ML20087J799
| ML20087J799 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 03/20/1984 |
| From: | MISSISSIPPI POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20087J781 | List: |
| References | |
| NUDOCS 8403230092 | |
| Download: ML20087J799 (11) | |
Text
-. _ _ .. __
e
- 1. (MP&L P/L Item No. 001)
SUBJECT:
Automatic Depressurization System (ADS) Valve Operability Requirements Technical Specification 3.5.1 and Bases 3/4.5.1 and 3/4.5.2.
DESCRIPTION Revisions to Technical Specifications 3.5.1.a.3 and 3.5.1.b.2 0F CHANGE: and Bases 3/4.5.1 and 3/4.5.2 are proposed to achieve consistency between the technical specifications and the plant's design and accident analyses.
- 1. Specifications 3.5.1.a.3 and 3.5.1.b.2 should be revised to require eight (8) operable ADS valves instead of seven (7). (Page 3/4 5-1.)
- 2. Bases 3/4.5.1 and 3/4.5.2 should be revised to indicate that although the ADS controls eight selected valves, the safety analysis takes credit for seven of these valves.
(Page B 3/4 5-2.)
In addition, in the same Bases section, an editorial revision is proposed to correct a typographical error.
(Page B 3/4 5-1.)
JUSTIFICATION: In the Grand Gulf design the ADS controls eight safety-relief valves. If the High Pressure Core Spray (HPCS) system fails to function properly after a small break loss-of-coolant accident (LOCA), the ADS automatically causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than the maximum value allowed by 10 CFR 50.46, i.e., 2200*F.
Significant input parameters for the Grand Gulf LOCA analyses are presented in FSAR Tabic 6.3-2. This table indicates that eight ADS valves are assurmd operable for the LOCA analyses.
The plant's response to the most limiting small break LOCA is discussed in the FSAR, in response to NRC Question 212.24 (as amended, November, 1981). This particular accident analysis series postulated a HPCS line break coincidental with the worst case single active failure, namely, loss of the Division I diesel generator. Under these assumptions all credit for core spray cooling is removed. The maximum peak cladding temperature (PCT) in these analyses was determined to be approximately 1824*F and occurred at a break size of approximately 0.01 ft2 In that these analyses assumed eight ADS valves to be operable, Technical Specification 3.5.1 should be revised, as discussed above, to be consistent with this assumption, i.e., require eight valves to be operable as a Limiting Condition for Operation. Additional, more restrictive analyses were performed in support of the proposed changes to Bases 3/4.5.1 and 3/4.5.2. These analyses assumed one ADS valve _ inoperable and are discussed below.
8403230092 840320 PDR ADOCK 05000416 p PDR N1sd1
s Action Statement e.1 of the subjnct specification permits one ADS valve to be inoperable for a duration of up to 14 days prior to requiring a plant shutdown. In order to ensure safe operation during this extended period of time with one ADS valve out of service, a series of small break analyses were performed utilizing assumptions identical to those discussed above except that only seven ADS valves were assumed operable.
The maximum PCT developed in these analyses was approximately 2064*F, which occurred at a break size of 0.015 fta. This maximum PCT meets the acceptance criteria of 10 CFR 50.46 (2200*F) and is also below the maximum PCT determined in Grand Gulf's most limiting LOCA analysis, which considers the entire break spectrum (2098*F, FSAR Table 6.3-3).
Bases 3/4.5.1 and 3/4.5.2 should be revised to indicate that although the ADS controls eight safety-relief valves, the safety analyses support an assumption of seven operable ADS valves, consistent with the above discussion. The FSAR will be revised, specifically Section 6.3.3 and the response to NRC Question 211.24, to reflect the above analyses and results associated with the assumption of one ADS valve out of s e rvi ce.
The change to page B 3/4 5-1 is proposed to correct a typo-graphical error and is purely administrative in nature.
SIGN 1FICANT HAZARDS CONSIDERATION:
The proposed changes to the technical specifications have been evaluated to involve no significant hazard, as defined in 10 CFR 50.92. The changes have been proposed to render the technical specifications consistent with the safety analyses and the plant design. The ADS system controls eight safety-relief valves; however, safety analyses have been conducted to demonstrate that plant performance with seven ADS valves operable meets the applicable NRC acceptance criteria, in particular, 10 CFR 50.46.
