ML20043F980
| ML20043F980 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 06/06/1990 |
| From: | Odell W GULF STATES UTILITIES CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20043F981 | List: |
| References | |
| GL-87-09, GL-87-9, RBG-32951, NUDOCS 9006180381 | |
| Download: ML20043F980 (44) | |
Text
{{#Wiki_filter:. GULF STATES UTILITIES COMPANY 10 1 R DF 'O ST A h0N PD51 04 K1 boy;?O ST f n ANCGylLLE LOufuaN4 7/'?b M6 ItCSI
^
ATH A cot >[ ti i4 0 3f. 6994 June 6 , 1990 Rac-32951 File No. G9.5, G9.42 i U. S. Nuclear Regulatory Canmission Docunent Control Desk Washington, D.C. 20555 Gentiment River Bend Station - Unit 1 Docket No. 50-458 Please find attached Gulf States Utilities Company's (GSU) response to your request for additional information dated July 31, 1989 regarding GSU's anendment request dated September 30, 1988 (reference RBG-28910) . GSU's letter requested the amendment of Facility Operating License NPF-47, Attachment A, 'Ibchnical Specifications, to imp 1 ment the NRC Generic Letter 87-09. This sutmittal provides GSU's evaluation of the new flexibility, ide~ ms those node changes will not be nade while depending on the a lon and the administrative controls to restrict the use of this change. As requested by the NRC the justifications for proplanned use of the proposed specification 3.0.4 will be subnitted separately but will depend on the information contained ' in this letter and approval of this request. This subnittal also' , contains a revised mark-up of the affected technical specifications. . Please contact Mr. L. A. F2igland at (504) 381-4145 if additional information-is required. Sincerely,
. . 'e1 I Manager-Oversight River Bend Nuclear Group 5'/ /pg Attachment 9006180381 900606 Op[
~
em nDocx osooo ee P PDC m lll
e ' .. l cc: U. S. Nuclear. Regulatory Cm mission-4 611.Ryan Plaza Drive, Suite 2000 ;
, Arlington, TX 76011 'i Mr. Walt Paulson.
-U. S. Nuclear Regulatory Cm mission !
[ ~ I
> i One White Flint' North. '
11555 Rockville Pike
- l. Rockville, ND-~: 20852 ,
a t l-F .NRC Resident Inspector .p L Post Office Box 1051 i f,; St.' Francisville, IA 70775 ! I 'Mr. William H. Spelli Administrator L
' Nuclear. Energy Division p;
J touisiana Department of Environmental Quality ! Post Office Box 14690 ; Baton Rouge, IA 70898 j h p s + 4 1 j.; i a i i e 1 1 P , 4 g.
}
?
t 7 i. I
-y k -. ,. . _ , - - , . . , -
~;d
! t UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION STATE OF LOUISIAFA ) t PARISH OF WEST FELICIANA ) Docket No. 50-458 In the Matter of ) GULF STATES UTILITIES COMPANY ) (River Bend Station - Unit 1) AFFIDAVIT W. H. Odell, being duly sworn, states that he is a Manager
- Overnight for Gulf States Utilities Con.pany ; that he is authorized on the part of said company to si.gn and file with the N2 clear Regulatory Commission the docuhants attached hereto; and that all such documents are true and correct to the best of his knowledge, information and belie!.
j{
- W. H.' Odell Subscribed and sworn to before me, a Notary Public in and for the State and Parish above named, this [p Yb' day of O LLAtl , 1990 . My Commission expires with Life.
O b , /dl Claudia F. Hurst Notary Public in and for West Feliciana Parish, Louisiana
~
f b .. 3
?
1. ATDateNr >
.As requested by - the ' NRC Staff by letter dated July 31, 1989, Gulf States I v >
Utilities Campany (GSU) has reviewed all River Bend Station -(RBS) Technical '
- - Specifications. (TS) and identified those TS action stataments for which the -
G proposed change to TS 3.0.4 will provide new operating flexibility. .Those
- action statements are discussed on the following pages. The TS action statcznents listed below .are those where there was not. sufficient justification to .show " equivalent level of safety", and therefore are those ,
where GSU will> not change nodes while depending on the action statements. i f 3.1.1
- . 3.3.7.6-
- , 3.5.2 ACTIm b 3.5.3 ACTION b i
3.7.1.2 3.8.1.2 ACTION a [ 3.8.2.2 ACTICN a 3.8.3.2 ACTION a.1 and b.1 i
- 3. 9. 2 - 1 3.9.11.1 ACTION b 3.10.4 a GSU will address future changes to the positions subnitted in this response to this NRC request for additional infonnation as follows:
.If a' new change to the exclusion list of this subnittal is identified,
- the NRC will be informed prior to use.
?
If the change was not part of this subnittal then it will be reviewed using 10CFR50.59 criteria including. GL87-09, the RBS Updated Safety Analysis Report and the RBS Safety Evaluation Report (SFR) (including the SER on this request) . The method to control this new flexibility will be a review by the. RBS ; Pacility Review Cmin2ttee (FRC) of those action statanents identified in this a subnittal being rehM upon during a. significant change in operating condition (OC) such cs during startup or before fuel novements/ reloading operations in the containnent or the fuel building. The FRC will review s these actions, including those in OC-3, 2, 1 or OC-S,
- which will be. entered or passed through, upon exiting OC-4 to ensure any required action is taken i' prior. to power or refueliag operations. At this point any further reliance on this flexibility would not be preplanned and OC changes can be made while depending on the new flexibility. When reviewing these actions applicable criteria such as the following will be utilized:
Does the plant configuration agree with the positions presented in this response? ' 1 f
' ~
Are other conditions present which were not considered in the SAR or'this t response? Was this condition previously known and the change in OC preplanned? { Was the condition known but not able to be fixed prior to the DC change , for reasons other than schedule? The 'IPC will use these criteria to make rchations to the Plant Manager if the operational condition of the plant should change or other action be taken. > The following discussions are provided for each of the TS Action Statements for which nonroutine use of the proposed change to TS 3.0.4 will provide new operating flexibilityt
- 1. TS 3.1.4.2 - Rod Pattern Control Systm The function of the Rod Pattern Control Syste (RPCS) .is to provide a backup to administrative procedures to limit the _ consequences of a ,
control rod drop accident below the low power setpoint (approximately 20% :
- of rated thermal pmer .(RPP)) and a control rod withdrawal error above the low pwer setpoint. 2his function is accmplished by restricting the patterns of control rods that can be established below the low pwer ;
setpoint to predetermined sets. Above the low power setpoint, restrictions on control rod mvment to minimize the- consequences of a-control rod drop accident are no longer imposed. The low power setpoint is well above the point at which the control rod drop accident I consequences are- no longer limiting. Above the low power setpoint, control rod withdrawals are restricted to prevent excessive change in the heat flux rate due to a control rod withdrawal error. Between the low ' power notpoint and the high power setpoint (approximately 70% of RTP), control rod motion is limited to four notches (2 feet) . Above the high power setpoint, control rod mtion is limited to two notches (1 foot) . Action'at This action statement applies with the RPCS inoperable while in Operational Condition 1 or 2. With the RFCS inoperable, control rod movment is restricted based upon reactor power level. Above 20% of RTP, control rod withdrawal is prohibited. Below 20% of RTP, control rod' movement is prohibited except by scram. Therefore, application of the proposed change .to TS 3.0.4 wiH provide operational flexibility by , allowing the plant to enter Operath.aal Condition 1 or 2 with the RPCS inoperable while emplying with the control rod mvement restrictions of this action statement. Cmpliance with this action statement will ensure ; that control rod mvement will rcmain within the constraints of the predetermined sets of the RFCS. Therefore, the function of the RPCS will lI be pnrformed by limiting the consequences of a control rod drop accident or a control rod withdrawal error. As a result, compliance with this
. action will ensure that an equivalent level of safety will be maintained- 3 during any modo changes into Operational Condition 1 or 2 with the RPCS !
inoperable, j l 2 l i 4 l u
W;# ( a , Action b: With up to eight inoperable. control mds bypassed in the Rod I Action Control Systm (RACS), Action Statements b.2 and b.3 allow plant operation to continue, pmvided each bypassed control rod is inserted and , disarmed, all inoperable control rods are separated .frm all other inoperable control rods by at least two control cells in all directions, there arc no more than three inoperable control rods in any RPCS group, and the position and bypassing of the inoperable control rods is verified by a second licensed operator or other technically qualified member of- ,. the unit technical staff. Therefore, application of the proposed change ' to TS 3.0.4 will provide operational flexibility by allowing the plant to. . enter Operational Condition 1 or 2 with up to eight inoperable control rods bypassed in the RACS while emplying with the above action _; requirements. C m pliance with this action will ensure that the safety . analysis for reactor scram reactivity and the constraints of the RICS are ! being. complied with. Purther, TS 3.1.3.1 currently allows continued , plant operation and modo changes with up to eight inoperable control rods inserted and disarmed. Werefore, empliance with the above restrictions will provide adequate assurance that plant operation will remain within - the USAR conclusions. As a result, cmpliance with this action will ensure that an equivalent level of safety will be maintained during any l
, mode changes into operational Cordition 1 or 2 with up to eight control ;
rods bypassed in the RACS. '
- 2. TS 3.3.1 - Reactor Protection Syst m Instrumentation
%e reactor protection systm (RPS) autmatically initiates a' reactor scram tot
- 1) Preserve the integrity of the fuel cladding,
- 2) Preserve the integrity of the reactor coolant system, '
- 3) Provent inadvertent criticality.
This TS provides the mininum requirements necessary to preserve the ability of the .RPS to perfom its intended function, even during periods i when instrument channels may be out of service because of maintenance or l to conduct required surveillances. l The RPS is made up of two independent trip systms. There are normally , four channels to monitor each parameter, with two channels in each trip . system. The outputs of the channels in a trip system are cmbined in a i one-out-of-two logic so that either channel will trip that trip system. The tripping of both trip systms will produce a reactor scram. Action b: With the number of operable channels less than the minimum ; operable channels per trip syst m requirement for both trip systems, this
. action statment requires that the trip system with the nost inoperable l channels be placed in the tripped condition and the action statement '
L required by Table 3.3.1-1 be entered. The trip system is not required to be placed in the tripped condition when this would cause the trip function to occur. This allowance is provided in order to avoid a plant transient and subsequent challenc., to safe shutdown systems. Pecept for-3 i
V a
.f-
. o the situation where placing the trip system in the tripped condition is not required, this action' statement requires the safety function of the inoperable channel be positively performed. When placing the channel in the tripped condition is not required, the action requirements of Table 3.3.1-1 require a safe final plant condition while allowing sufficient time to perform the evolution in a controlled manner. Application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to make node changes while emplying with those action statements that allow continued plant operation for an unlimited period of time. Restriction on node changes will be dictated by the applicable action statements of Table 3.3.1-1. Since the conditions required by the applicable action statement of Table 3.3.1-1 must be established prior to making the mode change, ccrupliance with ! Action b- will ensure that an equivalent level of safety is maintained during any mode changes while ccruplying with those action statemnts that allow continued plant operation for an unlimited period of time, as discussed below for the respective action statement. Action 2: This action statament applies to Table 3.3.1-1, Items 1.a, 1.b, 2.a, and 2.d while in operational Condition 3 or 4. This action requires all insertable control rods be verified to be inserted into the core and s the reactor mode switch be locked in the shutdown position within one ~ hour.' Therefore, application of the proposed change to TS 3.0.4 will proviue operational flexibility by allowing the plant to enter operational Condition 3 and 4 with all insertable control rods inserted and the reactor mode switch locked in the shutdown position. Compliance with this action will ensure that the scram function is not required since the control rods would be inserted. Additionally, with the mode switch in the shutdown position, a control rod block signal would be present to prohibit control rod withdrawal. Therefore, cmpliance with this action will ensure that an em '"alent level of safety is maintained : during any node changes into operata "ondition 3 or 4 with these I neutron flux nonitor scram functions 1., - m ble. is Action 3: This action statcsent applies to Table 3.3.1-1, Items 1.a,1.b, 2.a, 2.d, 9.a, 9.b, and 12 while in operational Condition 5. This action requires all core alterations be suspended and all insertable control l rods be inserted within one hour. Therefore, application of the proposed j change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter operational Condition 5 bv detensioning the reactor vessel i head with all insertable control rods inserted rad core alterations i suspended. Cmpliance with this action will ensure that the scram 1 function is not required since the control rods would be inserted and the ' core reactivity cannot be changed due to core alterations. Therefore, cmpliance with this action will ensure that an equivalent level of safety is maintained during any node changes into operational Condition 5 , by detensioning the reactor vessel head with these neutron flux monitor, ' scram discharge volume or reactor mode switch position scram functions inoperable. l 4 l l
r i
- - Action 5
- Wis action statment applies to Table 3.3.1-1, Itm 7 while in ;
operational Condition 1 or 2. m is action requires the reactor to be in Operational Condition 2 with the main steam lines closed within 6 hours , or the ' reactor is required to be in Operational Condition 3 within 12 hours. Therefore, application of the proposed change to TS 3.0.4 will i provide operational flexibility by allowing the plant to enter Operational Condition 2 with the main steam lines closed. j i We main steam line radiation instrument channels are provided to detect ,
. gross failure of fuel cladding and initiate a reactor scram to reduce continued failure. These instruments also provide signals to the prinary containment isolation system to contain the release of fission products.
