
| From: | Crouse R TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML19340E168 | List: |
| References | |
| TAC-43333, TAC-43335, TAC-48832, TAC-51387, TAC-52156, TAC-54259, NUDOCS 8101060588 | |
| Download: ML20151B360 (2) | |
Text
{{#Wiki_filter:# a . l APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NO. NPF-3 FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT NO. 1 Enclosed are forty-three (43) copies of the requested changes to the Davis-Besse Nuclear Power Station Unit No. 1 Facility Operating License No. NPF-3, together with the Safety Evaluation for the requested change. The proposed changes include:
- 1. Changes in Section 3.3.3.8, 3.3.3.8.2, 3.7.9.1, 3.7.9.2, 3.7.9.3, 3.7.9.4 and Tables 3.4-4, 3.7-4 and 3.7-5; (Fire Protection Technical Specifications)
- 2. Changes in Section 3.'4.1.1, 3.4.1.2, 3.9.8.1, 3.9.8.2 and Bases; (Decay Heat Removal Capabilities)
- 3. Changes in Section 3.5.3, 3.4.2, 4.4.2, Figures 3.4.2-a, 3.4.2-b, and Bases; (Iow Temperature RCS Over Pressure Event)
~
- 4. Changes in' Sections 4.1.3.9; (APSR Insertion Limits) and
- 5. Changes in Section 6.5.2.2 and Figures 6-2 and 6-3 (Reorgan-ization at Toledo Edison).
By -[ Vice President, Nuclear Sworn and subscribed before me this 26 day of December, 1980. a -- c c/ bbbdP f) t Publig/ bERRYLENym l
*
- M e - nsn or m l
~ ~ m.n 1 l
1
P'..' . Docket No. 50-346 License No. NPF-3 Serial No. 669 Attachment 1
- 1. Change to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications 3.3.3.8, 4.3.3.8.2, Table 3.3-14, Technical Specifications 3.7.9.1, 3.7.9.2, 3.7.9.3, 3.7.9.4, Table 3.7-4 and Table 3.7.5 concerning Fire Detection and Fire Suppression Systems.
A. Time Required to Implement This change is to be effective upon NRC approval B. Reason for Change (Facility Change Request 80-263) This is to comply with R.' Reid's letter dated September 23, 1980, Log No. 609 C. Safety Evaluation The proposed revision to Appendix A, Technical Specifications 3.3.3.8, 4.3.3.8.2, 3.7.9.1, 3.7.9.2, 3.7.9.3, 3.7.9.4, Tables 3.3-14, 3.7-4 and 3.7-5 incorporate changes made in conjunction with the Fire Protection System Upgrade and to comply with the Nuclear Regulatory Commission's Technical Specification format. These chan'ges"have been evaluated as an integrated part of the Davis-Besse Nuclear Power Station Unit No. 1, Fire Hazards Analysis Evaluation and Report. The report was submitted to the NRC for their review and was endorsed in Amendment 18 to the Davis-Besse Nuclear Power Station Unit No. 1, License No. NPF-3. mj d/8
INSTRUMENTATION FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATOR 3.3.3.8 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-14 shall be OPERABLE. APPLICABILITY: Whenever equipment in that fire detection zone is required to be OPERABLE. ACTION: With the number of OPERABLE fire detection instrument (s) less than the minimum number OPERABLE requirement of Table 3.3.14.
- a. Within 1 hour establish a fire watch patrol to inspect the accessible zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment or, annulus, then inspect the containment at least once per 8 hours or monitor the containment air temperature at least once per hour at the locations listed in Specifications 4.6.1.5.
- b. Restore the inoperable instrument (s) to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission i pursuant to Specification 6.9.2 within the next 30 days outlining j the action taken, the cause of the inoperability and the plans and i schedule for restoring the instrument (s) to OPERABLE status. 1
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. ;
l SURVEILLANCE REQUIREMENTS. . l c-4.3.3.8.1 Each of the above required [ accessible fire detection instruments shall
~
be demonstrated OPERABLE at Telst'one'e per 6 months byderformance. of a CHANNEL FUNCTIONAL TEST. Each of the above required (insecessible/ fire detection instruments shall be demonstrated OPERABLE Et'leaAc'once per 18 months by perfonnance of a CHANNEL FUNCTIONAL TEST. 4.3.3.8.2 The NFPA Standard 72D, Class A, supervised circuits supervision associated with the detector alarms of each of the above required fire ; detection instruments shall,be, demons.trated OPERABLE at least once per ) 6 months. The ginaccessible detector pupervised circuit will be l checked only at thC1Fcal' panel.rj 4.3.3.8.3 T'..e non-supervised circuits, associated with detector alarms, between the local planels and the control room shall be demonstrated OPERABLE at least once per 31 days.
," d m' ' ,l ~ f
, i _
DAVIS-BESSE, UNIT 1 3/4 3 .52 Amendment No.
S'h3et'l of 4 j TABLE 3.3-1k j
/ y; FIRE DETECTION INSTRUMENTS ;j INSTRUMENT LOCATION /
y
- NUMBER MINIMUM INSTALLED INSTRUMENTS OPERABLE HEAT FLAME SMOKE
- 1. Containment fa. FDZ-RCP 1 Elev. 6038 2 0 0 1*
#b. FDZ-RCP 2 Elev. 603' 2 0 0 1*
fc. FDZ-KCP 3 Elev. 603' 2 0 0 1* dd. FDZ-RCP 4 Elev. 603' 2 0 0 1*
#e. FDZ-PZR Elev. 603 ' 2 0 0 1*
#f. FDZ-214 Core Flooding Tank Area Elev. 565' 4 0 0 3*
#g. FDZ-215 Ctat. Letdown Cooler Area Elev. 565' 3 0 0 2*
#h. FDZ-220 Incore Instr. Trench Area Elev. 565' 6 0 0 4*
#1. FDZ-317 Hatch Area alav. 565' 26 0 0 20
- fj . FDZ-410 East Passage Elev. 603'/657' 13 0 0 9*
- 2. Containment Annulus fa. FDZ-A208 Elev. 590' 11 0 0 10 ib. FDZ-236H Elev. 774' 4 0 0 3 dc. FDZ-236L Elev. 590' 10 0 0 9
- 3. Auxiliary Building . -
- a. FDZ 402 #1 Electrical Penetration Rn.
Elev. 603' 14 0 0 12
- b. FDZ 405 Auxiliary Building Storage Ra.
