ML18066A506: Difference between revisions

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This change is coruiiste with NUREG-1432 as modifi d by industry own 's rou eneric chan e TSTF-30. M.2 CTS 4.S.3d specifies requirements for ensuring that all locked-closed manual containment isolation valves are closed and locked except those open under administrative control. The proposed ITS addresses manual containment isolation valves and blind flanges for both outside and inside containment in proposed SR 3.6.3.2 and SR 3.6.3.3 respectively.
This change is coruiiste with NUREG-1432 as modifi d by industry own 's rou eneric chan e TSTF-30. M.2 CTS 4.S.3d specifies requirements for ensuring that all locked-closed manual containment isolation valves are closed and locked except those open under administrative control. The proposed ITS addresses manual containment isolation valves and blind flanges for both outside and inside containment in proposed SR 3.6.3.2 and SR 3.6.3.3 respectively.
Blind flanges are specified in the proposed ITS since they are also isolation devices which must be in the closed position to function properly.
Blind flanges are specified in the proposed ITS since they are also isolation devices which must be in the closed position to function properly.
The CTS requires that this verification be performed prior to "the reactor going critical after a refueling outage." The proposed ITS SR 3.6.3.3 revises this to state prior to "entering MODE 4 from MODES if not performed within the previous 92 days." The proposed revision is needed to match the Applicability of Specification  
The CTS requires that this verification be performed prior to "the reactor going critical after a refueling outage." The proposed ITS SR 3.6.3.3 revises this to state prior to "entering MODE 4 from MODES if not performed within the previous 92 days." The proposed revision is needed to match the Applicability of Specification 3.6.3 which is MODES 1-4. This verification is modified by the statement "if not performed within the previous 92 days" to ensure that for shutdowns after extended times in MODES 1-4, that the valve status is reverified.
 
====3.6.3 which====
is MODES 1-4. This verification is modified by the statement "if not performed within the previous 92 days" to ensure that for shutdowns after extended times in MODES 1-4, that the valve status is reverified.
The proposed ITS SR 3.6.3.2 addresses the manual containment isolation valves and blind flanges* which are outside containment and requires that they be verified closed every 31 days. The addition of the 31 day frequency is to address the fact that since these valves are outside containment, they are in a location which would be more susceptible to inadvertent mispositioning than the devices inside containment and therefore need to be verified more frequently.  
The proposed ITS SR 3.6.3.2 addresses the manual containment isolation valves and blind flanges* which are outside containment and requires that they be verified closed every 31 days. The addition of the 31 day frequency is to address the fact that since these valves are outside containment, they are in a location which would be more susceptible to inadvertent mispositioning than the devices inside containment and therefore need to be verified more frequently.  
,, Palisades Nuclear Plant Page 7 oflO 01/20/98 /;;J-b   
,, Palisades Nuclear Plant Page 7 oflO 01/20/98 /;;J-b   
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APPLICABILITY:
APPLICABILITY:
MODES l, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. Containment A.1 Restore containment inoperable.
MODES l, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. Containment A.1 Restore containment inoperable.
to OPERABLE status. B. Required Action and B.1 Be in MODE 3. associated Completion Time not met . AND B.2 . Be in MODE 5. SURVEILLANCE REQUIREMENTS SR 3. 6 .1.1 SURVEILLANCE Perform required visual examinations and Type A leakage rate testing in accordance with the Containment Leak Rate Testing Program. Containment  
to OPERABLE status. B. Required Action and B.1 Be in MODE 3. associated Completion Time not met . AND B.2 . Be in MODE 5. SURVEILLANCE REQUIREMENTS SR 3. 6 .1.1 SURVEILLANCE Perform required visual examinations and Type A leakage rate testing in accordance with the Containment Leak Rate Testing Program. Containment 3.6.1 COMPLETION TIME 1 hour 6 hours 36 hours
 
====3.6.1 COMPLETION====
 
TIME 1 hour 6 hours 36 hours
* FREQUENCY In accordance with the Containment Leak Rate Testing Program Palisades Nuclear Plant 3.6.1-1 Amendment No. 05/31/99   
* FREQUENCY In accordance with the Containment Leak Rate Testing Program Palisades Nuclear Plant 3.6.1-1 Amendment No. 05/31/99   
* *
* *
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* Perform required Type B and C leakage rate testing, except for containment air lock testing, in accordance with 10 CFR 50, Appendix J, Option A, as by approved exemptions.
* Perform required Type B and C leakage rate testing, except for containment air lock testing, in accordance with 10 CFR 50, Appendix J, Option A, as by approved exemptions.
The leakage rate acceptance criterion is 1.0 La. However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakage rate acceptance criteria are < 0.6 La for the Type B and Type C tests
The leakage rate acceptance criterion is 1.0 La. However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakage rate acceptance criteria are < 0.6 La for the Type B and Type C tests
* Containment  
* Containment 3.6.1 FREQUENCY In accordance with the Containment Structural Integrity Surveillance Program ------NOTE--'---*
 
====3.6.1 FREQUENCY====
 
In accordance with the Containment Structural Integrity Surveillance Program ------NOTE--'---*
SR 3.0.2 is not applicable.
SR 3.0.2 is not applicable.
In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions Palisades Nuclear Plant 3.6.1-2 Amendment No. 05/31/99 I i
In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions Palisades Nuclear Plant 3.6.1-2 Amendment No. 05/31/99 I i
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* SURVEILLANCE Perform a system functional test for each hydrogen recombiner.
* SURVEILLANCE Perform a system functional test for each hydrogen recombiner.
Visually examine each hydrogen recombiner enclosure and verify there is no evidence of abnormal conditinns.
Visually examine each hydrogen recombiner enclosure and verify there is no evidence of abnormal conditinns.
Perform continuity and a resistance to ground test for each heater phase . Hydrogen Recombiners  
Perform continuity and a resistance to ground test for each heater phase . Hydrogen Recombiners 3.6.7 FREQUENCY 18 months 18 months 18 months Palisades Nuclear Plant 3.6.7-2 Amendment No. 05/31/99   
 
====3.6.7 FREQUENCY====
 
18 months 18 months 18 months Palisades Nuclear Plant 3.6.7-2 Amendment No. 05/31/99   
* *
* *
* Containment B 3.6.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.l Containment BASES BACKGROUND The containment consists of a concrete structure lined with steel plate, and the penetrations through this structure.
* Containment B 3.6.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.l Containment BASES BACKGROUND The containment consists of a concrete structure lined with steel plate, and the penetrations through this structure.
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.(&sect; All automatic containment.
.(&sect; All automatic containment.
i,Olation valves are OPERA E or are locked closed. .
i,Olation valves are OPERA E or are locked closed. .
* The uncontro ed containment leakag satisfies Specif1ca on 4.5. CONTROL RODS CONTROL RODS shall be shutdown and regulating rod . ent that provides cycle specific parame er limits ad cycle. These cycle specific paramet shall be determine for each reload cycle in accordance wi Specification  
* The uncontro ed containment leakag satisfies Specif1ca on 4.5. CONTROL RODS CONTROL RODS shall be shutdown and regulating rod . ent that provides cycle specific parame er limits ad cycle. These cycle specific paramet shall be determine for each reload cycle in accordance wi Specification 6.6.5. Plant operation within these limits is addressed in individual SP. cifications.
 
====6.6.5. Plant====
operation within these limits is addressed in individual SP. cifications.
ENT 1-131 shall be that concentration o 1-131 (&#xb5;Ci/gm) which ala would produce the same thyroid dose as he quantity and isotopic ixture of I-131, I-132, I-133, I-134 an I-135 actually present The thyroid dose conversion factors us for this calculation shall e those listed in Table III o'f TID-14844 "Calculation of Dist ce Factors for Power and Test Reactor Si *1-2 Amendment No. 3*, i4, i+, ii. Hi Revised 05/31/99 See 3.ta.5) See. J_!,._\/
ENT 1-131 shall be that concentration o 1-131 (&#xb5;Ci/gm) which ala would produce the same thyroid dose as he quantity and isotopic ixture of I-131, I-132, I-133, I-134 an I-135 actually present The thyroid dose conversion factors us for this calculation shall e those listed in Table III o'f TID-14844 "Calculation of Dist ce Factors for Power and Test Reactor Si *1-2 Amendment No. 3*, i4, i+, ii. Hi Revised 05/31/99 See 3.ta.5) See. J_!,._\/
4.4
4.4
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le . At le'ast one door in each a;; lock is properly clo{ed and. sealed. 3.1..:1 .. )  lffi le. uncontrolled leakage satisfies sr-cification 4.5.
le . At le'ast one door in each a;; lock is properly clo{ed and. sealed. 3.1..:1 .. )  lffi le. uncontrolled leakage satisfies sr-cification 4.5.
1------------------------------***
1------------------------------***
CONTROL ROOS CONTROL ROOS shall -length shutdown and regulating rods. The COLR is the docume that provides cycle specific parameter imits for the current reloa cycle. These cycle specific parameter mits shall be determined or each reload cycle in accordance with Specification  
CONTROL ROOS CONTROL ROOS shall -length shutdown and regulating rods. The COLR is the docume that provides cycle specific parameter imits for the current reloa cycle. These cycle specific parameter mits shall be determined or each reload cycle in accordance with Specification 6.6.5 Plant operation within these limits is ddressed in individual Spe fications.
 