Conpliance with this action protects against the potential release of radioactivity to the environment while the high main steam line radiation scram is not available. % e unavailability of the high main steam line radiation scram is not considered to be a reduction in safety. In fact, reactivity control failure frequency and core damage frequency . analyses have been- performed which indicate that the deletion of the high main e steam line radiation scram actually represents a 0.3% reduction in core , ! damage frequency (NEDO-31400, February 1987) . Therefore, cmpliance with ' this action will protect against the potential release of radioactivity to the environment and will ensure that an equivalent level of safety is i maintained during any mode changes into operational Condition 2 with this i main steam line radiation scram function inoperable. Action 6: This action statment applies to Table 3.3.1-1, Itms.10 and 11 while in Operational Condition 1. This action requires the inoperable channels in both trip systems be placed in the tripped condition with one hour or main turbine first stage pressure is required to be reduced below the autmatic scram bypass setpoint (40% of RTP) within two hours. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter Operational Condition 1 up to the autmatic scram bypass setpoint with these channels inoperabic and by allowing further power increase with the inoperable channels in the tripped condition. As stated in USAR Section 15.2.3, below 40% of RTP adequate protection is provided by the neutron flux and reactor pressure scram functions. Therefore, the main turbine stop valve closure and main turbine control valve fast closure scram functions are bypassed below 40% of RTP. Compliance with this action will ensure that the function of this inoperable instrumentation is positively perfomed by requiring the channel to be tripped or that the plant is in a cordition in which this function is not required per the safety analysis. Therefore, cmpliance with this action will ensure that an equivalent level of safety is maintained during any node changes into operational Condition 1 with this main turbine stop valve closure or main turbine control valve fast closure scram function inoperable. Actionj: This action statment applies to Table 3.3.1-1, Item 12 while in Operational Condition 3 or 4. This action requires all insertable control rods be verified to be inserted into the core within one hour. W erefore, application of the proposed change to TS 3.0.4 will provide 5 i
V c o operational flexibility by allowing the plant to enter Operational condition 3 and 4 with the reactor node switch shutdown position scram function inoperable ,and all control rods verified to be inserted. Cmpliance with this action will ensure that this scram function is not required since the control rods would be inserted. Therefore, empliance with this action will ensure that an equivalent level of safety is maintained during any node changes into Operational Condition 3 or 4 with this reactor mode switch shutdown position scram function inoperable.
' Action 8: This action statement applies to Table 3.3.1-1, It s 13 while in operational Condition 3 or 4. This action requires the reactor node switch be locked in the shutdown position within one hour. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter Operational condition 3 and 4 with the manual scram function inoperable and the mode switch locked in the shutdown position. Placing the mode switch in the shutdown position will autmatically insert a reactor scram signal and a control rod block signal. Therefore, empliance with this action will ensure that the nanual scram function will not be required since the control rods would be inserted and with the mode switch in the shutdown position, a control rod block signal would- be present to prohibit control rod withdrawal.
Therefore, empliance with this action will- ensure that an equivalent level of safety is maintained during any node changes into Operational Condition 3 or 4 with this manual scram function Loperable. Action 9: This action statment applies to Table 3.3.1-3, Item 13 while in operational Cordition 5. This action requires all core alterations be suspended, all insertable control rods be inserted, and the reactor mode switch be locked in the shutdown position within one hour. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter Operational Condition 5 by detensioning the reactor vessel head with the manual reactor scram ! function inoperable, core alterations suspended, all insertable control rods inserted, and the reactor node switch locked in the shutdown position. Compliance with this action will ensure that the manual scram function is not required since the control rods will be inserted, and any l action which could cause a change in core reactivity is prevented by the l presence of a control rod block signal and the requiremert to suspend core alterations. Therefore, empliance with this action will ensure that an equivalent level of safety is maintained during any nodo changes L into Operational condition 5 by detensioning the reactor vessel head with j this manual scram function inoperable.
'l Action 10: This action statenont applies to Table 3.3.1-1, itm 6 while in Operational Condition 1. This action requires the inoperable main !
steam isolation valve closure channels in both trip systems be placed in j the tripped condition within one hour or the reactor is required to be in ; Operational Condition 2 within 6 hours. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter Operational Condition 1 with the inoperable channels in both trip systems in the tripped condition. By requiring all 6
C c .. inoperable channels to be placed in the tripped condition, the function of these instruments is being positively performed. Therefore, empliance with this Action will ensure that an equivalent level of safety is maintained during any node changes into operational Condition 1 with these main steam isolation valve closure channels inoperable.
- 3. TS 3.3.2 - Isolation Actuation Instrumentation
'Ihe isolatim actuation instrunentation provides for autmatic and manual actuation of isoladon functions for thet (1) primary containment, (2) main steam lines, (3) secondary containment, (4) reactor water cleanup (RNCU) systs, ($) reactor core isolation cooling (PCIC) systs, and (6) residual heat renoval (RR) systs. The safety function of this instrumentation is to provide isolation of the respective systm(s) in response to various initiating signals to assure that the overall plant response to a design basis accident (DM) follows the assumptions in the safety analyses. This TS provides the minimum requirements necessary to preserve the ability of these isolation actuation systems to perform their intended functions, even during periods when instrument channels may be out of service because of maintenance or to conduct required surveillances.
Action at In the event that a channel's trip setpoint is less conservative than the TS allowable value, this action statment requires declaring the instrunent channel inoperable. Thus, Action a has the effect of establishing inoperability of the affected instrumentation and then transferring the required action to Action Statement b or c. Action a does not allow' a reduction in safety, but rather establishes the instrumentation's operability in relation to the required actuation setpoint. Any restrictions on modo changes will be dictated by the
- applicable action requirements that resula frm declaring the instrument channel inoperable as required by this action. Since the conditions required by the applicable action statements that must be entered must be established- prior to making the mode change, application of the operational flexibility allowed by the proposed change to TS 3.0.4 will provide an equivalent level of safety during any node changes while emplying with those action statements that allow continued plant l operation as discussed below.
Action c: With the number of operable channels less than the minimum operable channels per trip system requirement for both trip systerns, this ; action statenent requires the trip system with the most inoperable 4 channels be placed in the tripped condition and the action statement ' required by Table 3.3.2-1 be entered. The trip system is not required to be placed in the tripped condition when this would cause the trip function to occur. This allowance is provided in order to avoid a plant transient and subsequent challenge to safe shutdown systms. Except for the situation where placing the trip system in the tripped condition is not required, this action statment requires the safety function of the inoperable channel be positively performed. When placing the channel in the tripped condition is not required, the action requirements of Table i 7 1 l i a
g 3.3.2-1 require a safe final plant condition while allowing sufficient tim to perfom the evolution in a controlled manner. Application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to make mode changes while emplying with those action statement that allow continued plant operation for an unlimited period of time. Restriction on node changes will be dictated by the applicable action statments of Table 3.3.2-1. Since the conditions required by the applicable action statment of Table 3.3.2-1 must be established prior to making the node change, empliance with Action c will ensure that an equivalent level of safety is maintained ! during any mode changes while complying with those action statements that allow continued plant operation for an unlimited period of tine, as oiscussed below for the respective action statement. Action 21: This action statment applies to Table 3.3.2-1, Item l'.c i while in Operational Condition 1, 2 or 3. This action requires the ! affected primary containment isolation valves be closed within one hour i or the reactor is required to be in Operational Condition 3 within the next 12 hours and Operational Condition 4 within the following 24 hours. Therefore, application of the propcaed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter Operational ; Conditions 1, 2 and 3 with the affected primary containment isolation valves closed. Since the function of these instruments is being positively perfomnd by requiring the affected isolation valves to be ; closed, c epliance with this action will ensure that an equivalent level J of safety is maintained during any node changes into Operational ! Condition 1, 2 or 3 with this primary containment isolation i instrumentation inoperable, i Action 23: This action statement epplies to Table 3.3.2-1, Itms 2.b, l 2.d, 2.f, 2.g, and 2.h while in Operational Condition 1, 2 or 3 and Item ; 2e while in Operational Condition 1, 2 or 3 with the reactor node switch i in run and/or any nain turbine stop valve open. This action requires the reactor to be in Operational Condition 2 with the affected main steam line isolation valves closed within 6 hours or the reactor is required to ; be in Operational Condition 3 within 12 hours and operational Condition 4 within the next 24 hours. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to i enter Operational Conditions 2 and 3 with the affected main steam line isolation valves closed. Since the function of these instruments is ! b>ing positively performed by requiring the affected isolation valves to be closed, cmpliance with this action will ensure thet an equivalent level of safety is maintained during any node changes into operational 1 Condition 2 or 3 with this main steam line isolation instrumentation ! inoperable. i Action 25: . This action statment applies to Table 3.3.2-1, Itms 3.a and { 3.b while ir Operational Condition 1, 2 or 3. This action requires i secondary ontainment integrity - operating be established with the j standby gas treatment and Puel Building ventilation syst ms operating in the emergency node within one hour. Therefore, application of the 8 l
I proposed change to TS 3.0.4 will pmvide operational flexibility allowing the plant to enter Operational Conditions 1, 2 and 3 with secondary containment integrity-operating established and the standby gas treatment-and fuel building ventilation systems operating in the emergency node.
~
Since the function of these instrumnts is being positively performed by rcquiring the standby gas treatment and fuel building ventilation systes i to be operating in the emergency mode, c m pliance with this action will ensure that an equivalent level of safety is maintained during any mode changes into operational condition 1, 2 or 3 with this secondary containment isolation instrumentation inoperable. Action 26: This action statement applies- to Table'3.3.2-1, It s 5.n- ! while in Operational Condition 1, 2 or 3. This action requires the manual isolation function be restored to operable status within 8 hours or affected isolation valves be closed within one hour and the affected , system be declared inoperable. Therefore, the effect of this action ' statment is to positively perform the function of the inoperable isolation instrumentation by requiring the affected isolation valves to be closed and then transfer the required actions to the applicable action statements of the affected system. Operability rquirements for the RCIC system are contained in the TS. Continued plant' operation for an unlimited period of time is not allowtd wich the RCIC system inoperable i in operation Condition 1, 2 or 3 with reactor pressure greater than 150 psig per TS 3.7.3. Therefore, the operational flexibility allowed by the i proposed change to TS 3.0.4 cannot be applied to modo changes with ; reactor- pressure greater than 150 psig with the ncIC isolation instxumentation inoperable. Application of the proposed change to TS 3.0.4 will, however, allow mode changes into operational condition 2 or 3 ; with reactor pressure less than 150 psig with the RCIC manual isolation- ; instrumentation inoperable, pmvided the affected isolation valves are l closed. Since no credit is taken for operation of the RCIC system below ] 150 psig and the function of the inoperable isolation instrumentation is being positively performed by requiring the affected isolation valves to be closed, cmpliance with this action will ensure that an equivalent level of safety is maintained during any mode changes into Operational i Condition 2 or 3 with reactor pressure less than 150 psig with this RCIC ' isolation instrumentation inoperable. 1 Action 27: This action statement applies to Table 3.3.2-1, Items 4.a, l 4.b, 4.c, 4.d, 4.e, 4 f, 4.g, 4.h, 5.a, 5.b, 5.c, 5.d, 5.e, 5.f, 5.g, 5.h, 5.1, 5.j, 5.k, 5.1, and 5.m while in Operational Condition 1, 2 or
- 3. This action requires the affected isolation valves be closed within one hour and the affected system be declared inoperable. Therefore, the effect of this-action statcment is to positively perform the function of L the inoperable isolation instrumentation by requiring the affected j isolation valves to be closed and then transfer the required actions to the applicable action statements of the affected system. Since the RWCU systm is not relied upon to perform a safety function, requirements for i operability of this system are not contained in the TS. However, l operability requirments for the PCIC syst m are contained in the TS. l Continued plant operation for an unlimited period of tim is not allowed 9 !