Elev. 603' 1 0 0' 1
- c. FDZ 427 #2 Electrical Penetration Rm.
Elev. 603' 9 0 0 7
- d. FDZ 303 #3 Mechanical Penetration Rs.
Elev. 585' 15 0 0 12
- e. FDZ 304 Corridor to Mach. Pent. Ras #3& 4 Elev. 585' 6 0 0 4
- f. FDZ 310 Passage to BA Mix Tank Elev. 585' ~ 11 0 0 8'
- g. FDZ 312 Spent Fuel Pool Pump Rs. Elev. 585' 7 0 0 4
- h. FDZ 314 d4 !!ach Pent Room Elev. 585' 22 0 0 17
- 1. FDZ 300 Fuel Handling Area Elev. 585' 7 0 0 5
- j. FDZ 209 Corridor to #1 Hech Pent Rm Elev. 565' 4 0 0 3
- k. FDZ 227 Boric Acid Evap Passageway Elev. 565' 8 0 0 6
- 1. FDZ 208 #1 Mach Pent Room Elev. 565' 8 0 0 6 ')
j
- 1) h (Fire Detectors in high radiation areas which are NOT accessible.
8 ad
Shast 2 of 4 TABLE 3.3-14 FIRE DETECTION INSTRUMENTS . INSTRUMENT LOCATION NUMBER MINIMUM INSTALLED INSTRUMENTS OPERABLE
- 3. Auxiliary Building (Continued) HEAT FLAME _ SMOKE .
- m. FDZ 231 Clean Waste Booster Pump Rs.
Elev. 565' 1 0 0 1
- n. FDZ 232 Primary & Deborating Demin Viv Rm. .
Elev. 565' 2 0 0 1
- o. FDZ 234 Boric Acid Evaporator Rm. 1-2 Elev. 565' 1 0 0 1
- p. FDZ 235 Boric Acid Evaporator Rs. 1-1 Elev. 565' 1 0 0 1
- q. FDZ 236 #2 Mechanical Penetration Rm.
Elev. 565' 6 0 0 4 ,
- r. FDZ 240 Boric Acid Addition Tank Rm.
Elev. 565' 7 0 0 5
- s. FDZ 241 Passage to B.A. Addition Tk Rs. ,
Elev. 565' i 3 0 0 2
- t. FDZ 101 Equipment and Pipe Tunnel Elev. 545' 2 0 0 1
- u. FDZ 105 ECCS Pump Room 1-1 Elev. 545' 6 0 0 4 ,
- v. FDZ 110 Corridor to Central Area of Aux Bldg.
i Elev. 545' 6 0 0 5
- v. FDZ 113 Decay Heat Exchanger Pit Elev. 545 - 4 0 0 1
- x. FDZ 115 ECCS Pump Room 1-2 , 1 Elev. 545' 3 0 0 2 j
- y. FDZ -124 Clean Waste Receiver Tank Rs.1-1 ;
Elev. 545' - 4 0 0 4 l
- 4. Auxiliary Building Fan Rooms
- a. FDZ 500 Radwaste & Fuel Handling Area and Air Supply Area Elev. 623' 26 0 0 20.
- b. FDZ 501 Radwaste Exhaust Equipmenc and Main 4
Station Exhaust Fan Room Elev. 623' 27 0 0 22
- c. FDZ 515 Purge and Exhaust Equipment Rs.
Elev. 623' 26 0 0 22
- d. FDZ 516 Non-rad Air and Exhaast Equip. Rs.
Elev. 623' 1 0 0 1
- 5. Control Room Complex l
- a. FDZ 505 Main Control Board cabinets Elev. 623' 25 0 0 3
- b. FDZ 505 Control Cabinet Room Elev. G23' 25 0 0 5 l l
- c. FDZ 505 Computer Room Elev. 623' 25 0 0 1 4 l
4
,Shset 3 of 4 TABLE 3.3-14 FIRE DETECTION INSTRUME!rfS INSTRUMENT LOCATION NUMBER MINIMUM '
INSTALLED INSTRUMENTS OPERABLE
- 6. Cable Spreading Room HEAT FLANE SMOKE 5 0 0 5
- a. FDZ 422A Elev. 613'
- 7. A/C Equipment Roo=
14 0 0 11 !
- a. FDZ 603 Elev. 643' ,
- 8. Diesel Generator Rooms i
**a. FDZ 318 Diesel Generator Rm.1-1 0 5 Elev. 585' 6 0
**b. FDZ 319 Diesel Generator Rs. 1-2
- - Elev. 585' 5 0 0 4 l
- c. FDZ 321A Diesel Generator Day Tank Rs. 1-1 1 0 0 1 ,
Elev. 5 ,
- d. FDZ 320A Diesel Generator Day Tank Rm. 1-2 .
Elev. 5 1 0 0 1 l i
- 9. Battery Rooms . .
FDZ 428A Battery Room B Elev. 603' 3 0 0 2 l a. FDZ 429B Battery Room A Elev. 603' 2 0 0 2 l b. i e - 10. Component Cooling Water Pump Room i 12 0' O 9
- s. FDZ 328 Elev. 585' i
l
- 11. Feed Pump Rooms
- a. FDZ 237 Auxiliary Feed Pump 1-1 4 0 0 3 Elev. 565'
- b. FDZ 238 Auxiliary Feed Pump 1-2 4 0 0 3 Elev. 565' i
- 12. Switchgear Rooms ,
- a. FDZ 324 CD.High Voltage Switchgear 4 0 0 3 Elev. 585'- -
j b. FDZ 325 A High Voltage Switchgear 10 0 0 8 Elev. 585'
- c. FDZ 323 B High Voltage Switchgear ,
14 0 0 11 Elev. 585'
, a i
, Shie,c,4 ef 4 . T/2LE 3.3-14 . -
, FIRE DETFg ION INCTRUMENTS 2 8 ' ,
INSTRUMENT LOCATION, , NUMBER MINIMUM I
, INSTALLED INSTRUMENTS OPERABLE ; ,, if.SwitchgearRooms(Continued) HEAT FLAME SMOKE
- d. FDZ 428 F High Voltage Switchgear i I
Elev. 603' ~ 15 0 0 12 .