====6.6.5 Plant====
operation within these limits is ddressed in individual Spe fications.
DOSE EQUIVA NT 1-131 shall be that concentration of -131 (&#xb5;Ci/gm) which alon would produce the same thyroid dose as e quantity and isotopic xture of 1-131, 1-132, I-133, I-134 and -135 actually present. The thyroid dose conversion factors use for this calculation shall those listed in Table III of TI0-14844, Calculation of Oista e Factors for Power and Test Reactor Sits." 1-2. Amendment No. i4, i+, ii, Revised 05/31/99 *   
DOSE EQUIVA NT 1-131 shall be that concentration of -131 (&#xb5;Ci/gm) which alon would produce the same thyroid dose as e quantity and isotopic xture of 1-131, 1-132, I-133, I-134 and -135 actually present. The thyroid dose conversion factors use for this calculation shall those listed in Table III of TI0-14844, Calculation of Oista e Factors for Power and Test Reactor Sits." 1-2. Amendment No. i4, i+, ii, Revised 05/31/99 *   
*
*
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* ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES MORE RESTRICTIVE CHANGES (M) M.1 Not used. M.2 CTS 4.5.3d specifies requirements for ensuring that all locked-closed manual containment isolation valves are closed and locked except those open under administrative control. The proposed ITS addresses manual containment isolation valves and blind flanges for both outside and inside containment in proposed SR 3.6.3.2 and SR 3.6.3.3 respectively.
* ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES MORE RESTRICTIVE CHANGES (M) M.1 Not used. M.2 CTS 4.5.3d specifies requirements for ensuring that all locked-closed manual containment isolation valves are closed and locked except those open under administrative control. The proposed ITS addresses manual containment isolation valves and blind flanges for both outside and inside containment in proposed SR 3.6.3.2 and SR 3.6.3.3 respectively.
Blind flanges are specified in the proposed ITS sinc.e they are also iso_lation devices which must be in the closed position to function properly.
Blind flanges are specified in the proposed ITS sinc.e they are also iso_lation devices which must be in the closed position to function properly.
The C_TS requires that this verification be performed prior to "the reactor goii:ig critical after a refueling outage. " The proposed ITS SR 3. 6. 3. 3 revises this to state prior to "entering MODE 4 from MODE 5 if not performed within the previous 92 days." The proposed revision is needed to match the Applicability of Specification  
The C_TS requires that this verification be performed prior to "the reactor goii:ig critical after a refueling outage. " The proposed ITS SR 3. 6. 3. 3 revises this to state prior to "entering MODE 4 from MODE 5 if not performed within the previous 92 days." The proposed revision is needed to match the Applicability of Specification 3.6.3 which is MODES 1-4. This verification is modified by the statement "if not performed within the previous 92 days" to ensure that for shutdowns after extended times in MODES 1-4, that the valve status is reverified.
 
====3.6.3 which====
is MODES 1-4. This verification is modified by the statement "if not performed within the previous 92 days" to ensure that for shutdowns after extended times in MODES 1-4, that the valve status is reverified.
The proposed ITS SR 3.6.3.2 addresses the manual containment isolation valves and blind flanges which are outside containment and requires that they be verified closed every 31 days. The addition of the 31 day frequency is to address the fact that since these valves are outside containment, they are in a location which would be more susceptible to inadvertent mispositioning than the devices inside containment and therefore need to be verified more frequently.
The proposed ITS SR 3.6.3.2 addresses the manual containment isolation valves and blind flanges which are outside containment and requires that they be verified closed every 31 days. The addition of the 31 day frequency is to address the fact that since these valves are outside containment, they are in a location which would be more susceptible to inadvertent mispositioning than the devices inside containment and therefore need to be verified more frequently.
These changes are considered to. be More Restrictive changes since the CTS only requires the verification to be performed prior to going critical, does not require verification between refueling outages, and does not address blind flanges. This change is consistent with NUREG-1432.
These changes are considered to. be More Restrictive changes since the CTS only requires the verification to be performed prior to going critical, does not require verification between refueling outages, and does not address blind flanges. This change is consistent with NUREG-1432.
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\ \ 3.6-6 SR 3.0.2 is not applicable In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions Revised 05/31/99   
\ \ 3.6-6 SR 3.0.2 is not applicable In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions Revised 05/31/99   
* * *
* * *
* Containment Air Locks and  
* Containment Air Locks and 3.6.2 SURVEILLANCE RE UIREMENTS continued SR 3.6.2.2 CEOG STS SURVEILLANCE FREQUENCY  
 
====3.6.2 SURVEILLANCE====
 
RE UIREMENTS continued SR 3.6.2.2 CEOG STS SURVEILLANCE FREQUENCY  
-----------
-----------
-------NOTE--
-------NOTE--
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Only requ* ed to be perf rmed upon entry or exit thr gh the conta' ment air lock. Verify only one door in the air lock can be 1 'i' opened at a time . \I 3.6-7 Rev 1, 04/07 /95 Revised 05/31/99   
Only requ* ed to be perf rmed upon entry or exit thr gh the conta' ment air lock. Verify only one door in the air lock can be 1 'i' opened at a time . \I 3.6-7 Rev 1, 04/07 /95 Revised 05/31/99   
* *
* *
* Containment Isolation Valves /(Atmo,{Dheric  
* Containment Isolation Valves /(Atmo,{Dheric 3.6 CONTAINMENT SYSTEMS 3.6.3 Containment Isolation Valves j(Atmofa'pheric and LCO 3.6.3 Each containment isolation valve shall be OPERABLE.
 
===3.6 CONTAINMENT===
 
SYSTEMS 3.6.3 Containment Isolation Valves j(Atmofa'pheric and LCO 3.6.3 Each containment isolation valve shall be OPERABLE.
APPLICABILITY:
APPLICABILITY:
MODES 1, 2, 3, and 4. ACTIONS 0.f'J. IL ; ,;..: ,. t'OO""'
MODES 1, 2, 3, and 4. ACTIONS 0.f'J. IL ; ,;..: ,. t'OO""'
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Rev 1, 04/07 /95 Revised 05/31/99   
Rev 1, 04/07 /95 Revised 05/31/99   
* *
* *
* Containment Isolation Valves l(Atijl(}spheric  
* Containment Isolation Valves l(Atijl(}spheric 3.6.3 ACTIONS continued CONDITION C.
 
====3.6.3 ACTIONS====
continued CONDITION C.
Only applicable to penetration flow paths with only one containment isolation valve and a closed system. C.l One or more AND penetration flow paths with one containment C.2 @ isolation valve inoperable.
Only applicable to penetration flow paths with only one containment isolation valve and a closed system. C.l One or more AND penetration flow paths with one containment C.2 @ isolation valve inoperable.
L Seco dary containment byp ss leakage not wi in limit. ,. One or mo penetrat with contat t purge valv .110t within pu valve leakage 1111 s. CEOG STS 0.1 E. l ,, REQUIRED ACTION Isolate the affected penetration flow path by use of at least one closed and de---activated automatic valve, closed manual valve, or blind flange. --------NOTE---------
L Seco dary containment byp ss leakage not wi in limit. ,. One or mo penetrat with contat t purge valv .110t within pu valve leakage 1111 s. CEOG STS 0.1 E. l ,, REQUIRED ACTION Isolate the affected penetration flow path by use of at least one closed and de---activated automatic valve, closed manual valve, or blind flange. --------NOTE---------
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Rev 1, 04/07 /95 Revised 05/31/99   
Rev 1, 04/07 /95 Revised 05/31/99   
* *
* *
* c.1s J.4.1 Canta i nment Cooling Systems J (Atijl6spheri c ajj'd Dua ffi-10  
* c.1s J.4.1 Canta i nment Cooling Systems J (Atijl6spheri c ajj'd Dua ffi-10 3.6 CONTAINMENT SYSTEMS 3.6.681 LCO 3. 6. 66\ Two containment lsprji trains and/two containffiintl cooling trains shall be OPE BLE. APPLICABILITY:
 
===3.6 CONTAINMENT===
 
SYSTEMS 3.6.681 LCO 3. 6. 66\ Two containment lsprji trains and/two containffiintl cooling trains shall be OPE BLE. APPLICABILITY:
HODES 1, Z,  
HODES 1, Z,  
/and 4f.j ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME c.:\t 3.'\.1. <:.T'S cq.,-,.,.,::o.-e..
/and 4f.j ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME c.:\t 3.'\.1. <:.T'S cq.,-,.,.,::o.-e..
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I CD G) & APPLICABILITY:
I CD G) & APPLICABILITY:
MODES l and 2. ACTIONS CONDITION A. One hydrogen A.l recombiner inoperable.
MODES l and 2. ACTIONS CONDITION A. One hydrogen A.l recombiner inoperable.
: 8. Two hydrogen recombiners inoperable.  
: 8. Two hydrogen recombiners inoperable.
 
8.1 REQUIRED ACTION N).._>) --------NOTE---------
===8.1 REQUIRED===
ACTION N).._>) --------NOTE---------
LCO 3.0.4 is not applicable.
LCO 3.0.4 is not applicable.
Restore hydrogen recombiner to OPERABLE status . COMPLETION TIME 30 days Verify by l hour administrativ.
Restore hydrogen recombiner to OPERABLE status . COMPLETION TIME 30 days Verify by l hour administrativ.
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: 6. (continued)
: 6. (continued)
: 7. 8. In addition, this condition is similar to that in NUREG-1432 Section 3.5.2 which would allow one or more trains to be inoperable as long as at least 100 % of the ECCS flow equivalent to a single OPERABLE ECCS train is available.
: 7. 8. In addition, this condition is similar to that in NUREG-1432 Section 3.5.2 which would allow one or more trains to be inoperable as long as at least 100 % of the ECCS flow equivalent to a single OPERABLE ECCS train is available.
Specification  
Specification 3.6.6 requires two trains of containment cooling to be OPERABLE with the components for each train discussed above. The trains alignments are dictated by the associated diesel generator which would provide electrical power in the event of a loss of offsite power. However, from a heat removal standpoint, any' containment spray pump can provide sufficient cooling in conjunction with the containment air coolers to satisfy the analytical requirements.
 