1
- j. 4-3 i
with the RCIC system -inoperable in Operation Condition 1, 2 or 3 with reactor pressure greater than 150 psig per TS 3.7.3. Therefore, the operational flexibility allowed by the pmposed change to TS 3.0.4 cannot be applied to nose changes with reactor pressure greater than 150 psig with the RCIC isolation instrumentation inoperable. Application of the - proposed change to TS 3.0.4 will, however, allow node changes into operational Condition 2 or 3 alth reactor pressure less than 150 psig with the RCIC isolation instnmentation inoperable and into Operational Conditions 1, 2 and 3 with the RWCU isolation instrumentation inoperable, provided the affected isolation valves are closed. Since no credit is , [ taken for operation of the RCIC syst s below 150 psig and the function of the inoperable isolation instrumentation is being positively perfomed by l requiring the ai'fected isolation valves to be closed, cmpliance with this action will ensure that an equivalent level of safety is maintained during any node changes into Operational Condition 2 or 3 with reactor- i pressure less than 150 psig with this RCIC isolation instrumentation inoperable and during any mode changes into operational Conditions 1, 2 l or. 3 with this RWCU isolation instrunentation inoperable, Action 28: This action statement applies to Table 3.3.2-1, Its 3.c ; while handling drradiated fuel in the fuel building. This action
- requires the fuel building ventilation systs be placed into operation in the emergency node within one hour. Therefore, aoplication of the proposed change to TS 3.0.4 will provide operational ficxibility by allowing the plant to begin handling irradiated fuel in the fuel building with the ventilation system operating in the energency s;Se. This action requirement is exactly . the same as the Iro requirments for TS 3.6.5. 5 -
which require one fuel building ventilation system to be operating in the emergency mode when handling irradiated fuel in the fuel building. Since the function of this secondary containnent isolat . . instnmentation is being positively perfomed by operating the fuel building ventilation , system in the emergency mode, campliance with this Action will ensure that an equivalent level of safety is naintained upon comencing handling of irradiated fuel in the fuel building with this secondary containment isolation instrumentation inoperable. Action 29: This action statement applies to Table 3.3.2-1, Item 3.d while in Operational Condition 1, 2 or 3. This action requires the annulus mixing systm be initiated and aligned to at least one operating standby gas treatment systs train within one hour. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility.by allowing the plant to enter Operational Conditions 1, 2 ; and 3 with the annulus mixing system operating and aligned to at least one operating standby gas treatment systs train. Since the function of this secondary containment isolation instrumentation is being positively performed by requiring these systems to be in operation, cmpliance with this action will ensure that an equivalent level of safety is maintained during any node changes into operational Condition 1, 2 or 3 with this . secondary containment isolation instrunentation inoperable, f 10
_~ Action 30: mis action statement applies to Table 3.3.2-1, Items 6.a, L 6.b, 6.c, 6.d, 6.e, and 6.f while in Operational Cordition 1, 2 or 3. This action _ requires the affected system isolation valves be locked-closed and the affected system declared inoperable within one hour. Therefore, the effect of this action statement is to positively perfom the function of the inoperable isolation instrumentation by requiring the affected isolation valves to be locked closed and then transfer the required actions to the applicable action statments of the affected Mm system operating node. Since these instruments provide isolation signals' . to various portions of the Mm systs, and the mm systm itself performs ! various TS related functions (e.g., mergency core cooling, suppression pool cooling and shutdown cooling), the associated action statements for the M m system functions which hwe been made inoperable will dictate whether continued plant operation for an unlimited period of time is allowed and whether a modo change is allowed in accordance with the proposed change to TS 3.0.4. Since the isolation function of the MIR isolation instrumentation is being positively perfonned bre requiring the affected isolation valves to be locked closed and the conditions required by the applicable actions of the affected Mm function must be i established prior to making the node change, empliance with this action will ensure that an tativalent level of safety is maintained during any node changes into Operational mndition 1, 2 or 3 while complying with l those mm system action statements that allow continued plant operation l for an unlimited period of ti2rc with this Mm isolation instrumentation ! inoperable. [
- 4. TS 3.3.3 - Dnergency Core Cooling System Actuation Instrumentation The emergency core cooling systm (ECCS) is emprised of three divisions.
Division 1 consists of the low pressure core spray (LPCS) system, low pressure coolant injection (LPCI) train "A", and the automatic y depressurization system (ADS) as actuated by trip system "A". Division 2 consists of LPCI trains "B" and "C" and ADS as actuated by trip system "B". The high pressure core spray (HPCS) system emprises Division 3 of . the ECCS. The safety function of these subsystems is to maintain ! adequate core cooling in the event of a loss of coolant accident (LOCA), i over the full range of reactor operating conditions, to ensure that fuel _! cladding tmperature does not exceed 2200 degrees F. The autmatic H initiation of these systms is provided by the actuation instrumentation 3 listed on TS Table 3.3.3-1. This TS provides the rtinimum requirements ' necessary to preserve the ability of the ECCS actuation instrunentation 1 to perfom its intended functions, even during periods when instrument I channels may be out of service because of maintenance or to conduct I required surveillances. , i The ECCS actuation instrumentation is comprised of sensors that nonitor plant parameters which will change significantly in the event of a 10CA, such as reactor water level and drywell pressure. In order to ensure j ECCS operability, the ECCS actuation instrumentation also nonitors ' parannters associated with mergency system operation, such as reactor vessel pressure and pump discharge pressures. The ECCS actuation 11 i l l
instrumentation is separated into three divisions carresponding to the three divisions of the BIS and also includes trip functions associated with a loss of electrical power. Action a: -In the event that a channel's trip setpoint is less conservative than the TS allowable value, this action statement requires declaring the instrument channel inoperable. Thus, Action a has the effect of establishing inoperability of the affected instrunentation and then transferring the required action to Action Statement b. Action a does not allow a reduction in safety, but rather establishes the instrumentation's operability in relation to the required actuation setpcint. Therefore, restriction on mode changes will be dictated by the applicabic action requirements that result frcn declaring the instrument
- channel inoperable as required by this action. Since the conditions required by the applicable action statement that must be entered must be established prior to making the mode change, application of the operational flexibility allowed by the proposed change to TS 3.0.4 will provide an equivalent level of safety during any mode changes while complying with those action statements that allow continued plant operation for an unlimited period of time, as' discussed below for the respective action statement.
Action b With one or more ECCS actuation instrumentation channels inoperable, the requirements of this action statement merely transfer the required action to that identified for the trip function listed on Table 3.3.3-1. Therefore, the action statement listed on Table 3.3.3-1 will dictate whether continued plant operation for an unlimited period of time is allowed and whether a mode change is allowed in accordance with the proposed change to TS 3.0.4. Since the applicable action statements of Table- 3.3.3-1 will dictate whether a made change is allowed by the propostd change to TS 3.0.4 and the conditions required by the applicable action statment must be established prior to making the mode change, cmpliance with this action will ensure that an equivalent level of safety is maintained during any node changes while cmplying with those action statements that allow continued plant operation for an unlimited period of time with this ECCS instrumentation inoperable, as discussed below for the respective action statement. Action 32: This action statment applies to Table 3.3.3-1, Items A.l.d and B.1.c while in Operational Condition 4 or 5 when the applicable syst m is required to be operable per TS 3.5.2 or 3.5.3. This action requires the inoperable channels be placed in the tripped condition within one hour. Therefore, application of the proposed change to TS 3.0.4 will provide operational ficxibility by allowing the plant to enter Operational Condition 4 and 5 by tensioning or detensioning the reactor vessel head with inoperable low pressure ECCS injection valve permissive instrumentation channels in the tripped condition. With the incperable channels in the tripped condition, the function of this instrumencation will be positively perfomed. Therefore, cmpliance with this action will ensure that an equivalent level of safety will be maintained during any made changes into operational Condition 4 or 5 by tensioning or 12 l
w detensioning the reactor vessel head with this low pressure ECCS
. injection valve permissive instrumentation inoperable.
Action 37: This action statement applies to Table 3.3.3-1 Itm D.2.a Ij while in Operational Condition 1, 2 or 3,or during Operaticial Condition 4 or 5 when the Division 3 dier,el generator is required to be operable ! per TS 3.8.1.2.- This action statement requires the diesel renerator be ; declared inoperable. Therefore, this action statement merely establishes ; inoperability of the Division 3 diesel generator with retpect to the inoperability of this 4.16 KV standby bus sustained 'mdervoltage instrumentation and then transfers the required action to the appropriate [ action statements of TS 3.8.1.1 or TS 3.8.1.2, as applicable. Since TS J 3.8.1.1 requires all three diesel generators to be operable in Operational Conditions 1, 2 and 3 and provi6es shutdown actions with the Division 3 diesel generator inoperable, the- operational flexibility provided by the proposed change to TS 3.0.4 cannot be applied to mode changes into Operational Conditions 1, 2 or 3. However, TS 3.8.1.2 only requires the Division 3 diesel generator to be operable in Operational , Conditions 4 and 5 when the HPCS syst m is required to be operable per TS 3.5.2 or TO 3.5.3. Th e afore, in operational condition 4 or 5, Action 37 l merely requires empliance with Action b of TS 3.8.1.2 which dictates the appropriote action for the Division 3 diesel generator being . inoperable-when required. Since the conditions required by Action b of TS 3.8.1.2 must be estai.:lished prior to making the mode change, empliance with this ' action will ensure that an equivalent level of safety is maintained during any mode changes into operational Condition-4 or 5 by tensioning ,3 or detensioning the reactor vessel head with this 4.16 KV standby bus f sustained undervoltage Division 3 diesel generator actuation Lj instrumentation inoperable. L5. TS 3.3.4.2 - End-of-Cycle Recirculation Pump Trip System instrumentation i i The end-of-cycle recirculation pump trip (EDC-RPT) system is a part of the reactor protection syst m (RPS) and is an essential safety suppl ment to the reactor scram. The purpose of the EOC-RPT is to recover the loss of . thermal margin which occurs at the end-of-cycle. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor , system at a faster rate than the control rods add negative scram reactivity. Each EOC-RPT system trips both recirculation pumps, reducing coolant flow, in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the EOC-RPT protective feature will function are closure of the main turbine stop valves and fast closure of the main turbine control valves. A fast-closure sensor frcm each of two main turbine control valves provides input to the EOC-RPT system, a fast-closure sensor frm each of the other two main turbine control valves provides input to the second EOC-RPT system. Similarly, a position switch for .each of two main turbine stop valves provides input to _one EOC-EPT system. For each EOC-RPT system, the sensor relay contacts are arranged to form a 13
p (c . two-out-of-two logic for the fast closure of the main turbine control valves and a two-autm f-two logic for the main turbine stop valves, me operation of either logic will actuate the EOC-RPT system and trip both recirculation pumps. The requirements of this TS are applicable in Operational Condition 1 with thermal power greater than or equal to 40% , of RTP. ! Action at In the event that a channel's trip setpoint is less ! conservative than the TS allwable value, this action statement requires. , declaring the' instrunent channel inoperable. Thus, Action a has the ! effect of establishing inoperability of the affected instrumentation and ! then transferring the required action to Action Statement b or c. Action a does not all w a reduction in safety, but rather establishes the li instrumentation's operability in relation to the required actuation setpoint. Therefore, restriction on mde changes will be dictated by the applicable action requiremcmts that result from declaring the instrument channel inoperable as required by this action. Since the conditions required by the applicable action statements that must be entered must be established prior to making the mode change, application of the operational flexibility allowed by the proposed change to TS 3.0.4 will ,'
. provide an equivalent level of safety during any modo changes while cmolying with those action statements that allw continued plant operation for an unlimited period of time, as discussed belm for the respective action statement.