- e. FDZ 429 E High Voltage Switchgear' i Elev. 603' 8 0 0 6
- 13. Intake Structure
- a. FDZ 052 Diesel Fire Pump Room Elev. S76' 1 0 0 1
- 5. FDZ 052 Service Water Pump Room Elev. 5768' . 4 0 0 3
- c. FDZ 053 Service Water Viv. Room .
Elev. 565' s 8 0 0 6 4 4 1
*Th3 fire detection instruments located within the Containment are not required to be OPERABLE 'during the performance of Tf?u A Containneut Leakage Rate Tests.
c*Those detectors autoracically actuate fits supipression systems.
PLANT'S QT_EM_S
' 3/4.7.9 FIRE _SUrPRESSION SYSTEMS .
FIRE S*JPPR'SSION WATER SYSTEM LIMITING CONDITION FOR OPF. RATION 3.7.9.1 The fire supprettion water system shall be OPERABLE with: i s ) , )k ~#
- a. (Two) fire suppression pumps, each with a capacity of (2500) spa, .
> with their discharge aligned to the fire suppression header, 1
- b. Separate water supplies, each with a miniaua contained volume of !
250,000 gallons, and
- c. Ju OPERABLE flow path capable of taking suction from the Fire Water Storage Tank and the Intaka Forabay and transferring water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water flow alarm device on each sprinkler or '
hose standpipe, and the last valve ahead of the deluge valve on each deluge or spray system required to be OPERABLE per Speci-fications 3.7.9.2, 3.7.9.3, and 3.7.9.4. p; g) j' . \' - f?PLICABILITY: At all times; /(") hp' n _ ry %fl ACTION: - If*Arj y~
- a. With one pump and/or one water supply inoperable, restore the inoperable equipment to OPERABLE status within 7 days or, in lieu ,
of any other report required by Specification 6.9.1, prepare and , submit a Special Report to the Commission pursuant to Specifica-tied 6.9.2 within the next 30 days outlining the plans and pro-cedures to be used to restore the inoperable equipment to OPERABLE i statua or to provide an alternate backup pump or supply. The : provisions of Specifications 3.0.'3 and 3.0.4 are not applicable. *
- b. With the fire suppression water system otherwise inoperable:
Y 1. Establish a backup fire suppression vater system within 24 j , hours, and . 2. In lieu of any other report required by Specification 6.9.1, submit a Special Report in accordance with Specification 6.9.2: I a) By telephone within 24 hours, \ ! b) Confirmed by telegraph, mailgram, or facsimile transmission ! no later than the first working day following the event, ; and , ] i I D-3 Unit.1 3/4 7-38 Amendment No. f I ) - i !
- - - - . - - - - - , _ , . , , , . ,. -- - - - , _ - - - - . - , . . - , . . . , , - - - - - . - - . - - e -,-,._,,_,_,-,n _ - - -
l j PLANT.SYSTENS I l ACTION: (Continued)
. c) In writing within 14 days following the event, outlining the action taken, the cause of the inoperability and the l plans and schedule for restoring the system to OPERABLE status. l l
SURVEILLANCE REQUIREMENTS I 4.7.9.1.1 The fire suppression water system shall be demonstrated OPERABLE: i
- a. At least once per 7 days by verifying the contained water supply volume.
tf: ,
/ b. At least once per 31 days by starting the Electric Motor Driven g )\ pump and operating it for at least 15 minutes on recirculation !
U flow.
- c. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path is in its correct
/ position.
- d. At least once par si.x months by performance of a system flush.
\ e. At least once per 12 months by cycling each testable valve in the flow path through at least once complete cycle of full travel.
- f. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:
- 1. Verifyids that each automatic valve in the flow path actuates
- s. to its correct position, s\ c p 2. Verifying that each pump develops at least (2500) gym at a system head of (250) feet,
- 3. Cycling each valve in the flow path that is not cascable during plant operation through at least one complace cycle of full travel, and
- 4. Verifying that each fire suppression pump starts (sequentially) to maintain the fire suppression water system pressure greater than or equal to 95 psig,
- g. At least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Associrtion.
3/4 7-39 Amendment No. D-B Unit 1 t a 1
o ' PLANT SYS'TEMS SURVEILLANCE REQUIREMENTS (Continued) 4.7.9.1.2 The fire pump diesel engine shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying:
) 1. The fuel storage tank contains at least 300 gallons of fuel, i and
- 2. The diesel starts from ambient conditions and operate for at
'- least 30 minutes on recirculation flow.
- b. At least once per 92 days by verifying that a sample of diesel fuel ,
from the fuel storage tank, obtained in accordance with ASTM-D270-65, l 1s within the acceptable limits specified in Table 1 of ASTM D975-74 . when checked for viscosity, water and sediment. l
- c. At least once per 18 months, by subjecting the diesel to an inspection l in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.
4.7.9.1.3 The fire pump diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that:
- 1. The electrolyte level of each battery is above the plates, and
- 2. The overall battery voltage, when not discharging, is greater than or equal to 24 volts.
- b. At least o'n ce per 92 days by verifying that the specific gravity is appropriate for continued service of the battery.
- c. At least once per 18 months by verifying that:
i
- 1. The batteries, cell plates and battery racks show no visual ;
indication of physical damage or abnormal deterioration, and ! l
- 2. The battery-to-battery and terminal connection are clean, l l
tight, free of corrosion and coated with anti-corrosion material. l 1 D-B Unit 1 3/4 7-40 Amendment No. 1
)
l
n p,N PLANT SYSTEMS SPRAY AND/OR SPRINKLER SYSTEMS j t
. j L- i LIMITING CONDITION FOR OPERATION 3.7.9.2 The following spray and/or sprinkler systems shall be OPERABLE:
i
- a. Radvasta Exhaust and thin Station Exhaust Fan Rm. 623'
- b. #1 Electrical Penetration Rm. 603'
- c. #2 Electrical Penetration Rs. 603'
- d. #1 Mechanical Penetration Rm. 565' l
- e. #2 Mechanical Penetration Rm. 565'
- f. #3 Mechanical Penetration Rm. 585'
- g. d4 Mechanical Penetration Rm. 585'
- h. Corridor 304 to #3 & #4 Mechanical Penetration Rm. 585'
- i. Corridor 209 to #1 Mechanical Penetration Rm. 565'
- j. Passageway 227 to Boric Acid Evaporators 565'
- k. Passageway 310 and Mix Tank Area 585'
- l. Auxiliary Building Storage Room 405. 603'
- m. Clean Waste tecdiver Tank Rs. 1-1. 545'
- n. Cable Spreading Room 613' l
- o. Boric Acid Evaporator 1-1 565' (Water Curtain)
- p. f4 Mechanical Penetration Rs. 585' (Water Curtain) ,
- q. Component Cooling Water Pump Rm. 585'
- r. Diesel Generator Room 1-1. 385' .