MP&L considers the proposed changes to be similar to examples of proposed amendments that are not likely to involve a significant hazards consideration. These examples and other guidance were provided by the NRC in the Federal Register, dated April 6, 1983 (48 FR 14870). The specific examples noted are listed as follows: .
(ii) A change that constitutes an additional limitation, restriction, or control not presently included in the technical specification; and (vi) A change which either may result in some l increase to the probability or consequences of a l previously analyzed accident or may reduce in l some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan.
Nisd2
Example (ii) is applicable due to the more restrictive operability requirements imposed by the proposed change to Technical Specification 3.5.1. Example (vi) applies primarily to the analyses performed in support of the propeced changes to Bases 3/4.5.1 and 3/4.5.2. These analyses assumed only seven ADS valves to be operable and demonstrated that the maximum PCT for this series of accidents was clearly within the applicable NRC acceptance criteria.
In that the proposed changes promote consistency between the plant design, the safety analyses, and the technical specifications and in that the supporting safety analyses were carried out in accordance with 10 CFR 50.46 and Appendix K, the proposed changes are not considered to:
f
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2. Create the possibility of an accident of a type different from any evaluated previously; or
- 3. Involve a significant reduction in a margin of safety.
Therefore, the proposed changes to the technical specifica-tions were determined to involve no significant hazards considerations.
P N1sd3
3/4.5 EMERGENCY _ CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 ECCS divisions 1, 2 and 3 shell be OPERA 8LE with:
- a. ECCS division 1 consisting of:
- 1. The OPERABLE low pressure core spray (LPCS) system with a flow path capable of taking suction from the suppression pool and transferring the water through the spray sparger to the reactor vessel.
- 2. The OPERA 8LE low crossure coolant injection (LPCI) subsysten "A" of the RHR systes,with a flow path capable of taking suction from the suppression pool and transferring the water to ths reactor
. b. ECCS division 2 consisting of:
- 1. The OPERABLE low pressure coolant injection (LPCI) subsystems "S" and "C" of the RHR system, each with a flow path capable of taking suction from the suppression pool and transferring the wa r
- 2. L..;yothereactorvusel.
- t 7 OPERA 8LE ADS valves.
i l
- c. ECCS division 3 consisting of the OPERA 8LE high pressure core spray (HPCS) system with a flow path capable of taking suction from the suppression pool and transferring the water through the spray sparger to the reactor vessel. ,
, APPLICABILITY: OPERATIONAL CON 0! TION 1, 2* # and 3*.
ACTION: <
- 1. With the LPCS system inoperable, restore the inoperable LPCS systes to OPERA 8LE status within 7 days.
i l 2. With LPCI subsystem "A" inoperable, restore the inoperable LPCI subsystem "A" to OPERA 8LE status within 7 dayc.
- 3. With the LPCS system inoperable and LPCI subsystes "A" inoperable, restore at least the inoperable LPCI subsysten "A" or the inoperable LPCS system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- 4. Otherwise, be in at least NOT SHUTDOWN within the next 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
"The A05 is not required to be OPERA 8LE when reactor steam done pressure is less than or equal to 135 psig.
- $ee Special Test Exception 3.10.5.
l l GRAND GULF-UNIT 1 3/4 5-1
-- -- . .. - .. _ ~
3/4.5 EMERGENCY CORE COOLING SYSTEM BASES 1
3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN ECCS division 1 consists of the low pressure core spray system and low pressure coolant injection subsystem "A" of the RHR system and the automatic depressurization system (ADS) as actuated by trip system "A". ECCS division 2 consists of low pressure coolant injection subsystems "B" and "C" of the RHR system and the automatic depressurization system as actuated by trip system "B".
The low pressure core. spray (LPCS) system is provided to assure that the pygn3 core is adequately cooled ? h2; a loss-of-coolant accident and, together l with the LPCI system, provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the ADL The LPCS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.