Action b . This action statement applies if the number of operable channels is one less than the minimum operable channels per trip system requirement for one or both EOC-RPT trip systems. This action requires the inoperable channel (s) be placed in the tripped condition within one hour. Therefore, application of the proposed change to TS 3.0.4 will , provide operational flexibility by allming the plant to enter ; Operational Condition 1 with thermal power greater than 40% of RTP with i the inoperable channel (s) in the tripped condition. By requiring the inoperable channel (s) to be in the tripped condition, this action a requires that the function of the inoperable channel (s) be positively performed. Therefore, cmpliance with this action will ensure that an equivalent level of safety is maintained during any mode changes into i operational ~ Condition 1 with thermal power greater than or equal to 40% of RTP with one channel in one or both EOC-RPT trip syst m s inoperable.- , Action c.1: This action statement applies if two required channels in one EOC-RPT trip system are inoperable and they consist of one main turbine control valve channel and one main turbine stop valve channel. This action requires the inoperable channels be placed in the tripped condition within one hour. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by all w ing the plant to enter Operational Condition I with thermal power greater than 40% of RTP-with both inoperable channels in the tripped condition. By requiring the inoperable channels to be in the tripped condition, this action requires that the function of the inoperable channels be positively performed. Therefore, empliance with this action will ensure that an equivalent 14
I t
.- 1 level of safety- is maintained during any mode changes into operational- ,
condition 1 with thermal p wer greater than or equal to 40% of RTP with one- main turbine control valve channel and one main turbine stop valve channel in one EOC-RPT trip system inoperable. l
- 6. TS 3.3.5 - Reactor Core Isolation Cooling System Actuation Instrumenta-tion 1
The primary safety function of the reactor core isolation cooling (FCIC) j system is to provide adequate core cooling foll wing a control rod drop accident or a loss of feedwater transion.t without providing actuation of any of the mergency core cooling systes (ECCS) . Although not an ECCS, RCIC also acts as a backup to the high pressure core spray (HPCS) system.
'Ihe RCIC system is capable of autcmatically injecting to the reactor in time to maintain sufficient coolant in the reactor vessel so that the integilty of the fuel clad is not cmprmised without actuation of any of the ECCS. The RCIC actuation instrumentation provides the signals required to initiate TCIC system operation.
In addition to signaling the RCIC system to initiate, instrumentation is 7 provided to mnitor reactor vessel level, condensate storage tank level and suppression pool level. The RCIC system actuation signal is-initiated frm the reactor vessel . water level - l w 1 m level 2 trip ' function, which utilizes one-out-of-two twice logic. The operability of i the reactor vessel water level - low low level 2 trip function ensures : that the RCIC system receives an initiation signal in sufficient time to i mitigate the event. Action a: In the event that a channel's trip setpoint is less conservative than the TS allowable value, this action statement requires declaring the instrumnt channel inoperable. Thus, Action a has the effect of establishing inoperability of the affected instrumentation and then -transferring the required action to Action Statement b. Action a does not allow a reduction in safety, but rather establishes the instrumentation's operability in relation to the required actuation setpoint. Therefore, restriction on mode changes will be dictated by the applicable action requiremento that result fram declaring the instrument channel inoperable as required by this action. Since the conditions required by the applicable action statements that must be entered must be established prior to making the mode change, application of the operational flexibility allowed by the proposed change to TS 3.0.4 will provide an equivalent level of safety- during any mde changes while ' cmplying with those action statements that allow continued plant operation for an unlimited period of time, as discussed below for the respective action statement. Action b: With one or more RCIC system actuation instrumentation channels inoperable, the requirEEents of this action statment merely transfer the required action to that identified for the trip function listed on Table 3.3.5-1. Therefore, the action statment listed on Table 3.3.5-1 will dictate whether continued pitnt operation for an unlimited 15
period of tine is allowed and whether a nede change is allowed .in accordance with the proposed change to TS 3.0.4. Since the applicable action statemnts of Table 3.3.5-1 will dictate whether a nale enange is allowed by the proposed change to TS 3.0.4 and the conditions required by the applicable action statenent nust be established prior to making the node change, cmpliance with this action will ensure that an equivalent level of safety is maintained during any node changes while emplying ) with those action statements that allow continued plant operation for an unlimited period of time with this RCIC initiation instrunentation inoperable, as discussed below for the respective action statement. Action 52_: This action statement applies to Table 3.3.5-1, Itms 3 and 4 , while in Operational Condition 1, 2 or 3 with reactor pressure greater than 150 psig. This action requires at least one inoperable channel be placed in the tripped condition within one hour or the RCIC system must be declared inoperable. Application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter operational Cc,ndition 1, 2 or 3 with reactor pressure greater than 150 psig with one inoperable channel in the tripped condition. Since -TS 3.7.3 will not allow continued plant operation for an unlimited period of time with the RCIC system inoperable, at least one inoperable channel nust be placed in the tripped condition prior to making the mode change. By requiring the inoperable channel to be placed in the tripped condition, this action requires that the function of the inoinrable channel be positively perfomed. Therefore, cmpliance with this action will ensure that an equivalent level of safety is maintained during any node changes into Operational Condition 1, 2 or 3 with reactor pressure greater. than 150 psig with this RCIC actuation instrumentation s inoperable.
- 7. TS 3.3.6 - Control Rod Block Instrumentation: )
The control rod block instrumentation supplies input to the rod control and infomation s} stem which inhibits control rod selection and movement to prevent reactivity changes under the conditions specified in TS Table 3.3.6-2. This subsystem consists of two channels, each with multiple data inputs frm multiple instruments. The trip logic is arranged such that a trip in any one of the inputs will result in a control rod block. Action a: In the event that a channel's trip setpoint is less i conservative than the TS allowable value, this action statement requires declaring the instrument channel inoperable. Thus, Action a has the effect of establishing inoperability of the affected instrumentation and then transferring the required action to Action Statement b. Action a does not allow a reduction in safety, but rather establishes the instrumentation's operability in relation to the required actuation setpoint. Therefore, restriction on mode changes will be dictated by the applicable action requirements that result frm declaring the instrument channel inoperable as required by this action. Since the conditions required by the applicable action statement that must be entered must be established prior to making the mode change, application of the 16
0 c s operational flexibility allowed by the proposed change to TS 3.0.4 will provide an equivalent level of safety during any node changes wittle cmplying with .those action statements that allow continued plant operation for an unlimited period of time, as discussed below for the respective action statement. Action b: With the number of operable channels less than the minimum operable channels per trip function requirement, the requirements of this action statment merely transfer the required action to that identified J for the trip function listed on Table 3.3.6-1. Therefore, the action ! statenent listed on Table 3.3.6-1 will dictate whether continued plant operation for an unlimited period of time is allowed and whether a modo l change is allowed in accordance with the proposed change to TS 3.0.4 I Since the applicable action statements of Table 3.3.6-1 will dictate d d ether a node change is allowed by the proposed change to TS 3.0.4 and j tho x nditions required by the applicable action statement must be estetAlshed prior to making the mode change, cmpliance with this action 1 will ensure that an equivalent level of safety is maintained during any mcde . changes while emplying with those actions that allow continued plant operation for an unlimited period of time with this control rod block instrumentation inoperable, as discussed below for the respective-action statement. Action 60: This action statement applies to Table 3.3.6-1, Item 1.a ; while in Operational Condition 1 or 2 and Item 1.b while in Operational j Condition 1. This action requires the rod pattern control system to be , declared inoperable and the actions of TS 3.1.4.2 to be entered. Thus, j this action merely establishes inoperability of the rod pattern control system as a result of inoperability of this control rod block ! instrumentation and transfers the required action to the applicable action statement of TS 3.1.4.2. Action a of TS 3.1.4.2 specifies the < appropriate actions to be taken with the rod pattern controller f inoperable. Application of the proposed change to TS 3.0.4 will provide , operational flexibility by allowing the plant to make mode changes while ; cmplying with the actions of TS 3.1.4.2 that allow continued plant operation for an unlimited period of time. Since the conditions required l by the applicable action statements of TS 3.1.4.2 must be established : prior to making the mode change, empliance with this action will ensure } that an equivalent level of safety is maintained during any node changes f into Operational Condition 1 or 2 while cmplying with those actions that allow continued plant operation for an unlimited period of time with'the rod pattern control system high or low power setpoint instrumentation , inoperable. *
- 8. TS 3.3.7.5 - Accident Monitoring Systm l
The operability of the accident nonitoring instrumentation ensures that sufficient information is available on selected plant parameters to permit nonitoring and assessment of important variables following an accident. This capability is consistent with the recmmendations of ] Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear l 17
c Power Plants to Assess Plant Conditions During and Following an t Accident," Decmber 1975 and NUPJX3-0737, " Clarification of 7MI Action Plan Requirenents," November 1980._ These accident monitoring instnments provide monitoring capability only and do not perform' any autmatic functions to mitigate a DBA or transient. The requirements of this TS . are applicable in operational Conditions 1, 2 and 3. Action 81: This action statement applies with less than the required i primary containment or drywell arca radiation nonitors operable. This I action statement requires that an alternate preplanned method of mnitoring the appropriate parameter be initiated. This ensures that, in the event of a DBA or transient, an alternative nothod of nonitoring the parannter is available. This action statement also requires a special . report be prepared and subnitted to the NRC outlining the cause of the ! inoperability and the plans and schedule for restoring instnment operability. Since cmpliance with this action establishes alternate monitoring capability empliance with this action will ensure that an equivalent level of safety is maintained during any mode changes into. Operational condition 1, 2 or 3 with these primary containment or drywell area radiation nonitors inoperable. L
- 9. TS 3.4.3.2 - Operational Leakage This TS establishes the acceptable leakage rates for the reactor coolant system and the pressure isolation valves listed in TS Table 3.4.3.2-1.-
As stated in the bases for this TS, the allowable leakage rate frm the. reactor coolant- system was based on the predicted and experimentally observed behavior of cracks in pipes. The normally expeted background leakage due to equipnent design and the detection capability of the l instrumentation for determining system leakage was also considered. The evidence obtained frm experiments suggests that, for leakage scmewhat ! greater than that specified for unidentified leakage the probabilit,f is small that the imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if the _ leakage rates exceed the values specified or the leakage is located and knmn to be pressure boundary leakage, the reactor will be shut down to allow further investigation and corrective action. The leakage rate requirements for pressure isolation valves provide assurance of valve integrity, thereby reducing the probability of gross valve failure and consequent intersystm I/X'A. Therefore, the safety function of this TS is 'the prevention of a large break IOCA by the detection of small leaks. , Action c: If any reactor coolant system pressure isolation valve
. exhibits leakage above its limit while in Operational Condition 1, 2 or 3, this action statement requires the high pressure portion of tL affected system be isolated frca the low pressure portion. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter Operational Conditions 1, 2 and 3 with the high and low pressure portions of the affected system isolated. Establishing the conditions required by this action satisfies the safety function of the affected isolation valves in so far as TS 18
4 Oc 3.4.3.2 is concerned. These valves have other safety functions related to the_ . operability of- their respective systems. However, these additional safety functions are addressed by the applicable action atatments of the associated system. Since the conditions required by ; Action c satisfy the safety function of this TS and since any- additional-actions that may be required by the resulting syst m configuration are required to be established prior to making the mode change, compliance - # with this action will ensure that an equivalent level of safety is maintained during any mode changes into Operational Condition 1, 2 or 3 with a reactor coolant system pressure isolation valve inoperable..
- 10. TS 3.4.4 - Chemistry The water chmistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the primary coolant. Chloride limits are specified to limit stress corrosion cracking of the stainless steel. During cold shutdown and refueling L operations, the temperature necessary for stress corrosion to occur is not present. Thus, a higher chloride concentration is allowed.
Action c: With the chloride concentration exceeding the limits for mre than 24 hours and conductivity or pH exceeding the limits for mre than 72 hours while in Operational Condition 4 or 5, this action statement requires a detemination be made prior to entering Operational Condition 3 that the structural integrity of the reactor coolant system rmains acceptable for continued operation. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter Operational Condition 4 and 5 by tensioning or detensioning the reactor vessel head with the reactor coolant chemistry limits not being satisfied, provided a determination is made prior to entry into Operational Condition 3 which dmonstrates that the structural integrity of the reactor coolant system remains acceptable for continued operation. Operational Condition 4 is entered from Operational Condition 5 by tensioning the reactor vessel head. The temperatures
-associated with Operational Condition 3 are not pemitted in Operational Conditions- 4 or 5. Therefore, changing modes between Operational Conditions 4 and 5 does not result in the reactor coolant tmperature conditions of concern (i.e., power operation). Thus, compliance with this action will ensure that the structural integrity of the reactor coolant system remains acceptable for continued operation and hence, will ensure that an equivalent level of safety is maintained during any mode changes into Operational Condition 4 or 5 by tensioning or detensioning the reactor vessel head with the chemistry limits not satisfied.