- s. Diesel Generator Room 1-2. 585'
- t. Diesel Generator Day Tank 1-1 595'
- u. Diesel Generator Day Tank 1-2 595'
- v. Service Water Valve Room. 565'
- w. Service Water Pump Room. 575'
- x. Diesel Fire Pump Room. 575' D-B Unit 1 3/4 7-41 Amendment No.
O
_ . _ . _ _ _ ~_ _ .__ ._. . _ . _ _ _ ~ w l , PLANT SYSTEMS
)
[ LIMITIllG CONDITION FOR OPERATION (Continued) : f f APPLICABILITY: Whenever equipment protected by the'hpray/ sprinkler system is required to be operable. ACTION: 4
- a. With one or more of the above required spray and/or sprinkler systems ,
- inoperable, within one hour establish a continuous fire watch with ' backup fire suppression equipment for those areas in which redundant, ' systems or uomponents could be damaged; f or other areas, establish an hourly fire watch patrol. Restore the system to OPERABLE status
) within A4 days or, in lieu of any other report required by Specifica-sd tion 6.9.1, prepara and submit a Special Report to the Commission 9 pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status. ,
- b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.9.2 Each of the above required spray and/or sprinkler systems shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that nach valve (manual,
? power operated or automatic) in the flow path is in its correct position. . .
- b. At least once per 12 months by cycling each testable valve in the
-flow path through at least one complete cycle of full travel.
- c. At laast once per 18 months. ;
' 1. By performing a system functional test which .inclu. des simulated } automatic actuation of the system, and: , , a) Verifying that the automatic valves in the flow path actuate to their correct positions on a simulated test t signal, and i ! j b) Cycling es,ch valve in the flow path ,that is not testable j 2 during plant operation through at least one complete ; j eycle of full travel.
.2. By a visual inspection of the dry pipe spray and sprinkler i headers to verify their integrity, and l i , i l
D-B Unit 1 3/4 7-42 Amendsenc No. l , , 5 l
'. \
PLANT S'tSTDfS SURVEIL"JNCE REQUIJEMENTS (Continued)
- 3. By a visual inspection of each nozzle's spray area to verify '
,a the spray patter n is not obstructed. i
- d. Ar. least once per .') years by performing an air flow test through each open head spray / sprinkler header and verifying each open head )
spray / sprinkler nesszie is unobstructed. J, . t i ! a i
- l i
I i l l j . - l 1 l 1 l I J t l l 4 i l 3/4 7-43 Amendment No, 2 D-B Unit 1 - f l
- i
' 1
. PLAN'T SiSTEM5 ,
FIRE HOSE STATIONS
^
LIMITING CONDITION FOR OPERATION 3.7.9.3 The fire hose stations shown in Table 3.7-4 shall be OPERABLE. APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE. ACTION:
- a. With one or more of the fire hose stations shown in Table 3.7-4 inoperable, route an additional equivalent capacity fire hose to the unprotected area (s) from an OPERABLE hose station within 1 hour if the inoperable fire hose is the primary means of fire suppression; i otherwise route the additional hose within 24 hours. Restore the fire hose station to OPERABLE status within 14 days or, in lieu of a any other report required by Specification 6.9.1, prepara and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability, and plans and schedule for restoring the station to -
OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILI.ANCE REQUIREMENTS 4.7.9.3 Each of the fire hose stations shown in Table 3.7-4 shall be demonstrat,ed OPERABLE:
- a. At least once per 31 days by a visual inspection of the fire hose station to assure all required equipment ir at the station,
- b. At least once per 18 months ' by:
- 1. Removing the hose for inspection and re-racking, and i
- 2. Inspecting all gaskets and replacing any degraded gaskets in the couplings.
- c. At least once per 3 years by:
- 1. Partially opening each bose station valve to verify valve OPERABILITY and po flow blockage. .
- 2. Conducting a hose hydrostatic test at a pressure at least 50 psig greater than the maximum pressure available at any hose s ta tion.
DB Unit 1 3/4 7-44 Amendment No. 4%
- i, '. .
TABLE 3.7-4 i FIRE HOSE STATIONS i ID ELEV. LOCATION
- 1. HSC 21 565' Passageway 227 to B. A. Evaps
- 2. ESC 22 565' Corridor 209 to #1 Mech. Pent. Room (West End)
- 3. HSC 23 585' Hallway to Diesel Generator Rooms
- 4. HSC 24 585' Passageway 310 and Mix Tank Area
- 5. HSC 28 603' #2 Electrical Penetration Room
- 6. HSC 32 555' Walkway Above ECCS Pump Room No. 2
- 7. HSC 33 555' Walh ay Above ECCS Pump Roout No. 1
- 8. HSC 34 565' Passageway to B. A. Addition Tank Room
- 9. HSC 35 565' Corridor 209 to il Mech. Pent. Room (East End)
- 10. HSC 36 585' #4 Mechanical Penetration Room
- 11. ESC 37 585' Corridor 304 (Naxt to PWST Heat Exchanger)
- 12. HSC 38 585' #3 Mechanical Penetration Room
- 13. HSC 39 603' ,f1 Mechanical Penetration Room
- 14. HSC 45 585' Diesel Generator Room 1-2
- 15. HSC 46 585' Diesel Generator Room 1-1
- 16. HSC 47 565' #1 Machanical Penetration Room
- 17. HSC 48 565' #2 Mechanical Penetration Room
- 18. HRK 16 613' Cable Spread Room Stairwell
- 19. HRK 17 623' Control Room Stairwell
- 20. HRK 20 603' RACA Stairwell Entrance 21 HR 52 585' Intake Structure D-B Unit 1 3/4 7-45 Maendment No.
, , PLANT SYSTE!!S
~
. [
YARD FIRE HYDRANTS AND HYDRANT !!OSE HOUSES
< ;; i
[ LIMITING CONDITION FOR OPERATION
}
3.7.9.4 The yard fire hydrants and associated hydrant hose houses shown in
- Table 3.7-5 shall be OPERABLE.
APPLICABILITY: Whenever equipment in the areas protected by the yard fire hydrants is required to be OPERABLE.