====3.6.6 requires====
two trains of containment cooling to be OPERABLE with the components for each train discussed above. The trains alignments are dictated by the associated diesel generator which would provide electrical power in the event of a loss of offsite power. However, from a heat removal standpoint, any' containment spray pump can provide sufficient cooling in conjunction with the containment air coolers to satisfy the analytical requirements.
Likewise, any two containment spray pumps can provide sufficient cooling without reliance on the containment air coolers. (Although at least one air cooler fan is required for mixing of the containment atmosphere.)
Likewise, any two containment spray pumps can provide sufficient cooling without reliance on the containment air coolers. (Although at least one air cooler fan is required for mixing of the containment atmosphere.)
The Completion Time of 72 hours is reduced from the NUREG-1432 allowance of 7 days since the original design diversity does not exist. In addition, the NUREG-1432 Conditions B, C, D, E, and Gare not applicable to the Palisades design since complete redundancy between the containment spray system and the containment air coolers at the system level does not exist as discussed above. The appropriate changes are made to the TS and Bases to reflect these changes. This change is a plant specific change to reflect the Palisades Nuclear Plant design and analysis.
The Completion Time of 72 hours is reduced from the NUREG-1432 allowance of 7 days since the original design diversity does not exist. In addition, the NUREG-1432 Conditions B, C, D, E, and Gare not applicable to the Palisades design since complete redundancy between the containment spray system and the containment air coolers at the system level does not exist as discussed above. The appropriate changes are made to the TS and Bases to reflect these changes. This change is a plant specific change to reflect the Palisades Nuclear Plant design and analysis.

Revision as of 17:22, 5 May 2019

Forwards Responses to NRC Questions for ITS LCOs 3.6.3 & 3.6.6 of 990126 Submittal.One Editorial Change in Addition to Those Made in Response to NRC Comments & Conforming Changes Made to Associated Bases,Encl
ML18066A506
Person / Time
Site: Palisades Entergy icon.png
Issue date: 06/17/1999
From: HASKELL N L
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9906280246
Download: ML18066A506 (522)
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Text

{{#Wiki_filter:A CMS Energy Company June 17, 1999 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, Ml 49043 Tel: 616 764 2276 Fax: 616 764 2490 Nathan L. Haskell Director, Licensing DOCKET 50-255 -LICENSE DPR-20 -PALISADES PLANT -CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS-PARTIAL RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION -ITS SECTION 3.6, CONTAINMENT On January 26, 1998, Consumers Energy Company submitted a Technical Specifications Change Request (TSCR) to revise the Palisades Technical Specifications* to closely emulate the Standard Technical Specifications for Combustion Engineering Plants, NUREG-1432. On January 26, 1999, the NRC requested additional information regarding Sections 3.6, Containment, and 3. 7 Plant Systems, of that TSCR. This letter provides responses to the NRC questions for ITS LCOs 3.6.3 and 3.6.6 of our January 26, 1998 submittal. Responses to NRC comments on ITS Section 3.7 were submitted on March 30, 1999; responses to NRC comments on the remaining Section 3.6 LCOs were submitted on June 4, 1999. This letter also includes one editorial change in addition to those made in response to the NRC comments. That change revises the wording of ITS SR 3.6.3.2 and 3.6.3.3 from: " ... not locked, sealed, or otherwise secured, and ... "to I; I " ... not locked, sealed, or otherwise secured in position, and ... ". Conforming changes were also made to the associated Bases for these SRs. The revised wording makes the SR wording agree with the wording of all simiiar alignment SRs in the ( ITS. The revised page is included in Enclosure

2. Enclosure 1 to this letter contains:

a) answers to the Request for Additional Information (RAI) \ and, b) markups of the previously submitted pages to show where revisions have been made. 0 '. lt{ .fr1.v 1' --__ ....:___ _____ l) . 9906280246 990617 -----*--------*-----, PDR ADOCK 05000255 I P PDR -t (,, \ \ i I I .1v L {/I , \ ro,; u . I " . v'IW \ ' { 1[';.'J (1w Ul 1.t* I a ri\ I Enclosure 2 contains revised pages for Section 3.6, along with a list of revised pages and instructions for page replacement. These revised pages reflect changes resulting from our responses to the Section 3.6 RAI questions answered both in this letter and in our June 4, 1999 letter as well as the editorial change identified above. Each revised page is dated for identification. The NRC RAI of January 26, 1999 referred to an expected response date of March 31, 1999. Subsequently, in a telephone conversation with the NRR Project Manager for Palisades, Consumers Energy received permission to delay the Section 3.6 response to allow additional time for preparation and internal review.

SUMMARY

OF COMMITMENTS This submittal contains no new commitments and no revisions to existing commitments. CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRG Resident Inspector -Palisades Enclosures ... '. CONSUMERS ENERGY COMPANY PARTIAL RESPONSE TO JANUARY 26, 1999 RAI To the best of my knowledge, the content of this response to the NRG Request for Additional Information dated January 26, 1999 concerning Section 3.6 of our January 26, 1998 License Amendment request for conversion to Improved Technical Specifications, is truthful and complete. Director, Licensing Sworn and subscribed to before me this Mary Ann Engle, Notary4>ublic Berrien County, Michigan (Acting in Van Buren County, Michigan) My commission expires February 16, 2000 ', /7flv dayof 1999. 0

  • *
  • ENCLOSURE
  • 1 CONSUMERS ENERGY COMPANY PALISADES PL.ANT DOCKET p<l-255 CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS PARTIAL RESPONSE.*TO;:JANUARY 26, 1999 REQUEST FOR ADDITiONAL INFORMATION SECTION*

NTAINMENT ... \

  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 RE°i'.:'!UEST FOR ADDITIONAL INFORMATION SECTION 3.6,..CONTAINMENT
\' NBC REQUEST: 3.6.3 Containment Isolation Valves.<:

3.6.3-1 DOC A.1 DOC M.3 DOC LA.4 (ITS 3.6.1) JFD 2 CTS 1.0 CONTAINMENT INTEGRl:tY, Item e CTS 4.5.2.a.(4).(a) CTS 4.5.2.a.(4).(d) CTS 4.5.2.b.(1) CTS 4.5.2.d.(1) CTS 4.2 Table 4.2.2 Item 13.b .,_ STS SR 3.6.1.3.11 and Associated Bases (NUREG 1434) STS SR 3.6.1.3.14 and Associated* Bases (NUREG 1433) STS SR 3.6.3.6 and Associated: Bases ITS SR 3.6.3.5 and Associated Bases ITS B 3.6.3 Bases -References=-*-* See Comment Numbers 3.6.1-1, 3.6.1-7, and 3.6. 1"*8; *Comment: See Comment Numbers 3.6.1-1, :3.6.-1-7, and 3.6.1-8 . Consumers Energ,V Response: See response to RAl's 3.6.1-1, 3.6.1-7, and 3.6.1-8 Affected Submittal Pages: No page changes . 1

  • * * , CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-2 DOC A.1 DOCA.4 CTS 3.6.1.a ITS 3.6.3 APPLICABILITY and Associated Bases See Comment Number 3.6.1-2. Comment: See Comment Number 3.6.1-2.
  • Consumers Energy Response:

See response to RAI 3.6.1-2 Affected Submittal Pages: No page changes . 2 \ '\ . ;. l ;:'. . .

  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-3 DOC A.1 CTS 3.6.1 ACTION CTS 3.6.1 ACTION states that "With one or more containment isolation valves inoperable (including during the performance of valve testing), maintain at least one isolation valve OPERABLE in each affected penetration that is open ... ". The CTS 3.6.1 ACTION phrase "(including during the performance of valve testing)" is shown in the CTS markup as being deleted. This deletion was justified as an Administrative editorial change (DOC A.1 ). This is incorrect.*

While the change can be considered as an Administrative change in that it reflects current practice that is reflected in ITS LCO 3.0.2, it is not an editorial change. Comment: Provide a discussion and justification for the deletion of this phrase in CTS 3.6.1 ACTIONS. Consumers Enerqv Response: A new DOC (DOC A.17) has been added to discuss the deletion of the phrase "including during the performance of valve testing." Affected Submittal Pages: Att 3, CTS page 3-40, (ITS 3.6.3, page 1 of 5) Att 3, DOC 3.6.3, page 6 of .10 \\ 3 i ! RA I * ( Atib A Noi!. > <g> t...C..O 3.6

c. When pt>sitive reactivity es are made by boron di tion or CONTROL ROD motion (except* r testing one CONTROL R at a time}. ACTION: 1.5 psig when above COLD SHUT and 1.0 psig when in POWER OPE ION or HOT STANDBY. With containment internal pres re above the limit, r tore pressure t within the limit within 1 hou , or be in at least HO SHUTDOWN within the next 6 hours and in COLD HUTDOWN within the fo owing 30 hours. 3.6.4 Two ndependent containment hydroge recombiners shall be th plant is in POWER OPERATION or. OT STANDBY. With one ecombiner i operable, restor1 th1 inoperab recombiner to OPERABL status within r be in at least HOT S TDOWN within the next * (fj) <(AOI>

'S <ADD Aet"'I OrJ c. c.-k..{

  • *
  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES A.15 ({Al A* I f.g A.11 A, i<< tJe.u.J f\)t.t>.J

-Sc.c.. 1rJ3 ,,_ i *,...,s Cltt S<--<.. Palisades Nuclear Plant Action A.2 and D.2 which sta Ai+ I 3.b. 1-z. I -3 Rr+ 1 3.(,. 3 *S Page 6of10 01/20/98 3-b

  • I 'b./D.3 . (...C.O 3.6 3'".6.3 3.6.4 L.c.,o J. .... J ( At)t) (}run A Abe> ACTION: 1.5 psig when above COLD SHUT 1.0 psig when in POWER OPE and With containment internal pres re above the limit, r tore pressure t within the limit within 1 hou , or be in at least HO SHUTDOWN within the next 6 hours and in COLO HUTOOWN within the fo owing 30 hours.
  • *
  • ATTACHMENT 3 DISCUSSION CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES

......... A.15 ctlon A.2 and D.2 which sta A. l(g Ne.W -St.c.. 1Nl tf. i AA l 3.'7. 1-z A.11 rJr..1.>1 I -.3 A-Nt.,.,.; I RA-I 3.(,.3*5 Palisades Nuclear Plant Page 6of10 01/20/98 5-d

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  • DOC 3.6.3 -A.17 The Actions of CTS 3 .6.1 state, in part, "with one or more containment isolation valves inoperable (including during performance of valve testing) maintain at least one ... " In ITS 3.6.3, it is not necessary to included the parenthetical phrase "including during performance of valve testing" since LCO 3.0.2 stipulates that the Required Actions associated with a Condition be taken whenever the LCO is not met. Thus, excluding this phrase in the ITS is considered an administrative change since the actual requirements of the CTS remains unchanged .
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-4 DOC A.2 CTS 3.6.1 ACTION ITS 3.6.3 Condition A and Associated Bases CTS 3.6.1 ACTION states that "With one or more containment isolation valves inoperable (including during the performance of valve testing), maintain at least one isolation valve OPERABLE in each affected penetration that is open ... ". ITS 3.6.3 Condition A has a Note which makes the Condition applicable only if there are two Containment Isolation Valves (CIVs) in the penetration flow path which ensures one CIV remains OPERABLE while the other CIV may be inoperable.