- 11. TS 3.4.5 - Specific Activity As stated in the bases to this TS, the limitations on the specific activity of the primary coolant ensure that the two hour thyroid and whole body dose resultim from a main steam line failure outside primary containment during s+eady state operation will not exceed small fractions of the dose widelines of 10CFR100. The values for the limits on 19
g B specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters, such as site'_ boundary locations .and meteorological conditions, were not considered in this evaluation. Action at With the primrv coolant's specific activity greater than 0.2 microcuries per gram dose equivalent I-131 but less.than or equal to 4.0 microcuries per gram dose equivalent I-131, this action statement -allows continued operation for lintited time periods. This. limited time period is provided to acemnodate possible lodine spiking phenmena which .may occur follcwing changes in thermal power. With primary coolant activity greater than 0.4 microcuries per gram dose equivalent I-131 microcuries per gram 'or_ between 0.2 and 0.4 microcuries per gram dose equivalent I-131 for greater than 48 hours while in Operational Condition 1, 2 or 3, the reactor is reqdired to be in Operational Condition 3 with the main steam line isolation valves closed. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter Operational Condition 3 with the reactor coolant's specific activity greater than the limits, provided the main l steam line isolation valves are clesod. .Since closing the main steam isolation valves. will prevent the release of radioactivity .to the environment should a main steam line break occur outside containment, ccrapliance with this action will ensure that the radiological releases ! fram such a postulated event would be within the conclusions of the offsite dose analyses. Therefore, compliance with this action will ensure that an equivalent level of safety is maintained during any node l- changes into operational Condition 3 with the specific activity of the L primary coolant outside the limits of this TS. Action b, c: With the primary coolant's specific activity outside the ; limits of the LCO, these action statements provide requirements to perform additional sampling and analysis. These action statements do not require any changes to plant conditions as a result of the out of limit condition, they merely ensure that the- requirements of Action a are satisfied by requiring more frequent monitoring. Therefore, cmpliance with these actions will ensure that the requirements of Action a are satisfied and will maintain an equivalent level of safety during any mode changes into operational Condition 1, 2, 3 or 4 with the specific activity of the primary ecolant outside the limits of this TS.
- 12. TS 3.4.9.2 - Residual Heat Removal (RHR) - Cold Shutdown The requirements of TS 3.4.9.2 ensure long term core cooling capability during cold shutdown conditions. This TS requires that two loops of RHR shutdown cooling be operable in Operational Condition 4. This TS also requires one RHR shutdown cooling node loop to be in operation unless one recirculation pump is in operation. A single RHR shutdown cooling loop provides sufficient heat rcm3 val capability for removing decay heat and sufficient reactor coolant mixing to assure accurate temperature l indication. However, as discussed in the bases for this TS, single 20 l
I ] failure considerations required that two loops of Mm shutdown cooling be operabic. Sufficient time is available to align alternate heat removal capability such that no im:odiate threat is inposed because of inoperable mm shutdown cooling mode' loops during cold shutdown conditions. Additionally, this heat rmoval capability is not assumed in any DBA analysis, Action at This action statement applies with less than the rcquired Mm shutdown cooling node loops operable while in operational Condition 4. This action requires the operability of an alternate method capable of i decay heat renoval be denonstrated within one hour and every 24 hours thereafter for each inoperable mm loop. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by i allowing the plant to enter operational Condition 4 by tensioning the reactor vessel head with' one or both m m shutdown cooling mode-loops inoperable and an alternate decay heat removal method available for each ! inoperable loop. As stated in GSU's amendment request dated Dec mber 16, 1988 (reference RBG-29573), credit for meeting this action can only be taken if the alternate nethod(s) chosen has enough heat rmoval capacity for the plant condition at that tino. Therefore, compliance with this action ensures that the function of the inoperable mm shutdown cooling node loop will be performed while preserving the capability to cope with a single failure without loss of safety function. Mditionally, node changes into operational Condition 4 have no effect on decay heat generation or rmoval capability. As a result, empliance with this Action will ensure that an equivalent level of safety will be maintained < during any mode changes into operational Condition 4 by tensioning the reactor vessel head with a Rim shutdown cooling mode loop (s) inoperable. This operational flexibility was also found acceptable on a one-time-basis in Amendment 36 to the RBS TS. Action b: This action statement appl- with no RHR shutdown cooling , node loop or recirculation pump L. operation while in operational Condition 4. This action requires reactor coolant circulation be established by an alternate method within one hour and reactor coolant temperature and pressure be monitored at least once per hour. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter operational Condition 4 by tensioning the reactor vessel head with no mm shutdown cooling mode loop
- or recirculation pump in operation and an alternate method of coolant circulation established and reactor coolant temperature and pressure being nonitored. As stated above, coolant circulation is required to provide sufficient mixing to assure accurate temperature indication. As '
described in GSU's letter dated December 16, 1988 (reference RBG-29573), operation of the reactor water cleanup systs in conjunction with maximum control rod drive flow has been shown to provide adequate coolant circulation in order to prevent temperature stratification and permit accurate temperature nnnitoring. Thus, establishing an alternate method of coolant circulation will perform the safety function of the RHR shutdown coo:.ing node loop or recirculation pump which is not in operation. Therefore, empliance with this action will ensure that an 21
equivalent level of safety will be maintained during any nose changes into operational Condition 4 by tensioning the reactor vessel head with coolant circulation not being perfomed by an operating RHR shutdown cooling node loop or recirculation pump. This operational flexibility was also found acceptable on a one-tino basis in Annndment 36 to the RBS TS.
- 13. TS 3.5.1 - Emergency Core Coolina Systems - Operating The mergency core cooling systm (ECCS) is cmprised of three divisions.
Division 1 consists of the low pressure core spray (LPCS) system, low pressure coolant injection (LPCI) train "A", and the autmatic depressurization system (ADS) as actuated by trip system "A". Division 2 consists of LPCI trains "B" and "C" and ADS as actuated by. trip system "B". The high pressure core spray (HPCS) system cmprises Division 3 of the ECCS. The safety function of these subsystems is to maintain adequate core cooling in the event of a loss of coolant accident (IOCA) , over the full range of reactor operating conditions, to ensure that fuel cladding. temperature does not exceed 2200 degrees F. The autmatic initiation of these systes is provided by the mergency core cooling system actuation instrumentation. A requirement of the ECCS is that cooling water flow to the reactor vessel be initiated rapidly when the system is called upon to perfom its function. Be keeping the ECCS discharge lines filled with water, the time required for injection is minimized and the effects of water hanner are reduced. The function of the ECCS discharge line " keep filled" pressure alam is to notify the control rom operator that the ECCS discharge piping is not sufficiently filled to prevent water hanmer damage. ' Action f: With the " keep filled" pressure alarm instrumentation channel inoperable while in operational Condition 1, 2 or 3, this action requires
.perfoming Surveillance Requirement 4.5.1.a.1 at least once per 24 hours.
Therefore, application of the proposed change to TS 3.0.4 will provide. operational flexibility by allowing the plant. to enter operational Conditions 1, -2 and 3 with the " keep filled" pressure alam instrumentation inoperable and performing Surveillance Requirement 4.5.1.a.1. Perfomance of this surveillance is an alternative means of verifying that the injection line is adequately charged by venting the discharge piping at the high point vents. Perfomance of this surveillance will ensure that leakage has not drained the ECCS discharge piping. Therefore, cmpliance with this action will ensure that the discharge piping is sufficiently filled and that an equivalent level of , safety is maintained during any node changes into Operational Condition 1, 2 or 3 with this " keep filled" pressure alarm instrumentation' inoperable. Action g: This action statement is administrative in nature in that it is not required to be entered in response to the inoperability of an ECCS system, but rather requires an evaluation of the effects of an ECCS '; injection and the subnittal of a special report to the NRC within 90 days. Application of the proposed change to TS 3.0.4 will clarify that 22
ep F the plant can enter Operational Conditions 1, 2 and 3 prior -to subnittal of. the required.special report to the ime following an ECCS injection. Since entry 'into this action. statement does not alter the operability-requirements of the ICO, carpliance with this action does not involve a reduction in the level of safety and does not imply inoperability or a failure to meet the ICO. Therefore, an equivalent level of safety will be maintained during any mode changes into operational Cordition 1, 2 or 3 prior to subitittal of this special report to the NBC following an ECCS injection.
- 14. TS 3.5.2 - Emergency Core' Cooling Systems - Shutdown The LPCS, LPCI, and HPCS systans are required to be available to provide reactor vessel inventory makeup following any event which causes
-inadvertent draining of the reactor vessel when it contains irradiated i fuel. 'Ib satisfy this objective, at least two ECCS subsystems / systems j are required to be operable per this TS during Operational Conditions 4 l and 5, except when the reactor cavity is flooded to greater than 23 feet i above the reactor vessel flange, the reactor vessel head is rmoved and I the upper containrtent fuel pool gate is opened. Conformance with the conditions- of this exception to the applicability ensures that adequate time is available to restore reactor vessel makeup in the event of j accidental draining. a Action a_: This action statement applies when only one of the required ]j ECCs are operable. This action requires all operations with a potential for dr' tining the reactor vessel be suspended if the required ECCS is not j restorei to-operable status within four hours. Therefore, application of ^
the proposed change to TS 3.0.4 will provide operational flexibility by I allowing the plant to enter and exit Operational Condition 5 by detensioning or tensioning the reactor vessel head and to exit the j exception to the applicability with one ECCS inoperable and operations with a potential for draining the reactor vessel suspended. Compliance jl with this cction greatly reduces the possibility of an accidental 1 draining event bv requiring the suspension of all activities with a : potential for draining the reactor vessel. Since one ECCS would still be i operable, adequate- water injection capability would still exist, i Therefore, canpliance with this action will ensure that equivalent level j of safety is maintained during any mode changes into operational l Condition 4 or 5 by tensioning or detensioning the reactor vessel head or exiting the exception to the applicability with one of the required ECCS inoperable, ,
- 15. TS 3.5.3 - Suppression Pool The suppression pool provides a primary source of water to the energercy core cooling systans (ECCS) in the event of a IOCA. It also provides a heat sink for safety / relief valve operation and steam released frar a IOCA as well as limiting radioactive material release through this portion of primary containment. TS 3.5.3 requires a suppression pool level of at least 19 feet, 6 inches in Operational Conditions 1, 2 and 3 l
23 0
, a
F 1
-and at least 13 feet, 3 inches in operational Conditions 4 and 5, unless specific conditions are m t for draining the suppression pool.. These suppression pool levels are- sufficient to provide the requtred-water. ,
supply to the ECCS. This TS-is applicable in operational Cond!tions 1,- 2, 3, 4 and 5, except when the reactor cavity is flooded to greater than 23 feet above the reactor vessel flange, the reactor vessel head is removed and the upper containment fuel pool gate is opentd. This exception is provided to require suppression pool operability durang cold i shutdown only when required to support ECCS operability. At RBS, a suppression pool pumpback system (SPPS) is provided which is designed to maintain suppression pool level in the event of a passive ECCS failure. Action d.2: This action statment applies with both SPPS subsystems inoperable while in Operational Condition 4 or 5. This action requires one alternate pumpback method be demonstrated operable within 24 hours and every 24 hours thereafter. If this demonstration is not performed, the action further requires core alterations and operations with a potential for draining the reactor vessel be suspended, the reactor mode switch be locked in the shutdown position and prhnary containment l integrity-fuel' handling be established. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter and exit Operational Condition 5. by detensioning or tensioning the reactor vessel head and exit the exception to the applicability with no SPPS subsystems operable and an alternate pumpback method' available or core alterations and operations with a potential for draining the reactor vessel suspended, the reactor mode switch locked in the shutdown position and primary containment integrity
- fuel handling established. Establishing an alternate pumpback method will ensure that the function of the SPPS will be performed. The SPPS function is not required unless a IOCA occurs and there is leakage in an ECCS pipe in a break exclusion area between the containment and first
' isolation valve. Also only one SPPS subsystem is necessary. Therefore I the failure of the SPPS alone does not result in the failure of the -
suppression pool or ECCS. With no alternate pumpback method established, suspending core alterat. ions and operations with a potential for draining , the reactor vessel and locking the mode switch in the shutdown position greatly reduces the possibility of an accidental reactor vessel draining event and subsequent need for ECCS and its water source, the suppression pool. Requiring primary containment integrity - fuel handling to be established will further reduce the offsite radiological consequences of such.-an accidental reactor vessel draining event. These actions will ensure that the offsite radiological consequences of such an event will l remin within the bounds of the USAR events. Therefore, cmpliance with this action will ensure that an equivalent level of safety is maintained _ during any mde changes into Operational Condition 4 or 5 by tensioning or detensioning the reactor vessel head or exiting the exception in the applicability with both SPPS subsystems inoperable. 24 l
m O
.1
.9 16.- TS '3.6.1.4 - Primary Containment ~ Air Incks
. The primarv containment air locks form part of the primary containment pressure boundary. As such, the air 1cck's safety function is related to control of offsite radiological releases resulting from a DBA. Thus, the air lock's . structural integrity and leak tightness are essential to the successful mitigation of such an event. This TS requires the primary containment air locks to be operable in Operational Conditions 1, 2 and 3, when handling irradiated fuel in the primary containment and during core alterations and operations with the potential-for draining the reactor vessel. The air lock door inflatable seal system ensures the leak t{ghtnessoftheairlock. Air lock door seal pressure is normally monitored by installed pressure instrumentation.