~
ACTION:
- a. With one or more of the yard fire hydrants or associated hydrant hose houses shown in Table 3.7-5 inoperable, within 1 hour have sufficient additional lengths of 21s inch diameter hose located in an adjacent OPERABLE hydrant hose house to provide service to the unprotected area (s) if the inoperable fire hydraut or associated hydrant hose house is the primary means of fire suppression; otherwise,;
provide the additional hose within 24 hours. Restore the hydrant or hose house to OPERABLE status within 14 days or, in lieu of any
. other report required by Specification 6.9.1, prepara and submit a Special Report to the Consnission pursuant to Specification 6.9.2 vichin the next 30 days outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the hydrant or hose house to OPERABLE status.
- b. The provisions o'f Specification 3.0.3 and 3.0.4 are not applicable. ,
SURVEILLANCE REQUIRDCNTS .
- 4. 7. 9. 4 Each of the yard fire hydrants and associated hydrant hose houses shown in Table 3.7-5 shall be demonstrated OPERABLE:
- s. At least_ once per 31 days by visual inspection of the hydrant hose i house to assure all required equipment is at the hose house.
- b. At least once per 6 months (once during. March, April or May and once during September, October or November) by visually inspecting each yard fire hydrant and verifying that the hydrant barrel is dry and that the hydrant is not damaged.
l
- c. At least once per 12 mouths by 6 i 1.
Conducting a home hydrostatic test at a pressure at least 50 psig greater than the maximum pressure available at any yard fire hydrant. l
- 2. Inspecting all the gaskets and replacing say degraded gaskets in the couplings.
i 1
- 3. Performing a flow check of each hydrant to verify its OPERABILITT.
)
D-B Unit i 3/4 7-46 Amendment No. . V . i
7 .
-]
TABLE 3.7-5 YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES HYDRANT NUMBER , LOCATION
- 1. HH-4 Middle of South Side of Auxiliary Building
- 2. HH-11 SW Corner of Auxiliary Building -
- 3. HH-13 Westride of "02" Startup Transformer
- 4. HH-15 SW Corner of Service Building #1
- 5. HH-16 NE Corner Turbine Building Near Main Transformer 4
l D-B Unit 1 3/4 7-47 Amendment No.
1 4 Docket No. 50-346 , License No. NPF-3 Serial No. 669 Attachment 3
- 1. Change to Davis-Besse Nuclear Power Station Unit 1 Appendix A Technical Specifications 3.5.3, 3.4.2, 4.4.2, Figures 3.4.2-a, 3.4.2-b and Bases concerning Decay Heat Removal System Relief Valve in Modes 4 and 5.
A. Time Required to Implement This change is to be effective upon NRC approval B. Reason for Change (Facility Change Request 79-391) This is to comply with License Amendmer.t 28 requirements to Davis-Besse Nuclear Power Station Unit 1 License dated July 25, 1980 C. Safety Evaluation See attached l mj d/6 b I a t i t i
. s e s
, SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATIONS FOR DECAY HEAT REMOVAL SYSTEM RELIEF VALVE : AT DAVIS-BESSE NUCLEAR POWER STATION UNIT 1 (DB-1) f l
- 1. Introduction, On 10/1/76, the NRC requested that a low Temperature Overpressure Protection (LTOP) system be provided on the Davis-Besse Nuclear Power Station Unit. I which will prevent increasing pressure transients ;
from exceeding the applicable Technical Specification (3/4.4.9) l reactor coolant system pressure-temperature limits. The primary concern of this requirement is that during startup and shutdown conditions at low temperature, the reactor coolant system (RCS) pressure might exceed the reactor vessel prassure-temperature limits during an overpreseure transient. These limits in the Technical Specifications are established for protection agginst reactor vessel brittle failure. The inadvertent overpressurization i could be gener.ited by any one of a variety of transients. ! For overpressure protection, the Davis-Besse 1 LTOP system relies on the four-inch Decay Heat Removal System (DHRS) relief valve on the DHRS suction line from the RCS when the RCS is in modes 4 and 5 .
, as defined in the Technical Specifications. This is the safety function achieved by this relief valve. Consequently, when the RCS ;
is cooled down, this relief valve should be operable for low temper- < ature overpressure protection. There is only one DHRS relief valve. 1 The NRC has requested Toledo Edison to determine the corrective i action to be taken should this relief valve be inoperable when the ! RC temperature is below 280F (excluding mode 6). ,
- 2. Discussion , ;
Following overpressure transients were considered for the LTOP ; concern in regard ta the postulated inoperable DHRS relief ve.lve. i l
- 1. Core flood tank actuation.
i 2. HPI actuation. ;
- 3. Makeup control valve failing open.
- 4. Loss of DHRS cooling.
l l 2.1 Core Flood Tank Actuation Transient l ! This transient is not considered to be a credible transient i because 1) power will be removed from the core flood tank l
- isolation valve once it is closed upon plant cooldown and depressurization; or 2) the tank will be depressurized.
l 'I l l D . l
., y .
I
.2- - !
l 2.2 HPI Actuation Transient i The mitigation of this transient takes, credit for the DHRS relief valve. Consequently, without the relief valve operable, , the transient must be made incredible or an alternate relief ! mechanism provided. No alternate relief mechanisms are readily j l available; therefore, the transient should be made incredible. l This can be done by adminis'erative control requiring the HPI ! ! system be disabled and ty showing that the HPI system is not required in modes 4 and 5. Present Technical Specifications for DB-1 require that, as a ' minimum, one ECCS subsystem consisting of one HPI pump, one DH
- pump, and gne decay heat removal cooler be operabia in mode 4 l (Tavg <280 F). No requirement is imposed on operability of j one HPI pump in mode 5. Thun, no change is being proposed to j HPI pump operability in mode 5 by this Technical Specification change.
If a LOCA was to occur with RC Tavg <280F, the reactor coolant ' system would initially depressurize in a subcooled mode. When [ l the system pressure reaches the saturation pressure corresponding f to a 280F fluid tem;trature, (approximately 50 psia) the j system will flash and system depressurization will. slow. With
- the occurrence of saturated conditions, ECCS fluid must be i provided to mitigate the transient. Since the system pressure l
Vill decrease $c a low value and decay heat in this mode of plant operation is low, one DR pump will provide sufficient fluid to maintain core cooling. Through a separate license amendment request, changes to the Technical Specifications on j coolant loops required for decay heat removal are also provided . to ensure adequate redundancy in this capability.