DOC A.2 was prepared to justify the adding of ITS 3.6.3 Condition A Note; however, it was never used nor identified in the CTS markup. Comment: Revise the CTS markup to correct this discrepancy. Consumers Energy Re§JJonse: The CTS markup has been revised to show the addition of the Note to ITS Condition A which states "only applicable to penetration flow paths with two containment isolation valves." This change is described in DOC A:2.

  • Affected Submittal Pages: Att 3, CTS 3-40 (page 1 of 5)
  • 4
  • A tJoC > l.C..O 3.6 J.6.3 3.6.4 L::.oJ.i..J
  • c. When pt>sit1Ye reactivity ch,a ** es are made by boron di tion or CONTROL ROD motion (except* r testing one CONTROL R at a time). ACTION: 1.5 psig when above COLD SH 1.0 psig when in POWER OPE and With containment internal pres re above the limit, r tore pressure t within the limit within 1 hou , or be in at least HO SHUTDOWN within the next 6 hours and in COLD HUTOOWN within the fo owing 30 hours. Two nd1p1ndent containment hydroge recombiners shall be th plant is in POWER OPERATION or. OT STANDBY. With one ecombiner i operable, restore the inoperab recombiner to OPERABL status within r be in a least HOT S TOOWN within the next
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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS . RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-5 DOC A.7 JFD6 JFD 10 JFD 15 CTS 3.6.1 ACTIONS CTS 3.6.5 ITS 3.6.3 ACTIONS B, D and F and Associated Bases CTS 3.6.5 specifies that the containment purge exhaust and air supply isolation valves shall be locked closed whenever the plant is above COLD SHUTDOWN.

With one containment purge exhaust or air supply isolation valve not locked closed, the valve shall be locked closed within 1 hour *or a shutdown shall be started and completed in 36 hours. The corresponding ACTIONS in the ITS are ITS 3.6.3 ACTION D and F. CTS 3.6.1 ACTIONS which deals with inoperable CIVs have been modified to add a new Condition per ITS 3.6.3 ACTION B where the only two containment isolation valves in the same flowpath are inoperable. As stated in DOC A.7, under the CTS 3.6.1 ACTIONS, the same situation would have resulted in entering CTS 3.0.3. ITS 3.6.3 ACTION B is not applicable to purge exhaust valve or air* room supply valve not locked closed or leakage. Based on the descriptions and discussions in ITS B 3.6.3 Bases, JFD 6, JFD 10 and JFD 15, the purge exhaust system and the air room supply system consist of two valves in series located outside containment for each penetration. Thus if both valves in the penetration are not locked closed, the CTS does not specify an ACTION which would result in entry into CTS 3.0.3. Since ITS 3.6.3 ACTION D only addresses one valve per penetration being inoperable and ITS 3.6.3 ACTION B excludes these valves, no ACTION is provided for when both valves in either the purge exhaust or the air room supply penetrations .are not locked closed. Comment: Revise the CTS/ITS markup to address this situation. Provide the appropriate discussions and justifications for this change. It should be noted that this change may result in the re-arrangement of ITS 3.6.3 ACTIONS. Consumers Energy Response: ITS 3.6.3 Condition D has been revised to address "one or more purge exhaust or air room supply valves not locked closed." A new DOC (DOC A.18) has been provided to justify this change. Conforming changes have been made to the Bases and CTS markup. Affected Submittal Pages: Att 1, ITS 3.6.3, page 3.6.3-3 Att 2, ITS 3.6.3, page B 3.6.3-9 Att 3, CTS page 3-40 (ITS 3.6.3, page 1 of 5) Att 3, DOC 3.6.3, page 6of10 Att 5, NU REG 3.6.3, page 3.6-10 Att 5, NUREG 3.6.3, page B 3.6-26

  • I 5
  • ACTIONS CONDITION
c. ---------NOTE---------

C.1 Only applicable to penetration flow paths with only one containment isolation valve and a closed system. ---------------------- One or more A.till penetration flow paths with one containment C.2 isolation valve inoperable.

  • or rnore. D. OneApurge exhaust or D.1 air room supply valves not locked closed.
  • Palisades Nuclear Plant Containment Isolation Valves 3.6.3 REQUIRED ACTION COMPLt TION TIME -Isolate the affected 72 hours penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. --------NOTE---------

Isolation devices in high radiation areas may be verified by use of administrative means. --------------------- Verify the affected Once per 31 days penetration flow path is isolated. Lock closed the 1 hour affected A \ 3.6.3-3 5-c.-Amendment No. 01/20/98 RAI . .5' ACTIONS (continued)

  • * . D"'-htiur IS Pmvd-tJ -fo Jocr: CJ +he. : 0. H'cc-b:i VC.S. rhc. I hwr CmPk11o,, lirn<. fV'Tiv1dcs
a. pw-1 oJ of *hrrK fo C4rrc.ct -t+l.,

w 1-fh 1 o{ ffk.O<-Vh \ \1-ur to -mtRi i aRse valve-. \In the event one or more ntainment purge valves in one or penetration flow pat s are not within the purge va ve jleakage limits, purge va ve leakage must be restored t limits, or the ected penetration must be is ated. he method of isolatio must be by the use of at leas one isolation barrier that cannot be adversely affected y a single active failure Isolation barriers that mee this . riterion are a clos and de-activated automatic alve with resilient seals, or blind flange. A purge valv with seals uti ized to satisfy Required Acti n E.l must ave been demonstr ed to meet the leakage requi ements of R 3.6.3.5. The ecified Completion Time is r asonable, fonsidering that ne containment purge valve r ains closed so that a gross reach of containment does no exist. In accordance '"th Required Action E.2, this penetration !flow path mus be verified to be isolated a periodic basis. The p riodic verification is necessary to ensure that contain ent penetrations required to be isolated following a accident, which are no longef capable of being automatica y isolated, will be in the position should an vent occur. This Required Aftion does not require y testing or valve manipulat'on. Rather, it involves verification, through a syst walkdown, that those isolati n devices outside containmen capable of being mispos* ioned are in the correct po tion. For the isola on devices inside containmen , the time period Ii speci ied as "prior to entering MO E 4 from MODE 5 if not perf rmed within the previous 92 ays" is based on 1

  • eng neering judgment and ts reasonable in view of I th inaccessibility of the isol ion devices and other ! a inistrative controls that wi l ensure that isolation
device misalignment is an unli ely possibility
  • Palisades Nuclear Plant B 3.6.3-9 01/20/98 5-b
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  • BASES ACTIOMS Containment Isolation Valves ((AtllJOspherii' and e J.s.rl.LY C.1 and C.2 (continued) path must be isolated.

The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate Required Action C.l must be completed @ \...!_!;) Completion Time. The specified time I \1 period is reasonable, considering the relative stability of th* closed syste* (hence, reliability) to act as a penetration isolation boundary and the relative importance of supporting containment OPERABILITY during MODES 1, 2, 3, and 4. In the event the affected penetration is isolated in accordance with Required Action C.l, the affected penetration flow path must be verified to be iso1ated on a ;'"'S periodic basis. This 1s necessary to assure leak tightness of containment and that containment penetrations requiring hoht*ion following an accident are isolated. The Completion Time of once per 31 days for verifying that each affected penetration flow path is isolated is appropriate considering the valves are operated under administrative controls and the probability of their misalignment is

  • Condition C is modified by a Note indicating that this Condition is only applicable ta those penetration flow paths with only one containment isolation valve and a closed system. Thts Note is necessary stnc1 this Condition is written to specifically address those penetration flow paths in a closed system. Required Action C.2 is modified by a Note that applies to valves and blind flanges.located.1n high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by adllin1strat1v1 means 1s considered acceptable, since access ic i,\i. i; t*o these areas is typically restricted.

Therefore, the /G'\ .... , l\c d these once they have\& X ,.;i * ; -r 1 .. 1 ....e. .been verified \.o be 1n the proper o is small * . el ... tht condary conta1nman bypass leakage ra not J within 1 it, the assumption of the safety anal sis are not (continued) \ CEOG STS ' B 3.6-26 Rev 1, 04/07/95 houf IS Pfbvid.cd +o locS cl.o$c:.d fhc a. f'.fc.t.td lla./IAJ, the ( hovr C...o"-r/..+.,,. -t"'"' Prou1des o. f'r..r1 aJ I)+ to C.oact tht f robl<.M U,m rflc-l\S<Jro.i1t C.-.. th(.. lt'rd'Qria.rc..(...

  • O 1'Ma.1n+o-1n1nJ C./oScd.
  • 5-9c_ ---------
  • * @ * .:.TS .3.b.S
  • ACTIONS continued CONDITION C.