The surveillance requirements provide adequate assurance that the LPCS system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hansier damage to piping and to start cooling at the earliest moment.
The low pressure coolant injection (LPCI) mode of the RHR system is provided to assure that the core is adequately cooled following a loss-of-coolant accident. The LPCI system, together with the LPCS system, orovide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS.
The surveillance requirements provide adequate assurance that the LPCI system vill be OPERA 8LE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling. at the earliest moment.
ECCS division 3 consists of the high pressure c' ore spray system. The I high pressure core spray (HPCS) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the eveat of a small break in the reactor coolant system and loss of coolant which doec not result in rapid depressurization of the reactor vessel. The HPCS system permits the reactor to be shut down while maintaining sufficient reactor
- vessel water level inventory until the vessel is depressurized. The HPCS system operates over a range of 1160 psid, differential pressure between reactor vessel and HPCS suction source, to O psid.
The capacity of the system is selected to provide the required core cooling.
The HPCS pump is designed to deliver greater than or equal to 1440/5010 gpa at i
i differential pressures of 1160/200 psi. Initially, water from the condensate storage tank is used instead of injecting water from the suppression pool GRAND GULF-UNIT 1 B 3/4 5-1
e 3/4.5 EMERGENCY CORE COOLING SYSTEN s BASES l
ECCS-OPERATING and SHUTDOWN (Continued) -
into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.
With the HPCS system inoperable, adequate core cooling is assured by the OPERASILITY of the redundant and diversified automatic deoressurization system and both the LPCS and LPCI systems.' . In addition, the rea: tor core isolation cooling (RCIC) system, a system for vhich no credit is ta4en in the safety analysis, will automatically provide'iiiakeup at reactor operating pressures on a reactor low water level condition. The HPCS out-of-service period of 14 days is based on the demonstrated OPERASILI1Y of redun:! ant ind diversified lowpressurecorecoolingsystems.3-The surveillarte requirements provide adequate assurtnce that the HPCS system will be OPERA 8LE when required. Although all active conponents are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor shutdown, The pump discharge alping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.
Upon failure of the HPCS sysdem to function properly after a small break loss-of-coolant accident, the automatic depressurization rystes- (ADS) auto-matica11y causes selected safety relief valves tc open, ci: pressurizing the reactor so that flow from the , low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less ti.an 2230*F. ADS is conservatively required to be OPERA 8LE wnenever reactor veel 3ressure exceeds 135 psig even though low pressure core cooling systems pre. Je adequate core cooling up '.o 350.psig. .
eght ADS automatically controls r;= s, elected safety-relief valves although the safety analysis cnly takes credit f( 1 valves. It is therefore appro-priate to permit one valve to be out-of-serv for up to 14 dajs without materially reducing systen reliability. Seven 3/4.5.3 SUPPRESSION POOL The supression pool is required to be OPERABLE as part of the ECCS to
' ensure that a sufficient supply of water is available to the HPCS, LPCS and
}' LPCI systems in the event of a LOCA. This limit on suppression pool minimum water volume ensures that sufficient water is available to permit recirculation cooling flow to the core. The OPERA OPERATIONAL . CONDITIONS 1, 2 or 3 is ,8ILITY required of the suppression by Specification pool in 3.6.3.1.
Repair work might require making the suppression pool inoperable. This specification will permit those repairs tosbe made and at the same time give 1ssurance that the irradiated fuel has art adequate cooling water supply when ,
the suppression pool must be made inoperable, including draining, in OPERATIONAL
~
COMDITION 4 or 51 {
In OPERATIONAL CONDITION 4 and 5 the suppression chamber minimum required water volume is reduced because the' reactor coolant is maintained at or below
~
1 200*F. Since piessure suppression is not required below 212*F, the minimum l
required water volume is based on NPSH, rectreulation volume, and vortex preven-tion plus a l'2" safety margin for conservatism.
GRAND GULF-UNIT 1 s B 3/4 5-2
- 2. (MP&L P/L Item No. 005)
SUBJECT:
Technical Specification Tables 3.3.2-1 and 4.3.2.1-1 ftem 4.h.
pages 3/4 3-12, 3/4 3-14, 3/4 3-22, and 3/4 3-23a.
