ML20246D380
| ML20246D380 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/21/1989 |
| From: | TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML20246D360 | List: |
| References | |
| 1661, NUDOCS 8908280025 | |
| Download: ML20246D380 (12) | |
Text
._ - __ - - _ _ _ _
- INDEI
- s
't,IMITING CCCITIONS FOR OPEPATION AND SURVEILLANCE REQUIREMENTS e>mee
$"RW8 PAGF *$$$E
.SECTION -
- i ~ $ " l
-$Fz!
3/4.2 POWER DISTRIBtJTION LIMITS " ?r 0 ?r 3/4 2-1 '~ 4 I 4 3/4.2.1 - AII AL POWER IMBALANCE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - u,
'fuCLEAR HEAT FLUX HOT g:ig ?
3/4.2.2 3/4 2-5 'F W CMANNEL FACTOR - F . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ww !
Q 3/4.2.3 . NUCLEAR ENTHALPY RISE y HOT CHANNEL FACTOR e 3/42-7 t.H*"*'******""*"*""*"
3/4.2.4 QUADRANT POWER TILT..................................
3/42-9 3/4.2.5 DN5 PARAMETERS....................................... 3/4 2-13 3/4.3 INSTEUMENTAT70N REACTOR PROTECTION SYSTEM INSTRUMENTATION............ 3/4 3-1 3/4.3.1 SAFETT SYSTDiS IiWidjMENTATION 3/4.3.2 Safe ty Features Actua ti on Sys tem. . . . . . . . . . . . . . . . . . . . . 3/4 3-9 S te am a nd Feed Rupture Control Sys tem. . . . . . . . . . . . . . . . 3/4.3-23 Anti cipatory Reactor Trip Sys tem . . . . . . . . . . . . .. . . . . . 3/4 3-30a l
3/4.3.3 MONITORING INSTRUMDiTAT:CN Radi ati ca Moni tori ng I ns tr.=scu ti on. . . . . . . . . . . . . . s . . 3/4 3-31 Incore Detectors. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~ 3/4 3-35 Sei s=ic Instra.entation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-37 Meteorol ogi cal Instr.:menta ti on. . . . . . . . . . . . . . . . . . . . -. . 3/4 3-40 Remote Shutdown Ins trumenta ti on. . . . . . . . . . . . . . . . . . . . . . 3/4 3-43 cniturms Post-AccidentA instrumentation........................ 3/4 3-45 0.l ori ne Detecti on 5ystens . . . .. . . . . . . . . . . . . . . . . . . . . . . . 3/4 3- 51 Fi re Dete cti on Instrumentati o n. . . . . . . . . . . . . . . . . . . . . . . . 3/4 3- 52
-Radioactive Liquid Effluent Monitoring Instrumentation 3/4 3-57 J
3/4.4 REACTDR CDOLANT SYSTIM
$o 3/4.4.1 COOLANT LOOPS AND CDOLANT CIRCULATION
'.-mNofL a
mo Startup an d Power 0pe rati on. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-1 oo
- "g8 Shutdown and Ho t Standby . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/44-2 SU 3/4.4.2 SAFETY VALES - SHUTD0WN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4- 3 oo
' on p 3/4.4.3 SAFETY YALYES AND ELECTRO)% TIC RELIEF YALVE - OPERATING 3/4 4-4 m
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&Q m t'fk DAVIS-BESSE, UNIT 1 IV Amendment No. II* 86
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N Docket Number 50-346
. *J , License Number NPF r l Serial. Number 1661L
- Attachment'1 -
Page 19 ic;l l' INSTRUMENTATION
' POST-ACCIDENTkNTRUENTTON-LIMITING CONDITI'ON FOR OPERATION 3.3.3.6 The post-accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.
APPLICABILITY: MODES 1,!2 and 3.
. ACTION:
the Anmm Cknnek O?F.RA6W ,nguten},j[rn
- a. With the number of 0PERABLE post-accident monitoring channels l ess.tha required by Table 3.3-10, either restore the inoperable channel to OPERABLE status within 30 days, or be in HOT. SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.. The provisions of Specification 3.0.4.are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.6 ~Each post-accident monitoring instrumentation channel'shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-10.
DAVIS-BESSE, UNIT 1 3/4 3-46
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3/4 3-50 Amenament .'io._gX,78 QMIS BESSE, UNIT ) i f
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Docket Number 50-346 License Number NPF-3
'o' Serial Number 1661
- c, Attachment 1-Page 26 3/4.3 INSTRUMENTATION BASE 3 -
1.
REMOTE SMUTDCWN INSTRUMENTATION (Continued)
NOT 57ANDSt of the facility from locations outside af the control room.-
This capatt11ty is required in the event control room habitability is lost.
McAlTb2l%
3/4.3.3.6 POST-ACCIDENT fM57RUMEN,'47 4
.xcMt TIM The OPERA 8!LITY of the post-accident 4:nstrumen)tation ensures that sufft-cient information is available on selected plant parameters to monitor and assess these variables following an accident. ;
3/4.'3.3.7, CMLORINE DETECTION SYSTEM 3 The OPERABILITY of the chlorine detection systems ensures that an acci-dental chlorine release will be detected promptly and the control room i 4111 be isolated automatically. The control room ventilation system will be started manually in the recirculation mode to provide the required protection. The chlorine detection systems recuired by this j specification are consistent with the recommendations of R69ulatory Guide 1.95. " Protection of Nuclear Power Plant Centrol Room Operations Against an Accicental Chlorine Release." February 1975.