Only applicable to penetration flow paths with only one containment isolation valve and a closed syste*. ---------------------; On* or more penetration flow paths with one containment isolation valve inoperable

  • E Seco dary containment byp ss leakage not wi in limit. £,. One or more pen*tratii:aA flow*p with ou or mare conut-t*pu vahff-*t wit n purge valve. 1 akag1 l i*its. CEOG STS 'O. o ** -e:p...., 1 c

... c+ o ... .:t.: .. roo-\; ... e,*1 ..io.l.tt. ,..+ loc.\:.eJ Containment Isolation Valves(fulil.§spheric

3. 6. 3 0.1 E.J Isol1t1 the affecte penetration flow th by use of at lu onl* [closed an de-activated automatic va 1 with res11 ient s als, closed man al valve with ient seals, or blin flange). 3.6-10 "l>.\ L iac.k. c..\ose'4 "c..\,,._..f -J-24 hours j (continued)

Rev 1, 04/07/95 I __ J

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.3-6 DOC A.8 JFD6 CTS 3.6.1 ITS 3.6.3 ACTIONS Note 1 The CTS markup of CTS 3.6.1 has a *Footnote associated with it. The footnote states that "Penetration flow paths may be unisolated intermittently under Administrative control." A Technical Specification (TS) Amendment Request dated March 26, 1997, proposes to add this footnote to the Palisades CTS which is Less Restrictive (L) change to the CTS. This portion of the Amendment Request has not been approved at this time. The CTS markup modifies this footnote to exclude the 8 inch purge exhaust valves and the 12 inch air room supply isolation valves. This change is designated as an Administrative change (DOC A.8). This is incorrect.

As proposed in the March 1997 Amendment Request the footnote would allow the purge exhaust and the air room supply isolation valves to be opened intermittently under Administrative control. Thus the change proposed in the CTS markup would be a More Restrictive change since it would restrict this allowance. Comment: The licensee should do one of the following: a . b. c. Revise the CTS markup and provide additional discussion and justification to show that the exception is a More Restrictive change; Revise the March 26, 1997, Amendment Request to correctly reflect the intent that the purge exhaust and air room supply valves are excluded from the footnote. Revise the CTS accordingly; or Delete this portion of the footnote from the March 26, 1997, Amendment Request and revise the CTS markup to show the addition of the corrected . footnote as a Less Restrictive (L) change. Provide the appropriate discussion and justifications for this Less Restrictive change. Consumers Energy Response: The Technical Specification Amendment Request dated March 26, 1997 which requested the addition of footnote"*" to CTS 3.6.1 has been approved in Amendment 184 to the Facility Operating License for the Palisades Plant. CTS 3.6.1 and Footnote "*" establish the requirements for containment penetrations whose status ensures containment integrity. In addition to the requirements of CTS 3.6.1, further restrictions are placed on the containment purge exhaust and air room supply isolation valves as stated in CTS 3.6.5. Since Footnote "*" does not modify the requirements of CTS 3.6.5, the ability to unisolate penetration flow paths under administrative control does not apply to the containment purge exhaust and air room supply isolation valves. As such, characterizing the \ change to Footnote "*" as "Administrative" as described in DOC A.8 is appropriate since there is no change in requirement between the CTS and ITS . Affected Submittal Pages: No page changes. 6

  • DOC 3.6.3 .:A.18 CTS 3.6.5 provides the corrective actions when one containment purge exhaust or air room supply isolation valve is not locked closed. CTS 3.6.5 requires the valve be locked closed within 1 hour, or be in at least Hot Standby within the next 6 hours and in Cold Shutdown within the following 30 hours. CTS 3.6.5 does not provide corrective actions if two or more containment purge exhaust or air room supply isolation valves are not locked closed. Thus, the actions of LCO 3.0.3 must be invoked. The actions ofLCO 3.0.3 are equivalent to the actions of CTS 3.6.5. That is, both specifications require that compliance with the LCO be established within 1 hour, or the plant must be placed in Cold Shutdown with the next 36 hours. In ITS 3.6.3, containment purge exhaust or air room supply isolation valve inoperability is addressed by Condition D. Condition D has been further modified from the requirements of CTS 3.6.5 to address "one or more" containment purge exhaust or air room supply isolation valves not locked closed. This change has been characterized as an Administrative change on the basis that the-corrective actions of CTS 3.6.5 for one valve not locked closed are equivalent to the actions of LCO 3.0.3 for multiple valves that are not locked closed. As such, the requirements of the CTS are equivalent to the requirements of the ITS for inoperable containment purge exhaust or air room supply isolation valves . 5-e__
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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-7 DOC A.9 CTS 4.5.2.c(2)

ITS 3.6.3 ACTIONS Note 4 and Associated Bases CTS 4.5.2.c(2) has been modified to add ITS 3.6.3 ACTIONS Note 4 which states "Enter applicable conditions and Required Actions of LCO 3.6.1 "Containment" when leakage results in exceeding the overall containment leakage rate acceptance criteria". DOC A.9 states that this Note provides guidance on use and application of the ITS where it wasn't explicitly addressed as part of the CTS. DOC A.9 only addresses the ITS aspects, application and interpretation of this Note and does not provide sufficient information with regards to how the CTS is consistently interpreted in this area so that the staff can determine whether this change is Administrative, More Restrictive, or Less Restrictive (L). Comment: Provide additional discussion and justification to show how Consumers Energy interprets the CTS with regards to this change. Consumers Energy Response: DOC A. 9 has been revised to clarify how the CTS addresses individual containment isolation valve leakage relative to the overall containment leakage limit. Affected Submittal Pages: Att 3, DOC 3.6.3, page 4of10 7

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  • A.9 A.10 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES In the proposed ITS, ACTION Note 4 is added which is not included in the CTS. Note 4 states "Enter applicable Conditions and Required Actions of LCO 3. 6 .1 "Containment," when leakage results in exceeding the overall containment leakage rate acceptance criteria." This note provides guidance in accordance with LCO 3.0.6 to specify when other TS should be followed.

This note is considered to be administrativP. iii nature in that it provides guidance on the use and application of the ITS where it wasn't explicitly addressed as part of the CTS. These changes are consistent with NUREG-1432.1' See Jiji[.(i A In proposed ITS, ACTION Notes 2 and 3 are added which are not included in the CTS. These notes involve the application of TS usage rules. Note 2 states "Separate Condition entry is allowed for each penetration flow path. " This allows a separate "clock" for each penetration flow path in accordance with the usage rules of NUREG-1432. This would allow more than one flow path to have an inoperable valve. However, since it is only for a limited period of time before each flow path must be: isolated, this allowance is acceptable.1' Note 3 states "Enter applicable Conditions and Required Actions for systems made inoperable by containment isolation valves." Following the philosophy of ITS LCO 3 .0.6, "cascading" is not required unless specified in the individual specifications. If a containment isolation valve is closed, it most likely is going to have an impact on the system it is isolating. Therefore, the note requires that if a system is made inoperable by a containment isolation valve the applicable conditions and actions should be entered.' These changes retleet the iisage Riies ef fl'UIU!iC 143:2 aail are adiriiriistrative in nature. The changes made are consistent with NUREG-1432. ' Ri't l *i L A.11 CTS 3.6.1 ACTION a states "Restore the inoperable valve to OPERABLE status within 4 hours." This statement is not included in the proposed ITS because it is not necessary. The option to restore a component to OPERABLE status is already provided by LCO 3.0.2 and is redundant to specify it in the ACTIONS. This is an administrative change to reflect the usage rules of NUREG-1432. This change is consistent with NUREG-1432. Palisades Nuclear Plant \ I Page 4of10 /-o-01/20/98 I

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  • DOC 3.6.3 -*A.9 . .. The intent ofITS 3.6.3 Action Note 4 is to provide guidance to the ITS user to take the appropriate Actions of Specification 3.6.1 when the containment leakage rate acceptance criteria *is exceeded.

Note 4 does not imply any additional requirements, but simply provides a cross-reference between two specifications. Thus, the addition of Note 4 can only be characterized as an "Administrative" change. The corresponding requirements in the CTS that would be cross-referenced are; CTS 4.5.2c(2} which requires that corrective actions be initiated "if at any time it is determined that the total containment leakage rate exceeds La", and CTS 4.5.2d(l) which requires individual penetrations and containment isolation valves to be leak tested. As such, containment isolation valves with excessive leakage are treated consistently between the ITS and CTS. That is, total leakage must be evaluated against the overall containment leakage limit.

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-8 DOCA.10 CTS 3.6.1 ACTIONS ITS 3.6.3 ACTIONS Notes 2 and 3 and Associated Bases CTS 3.6.1 ACTIONS have been modified to add ITS 3.6.3 ACTIONS, Notes 2 and 3. DOC A. 1 O states that these Notes provide guidance on use and application of the ITS where it wasn't explicitly addressed as part of the CTS. DOC A.1 O only addresses the ITS aspects, application and interpretation of these Notes and does not provide sufficient information with regards to how the CTS is consistently interpreted in this area so that the staff can determine whether this change is Administrative, More Restrictive, or Less Restrictive(L).

Comment: Provide additional discussion and justification to show how Consumers Energy interprets the CTS with regards to this change. Consumers Energy Response: DOC A.10 has been revised to provide additional justification to show how the addition of ITS 3.6.3, Actions Notes 2 and 3 relate to the requirements of the CTS. Affected Submittal Psqes: Att 3, DOC 3.6.3, page 4 of 10 8

  • A.9 A.10 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES In the proposed ITS, ACTION Note 4 is added which is not included in the CTS. Note 4 states "Enter applicable Conditions and Required Actions of LCO 3.6.1 "Containment," when leakage results in ef{ceeding the overall containment leakage rate acceptance criteria." This note provides guidance in accordance with LCO 3.0.6 to specify when other TS should be followed.