Action d: This action statcutent applies with one prinary containment air lock . door inflatable seal system air flask pressure instrumentation channel inoperable. This action requires the air flask pressure be verified to be greater than or equal to 9(; psig at least once per 12 , hours'if the pressure instrumentation channel is not restored to operable status within seven days. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter Operational Condition 1, 2 or 3 or c rmience core alterations, handling of irradiated fuel in the primary containment or conduct operations with a potential for draining the reactor vessel with this pressure instrumentation inoperable and the air flask pressure verified to be greater than or equal to 90 psig. Since the air lock is equipoed p with air accumulators which have been sized to keep the seals inflatad i for 30 days ~ in the event of a loss of air supply, the 12 hout surveillance frequency is more than frequent enough to ensure adequate pressure to the door seals. Therefore, canpliance with this action will ensure that an equivalent level of safety is maintained during any ~ node changes with this primary containment air lock door inflatable seal-system air flask pressure instrunentation inopere.ble. !
- 17. TS 3.6.2.3 - Drvwell Air locks The drywell air -lock forms part of the drywell pressare boundary. As such, drywell air lock operability is necessary for drywall integrity to i ensure that the steam released for the full spectnm of drywell pipe '
breaks is condensed inside prinary containment by the suppression pool. By utilizing the suppression pool as a heat sink, energy released to the primary containment is minimized, the severity of the transient is reduced, and primary containment integrity is thereby maintained. Since the air lock's structural integrity and leak tightness are essential to the successful mitigation of these events, the drywell air lock's safety function requires that it be closed and its leakage be within design limits. The requirements of this TS are applicable in Operational Conditions 1, 2 and 3. The air lock door inflatable seal system ensures
'the leak tightncss of the air lock. Door seal pressure is normally nnnitored by installed oressure instrumentation.
25 i
r
+
a. Action c: This action statement applies with one drywell air lock door inflatable scal system air flask pressure instnnentation channel inoperable. This action requires the air flask pressure be verified to be greater than or equal to 75 psig at least once per 12 hours if the l pressure instrumntation channel is not restored to operable status ! within. seven days. Therefore, application of the proposed change to TS j i 3.0.4 will provide operational flexibility by allowing the plant to enter operational Condition 1, 2 and 3 with this pressure instrumentation inoperable ami the air flask pressure verified to be greater than or equal to 75 psig. Since the air lock is equipped with air accumulators 3 which have been sized to keep the seals inflated for 30 days in the event of a loss of air supply, the 12 hour surveillance frequency is nore than ; frequent enough to ensure adequate pressure to the door seals. Therefore, cmpliance with this action will ensure that an equivalent level of safety is maintained during any mode changes into Operational Conditions 1, 2 or 3 with this drywell air lock door inflatable seal ! system air flask pressure instrumentation inoperable.
- 18. TS 3.6.3.1 - Suppression Jool Operability of the suppression pool ensures that the drywell and primary containment pressures will not exceed their design pressures during primary system blowdown frm full operating pressure. The suppression pool also provides a heat sink for safety / relief valve operation and condenses steam released during a LOCA to limit radioactive material release through this portion of primary containment. TS 3.6.3.1 requires a minimum suppression pool level of 19 feet 6 inches with a maximum of 20 feet 0 inches to ensure the suppression pool volume is within the-required range to meet the primary containment structural design constraints imposed during accident conditions. The average temperature of the suppression pool is also limited, to 95 degrees F average during normal plant operations with a maximum allowable of 120 degrees F with the main steam isolation valves shut following a reactor scram, to ensure that the peak pool tmperature does not exceed post-ILCA design requirements. Installed instrumentation provide control rom operators visual confirmation of suppression pool level and toq;erature.
Action c: This action statement applies with only one suppression pool water level indicator and/or seven suppression pool temperature indicators in separate locations operable wnile in Operational Condition 1, 2 or 3. This action requires the suppression pool level and/or temperature be verified to be within the limits once per 12 hours if the inoperable indicators are not restored to operable status within seven days. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter Operational Conditions 1, 2 and 3 with reduced indication of suppression pool level and/or temperature and the affected parameter being verified to be within its ljnits. In this condition, adequate instrumentation remains axilable to determine suppression pool level and average temperature. Additionally, failure of this instrumentation does not, in itself, cause the suppression pool to becme unable to perform its safet9 26
t
'L !
-function. Therefore, empliance with tnis action will ensure that an equivalent level of safety will be maintained during any node changes into operational' CbMition 1, 2 or 3 with on?v one suppression pool level 3 aM/or - seven suppi ,ssion pool tmperature irdicators in separate !
locations operable. l i
- 19. TS 3.7.1.1 - Standbv Sett ce i Water System ;
l The standby service weter NSW) system removes the decay heat ganerated .f by the reactor core and heat fim plant auxiliaries which require cooling ' water during an emergency shutdown of the plant. The SSW tystem i transfers this heat to the ultimate heat sink. Operability of .the SSW j system' and ultimate heat sink ensure that sufficient cooling capacity is 2 available for continued operation of safety-related equipnent during l normal and accident conditions. i 1 The SSW system consists of two subsystems, each containing two 3SW system pumps. Division I of the SSW system serves Division I plant loads with ! h' the "A" and "C" SSW oumps and Division II of the SSW system cerves Division II plant loads with the "B" and "D" SSW pumps. The "A" SSW prnp a is supplied energency onsite AC power frm the Division I diesel generator, the "B" and "D ' SSW pumps are supplied mergency onsite AC . power frm the Division II diesel generator and the "C" SSW pump is 3 i supplied emergency onsite AC power frm the Division III diesel generator. The Division III diesel generator can be cooled by either division of the SSW system. This arrangment assures adequate cooling to the Division III diesel generator in the event of a failure of either diesel generator following a loss of offsit: pc.?r.
-In Operational Conditions 1, 2 and 3, both SSW subsyste m are required to I be operable. In Operational Conditions 4 and 5 and when handling irradiated fuel in the primary containment or Fuel Building,. the .!
subsystem (s) associated with the required RHR shutdown cooling mode- I loop (s), ECCS and diesel generators is required to be operable. Action a: This action statment applies with the SSW system flev path to j one _or nore systems or cmponents inoperable. This action requires the $ associated systems or emponents ba declared inoperabk. Thus, this Q action merely establishes system /cmponent operability with respect to j SSW system flowpath operability and then transfers the required action to the appropriate action statements for the affected system. Application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to make node changes while cmplying with those systs actions that allow continued plant operation for an unlimited period of time. These system actions will ensure that the appropriate plant response is taken. These system actions will, therefore, determine whether continued plant operation for an unlimited period of time is , allowed and whether a node change is allowed by the proposed change to TS ' 3.0.4. Since the conditions required by the applicable action statements for the affected system must be established prior to making the mode change, empliance with this action will ensure that an equivalent level l ,. 27 l I
V
}
of_' safety is maintained during any mode changes while cmplying with those system actionc. that allow ' continued plant operation for an unlimited period of time with the SSW system flow path to one or more j systems or camponents inoperable. Action c: This action statement applies with only one SSW pump and its - associated flow path operable while in operational Condition 4 or 5 or j when handling irradiated fuel in the primary . containment or -fuel j building. .This action recraires, in Operational Condition 4 or 5, the ; associated equipment be declared inoperable. Thus, this action merely i'
- establishes system operability with respect to SSW system operability and l
then transfers the r< quired action to the appropriate action statement i for the affected systar,. Application of the proposed change to TS 3.0.4 I will_ provide operational flexibility by allowing the plant to enter Operational Condition 4 and 5 by tensioning or detensiondag the reactor vessel head while emplying with those system actions that allow continued plant operation for an unlimited period of time. These system actions will ensure that the appropriate plant resrunse is taken. These ' system actions will, therefore, determine whether continued plant operation for an unlimited period of time-is allowed and whether a mode j L change is allowed by the proposed change to TS 3.0.4. v Action c also requires verifying adequate cooling for the required diesel l generators is available while handling irradiated fuel in the primary ' containment or fuel building. If adequate cooling.is not available, - the diesel generators are required be declared inoperable. Therefore, this action merely establishes diesel generator operability with respect to , cooling water availability and then transfers the required action to the ' applicable action' statements of TS 3.8.1.2. Application of the proposed o change to TS 3.0.4 will provide operational flexibility by allowing the plant to cannence handling of irradiated fuel in the primary containment 1 and fuel building if _ adequate cooling is available for the diesel generators or while emplying with those system actions that allow _l continued plant operation for an unlimited period of tine. Cmpliance with the applicable actions of TS 3.8.1.2 will ensure that appropriate ! plant response is taken. The actions of TS 3.8.1.2 will, therefore, ! determine whether continued plant operation for an unlimited period of , time is allowed and whether a mode change is allowed by the proposed .l change to TS 3.0.4. 4 Since the conditions required by the applicable action statements for the affected system must be established prior to making the mode change, compliance with these actions will ensure that an equivalent level of { safety is maintained during any node changes into Operational Condition 4 " or 5 by tensioning or detensioning the reactor vessel head or cmmencing handling of irradiated fuel in the primary containment or fuel building while emplying with those system actions that allow continued plant operation for an unlimited period of time with only one SSW pump and its associated flow path operable. 28 I d 1
7 1 . j
]
- 20. TS 3.7.2 - Mah Control Rom Air Conditioning System The . main control room mergency filtration system is a subsysts of the f main control room ventilation system consisting of HEPA filters, charcoal j beds, heaters and fans. The operability of- the main control rom air. b conditionini system ensures that the ambient main control rom air temperature does not exceed the allowable temperature for-continuous duty rating for the equipnent and instrumentation cooled by this system and that the main control rom will rmain habitable for operations personnel during and follming all design basis accidents.- The operability of this system, in conjunction with control rom design provisions,. is based on limiting the radiation exposure to personnel occupying the main control rocm to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General. Design Criterion 19 of Appendix A to 10CFR50.