DESCRIPTION This proposed change to Technical Specitication Table 3.3.2-1 0F CHANGE: iten 4.h, consists of three parts:
- 1. Changing the minimum operable channels per trip system from "NA" to "1" for the SLCS initiation RWCU isolation function;
- 2. Replacing Operational Condition 3 with Operational condition 5 and adding footnote "##" which requires the SLCS initiation RWCU isolation function to be operable iu Operational Condition 5 only when control rods are withdrawn;
- 3. Replacing present Action 27 for the SLCS initiation RWCU isolation function with new Action 30 on Table 3.3.2-1, which requires the affected SLCS pump to be declared inoperable whenever the associated SLCS initiation instrumentation is inoperable.
The proposed change described in 2. above shculd also be provided in Technical Specification Table 4.3.2.1-1, item 4.h.
JUSTIFICATION: In order to enter Action b or e of Technical Specification 3.3.2, the number of operable channels must be less than that required by the Minimum Operable Channels per Trip System in Table 3.3.2-1. Presently no channels are required to be operable for the RWCU isolation function upon SLCS initiation.
Therefore, the action statements for Technical Specification 3.3.2 can never be entered for this RWCU isolation function.
The SLCS initiation RWCU isolation function design consists of one channel per trip system. Requiring one channel per trip system to be operable will reflect system design and require entering appropriate action statements for inoperable channels.
The present applicable operational conditions for the SLCS in Technical Specification 3/4.1.5 are not the same as the ones for the SLCS initiation RWCU icolation function in Tables 3.3.2-1 and 4.3.2.1-1. Technical Specification 3/4.1.5 requires the SLCS to be operable in Operational Conditions 1, 2, and 5* where the "*" footnote applies with any control rod withdrawn but is not appif cable to control rods removed per Technical Specification 3.9.10.1 or 3.9.10.2. The applicable operational conditions for the SLCS initiation isolation funecion in Tables 3.3.2-1 and 4.3.2.1-1 are 1, 2, and 3. The applicable operational conditions for these specifications shculd be identical since the specifications involve the same l SLC system. Also, since control rods cannot be pulled in l Operational Condition 3 the SLCS and the associated RWCU l isolation function are not required to be operational. This part of the proposed Technical Specification change deletes L/N2sd1
v w.
\t t .
5
> l Operational Conditioak3 and adds 5ff to item 4.h of Tables 3.3.2-1 and 4.3.2.1-1 (the "*" footnote from Technical Specification 3.1.5 and the new "##" footnote are identical).
Present Action 27 for the SLCS initiation RWCU isolation function in Table 3.3.2-1 requires the RWCU isolation valves to be closed and the RWCU system to be declared 1.. operable when-ever the associated SLCS initiation instrumentation is inoperable. This action can have .in adverse impact on reactor water quali;y at power. The new proposed Action 30 for the SLCS initiation function will require that the affected SLCS
_ pump be declared inoperabic. This new Action 30 will then l require entry into Action a.1 or b.1 of Technical Specificacion 1.1.5 to determine the appropriate action requirements for an inoperable SLCS pump. There are two SLC systems in the Grand
.Culf design. SLCS "A" initiation will close RWCU outboard isolation valve lG33-F004 SLCS "B" initiation will close RWCU inboard isolation valves G33-T001 and outboard isolation valve G33-F251. Initiation of either SLCS "A" or "B" will cause iselation of the RWCU system; therefore, one SLCS pump can be
, declared inoperable without adversely af fecting the isolation capability of the RWCU, since the RWCU system will isolate if the remaining SLC system is initiated.
SICNIFICANT HAZARDS CONSIDERATION:
Requiring one minimum operabit channel per trip system for the SLOS initiation isolation function constitutes an additional limitation not presently in the Technical Specifications.
The change to the applicable operational conditions is made to i promote consistency among Technical Specification 3.1.5, Table 3.3.2-1, and Table 4.3.2.1-1.