! i Pursuant to the above, it is concluded that operability of the HPI pump is sot required in modes 4 and 5.
2.3 Makeup Control Valve Failinz Open Trancient i } The mitigation of this tr.nsient takes credit for the DERS i relief valve. This transient can not be eliminated since the j makeup system can not be isolated. It is needed for RC pump l seal injection and reactor coolant inventory control. Conse- l l qusntly, a different means of mitigating the transient mast be ! provided if the DERS relief valve is inoperable. The corrective i 1 action for this transiunt has been analytically determined by l } B&W and is discussed in Section 3. I 1 l I i .
- j. .
4 l i
.. -3 2.4 Loss of DHRS Cooling This transient is mitigated by the DHRS relief valve and should be accounted for in this safety evaluation. A separate license amendment request is being made to the NRC in response to their letter dated June 11, 1980. This change requires operability of at least two of the following ecolant loops in modes 3, 4 and 5.
- 1. Reactor Coolant Loop 1
- 2. Reactor Coolant Loop 2
- 3. Decay Heat Loop 1
- 4. Decay Heat Loop 2 Thus, if DERS cooling is lost, steam generators will be available for decay heat removal. Moreover, loss of DHRS cooling in addition to the inoperability of the DHRS relief valve assumes more than a single failure of a safety system and thereby exceeds the design criteria for this system.
Based on the above, loss of DHRS cooling transient need not be considered in the present case.
- 3. proposed Corrective Action From the descript. ion provided in section 2 above, it is concluded that the worst case increasing pressure transient is the addition of water to the RCS by the makeup line with the makeup valve full open.
A calculation was performed by B&W to determine the combination of initial pressurizar level and reactor coolant system pressure which would prevent the reactor coolant pressure from. exceeding tha technical specification pressure-temperature limits assuming the entire makeup tank water inventory is emptied into the RCS. The pressure-tenparature limit used in the calculation is the cooldown limit from the Davis-Besse Unit 1 Technical Specifications (Figure 3.4-3). The calculations were made for two different pressure limits from the limit curve. They are 524 psig limit at 190F and 825 psig limit at 194F. The 524 psig limit is applicable to mode 5 and the 825 psig limit is applicable to mode 4. The results are shown in Figures 3.4.2-a and 3.4.2-b. Figure 3.4.2-a shows the allowable
- initial pressurizer level and pressure for 825 psig limit. Figure
- 3.4.2-b is a similar curve for 524 psig limit. The above curves
! are based on the assumption that the Make Up Tank level is less ! than or equal to 73 inches and make up pump suction is not switched to the BWST on low make up tank level. 1
; ,X -
! . i; The proposed method requires limiting RCS pressure and pres-surizer level within the bounds of attached figures so that pressure-temperature limits are not violated on a worst case increasing pressure transient. In addition, to limit the amount of make up water, the ' interlock which transfers make up pump suction to the borated water storage tank on low make up cank level must also be defeated. f This method also requires that the HFI system be isolated so r that the most severe transient would be due to an operating MU pump which adds seal injection and maximum make up water to the RCS. The justification for isolating the RPI system ' (in modes 4 and 5) was provided .in the section 2.2 above. , 4. Conclusion t t t Twc changes to the Technical Specification Appendix A are being ] proposed by this License Amand= mat Request. ; i First, the requirement of having a minimum of one HPI pump operable ! in mode 4 is being deleted from Technical Specification 3.5.3. The justification for this is provided in Section 2.2 of this safety j evaluation. The action statement is also being changed accordingly. ! In addition, to correct the terminology in this section, "Low i i Pressure Injection" in being changed to the more appropriate "Decay Heat". 3 N Through this license ==andmaat request, a new Technical Specification
- is being provided for the Decay Heat Removal System relief valve (DH-4849). The operability requirement of pressuriser code safety ;
valves in modes 4 and 5 is being deleted since DB-4849 provides the
, overpressure protection in these modes. The capability of DB-4849 i to provide LTOP is accomplished by opening the DER isolatio.1 ,
valves DB-11 and DB-12 and res;oving power from thefx motor operators.' 1 Plant cooldown and depressurisation is achieved with DH-11 and DH-12 open and incapable of inadvertent closure. When pressure is decreased to less than 30 pois, the pressuriser steam bubble is . l replaced with nitrogen. d l During plant heatup and repressurization, a steam bubble is drawn ! is l j in the pressuriser greater and than 50 pais. the When nitrogen is vented when RCS temperature is greater RCSthanpressure,l' 280 , J power is restored to the motor operators of DR-11 and DB-12 and the ! valves are closed. j To provide administrative control for the above, the limiting
- condition for operation requires that valves DH-11 and DB-12 be j open in modes 4 and 5 with control power to their operators removed. ;
j If DH-11 and/or DH-12 is not open, the bypass valves DH-21 and DH- t j 23 should be opened within one hour to ensure the operability of a ,
- relief path through the relief valve in order to accomplish over-j pressure protection. l
! l l
- - - . - _ - - - = .-- .- _ _ _ _ _ _ . -- a
8 -
~5-In the event that the relief valve is inoperable, the capability of both high pressure injection (HPI) pumps to inject water into RCS should be disabled. As per Section 2.2 above, it is concluded that disabling this capability in mode 4 (and 5) does not involve an unreviewed safety issue.
In addition to disabling the RPI system, RCS pressure and pressurizer level should be brought within the limits of the applicable curve attached and the make up tank level should be reduced to <73 inches. Also, the interlock which automatically transfers the make up pump suction to the borated water storage tank on low make up tank level '. should also be defeated. A description of this approach was provided in Section 3 of this safety evaluation. It is inferred that if this approach is followed, reactor coolant syntes pressure-temperature limits will not be violated. In addition, the surveillance requirements imposed on DH-4849 and DH-11 and DH-12 (DH-21 and DH-23) are considered to be adequate to ensure the operability of the DB-1 low temperature overpressure protection system. Pursuant to the above, it is concluded that the technical specifica- l tion changes proposed by this license amendment request do not constitute an unreviewed safety question. bt bc/11-15 1 l l 1
! l I
I l i I
. , l. - ?. , EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T;yg < 280'F , i.' LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS sub' system comprised of the following shall be OPERABLE:
. 5: ?:S:'.: M ;h c.
n-. - l c.j- t' R 0I U r -r . CICCC f OneOPERABLE!n'r~,y,'~k*hN et :,, , t, , pump, a)r. s b/. One OPERABLE decay heat cooler, and c f. An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) and transferring suction to
+5e containment emergency sump,.