3/4.3.3.8 FIRE DE*E TION INSTRLHENTATION Operability of the fire detection instrumentation ensures that adecuate warning capability is available for the prompt detection of fires.
This capaci11ty is required in order to detect and locate fires in their early stages. Promet detection of fires will reduce the potential for damage to safety related etutament and is an integral element in the overall facility fire protection pro' gras.
In the event that a portion of the fire de' taction instrumentation is inopersole, the establishment.of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable ;
instrumentation.fs restored ts CPERASILITY.
I l
DAVIS-BESSE. UNIT 1 2 3/4 3-3 Amenchnent No.,3$R,58-
o
.i?.'- Docket Number 50-346 o- License Number NPF-3 Serial Number 1661
- '"""bONTAINMENTSYSTEMS e 7 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN ANALYZERS LIMITING CONDITION FOR OPERATION 3.6.4.1 Two independent containment hydrogen analyzers shall be OPERABLE.
APPLICABILITY: MODES 1 and 2.
ACTION:
With one hydrogen analyzer. inoperable, restore the inoperable analyzer to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS
-7 4.6.4.NEach hydrogen analyzer shall be demonstrated OPERABLE at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gases containing:
N _ __ - - _ ~ ~ _ - ~ _
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gf x CARudEL CH6CK, DAVIS-BESSE, UNIT 1 3/4 6-23
, ' Docket Number 50-346 License Number NPF-3 T
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<' , Serial Number 1661 Attachment 1 l l00 n -
CONTAINMENT SYSTEMS BASES 3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the Hydrogen Analyzers, Containment Hydrogen Dilution System and Hydrogen Purge System ensures that this equipment l will be available to maintain the maximum hydrogen concen'tration within the containmen't vessel at or 'below three volume percent following a LOCA.
The two redundant Hydrogen Analyzers determine the content of hydro-gen within the containment vessel.
The Containment Hydrogen Dilution (CHD) System consists of two full capacity, redundant, rotary, positive displacement type blowers to supply air to the containment. The CHD System controls the hydrogen concentra-tion by the addition of air to the containment vessel, resulting in a pressurization of the containment and suppression of the hydrogen volume fraction.
The Containment Hydrogen Purge System Filter Unit functions as a backup to the CHD System and is designed to release air from the con-tainment atmosphere through a HEPA filter and charcoal filter prior to discharge to the station vent.
3/4.6.5 SHIELD BUILDING 3/4.6.5.1 EMERGENCY VENTILATION SYSTEM The OPERABILITY of the emergency ventilation systems ensures that containment vessel leakage occurring during LOCA conditions into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is neces-sary to meet the assumptions used in the safety analyses and limit the site boundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions.
I
' DAVIS-EESSE. UNII' 1 B 3/4 6 4 Amendment No. 56
Docket Numbar 50-346 l License Number ~NPF-3 E Serial Number'1661 0 '
.' Attachment 2 i Page 1 SIGNIFICANT HAZARDS CONSIDERATION Description of Proposed Technical Specification Changes The. purpose of this significant hazards consideration is to review proposed changes to the Davis-Besse Nuclear. Power Station Technical Specifications (TS)-
, 3/4.3.3.6 (Post-Accident Instrumentation), Table 3.3-10 (Post-Accident Monitoring Instrumentation), Table 4.3-10 (Post-Accident Monitoring Instrumentation Surveillance Requirements), Technical Specification 3/4.6.4.1~
(Combustible Gas Control - Hydrogen Analyzers), Bases 3/4.3.3.6 (Post-Accident Instrumentation) and Bases 3/4.6.4 (Combustible Gas Control). This request proposes the removal of all line items using the term " status" in Tables 3.3-10 and 4.3-10. To eliminate an inconsistency in.the mode applicability
'for the hydrogen analyzers, the application proposes the removal of the line item on_ containment. vessel hydrogen from the two tables and the addition of a monthly Channel Check'to Specification 3/4.6.4.1. In addition, several administrative changes have been made for clarification purposes. The Technical Description (Attachment 1) discusses these changes in detail.-
Significant Hazards Consideration The Nuclear' Regulatory Commission has provided standards in 10CFR50.92(c) for determining whether a significant hazard exists. A proposed amendment to an Operating License for a facility involves no significant hazards if operation of the. facility in accordance.with the proposed changes.vould not: 1) Involve
.a significant increase in the probability or consequences'of an accident previously evaluated; 2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) Involve a significant L
reduction-in a margin of safety.
The proposed changes do not involve a significant hazards consideration because the operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in accordance with these changes vould:
- 1. Not involve a significant increase in the probability or consequences of an accident previously evaluated because no hardware changes are involved and.all accidents remain bounded by previous analyses and no new malfunctions have been created. [10CFR50.92(c)(1)]
- 2. Not create the possibility of a new or different kind of accident from any accident previously evaluated because the accident conditions and assumptions are not affected since no hardware changes are being made. On matters related to nuclear safety, all accidents are bounded by previously analysis and no new malfunctions are involved. [10CFR50.92(c)(2)J l
- 3. Not involve a significant reduction in a margin of safety since the information provided by the line items being deleted is obtained by.
l available instrumentation or annunciation and, therefore, the information available has not been reduced. [10CFR50.92(c)(3)]
Conclusion Based on the discussion above, it is concluded that the proposed changes do not involve a significant hazards consideration.
l'
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