When an actuation signal is present, resulting fran a sensed loss of coolant accident (as indicated by high drywell pressure or low reactor vessel water level), the main control room ventilation sys:em enters into the emergency mode of operation. In the mergency mode, tie main control rom is 1solated and the emergency filtration systal starts. The emergency filtration system circulates main control rom recycle air and fresh makeup air through the charcoal beds and HEPA filters to remve any radioactive airborne particulate prior to releasing it to the main control rom envelope. Separate outside air intakes are furnished to ' provide alternative sources of makeup air with the capability of ; selecting either source fram the main control rom. Redundant radiation mnitors in the main control rom air supply duct system monitor .the radioactivity level in the outside supply air. With this systs in operation, main control roam habitability is maintained since the main control room air is continuously filtered to remve airborne particulate prior to entering the control rom envelope. The main control rom air conditioning system is required to be operable in all operational conditions and while handling irradiated fuel in the primary containment or fuel building. Action b.1: This action statement applies with one main control rom air conditioning air handling unit / filter train-subsystem inoperable while in Operational Condition 4 or 5 or while handling irradiated fuel in the primary containment or fuel building. This' Action requires the operable subsystem be placed into operation in the emergency mode if the inoperable subsystem is not restored to operable status within seven days. By requiring the operable subsystem to be operating in the mergency mode, this action ensures that the safety-related function of the systm is being positively performed. Application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to enter or exit Operational Condition 5 by detensioning or tensioning the reactor vessel head or catmonce handling of irradiated fuel in the primary containment or fuel building with one subsystem inoperable and the operable subsystem operating in the emergency mode. 29
p u A Since placing the operable subsystem into operation' in the emergency node - ensures that the safety. function is being positively perfomed, empliance with this action will ensure that an equivalent level of safety is maintained during any node changes into Operational Condition 4 or 5.by tensioning or detensioning the reactor vessel head or ccnmencing handling of irradiated fuel in the primary containment or fuel building with one main control rocm air conditioning air handling unit / filter train subsystem inoperable. This operational flexibility was also found acceptable on a one-time basis in Amendment 36 to the PSS TS. Action b 2: This action statement applies with both main control rocm air conditioning air handling unit / filter train subsystems inoperable while in Operational Condition 4 or 5 or while handling irradicted fuel in the primary containment or fuel building.- This action requires core alterations, handling of irradiated fuel in the primary containment and fuel building and operations with a potential for draining the reactor vessel be suspended. - Conpliance with this action removes the conditions
' for which an accident is assumed to occur, and for which the main control rocm mergency filtration systoni is required to operate. Application of the proposed change to TS 3.0.4 will provide operational flexibility by '
allwing the plant to enter or exit Operational Cordition 5 by detensioning or tensioning the reactor vessel head with both main control ! .Ioam air conditioning air handling unit / filter train- subsystems inoperable. Perfornance of these mode changes do not provide any additional conditions during which an accident is postulated to occur. Theafore, cmpliance with this action will ensure that an equivalent level of safety is maintained during any mode changes into Operational Condition'4 or 5 by tensioning or detensioning the reactor vessel head with both main control rocm air conditioning air handling unit / filter train subsystems inoperable. This operational flexibility was also found acceptable on a one-time basis in Amendment 36 to the RBS TS.
- 21. TS 3.7.4 - Snubbers Snubbers are required to be operable to ensure that the structural integrity of the reactor coolant system and all'other safety-related systems is maintained during and after a seismic or other event that initiates dynamic loads. Therefore, the snubber's safety function is to ,
contribute to the operability of the system to which it is attached. The requirements ;of TS 3.7.4 are applicable while in Operational Conditions 1, 2 and 3 and while in Operational Condition 4 and 5 for snubbers located on systems required to be operable in those operational conditions. Action: The action statement for this TS requires the inoperable snubber (s) to be replaced or restored to operable status and an engineering evaluation of the attached cmponents be cmpleted within 72 hours. If these actions are not completed, the affected system is required to be declared inoperable and the appropriate action for that systm followed. This action has the effect of establishing operability of the attached system or ccuponent with respect to snubber operability 30
g 4 and then ' transferring the required action to the appropriate action statement of the attached syst m. This action does not allow a reduction , in safety, but rather establishes a system or cmponent's operability in ' relation to operability of its snubbers. 1 Application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing plant operation under virtually any condition, provided the inoperable snubber (s) has been restored or replaced and an li engineering evaluation has been cmpleted showing that the attached -l cmponent remains capable of meeting its designed service. 'With the j inoperable snubber repaired and a detennination made that no damage has , -l been done, the system remains fully operational, no 32npact on safety j function 'is possible and empliance with the Ico is restored. It is :! noted.that the inoperable snubber would be repaired or replaced and the ! detennination empleted (thus restoring cmpliance with the IID) before ! changes frm one operational condition to another would be allowed. 1 Alternately, the attached system must be declared inoperable and the appropriate Action Statements for the affected system must be followed.
.In this case, the applicable action statement for the affected system ;
will detennine whether continued plant operation is allowed for an _i unlimited period of time and whether a mode change is allowed by the l proposed change to 3.0.4. Since the conditions required by the l applicable Action Statment for the affected system must 'be established l prior to making the mode change, cmpliance with this action will. ensure o that an equivalent level of safety is maintained during any mode changes ' with inoperable snubber (s), i
- 22. TS 3.7.6.1 - Fire Suppression Water Syst s The operability of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish - fires l occurring in any portion of the plant where safety-related equipnent is k located. The fire suppression systems consist of the water system, spray [
and/or sprinkler systems, halon systems and fire hose-stations. As stated in the bases for this TS, the collective capability of the fire suppression systems is adequate to minimize potential damage to ! safety-related equipnent and is a mjor element in the plant fire l protection program. The requirennnts of TS 3.7.6.1 are applicable at all : times. . Action b: This action statement applies with the fire suppression water f system inoperable due to having more than one pump and/or mre than one water supply inoperable. This action requires a backup fire suppression water systs be established within 24 hours. Therefore, application of , the proposed change to TS 3.0.4 will provide operational flexibility by ?
. allowing modo changes with more than one fire suppression system pump and/or more than one water supply inoperable and a backup supply established. Cmpliance with this action ensures that the safety function of this system will be performed if required. Additionally, node changes have no effect on whether or not this system would be required to perfonn its safety function. Therefore, cmpliance with this 31 l
l 0
] .. i
'4 1 action will ensure -that. an equivalent level of safety is maintained , during- any node changes with the fire suppression water system ! inoperable.- l
- 23. TS 3.7.8 - Area Temperature Monitoring
'Ihe area tmperature limitations ensure that safety-related equipmnt will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures 'may degrade equip mnt and can cause loss of its operability. The tmperatures are required to be within the limits shwn in TS Table 3.7.8-1 whenever the equiprent in the associated area is required to be operable. ,
i Action a: This action statement applies with one or nere areas exceeding the temperature limit (s) shown on Table 3.7.8-1 by less than 30 degrees F for more than eight hours. This action requires a -special report be subnitted to the NRC to provide a record of the amount by which and the cumulative time that the temperature in the affected area exceeded its limit and to provide an analysis which demonstrates continued operability F of the affected equipent. Entry into this action is an administrative-response to exceeding the temperature lintits of one or more of the areas indicated in Table 3.7.8-1. This Action requires that an evaluation of the effects of exceeding taiperature limits be performed. Cmpliance with this action either dmonstrates that the affected equipent is operable, thereby ensuring the level of safety provided by the Loo is maintained, or establishes that the affected equipmnt is inoperable which transfers the required action to the applicable action statements of the affected equipnent. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing mode changes-with temperatures in one or more areas less than 30 degrees F above the limit (s) of Table 3.7.8-1 following an evaluation of the effects of the out of limit temperature. If the equipent is determined to be inoperable,.the applicable actions of the affected equipnent will determine whether continued plant operation is allowed for an unlimited period of time and whether a mode change is allowed by the proposed-change to TS 3.0.4. Since the conditions required by the actions of the affected system must be established prior to making the mode change,. cmpliance with this action will ensure that an equivalent level of safety is maintained during any mode changes with one or more area temperatures less than 30 degrees F in excess of the limits of Table 3.7.8-1. Action b: If the temperature in one or nore areas exceeds the limit . shown in Table 3.7.8-1 by more than 30 degrees F, this action requires the equipmnt in the affected area I4 be declared inoperable. This action has the effect of definina operability of a system or cmponent with respect to area tempereture and then transferring.the required action to the applicable action statements of the affected equipnent. This action does not allow a reduction in safety, but rather defines a system or cmponent's operability in relation to the area temperature. 32 J
E , j s;
~'
l As such,- the applicable actions of the affected system will determine whether continued plant operation is allowed for an unlimited period of 3 time and whether a mode change is allowed by the proposed change to TS 3.0.4. Since the conditions required by the applicable action statements of the affected equipnant must be established prior to mking the mde change, ccupliance with this action will ensure that an equivalent level of safety is maintained during any mode changes with the temperature in one or mre areas more than 30 degrees F in excess.of the limits of Table 3.7.8-1. 24.'rS 3.8.1.2 - AC Sources - Jhutdown The operability of the minimum AC electrical power sources specif.!ed during shutdown' and refueling is required to ensure that: (1) the facility can be maintained in the shutdown or refueling condition for an extended period of time, and (2) sufficient instrumentation and contrc.1 , capability in available for mnitoring and mintaining the unit status. l The requiremnts of this TS are applicable in operational conditions 4 ! ( and 5 and when handling irradiated fuel in the primary containment or fuel building. Action b: This action statement applies when the Division III diesel generator is inoperable when required to support operability of the high .I pressure core spray (HPCS) system per TS 3.5.2 or TS 3.5.3. This action requires the Division III diesel generator be restored to operable status within 72 hours or the HPCS and "C" standby service water (SSW) system i pump be declared inoperable. Thus, this action merely establishes . operability of the HPCS and "C" SSW pump with respect to Division III ' diesel generator operability and then transfers the required action to the _ applicable action statem nts of the associated TS. The action statements of the associated TS provide the appropriate actions to be taken and will determine whether continued plant operation is allowed for an unlimited period of time and whether a mde change is allowed in accordance with the proposed change to TS 3.0.4. Since the conditions required by the applicable action statements of the TS for the HPCS and "C" SSW pump must be established prior to making the mode change, compliance with this action will ensure that an- equivalent level of safety' is maintained during any mode changes while ccmplying with those action statements that allow continued plant operation for an unlimited period of time with the Division III diesel generator inoperable.
- 25. TS 3.8.2.2 - DC Sources - Shutdown The operability of the minimum DC electrical power sources specified during' shutdown and refueling is required to ensure that: (1) the facility can be maintained in the shutdown or refueling condition for an extended period of time, and (2) sufficient instrumentation and control capability is available for mnitoring and maintaining the unit status.
The requirements of this TS are applicable in Operational Conditions 4 and 5 and when handling irradiated fuel in the primary containment or fuel building. 33 i l'
1 Action b: This action statement applies with' the Division III DC. 't electrical power source inoperable when required to support operability of the HPCS per TS 3.5.2 or TS 3.5.3. This action requires the HPCS and "C" SSW-system pump to be declared inoperable. Thus, this action merely establishes operability of the HPCS and "C" SSW pump with respect to : Division III DC electrical power source operability and then transfers i the required action to the applicable action statements of the associated , TS. The action statements of the associated TS provide the appropriate ! actions to be taken and will determine whether continued plant operation is allowed for an unlimited period of time and whether a mode change is ; allowed in accordance with the proposed change to TS 3.0.4. Since the ! conditions. required by the applicable action statenents-of the TS for the i HPCS and "C" SSW pump must be established prior to. making the mode change, empliance with this action will ensure than an equivalent level ! of safety is maintained during any node changes into Operational Condition 4-or 5 by tensioning or detensioning the reactor- vessel head while emplying with those action statenents that alloa continued plant operation for an unlimited period of time with the Division III DC electrical power source inoperable, i
- 26. TS 3.8.3.2 - Distribution - Shutdown The operability of the AC and DC electrical power distribution systems ;
during shutdown and refueling is required to ensure that: (1) the facility can be maintained in the shutdown or refueling condition for an extended period of time, and (2) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status. The requirements of this TS are applicable in Operational Conditions 4 and 5 and when handling irradiated fuel in the primary containment or fuel building. Action a.2: This action statement applies with the Division III AC electrical power distribution system de-energized when- required to support operability of the HPCS per TS 3.5.2 or TS 3.5.3. This action : requires the HPCS and "C" SSW system pump be declared inoperable. Thus, this action merely establishes operability of the HPCS and "C" SSW pump with respect to Division III AC electrical power distribution system ' status and then transfers the required action the applicable action statements of the associated TS. The action statenents of the associated i TS provide the appropriate actions to be taken and will determine whether continued plant operation is allowed for an unlimited period of time and 1 whether a made change is allowed in accordance with the proposed change to TS 3.0.4. Since the conditions required by the applicable action
, statements of the TS for the HPCS and "C" SSW pump must be established prior to making the mode change, compliance with this Action will ensure that an equivalent level of safety is maintained during any node changes into Operational Condition 4 or 5 by tensioning or detensioning the reactor vessel head while cmplying with those action statements that allow continued operation for an unlimited period of time with the Division III AC electrical power distribution syetem de-energized.