This note is considered to be in nature in that it provides guidance on the use and application of the ITS where it wasn't explicitly addressed as part of the CTS. These changes are consistent with NUREG-1432.1' In proposed ITS, ACTION Notes 2 and 3 are added which are not included in the CTS. These notes involve the application of TS usage rules. Note 2 states "Separate Condition entry is allowed for each penetration flow path." This allows a separate "clock" for each penetration flow path in accordance with the usage rules of NUREG-1432. This would allow more than one flow path to have an inoperable valve. However, since it is only for a limited period of time before each flow path must be isolated, this allowance is acceptable.1' Note 3 states "Enter applicable Conditions and Required Actions for systems made inoperable by containment isolation valves." Following the philosophy of ITS LCO 3.0.6, "cascading" is not required unless specified in the individual specifications. If a containment isolation valve is closed, it most likely is going to have an impact on the system it is isolating. Therefore, the note requires that if a system is made inoperable by a containment isolation valve the applicable conditions and required action.S should be entered., These changes refleet th11 iisage R:1li6 8f W\JIWC 143:2 a&a are adIIiiriistrative in nature. The changes made are consistent with NUREG-1432. A.11 CTS 3.6.1 ACTION a states "Restore the inoperable valve to OPERABLE status within 4 hours." This statement is not included in the proposed ITS because it is not necessary. The option to restore a component to OPERABLE status is already provided by LCO 3.0.2 and is redundant to specify it in the ACTIONS. This is an administrative change to reflect the usage rules of NUREG-1432. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 4of10* 01/20/98 8 -Q._..,

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  • DOC 3.6.3 -"A.IO INSERT A ... The application of Note 2 is consistent with the approach used in the CTS to address inoperable CIV s in multiple penetrations.

That is, corrective actions are taken on a per penetration basis with the allowed outage time for each inoperability beginning at the time of discovery of the inoperable valve. INSERTB ... This philosophy is consistent with the CTS as specified in LCO 3 .0.1 (i.e., take the corrective actions for the inoperable component). As such, Note 3 does not impose any additional requirement to, or relaxation from, the CTS but simply clarifies that support system Conditions and Required Actions must be entered and that reliance exclusively on the support system Required Actions is not permissible .... I \ 8-h

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-9 DOC A.15 CTS 3.6.1 ACTION CTS 3.6.3 CTS 4.5.3.d ITS 3.6.3 ACTIONS A, C, and E and Associated Bases ITS SR 3.6.3.2 and SR 3.6.3.3 and Associated Bases CTS 3.6.1 ACTION has been modified by the addition of ITS 3.6.3 Require Action (RA) A.2 and its associated Note which states "Isolation devices in high radiation areas may be verified by the use of administrative means." The proposed addition of this ITS Note is acceptable; however, DOC A.15 is incorrect and is not the appropriate justification for this change. To start with, DOC A.15 states that the Note is added to ITS 3.6.3 RA A.2 and D.2. There is no ITS 3.6.3 RA D.2 in the ITS markup. Furthermore, the Note is also added to ITS 3.6.3 RA C.2, E.2, SR 3.6.3.2, and SR 3.6.3.3. Depending on the change, the Note either modifies a CTS requirement or does not directly modify a CTS requirement but modifies a new added requirement.

ITS 3.6.3 RA A.2 is by the CTS markup a More Restrictive change, thus the addition of the Note to this change is also a More Restrictive change. ITS 3.6.3* RA C.2 is. a Less Restrictive (L) change (See Comment Number 3.6.3-11), thus the addition of the Note to this change is a Less Restrictive (L) change. ITS 3.6.3 RA E.2 is a Less Restrictive (L) change (See Comment Numbers 3.6.3-28 and 3.6.3-29), thus the addition of the Note to this change is a Less Restrictive (L) change. CTS 3.6.3 specifies that prior to the reactor going critical after a refueling outage an administrative check will be made to confirm that all "locked closed" manual CIVs are closed and locked. A TS Amendment Request dated March 26, 1997, proposes to modify CTS 3.6.3 to change the "administrative check" to a "visual check" and renumber the specification CTS 4.5.3.d. Even though this Amendment has not been approved by the staff, the CTS conversion submittal is presented as though it has been approved. The corresponding ITS SRs are ITS SR 3.6.3.2 and SR 3.6.3.3. The addition of this Note to these SRs would be considered as an Administrative change if the original pre-March 1997 Amendment Request CTS SR (CTS 3.6.3) is used. If the Amendment Request CTS SR (CTS 4.5.3.d) is used then the addition of the Note is a Less Restrictive (L) change since the Note would allow an administrative check rather than a visual verification in high radiation areas. Comment: Revise the CTS markup and provide the appropriate discussions and justifications for these Administrative, More Restrictive, and Less Restrictive (L) changes. See Comment Numbers 3.6.3-11, 3.6.3-28 and 3.6.3-29 . (continued) 9

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.3-9 (continued)

Consumers Energy Respoase: The following changes have been made to reflect the addition of the ISTS Note that allows isolation devices in high radiation areas to be verified by the use of administrative means. 1) DOC A.15 has been deleted in its entirety. The Note formerly discussed in this change is now addressed by DOC M.4 and DOC L.2. 2) DOC M.4 has been revised to include the Note to Required Action A.2. 3) A new DOC (DOC L.2) and NSHC (NSHC L.2) were provided in response to RAI 3.6.3-11 to justify the less restrictive aspects of ITS Condition C. That justification also includes the Note to Required Action C.2. 4) The Note previously presented in ITS Required Action E.2 is no longer required and has been deleted. See response to RAI 3.6.3-29.

5) A new DOC (DOC L.3) and NSHC (NSHC L.3) have been provided to justify the less restrictive aspects of the Note *included in ITS SR 3.6.3.2 and SR 3.6.3.3. This change is based on the TS Amendment Request dated March 26, 1997 that has been approved by the staff. Affected Submittal Pages: Att 3, CTS, page 3-40 (ITS 3.6.3, page 1 of 5) Att 3, CTS, page 4-21 (ITS 3.6.3, page 4 of 5) Att 3, DOC 3.6.3, page 6of10 Att 3, DOC 3.6.3, page 8 of 10 Att 3, DOC 3.6.3, page 10of10 Att 4, NSHC 3.6.3, page 2 of 2 10 Ari I
  • ( Atib A Nor!..> (...(..0 3.6
f.6.3 3.6.4 l..c..oJ.L..J c. When pt>sitive reactivity es are made by boron di tion or CONTROL ROD motion (excep(
  • r testing one CONTROL R at a time). ACTION: 1.5 pstg when above COLD SHU and 1.0 pstg when tn POWER OPE ION or HOT STANDBY. With containment internal pres rt above the limit, r tort pressure t within the limit within 1 hou , or be in at least HO SHUTDOWN within the next 6 hours and in COLD HUTDOWN within the fo owing 30 hours. Two ndependent containment hydroge recombiners shall be th plant ts 1n POWER OPERATION or. OT STANDBY. With one ecombiner 1 operable, restore the inoperab recombiner to OPERABL status within r be tn a least HOT S TDOWN within the next c.+kt (RDO
  • *
  • 4.5 4.5.Z

+:: c.c.+: 0.._ 1 {,. s ' *(continued) (b) A full air 1 k penetration test shall be performe at six-month i 1rvals. During the period between t e six-month 1sts when CONTAINMENT INTEGRITY 1s re u1red, a reduc1d p ssur1 t1st for th1 door seals or a f l air lock pen ration test shall be perfonnld with 72 hours , after e th1r each air lock door opening or t first of a series of openings. I Sc.c.

14. 5.3 I Containment Isoltt1ofV'y.1yss 1
l. 5 Th.* isolation valves shall be demonstrated OPERABLE by p1rfonunce I of a cycling test and verification of isolation time for auto ----L-isolation valves prior to declaring the valve to b1 OPERABLE after I maintenance, repair, or replacement work is p1rfor1111d on th* valve or its associated actuator, control, or power circuit.
  • 21
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  • ATTACHMENT 3 DISCUSSION CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES ..............

....--.-,_.,.....,....,..._ __ ....-....- A.15 rJe.u.J -Sc..c.. i A.11 ('le.VI A. i'? f\)&. t>J I Palisades Nuclear Plant AA I 3."-1-z. Rf+I \ \ Page 6of10 JO-C-ction A.2 and D.2 which sta 01/20/98

  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES M.2 (continued)

M.3 These changes are considered to be Restrictive changes since the CTS only requires--the verification to be perfonned prior to going critical, does not require verification between refueling outages, and does not address blind flanges. This change is consistent with NUREG-1432. requires that the containment purge and v ntilation isolation valves be leak tested every 6 months to verify adequate closure. The proposed ITS SR 3.6.3.5 re ires the six month frequency (184 days) p sit requires that they be leak tested wi 92 days after opening the valves. The 9 day requirement exists becaus the valves have resilient seals that may inc additional degradation beyond that hich would occur if it was not opened. se valves should not normally be opened MODES 1-4, but this additional require ent is added if for some reason they mus opened. Since the Palisades CTS does ot have this requirement the addif n of the 92 day test after the valves have en opened is considered a More estrictive chan e. This chan e is consiste wi UREG-1432. ITS includes Required Action A.2 which states "Verify the affected penetration flow path is isolated." This verification is not required in CTS 3.6.1 Actions. This verification is necessary to ensure that the flow path with the inoperable valve remains isolated until the inoperable valve is restored\, The Completion Time for /ld'.M. A.t this aetia11 is "Once per 31 days.for isolation devices outside containment AND Prior to entering MODE 4 from MODE 5 if not perfonned within the previous 92 days for isolation devices inside containment." This verification is not included in the CTS and therefore this is considered to be a more restrictive change. This change is consistent with NUREG-1432 .