APPLICABILITY: MODE 4. ACTION: ..
- a. Wih no Eccs salos3skm optAABLE. bemaa e/. +he I mcpvaWlib ef the, bH pump, +he .bH cooler or t:he, l
- f. toe path Aem .the bevate.1 werter stovaye 4anK.
Yes, tom at \ east one. ec.cs sutosyste.m tb o etRAatt ! StoitAs Wifh'th One hour of M4inith6 % Rectoy'coolenb ! 5pte Ty les.s. .bn Leo *F b,3 vat. eqt al.ter nait, heat y:e.mowd rAe4ods.
- b. pt In the event the ECCS is actuated and injects water into the ((.
reactor coolant system, a Special Report shall be prepared and submitted to the Cormission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. SURVEILLANCE REOUIREMENTS ;, 4.5.3 The ECCS subsystems shall be demonstrated OPERABLE per the - apolicable Surveillance Requirements of 4.5.2.
. DAVIS-BESSE, UNIT 1 . 3/45-6
}
REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN 1 LIMITING CONDITION FOR OPERATION i l 1 3.4.2 Decay Heat Removal System relief valve DH-4849 shall be OPERABLE with a lif t setting of ( 330 PSIG* and isolation valves DH-11 and DH-12 open and control power to their valve operators removsd. l APPLICABILITY: MODES 4 and 5. ACTION: A. With DH-4849 not OPERABLE: l
- 1. Make the valve OPERABLE within eight hours; or
- 2. a. Within next one hour, disable the capability of both high pressure injection (HPI) pumps to inject water into the reactor coolant system; and
- b. Within next eight hourst
- 1. Disable the automatic transfer of makeup pump suction to the borated water storage tank on low makeup tank 1 level; and I
\
- 2. Reducemakeuptankleveltof73inchesandreduce reactor coolant system pressure and pressurizar level within the acceptable region on Figure 3.4.2-a (in MODE 4)and 3.4.2-b (in MODE 5).
B. With DH-11 or DH-12 closed, open DH-21 and DH-23 within one hour. C. With the control power not removed from DH-11 and DH-12, reamve the power to the valve operators at the Motor Control Centers within one hour. SURVEILLANCE REQUIREMENTS l 1 4.4.2. Decay Heat Removal System relief valve DH-4849 shall be deter-mined OPERABLE
- a. per the surveillance requirements of Specification 4.0.5.
- b. at least once per 24 hours by verifying eithert
, 1. 1 solation valves DH-11 and DH-12 open with control power l removed from their valve operators; or
- 2. valves DH-21 and DH-23 open.
I. The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. DAVIS-BESSE UNIT 1 ' l
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Reactor Coolant System Pressure - Pressurize" Level Limits for inoperable Decay Heat Renoval System Relief Valve in M " 5 i Figure 3.4.P-4 I I I I f 1 Davis-Besse Unit 1 ) 4 i i i
3/4.4 REACTOR COOLANT SYSTEM 3ASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. With one reactor coolant pump not in operation in one loop. THERMAL POWER is restricted by the Nuctsar Overpower Based on RCS Flov and AXIAL POWER IMBALANCE and the Nuclear Overpower Based on Pump Monitors trip, ensuring that the DNBR will be maintained above 1.30 at the maximum possible TRERMAL POWER for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR equal to 22%, whichever is more restrictive. In MODES 3, 4 and 5, a single reactor coolant loop or DER loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two DHR loops to be OPERABLE. The operation of one Reactor Coolant Pump or one DER pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during. boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction vill, therefore, be within the capability of operator recognition and control. 3/4.4.2 and 3/4.4.3 SAFETY VALVES l The pressurizar code. safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psig. Each safety valve is
! designed to relieve 336,000 lbs. per hour of saturated steam at the valve's setpoint.
During operation, all pressurizar c' ode safety valves must be OPERABLE to pt c'.*' u t the RCS from being pressurized above its safety limit of 2750 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from any transient. 4
; The relief capacity of the decay heat removal system relief valve is adequate to relieve any overpressure condition which could occur during shutdown. In che event that this relief valve is not OPERABLE, reactor coolant system pressure, pressurizer level and make up water inventory is limited and the capability of the high pressure injection system to inject water into the reactor coolant system is disabled to ensure operation within reactor coolant system pressure - temperature limits.
Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. l DAVIS-BESSE UNIT 1 B 3/4 4-1 1
Docket No. 50-346 License No. NPF-3 Serial No. 669 Attachment 4 I. Change to Davis-Besse Nuclear Power Station Unit 1, Appendix A l Technical Specifications 4.1.3.9 concerning APSR insertion limits similar to those for Regulating Rods. A. Time Required to Implement ; This change is to be effective upon NRC approval l B. Reason for Change (Facility Change Request 80-229) This change will make the requirements for checking APSR insertion limits similar to those for regulating rods which will greatly simplify documentation requirements and sub-stantially reduce the likalihood of a missed surveillance test and the associated LER C. Safety Evaluation See attached ,
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v SAFETY EVALUATION FOR CRANGING THE TECHNICAL SPECIFICATION SURVEILLANCE REQUIRDtENTS FOR AXIAL POWER SHAPING RODS INSERTION LIMIT In the present technical specifications for Davis-Besce Nuc' lear Power Station Unit 1 (DB-1), the position of the axial power shaping rod (APSR) group is required to be determined within the APSR position limits at least once every four hours. This surveillance requirement (and the associated limiting condition for operation) was added to the technical specifications through License Amendment 33 for operation of , DB-1 Cycle 2. No surveillance of APSR position was required by technical specifications prior to the issuance of reload license (License Amendment No. 33). The change to the technical specifications proposed here provides for reducing the APSR insertion limit surveillance requirement from four hours to 12 hours when the APSR insertion limit alarm is operable. When this alarm is not operable, the exis ting f requency of four hours is to be observed.