34 (
t Action b.2: This action statenent ' applies with the Division III DC electrical power distribution system de-energized when required to support operability of the HPCS per TS 3.5.2 or TS 3.5.3. . This action requires the HPCS and "C" SSW system pump be declared inoperable. Thus, this action merely establishes operability of the HPCS and "C" SSW pump with respect to Division III DC electrical power distribution system status and then transfers the required action to the applicable action statenonts of. the associated TS. The action statomonts of the associated TS provide the appropriate actions to be taken and will determine whether , continued plant operation is allowed .for an unlimited period of time and l whether a mode change is allowed in accordance with the proposed change j to TS 3.0.4. Since the conditions required by the applicable action j statements of the TS for the HPCS and "C" SSW pump must be established. 1 prior to making the node change, cmpliance with this action will ensure ? that an equivalent level of safety is maintained during any node changes t into operational Condition 4 or 5 by tensioning or detensioning the reactor vessel head while ccmplying with those action statenents that lg allow continued plant operation for an unlimited period of tino with the '2 Division III DC electrical power distribution system de-energized. I
- 27. TS 3.9.11.1 - Residual Heat Removal and Coolant Circulation - High Water Invel 1
The requirements of TS 3.9.11.2 ensure long term core cooling during ! refueling conditions. This TS requires that at least one RHR shutdown ! cooling loop be in operation in Operational Condition 5 with the reactor cavity water level greater than 23 feet above the reactor vessel flange and irradiated fuel in the reactor vessel. The requirement that at least one RHR loop be operable and in operation ensures that: (1) sufficient i, cooling capacity is available to remove decay heat and maintain the water 1 in the reactor pressure vessel below 140 degrees F as required during l Operational Condition 5, and (2) sufficient coolant circulation wauld be i available through the reactor core to assure accurate temperature indication. This heat renoval capability is not assumed in.any DBA analysis. Action a: This action statement applies with no RHR shutdown cooling mode loop operable while in Operational Condition 5 with the reactor < cavity water level greater than 23 feet above the reactor vessel flange. l This action requires operability of an alternate method capable of decay 1 heat renoval be demonstrated within one hour and every 24 hours thereafter. If the alternate method cannot be demonstrated the ACTION requires containment integrity fuel handling be established and the requirements of ACTION 6 are followed. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility by allowing the plant to increase reactor cavity water level to greater than . 23 feet above the reactor vessel flange with no RHR shutdown cooling node j loop operable and an alternate method of decay heat removal available. l As stated in GSU's amendment request dated December 16, 1988 (reference RBG-29573), credit for meeting this action can only be taken if the alternate method (s) chosen has enough heat removal capacity for the plant 35 l k
~.
C' condition at that tino. Therefore, cmpliance with this action ensures that the function of the-inoperable M m shutdown cooling mode loop will be perforned. With the reactor vessel head rm oved and 23 feet of water above the reactor vessel flange, a large heat sink is available for core-cooling. Thus, in the event of a failure of the alternate nethod of
' decay heat removal, adequate tino is provided to initiate other alternate nethods of decay heat rmoval or energency procedures to cool the core.
Additionally, node changes by increasing the reactor cavity water level to greater than 23 feet above the reactor vessel flange hava no effect on decay heat generation or re oval capability. As & result, compliance with this - action will ensure that an equivalent level of safety will be naintained during any node changes into Operational Condition 5 with the reactor cavity water level greater than 23 feet above the reactor vessel flange with less than the required mm shutdown cooling node loop operable.
- 28. TS 3.9.11.2 - Residual Heat Renoval and Coolant Circulation - Im Water Level The requirements of TS 3.9.11.2 ensure long tenn core cooling during cold shutdown conditions. This TS requires two loops of RHR shutdown cooling be operable .while in Operational Condition 5 with the reactor cavity water level less than 23 feet above the reat.cor vessel flange. This TS also requires one RHR shutdown cooling node loop to be in operation. A single RHR shutdown cooling loop provides sufficient heat rmoval capability for rmaving decay heat and sufficient reactor coolant mixing to assure accurate temperature indication. However, as discussed in the bases for this TS, single failure considerations required that two loops of' RHR shutdown cooling be operable. Sufficient time is available to align alternate heat renoval capability such that no imnediate threat is imposed because of inoperable RHR shutdown cooling mode loops during cold-shutdown conditions. Additionally, this heat removal capability is not assumed in any DBA analysis.
Action at This acticr. statement applies with less than the required RHR shutdown cooling mode loops operable while in Operational Condition 5 with the reactor cavity water levr1 less than 23 feet above the reactor vessel flange. This action requires operability of an alternate method capable of decay heat removal be demonstrated within one hour and every 24 hours thereafter for each inoperable RHR loop. Therefore, application of the proposed change to TS 3.0.4 will provide operational flexibility. by allowing the plant to enter Operational condition 5 with the reactor cavity water level less than 23 feet above the reactor vessel flange by detensioning the reactor vessel head or draining the reactor cavity with one or both RHR shutdown cooling node loops inoperable and an alternate decay heat removal method available for each inoperable loop. As stated in GSU's amendment request dated December 16,1988 (reference RBCr 29573), credit for meeting this action can only be taken if the alternate method (s) chosen has enough heat removal capacity for the plant condition at that time. Therefore, cmpliance with this action ensures that the function of the inoperable RHR shutdown cooling mode loop will be 36
n = a i i
)
performd while preserving the capability to cope with a single failure f without loss of ' safety function. Additionally, mode changes to } Operational Condition 5 with the reactor cavity level less than 23 feet ,l above the reactor vessel flange by draining the reactor cavity or ) detensioning the reactor vessel head have no effect on decay heat j generation or remval capability. As a result, ccmpliance with this ! action will ensure that an equivalent level of safety will be maintained } during any mde changes into operational Condition 5 with the reactor d cavity water. level less than 23 feet above the reactor vessel flange with a M m shutdown cooling node loop (s) inoperable. This. operational. flexibility was also found acceptable on a one-tim basis in Amndment 36 to the RBS TS. ! Action b: This action statemnt applies with no MR shutdown cooling { mde loop in operation while in Operational Condition 5 with the reactor ' cavity water level less than 23 feet above the reactor vessel flange.- , This action requires reactor coolant circ Ration be established by an [ alternate method within one hour and reactor coolant tmperature be i monitored at least once per hour. Therefore, application of the proposed ; change to TS 3.0.4 will provide operational flexibility by allowing he l plant to enter Operational Condition 5 with the reactor cavity water- j level less than 23 feet above the reactor vessel flange by detensioning ; the reactor vessel head or draining the reactor cavity with no m m. ! shutdown cooling mode loop in operation and an alternate method of W coolant circulation in operation. As stated above, coolant circulation ; is required to provide sufficient mixing to assure accurate temperature indication. As described in GSU's letter dated December 16, 1988 ] (reference RBG-29573), operation of the reactor water cleanup system in J conjunction with maximum control rod . drive flow or operation of a i recirculation pump have been shown to provide adequate coolant j circulation in order to prevent temperature stratification and pemit accurate temperature monitoring. Thus, establishing an alternate method i of. coolant circulation will perform the safety function of the RHR y shutdown cooling mde loop which is not in operation. Therefore, compliance with this action will ensure 'that an equivalent level of safety will be maintained during any mde changes into operational Condition 5 with the reactor cavity water level less than 23 feet above the reactor vessel flange with coolant circulation not being performed by an operating Ma shutdown cooling mode loop. This operational
' flexibility was also found acceptable on a one-time basis in Amendment 36 g to the RBS TS. 1 l
l' 29. TS 3.9.12 - Inclined Fuel Transfer Systm The requiremnts of this TS ensure the operability of barriers prohibiting personnel access to the inclined fuel transfer system (IETS) 1 area are in place prior to transferring irradiated fuel or control rods I I through IFIS. This restriction provides protection for plant personnel frcm high radiation areas in addition to these required in TS 6.12.2. j Therefore, the safety function of this equipmnt is to prevent 'q uncontrolled personnel access to high radiation areas. l l 37 l
9 Js , n --.
# . y '
a, , e '
.s
;g. ,. s , < >
' l9 -
s t Action:a: This. action stat m ent applies when the barriersJ to .the> IFTS '
"iG. area are inoperable during periods when the systm -is;in use. 'Ihe Action.-
w requires: entry- to- thef area be.. prohibited,,a- continuous . watch be' established and .the area.be conspicuously posted. These actions lare.in excess'of those"in TS 6.12.2, which require the area be L' conspicuously : -)
', posted, roped offmand a' warning device.be activated but access may be- ,
? allowed. The. effect of 1 implmenting c this ' action ; would- .' continue to' . J w provide- the functions required .by .the I40 thereby conplying with the intent of the restrictions. As a result, 'cmpliance l with this.., action.
'. will' ensure, that an equivalent level 'of-
. safety .is ' maintained by-
- 7. : precluding inadvertent ~ entry into1IPTS . areas.:duringH fuel transfer ,
l Q 1 4 s 3 , i I l
-{i l
l li; s
'I'. s 'f f 3 n <
7
.k
; {'
-ij, 1 ;
1 e 'I l:i '_ ']- i' ;! i- : ._c., . , [ '. >>, 38 i I i i 3 . , = . - . .
1-I
'mBLE - 1 i
Twhnical Specification (TS) Actions for which the pmposed change to TS 3.0.4 . will provide new operating flexibility:
- 1. TS 3.2.4.? - Rod Pattern Control Syst m - ;
Action a.1 i Action a.2 j Action b.2 ' Action b.3 .;
- 2. TS 3.3.1 - Reactor Protection Systm Instrumentation -L Action b. i Action 2 'I
. Action 3- j Action 5 Action 6 Action-7 -l Action 8 i Action 9 !
Action 10 !
'3. TS 3.3.2 - Isolation Actuation Instrumentation '!
Action a ' Action c l Action 21 i Action 23 )
' Action 25 Action 26 Acthn 27 '
Action 28 Action 29 i Action 30 -l
- 4. TS 3.3.3 - Emergency Core Coolina System Actuation Instrumentation -l Action a Action b Action 32 -I Action'37 j
- 5. TS 3.3.4.2 - End-of-Cycle Recirculation Pump Trip Syst m Instrumentation Action a- I Action b Action col ;
- 6. TS 3.3.5'- Reactor Core Isolation Cooling System Actuation Instrumentation ,
Action a ' Action b Action 52 Page 1 of 3 l o
.- . d c
-4 T
[. . ' l[ ,
- 7. TS 3.3.6 - Control %1 Block Instrumentation
. Action 6 l 3 Action b Actir/1 60.
- 8. - 'IS 3.3'.7.5 - Accident Wmitoring Syste Action 81
- 9. TS 3.4.3.2 - operational IAykage
- Action c
- 10. TS 3.4.4 - Ch mistr ,y Action c.1 L 11. TS 3.4.5 - Specific Activity Action a la Action b Action c
- 12. TS 3.4.9.2 - Residual Heat Removal - Cold Shutdown Action a Action b
- 13. TS 3.5.1 - nnergency Core Cooling syst e s - operating Action f e-Action g
- 14. TS 3.5.2 - nnergency Core Cooling Systes - Shutdown 7 Action a
- 15. TS 3.5.3 - Suppression Pool Action d.2
. 2 6. TS 3.6.1.4 - Trimrj containment Air Incks Action d
- 17. TS 3.6.2.3 - Drywell- Air Inck Action c
- 18. TS 3.6.3.1 - Suppression Pool.
Action c !
- 19. TS 3.7.1.1 - Standby Service Water Systm Action a Action c
' 20._ . TS 3.7.2 - Main control Room Air conditioning Systm Action b.1
- ~ ~
Action b.2 1 Page 2 of 3 ) i
- l M/
+
i
- 21. ' TS 3.7.4 - Snthrs
' Action
- 22. TS 3.7.6.1'- Fire Su pression Water System Action b
- 23. TS 3.7.8'- Area Tweerattr e Monitoring Action a-Action b
- 24. TS 3.8.1.2 - A. C. Sources - Shutdown Action b
- 25. TS 3.8.2.2 - D. C. Sources - Shutdown Action b
- 26. TS 3.8.3.2 - Distribution - Shutdown Action a.2 Action b.2
- 27. TS 3.9.11.1 - Residual Heat Renoval and Coolant Circulation - High Water-Invel Action a 28.' TS 3.9.11.2 - Residual 1! eat Rcunaval and coolant Circulation - Im Water level Action a-Action b'
- 29. TS 3.9.12 - Inclined Fuel Transfer System Action a l
i i Page 3 of 3
,, t
+}}