  • Palisades Nuclear Plant Page 8of10 01/20/98 10-d
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  • DOC 3.6.3 M.4 . .. Required Action A.2 is also modified by a Note which states "isolation devices in high radiation areas may be verified by use of administrative means." This allowance minimizes personnel exposure and recognizes that high radiation areas are usually restricted such that the probability of a valve misalignment is small. . .. ;o-e ' I
  • *
  • LA.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES . contains a reqmremen a t e contamme purge and entilation isolation
v. Ives be determined closed at least once every months "by erforming a leak r e test between the valves." The overall requi ment is that the alves be demons ated to be closed by performance of a leak test The method of ccomplishing s testing can be addressed by plant procedures hich are under the ontrol of the icensee. Therefore, the method of testing is not ncluded in the roposed IT and will be addressed by plant procedures.

Ch ges to these procedures are made i accordance with the plant change control proces . This change is G-1432. LA.5 LE by left channel test signal, pplicable isolatio valves actuate to their required positi during COLD f< r+ I HUTDOWN or t least once per refueling cycle." Th requirement to test the valves 3. "during COLD HUTDOWN" is not included in the oposed ITS. The proposed ITS ill test the v ves every 18 months which is the equ* alent to the "once per refueling ycle" in the roposed ITS. The requirement totes the valves during "COLD SHUTDO "will be addressed by plant proced es. Changes to plant procedure§ are made

  • accordance with the plant procedure ge process. This changeit' consiste with NUREG-1432.

LESS RESTRICTIVE CHANGES (L) L.1 CTS Table 4.2.2 item 13.a specifies that the Containment Purge and Ventilation Isolation Valves shall be determined closed "at least once per 24 hours." The proposed. ITS SR 3.6.3.1 changes the frequency for verifying the 8 inch containment purge valve and 12 inch air room supply valves are closed from 24 hours to 31 days. This change is reasonable because these valves are not allowed to be open in proposed MODES 1-4. These valves are required to be locked closed, either electrically or by some other means to de-activate the valve. Therefore, the likelihood of the valve being opened is very remote. The 31 day frequency is consistent with the surveillance requirements for other safety related systems which require valve position verification. This change is consistent with NUREG"".1432. t 3 l* s ...... RA-I \ ' Palisades Nuclear Plant Page 10of10 /Ml J1,.i .. N/is/16/l7 )0-L Sec.. (J ti I 5 * <o. 3-3 I 01/20/98

  • *
  • 3.6.3 DOC L.3 CTS 4.5.3d requires that prior to the reactor going critical after a refueling outage, a visual check will be made to confirm that all locked-closed manual containment isolation valves are closed and locked except for valves that are open under administrative controls.

In the ITS, manual valve position verification is required by proposed SR 3.6.3.2 and SR 3.6.3.3. SR 3.6.3.2 and SR 3.6.3.3 are modified by a Note which states "valves and blind flanges in high radiation areas may be verified by use of administrative means. " The allowance to verify valve position by use of "administrative means" is a relaxation from the CTS requirement to perform a "visual check." The proposed change allows special provisions for high radiation areas to minimize personnel exposure while still keeping track of containment isolation valve . status. This change is considered acceptable since a verification requirement still exists, and because high radiation areas are restricted such that the probability of valve misalignment is small. This change is consistent with NUREG-1432 . I I

  • *
  • 2. 3. L. t.. L, t..I L. 5 L.b ATTACH1\.1ENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Does the change create the possibility of a new or different kind of accident from* any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue to verify that the

  • purge exhaust and air room supply valves are still closed. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Does this change involve a significant reduction in a margin of safety? The proposed change increases the surveillance interval from 24 hours to 31 days. The margin of safety afforded by the purge exhaust and air room supply valves are that they are closed and thereby isolating a direct path from the containment atmosphere. Since the purge exhaust and air room supply valves are required to be closed in MODES 1, 2, 3 and 4, the likelihood of these valves being open is very small. Therefore, this change does not involve a significant reduction in a margin of safety. Se.c. lNS(Ai t\A I . . -1 I SLt. I:MS C,cf R.A-I 3 . (o. 2> -q Sec.. I.JJSf:,rt l 3 *b* 3 -'7-.. Sec. XNSu.1 R1+ 1 j.,.J -/t.; /1s /1t/17 I.Nsu1 /<.fl l 3

  • b-3 -3 I Palisades Nuclear Plant Page 2of2 01/20/98
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  • 3.6.3 N$HC L.3 CTS 4.5.3d requires that prior to the reactor going critical after a refueling outage, a visual check will be made to confirm that all locked-closed manual containment isolation valves are closed and locked except for valves that are open under administrative controls.

In the ITS, manual valve position verification is required by proposed SR 3.6.3.2 and SR 3.6.3.3. SR 3.6.3.2 and SR 3.6.3.3 are modified by a Note which states "valves and blind flanges in . . high radiation areas may be verified by use of administrative means." The allowance to verify valve position by use of "administrative means" is a relaxation from the CTS requirement to perform a "visual check. " The proposed change allows special provisions for high radiation areas to minimize personnel exposure while still keeping track of containment isolation valve status. This change is considered acceptable since a verification requirement still exists, and because high radiation areas are restricted such that the probability of valve misalignment is small. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change allows the periodic verification of manual containment isolation valves in high radiation areas using administrative means. The proposed change does not involve a change to any accident initiators or precursors . Therefore, the proposed change does not involve a significant increase in the probability of an accident. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change does not alter the assumed capability of the containment isolation function since the containment isolation valves in high radiation areas are still assured to be in their correct position. Thus, the consequence of an accident remain unchanged. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change still ensures that an OPERABLE containment isolation barrier is available whenever the plant is in a condition that requires containment isolation. Thus, th\s change does not create the possibility of a new or different kind of accident from any accident previously evaluated . ' ICC:.

  • 3 . *
  • Does. this change involve a significant reduction in a margin of safety? The margin of safety is a function of the overall containment leakage. The proposed change allows the periodic verification of manual containment isolation valves in high radiation areas using administrative means. This change does not relax the requirement to maintain containment integrity, but recognizes that access to high radiation areas are restricted such that the probability of valve misalignment is small. As such, this change does not result in any activity that would result in an increase in the amount of radioactive material released to the environment.

Therefore, the proposed change does not involve a significant reduction in a margin of safety . "

  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.3-10 DOC AX CTS 3.6.5 ITS LCO 3.6.3 ITS 3.6.3 APPLICABILITY CTS 3.6.5 specifies the OPERABILITY and APPLICABILITY for the containment purge exhaust and air room supply isolation valves. The CTS markup of 3.6.5 shows an Administrative change to the OPERABILITY and APPLICABILITY which states "The containment isolation valves shall be OPERABLE in MODES 1-4." The designation for this Administrative change has been cut-off by the edge of the page, such that it cannot be determined which DOC AX is applicable to this change. Comment: Correct this discrepancy.

Consumers Enerqv Re@oase: Corrected. DOC A 1 is referenced for this change. Affected Submittal Pages:

  • Att 3, CTS page 3-40 (ITS 3.6.3, page 1 of 5) * \ .. 11 P.Pi I '!i-/0.3 .
  • L,.C..O 3.6 3:6.3 13.6.4 Lc..oJ .. i..J ---* ( At)t\ A tVoe. > {{Al

.1-2.. c. When pt>s1t1ve reactivity ch,.a .** es are made by boron di ti on or CONTROL ROD motion (except* r testing one CONTROL R at a time). ACTION: and 1.0 ps1g when 1n POWER OPE With containment internal pres re above the 11m1t, r tore pressure t within the limit within l hou , or be in at least HO SHUTDOWN within the next 6 hours and in COLD HUTDOWN within the fo owing 30 hours. TwoAndependent containment hydroge recombiners shall be ERABLE when 1s in POWER OPERATION or. OT STANDBY. With one ecombiner 1 op1r1bl1, restor1 th1 inoperab recombiner to OPERABL status within r be in a least HOT S TOOWN within the next ... -H....,

  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-11 DOC M.1 CTS 3.6.1 ACTION ITS 3.6.3 ACTION C and Associated Bases CTS 3.6.1 ACTION is modified by the addition of ITS 3.6.3 ACTION C. The CTS markup shows this addition as a More Restrictive (DOC M.1) change, based on the justification that the CTS does not address single isolation valves on closed systems, thus it is a new requirement.

This is an incorrect justification for this change. One could interpret the CTS for single valve/closed system case in one of two ways. In the first case, based on the discussion in DOC M.4, since there is no remedial measures provided in CTS 3.6.1 ACTION, then CTS 3.0.3 should be entered. ITS 3.6.3 ACTION C is Less Restrictive than CTS 3.0.3. In the second case, it is the staffs position that CTS 3.6.1 ACTION does provide the appropriate remedial measures for the single valve/closed system case. Since an OPERABLE isolation valve cannot be maintained in this case per CTS 3.6.1 ACTION, CTS 3.6.1 ACTION c would be entered. Thus the addition of ITS 3.6.3 ACTION C is Less Restrictive than CTS 3.6.1 ACTION. In either case, the addition of ITS 3.6.3 ACTION C is a Less Restrictive (L) change not a More Restrictive change. See Comment Number 3.6.3-9. Comment: Revise the CTS markup and provide a discussion and justification for this Less Restrictive (L) change. See Comment Number 3.6.3-9 . Consumers Energy Response: A new DOC (DOC L.2) and NSHC (NSHC L.2) have been provided to justify the addition of ITS Condition C and its associated Required Actions and Completion Times. Previously, the addition of ITS Condition C was inappropriately characterized as a More Restrictive change as described in DOC M.1. As such, DOC L.2 replaces DOC M.1 which has been deleted in its entirety. In addition, conforming changes have been made to the CTS markup. Affected Submittal Pages: Att 3, CTS page 3-40 (ITS 3.6.3, page 1 of 5) Att 3, DOC 3.6.3, page 7 of 10 Att 3, DOC 3.6.3, page 10of10 Att 4, NSHC 3.6.3, page 2 of 2 '1 12

  • RA I ;.(D.3 -4 . L,.C..O 3.6 (

Ctrun A Not!> c. When pt>s1t1ve react1v1ty es are made by boron d1 tion or CONTROL ROD motion (except* r testing one CONTROL R at a time). ACTION: and 1.0 ps1g when 1n OPE With containment internal pres re above the limit, r tore pressure t within the limit within 1 hou , or be in at least HO SHUTDOWN within the next 6 hours and in COLD HUTDOWN within the fo owing 30 hours. !?.Al j.lo *1-2. 3.6.4 Two ndependent containment hydroge recombiners shall be th plant is in POWER OPERATION or. OT STANDBY. With one ecombiner

  • i operable, restore the inoperab recombiner to OPERABL status within r be in a least HOT S TOOWN within the next c..+l...t
  • *
  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES MORE RESTRICTIVE CHANGES (M) M.1 AA!