'Ihe technical specification on APSRs ensures that power peaking limits are not exceeded by APSRs being inserted beyond the rod position limits specified in technical specifications Figures 3.1-Sa through d. Thus to ensure operation within the acceptable power peaking limits, theAn APSR alarm position should be verified to be within the position limits.
exists on the computer which alarms the operator whenever the APSR position is not within the acceptable region. Per the tachnical specifi-cations, the operator is required to restore the (inoperable) rod to within limits within two hours. Thus, if the alarm is operable, the operator vill be inanediately apprised of the APSR being positioned out of limits and coYrective action can be taken in time. A 12 hour surveil-lance requirement is therefore cons.idered to be adequate for cases when the APSR position alarm is operable. However, with, the alarm inoperable, the surveillance should be performed every four hours. It is emphasized that the above change vill make the surveillance require-ments for APSR position identir.11 to the surveillance requirement for regulating rod groups position presently prescribed in the technical spectfication. Based on the discussion heretofore, it isconN1udedthattheproposed change to the technical specification does not involve an unreviewed safety que.stion. . t
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. ' kEACTIVITY CONTROL SYSTIMS AXI AL POVER SRA. PING ROD INSERTION LUi1TS .,
LD4ITING Cob"DITION FOR OPERATION 3.1.3.9 The axial power shaping rod group shall be limited in physical inser-tion as shown on Figures 3.1-Sa, -5b, -Se, and -5d. APPLICA3ILITY: MODES 1 and 2*. ACTIg: With the axial power shaping rod group outside the above insertion limits, either:
- a. Restore the axial power shaping rod group to within the limits within 2 hours, or '
- b. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL PokT.R which is allowed by the rod group position l using the above figures within 2 hours, or l
- c. Be in at least HOT STANDBY vithin 6 hours. .
SURVEILLANCE REQUIR. NS 4.1.3.9 The position of the axial povar shaping rod group shall be deter-cised to be within the insertion limits at least once every i ir-- 1 m hours e.xc.ep when h a.xial pow shapinj ved in.sertion Wmit alarm is (noxrable, fHen L veyP3 th4- c you.y to 6e. Wi+in fhe Macyfi'on imi' on cas once e_ve.g 4 hours.
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6 DAVIS-BESSE, UNIT 1 3/4 1-34 Amendment No I
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Attachment 5 { I. Change to Davis-Besse Nuclear Power Station Unit 1. Appendix A l Technical Specifications 6.5.2.2 and Figures 6-2 and 6-3. ! A. Time Required to Implement l This change is to be effective upon NRC approval B. Reason for Change (Facility Change Request 80-249) These changes reflect recent reorganization at Toledo Edison - and changes in the Company Nuclear Review Board. l C. Safety Evaluation This Technical Specification change request reflects only l organizational differences from those currently identifisd in ; the Technical Specifications. There are no physical changes I to the facilities and therefore reduces no margin of safety ; nor does it increase the probability of any accident or malfunction. i 1 mj d/9 j l l l 1
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'ACN NISiRAT1VE CONTR01.S C W OSITION .
6.5.2.2 The Company Nuclear Review Board shall be comeo' sed of the:
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,, .. ..-!.--3. W N.E C; . . ; '. - > ' 4 m n i Member: Vice Pre 3ident, Nuclear Member: Vice President, Energy Supply Mev.ber: General Superintendent, Trans:nission and Substations Mc.ber: Superin'tendent, Davis Sesse Station Member: Director, Nuclear Services Member:Mbay r, Nuclear 3 : .' : : '.r~~ D ;'-'- . "_ Engineering .
Member: Nuclear Engineer, rad"e ' : L'ulcaranyaer.na] Me.v.b e r : Director, Quelity Assurance Member: General Superintendent, fossil Generation facilities
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, a ALTERNAlts
- 6. 5. 2. 3 All alternate members shall be appointed in writing by the CN9S Chairman to serve on a temporary basis; however, no more than two alter-nates shall participate as voting tr. err.bers in CNRS activities at any one time.
CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the CNR8 Chairr.an. to provide expert advice to the CNRB. , 1
.EETING FRE0VENCY l 6.5.2.5 The CNRB. shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and a t least once per six rr.onths therea f ter.
OUORUM , 6.5.2.6 A quorum of CNRS shall consist of the Chairv.an or his designated alternate and at least half of the appointed CNRS mer.bers or their alternates. . No more than a rninority of the quorum shall have line responsibility for operation of the facility, kDthers as deemed advisable by the C.NRS chairman, who are appointed to the Ccr.pany Nuclear Review Soard shall have an aceds .ic degrte in an Engineering or Physical Science field; and in additien, shall have
, a r.inic.um of five years of technical experience, of which 3 minimum of three years shall be in one or more of the areas specified in Spec i fica t ion 6. 5. 2.1.
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I I i - I- ~~ i l ASSISTANT SHIFT C&HF QLHF MAINTmANCE 1&C SUPERVISOR FORDIAN FORDULN FOREMAN MECHANICS 4 i i ! 1 l l STAFF 1 , CL REAGOR TESTERS TES1ERS AyD a OPERATORS z e t_ ELECTRICIANS._ DAVIS-RESSE NUQ. EAR FOWER STATIOst E3JU F FMENT STATI(El ORGANIZATION OFERATORS FIGURE 6.2-2
L .e. ADMINISTRATIVE CONTROLS COMPOSITION 6.5.1.2 The Station Review Boar:d shall be composed of the: Chairman: Assistant Station Superintendent Member: Operations Engineer . Member: Technical Engineer Member: Maintenance Engineer Member: Lead Instrument and Control Engineer Member: Nuclear and Perfonnance Engineer Member: Chemist and Health Physicist u-_w nsaauwi6 e kJ E.31W i i s ss e f Member: Station Superintendent ALTERNATES 6.5.1.3 All alternate members sh'all be appointed in writing by the SRB Chairman to serve on a tempcrary basis; however, no more than two alternates shall participate as voting members in SRB activities at any one time. l MEETING FREQUENCY , 6.5.1.4 The SRB shall meet at least once per calendar month and as convened by the SRB Chairman or'his designated alternate. l l QUORUM - 6.5.1.5 A quorum of the SRB shall consist of the Chainnan or his designated alternate and four members including alternates. RESPONSIBILITIES 6.5.1.6 The Station Review Board shall be responsible for: i
- a. Review of l} all procedures required by Specification 6.8 and I changes thereto, 2) any other proposed procedures or changes thereto as detennined by the Station Superintendent to affect nuclear safety.
- b. Review of all proposed tests and experiments that affect a,, ,/
nuclearsafety.,lg . v c uj b) l-j" DAVIS-BESSE, UNIT 1 6-6 Amendment No.
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