Changing the action statement for the SLCS initiation function from the present Action 27 to the new Action 30 on Table 3.3.2-1 is made to ensure reactor water quality, by not isolat-ing the RWCU, and still retain the isolation function from the redundant SLCS system. Also, since the affected SLCS pump is declared inoperable by the new Action 30 the SLCS initiation function must be restored within the time constraints of the action statements of Technical Specification 3.1.5.
The proposed changes do not:
- 1. Involve significant increase in the probability or consequences of an accident previously evaluated; or
- 2. Create the possibility of an accident of a type different from any evaluated previously; or
- 3. Involve a significant reduction in a margin of safety.
Therefore, the proposed changes do not involve a significant hazards consideration.
l l
L/N2sd2
TABLE 3.3.2-1 (Continued)
ISOLATION ACTUATION INSTRUMENIAT'10N .
VALVE GROUPS MININUM APPLICA8tE OPERATED BY OPERA 8LE CHANNELS OPERATIONAL Q CONDITION ACTION g TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEN (b)
- 4. REACTOR WATER CLEANUP SYSTEM ISOLATION (Continued) ,
- f. Main Steam Line Tunnel 8 1 1,2,3 27 Ambient Temperature - High .,
- g. Main Steam Line Tunnel a '
Temp. - High 8 1 1,2,3 27
- h. SLCS Initiation 8 III 4W l 1, 2, t5 N-30 l
- 1. ManualInidiation 8 2 1,2,3 26 w 5. REACTOR CORE ISOLATION COOLING SYSTEN ISOLATION '
k KIC Steam Line Flow - High 4 1 1,2,3 27 w
a.
K IC Steam Supply 4, 9g*
! b.
Pressure - Low ,3 1 1,2,3 27
- d. RCIC Equipment Room Ambient ,
hemperature - High 4 1 1, 2, 3 27 !
i i
- e. RCIC Equipment Room a Temp.
- High 4 .,
1 1, 2, 3 27 Main Steam Line Tunnel f.
Ambient Temperature - High 4 1 1, 2, 3
- 27 i g. Main Steam Line Tunnel 27
- A Temp. - High 4 1 1,2,3
- 9 *
- h. Main Steam Line Tunnel ,
.27 Temperature Timer 4 1 1,2,3 i
9
INSTRWENTATIM TABLE 3.3.2-1 Is0LATT511 ETCHf5il (Ccctinued)
INSTauMENTATION ACTION
- Be in at least HOT SWTDOW within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOW :
ACTION 20 l within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.- l Close the affected system isolation valve (s) within one hour er:
ACTION 21
- a. In OPERATIONAL CONDITION 1, 2, or 3, be in at least NOT l $NUTDOW within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SWTDOW !
within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b. In Op cational Condition *, suspend CORE ALTERATIONS, handling of irradiated fuel in the primary containment and operations with a potential for draining the reactor vessel.
ACTION 22 -
Restore the manual initiation function to OPERA 8LE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDOW within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (
and in COLD SHUTDOW within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. l
- ACTION 23 -
Be in at least STARTUP with the sesociated isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least NOT SHUTDOWh within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> i and in COLD SHUTOOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 24 -
Se in at least STARTUP within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
ACTION 25 - Establish SECONDARY CONTAIMENT INTEGRITY with the staney gas treatment system operating within one hour.
ACTION 24. -
Restore the manual initiation function to OPERA 8LE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or close the affected system isolation valves within the next hour and declare the affected system inoperable.
Close the affected system isolation valves within one hour ACTION 27 -
and declare the affected system inoperable.
ALTION 28 - Lock the affected system isolation valves closed withir. one hour and declare the affected syytes inoperable.
ACTION 29 - Close the affected system ifolation valves within one hour and l
declare the affected sy the or component inoperable or:
- a. In OPERATIONAL TION 1, 2 or 3 be in at least HDT SHUTDOW within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWP within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b. In 0PERATIONAL CONDITION # su: pend CORE ALTERATIONS and opera-
- Deci.=.I'E"Nr="d'.**."*Nek'pu drajn(ng n reactor vesset g ACTION 30 NOTES
- When handling irradiated fue N the primary or secondary containment and durind CORE ALTERATLONS and operations with a potential for draining the reactor vessel.
- During CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
- (a) See Specification 3.6.4 Table 3.6.4-1 for valves in each valve group. .
(b) A channel may be placed in an inoperable status fer up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the trippad con-dition provided at least one other 0PERA8LE channel in the same trip system is monitoring that parameter.
(c) Also actuates the standby gas treatment system.
(d) Also actuates the control room emerg2ncy filtration system in the isolation mode of operation.
(e) Two upscale-Hi Hi, one upscale-Hi Hi and one downscale, or two downscale signals from the same trip system actuate the trip system and initiate i isolation of the associated containment and drywell isolation valves.
GRA M GULF-UNIT 1 3/4 3-14 j
w wm, m e.m.unoew mme Norerniumw m = 'am
- 3. 9. s. o.1. . at 3.T.Ic.2 Manoso Pun 5 reeir swr ew
TAP.LE 4.3.2.1-1 (Continued)'
c3 9
8 ISOLATION ACTUATION INSTRUMENTATION SURVEILEANCE REQUIREMENTS E CHANNEL OPERATIONAL Gi
' CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN %dHICH TRIP FUNCTION TEST _ CALIBRAT!DN SURVEILLANCE REQUIRED
! CHECK _
- 4. REACTOR WATER CLEANUP SYSTEM ISOLATION (Continued) [
[
- f. Main Steam Line Tunnel Ambient ~
Temperature - High S' , M R 1,2.3
- g. Main Steam Line Tunnel 4 a Temp. - High S M R 1, 2, 3
- h. SLCS Initiation NA M(b) M 1,2 M *
! 1. Manual Initiation MA MI ") NA 1, 2, 3 l 4
. i g .
- 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION b.
1, 2, 3 y a. RCIC Steam Line Flow - High S M R
- b. RCIC Steam Supply Pressure -
S M R 1, 2, 3
. Low
- c. RCIC Turbine Exhaust Olaphraga Pressure - High S M R 1, 2, 3
- d. RCIC Equipment Room Ambient .
Temp <.rature - High 5 M R 1, 2, 3
- e. RCIC Equipment Room a Temp. - '
High 5 M R 1,2,3
- 7
- f. Main Steam Line Tunnel Ambient Tmperature - High 5 M R ,
1, 2, 3 ,
- g. Main Steam Line Tunnel M R 1, 2, 3 i a Temp. - High 5
i TABLE 4.3.2.1-1 (Continued) t ISOLATION ACTUATION INSTRLeelTATION SURVEILLANCE REQUIREENTS E CHANNEL . OPERATIONAL q CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN tellCN CHECK TEST CALISRATION. ENIVQLLMIg_M g 1Rg1
! TRIP FUNCTION I
[ 6. RNA SYSTEN ISOLATION (Continued)
- e. Dryuell Pressure - High 5 M R 1. 2. 3
- f. Manual Initiation MA MI *) NA 1.2.3 l
i
' wn handling irradiated fuel in the primary or secondary containment and during CORE ALTERATION 6 and operations with a potential for draining the reacter vessel.
- The low condenser vacuum M51V closure may be manually bypassed during reacter SitiT00tel or for raector
} STARTUP when condenser vacuum is below the'?. rip setpoint to allow opeqing of the M51Vs. The manuel bypass shall be resoved when condenser vacuum exceeds the trip setpoint.
4 #0uring CORE ALTERATION and operations with a potential for draining the reacter vessel.
, y (a) Manual initiation switches shall be tested at least once per 18 months during shutdoun. All ether 1
circuitry associated with unual initiation shall receive a CHANNEL FUNCTIONAL TEST st least once per 31 days as part of cir.:ultry required to be tested for automatic system isolation.
- (b) Each train er logic channel shall he tested at least every other 31 days.
(c) Calibrate trip unit at least once per 31 days.
R** wT"*R*"! N
- r A PP a c ^ *> " ~ro e o n 'r a o u Ro*s
- ## sWsnou rn nur conmot-sc> ren Sreenrocerso n .xp. to. L o n 3.9.to.z . l f
l 1
i
- _ _ _ __ _ _ _