I/ Al Proposed ITS Action represents a new requirement for R lisades because the CTS does nofaddress sing isolation valves on closed system . Proposed ITS Required Action C.1 specifies a 72 hour AOT for inoperable con inment isolation valves in closed systems to i olate the affected penetration flow ath by use of at least one clos and de-activated tomatic valve, closed manual valv , or blind flange. This AOT

  • reasonable base on the low potential for loss of int grity of a closed system and i consistent wi the typical 72 hour AOT for safety elated systems. Required Ac on C.2 requires at the affected penetration flow p is verified to be isolated on per 31 days. T s ensures that the flow path remai isolated and is consistent wi the typical fre encies for ongoing valve lineup .v 'fication activities used throu out the proposed TS. This change is considered to e more restrictive since the P isades CTS do s not address the configuration for losed systems with single co inment . isolati n valves.

This change is coruiiste with NUREG-1432 as modifi d by industry own 's rou eneric chan e TSTF-30. M.2 CTS 4.S.3d specifies requirements for ensuring that all locked-closed manual containment isolation valves are closed and locked except those open under administrative control. The proposed ITS addresses manual containment isolation valves and blind flanges for both outside and inside containment in proposed SR 3.6.3.2 and SR 3.6.3.3 respectively. Blind flanges are specified in the proposed ITS since they are also isolation devices which must be in the closed position to function properly. The CTS requires that this verification be performed prior to "the reactor going critical after a refueling outage." The proposed ITS SR 3.6.3.3 revises this to state prior to "entering MODE 4 from MODES if not performed within the previous 92 days." The proposed revision is needed to match the Applicability of Specification 3.6.3 which is MODES 1-4. This verification is modified by the statement "if not performed within the previous 92 days" to ensure that for shutdowns after extended times in MODES 1-4, that the valve status is reverified. The proposed ITS SR 3.6.3.2 addresses the manual containment isolation valves and blind flanges* which are outside containment and requires that they be verified closed every 31 days. The addition of the 31 day frequency is to address the fact that since these valves are outside containment, they are in a location which would be more susceptible to inadvertent mispositioning than the devices inside containment and therefore need to be verified more frequently. ,, Palisades Nuclear Plant Page 7 oflO 01/20/98 /;;J-b

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  • LA.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES tem . con ms a requrremen e contamme purge and entilation isolation
v. Ives be determined closed at least once every months "by erforming a leak r e test between the The overall requi ment is that the alves he demons ated to be closed by performance of a leak test The method of ccomplishing s testing can be addressed by plant procedures hich are under the ontrol of the 1censee. Therefore, the method of testing is not eluded in the reposed IT and will be by plant procedures.

C ges to these procedures are made

  • accordance with the plant change control proces . This change is -1432. LA.5 LE by left channel test signal, pplicable isolatio valves actuate to their required positi during COLD f<Pr I . HUTDOWN or t least once per refueling cycle." Th requirement to test the valves 3." .'!>*Ir "during COLD HUTDOWN" is not included in the oposed ITS. The proposed ITS ill test the v ves every 18 months which is the equ* alent to the "once per refueling ycle" in the reposed ITS. The requirement totes the valves during "COLD SHUTDO " will be addressed by plant proced es. Changes to plant procedur;§ are made
  • accordance with the plant procedure ge process. This change 1s consiste with NUREG-1432.

LESS RESTRICTIVE CHANGES (L) L.1 CTS Table 4.2.2 item 13.a specifies that the Containment Purge and Ventilation Isolation Valves shall be determined closed "at least once per 24 hours." The proposed. ITS SR 3.6.3.1 changes the frequency for verifying the 8 inch containment purge valve and 12 inch air room supply valves are closed from 24 hours to 31 days. This change is reasonable because these valves are not allowed to be open in proposed MODES 1-4. These valves are required to be locked closed, either electrically or by some other means to de-activate the valve. Therefore, the likelihood of the valve being opened is very remote. The 31 day frequency is consistent with the surveillance requirements for other safety related systems which require valve position verification. This change is consistent with NUREG-1432. L2.. s 'I.t.).S lit f /(.A\ '?l-1 I /.... 3 {{Al ?1.(..3-9 L. L/ RA-I J,.<t,.?,-/ ')_ Palisades Nuclear Plant Page 10of10 01/20/98 l*5 xµsai,; /Ml Sec, WS'*.(.j I()..-c_ L*h /{ Ill . to. 3 -3 I

  • *
  • DOC 3.6.3 L.2 CTS 3.6.1 contains the corrective actions for one or more inoperable containment isolation valves. For penetrations associated with a closed piping system and one CIV, CTS 3.6.1 action "c" would require the plant to be placed in at least Hot Shutdown within 6 hours and in Cold Shutdown within the following 30 hours. In ITS 3.6.3, this same inoperability would be addressed by Condition C which would require the affected penetration be isolated in 72 hours, and verified isolated once per 31 days. The Required Actions of the ITS are less restrictive than the CTS since they allow 72 hours to isolate the affected penetration versus a forced shutdown.

The 72 hour period provides the necessary time to perform repairs on a failed CIV when relying on an intact closed system. A Completion Time of 72 hours is considered appropriate given the location of certain valves, the reliability of the closed system, and that 72 hours is typically provided for losing one train of redundancy throughout the ITS. If the closed system and associated CIV were both inoperable (as a containment boundary), the plant would be in LCO 3.0.3 since there is no specific Condition specified. Although ITS Required Action C.2 is an additional restriction on plant operations since it requires a verification that the affected penetration is isolated once per 31 days, the overall change related to the addition of ITS Condition C is characterized as Less Restrictive. Lastly, Required Action C.2 is modified by a Note which states "isolation devices in high radiation areas may be verified by use of administrative means." This allowance minimizes personnel exposure and recognizes that access to high radiation areas are restricted such that the probability of a valve misalignment is small. This change is consistent with NUREG-1432 as modified by TSTF-30 . I I Jdi-d

  • * -* 2. 3. L. i LIL.I l.5 L.b ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proiJ"osed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue to verify that the purge exhaust and air room supply yalves are still closed. Thus, this change does not create the possibility of a new or aifferent kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? The proposed change increases the surveillance interval from 24 hours to 31 days. The margin of safety afforded by the purge exhaust and air room supply valves are that they* are closed and thereby isolating a direct path from the containment atmosphere. Since the purge exhaust and air room supply valves are required to be closed in MODES 1, 2, 3 and 4, the likelihood of these valves being open is very small. Therefore, this change does not involve a significant reduction in a margin of safety . Se.t It.lSCA r S<-t. I:MS Set. I}'JS'c.rt Lusu 7 I ?l . <o. 5 .. I I /(.f+ I 3 . (o

  • 2> -'l l 3-b. 3-17... 1?1+ t :i.&.J -II../ /1s /1t/17 /(.,q ' 3
  • b* 3 -I \ ' Palisades Nuclear Plant Page 2of2 01/20/98 e_
  • *
  • NSHC 3.6.3 L.2 CTS 3.6.1 contains the corrective actions for one or more inoperable containment isolation valves. For penetrations associated with a closed piping system and one CIV, CTS 3. 6 .1 action "c" would require the plant to be placed in at least Hot Shutdown within 6 hours and in Cold Shutdown within the following 30 hours. In ITS 3.6.3, this same inoperability would be addressed by Condition C which would require the affected penetration be isolated in 72 hours, and verified isolated once per 31 days. The Required Actions of the ITS are less restrictive than the CTS since they allow 72 hours to isolate the affected penetration versus a forced shutdown.

The 72 hour period provides the necessary time to perform repairs on a failed CIV when relying on an intact closed system. A Completion Time of 72 hours is considered appropriate given the location of certain valves, the reliability of the closed system, and that 72 hours is typically provided for losing one train of redundancy throughout the ITS. If the closed system and associated CIV were both inoperable (as a containment boundary), the plant would be* in LCO 3.0.3 since there is no specific Condition specified. Although ITS Required Action C.2 is an additional restriction on plant operations since it requires a verification that the affected penetration is isolated once per 31 days, the overall change related to the addition of ITS Condition C is characterized as Less Restrictive. Lastly, Required Action C.2 is modified by a Note which states "isolation devices in high radiation areas may be verified by use of administrative means." This allowance minimizes personnel exposure and recognizes that access to high radiations areas are restricted such that the probability of a valve misalignment is small. This change is consistent with NUREG-1432 as modified by TSTF-30 . 1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated? Analyzed events are assumed to *be initiated by the failure of plant structures, systems or components. The proposed change extends the allowed outage time for an inoperable containment isolation valve in a penetration that consists of a closed pipmg system and one containment isolation valve. The proposed change does not involve a change to any accident initiators or precusors. Therefore, the proposed change does not involve a significant increase in the probability of an accident. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change does not alter the assumed mitigatory capability of the containment isolation function since the redundant isolation barrier formed by the closed system is required to be Operable. Thus, the consequence of an accident occurring during the time presently allowed in the CTS is the same as the consequences for an accident occuring during the * . time proposed in the ITS. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated .

  • 2 . 3. *
  • Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change still ensures that an OPERABLE containment isolation barrier is available whenever the plant is in a condition that requires containment isolation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? The margiri of safety is a function of the overall containment leakage. Piping systems that penetrate containment must be provided with isolation and containment capabilities having redundancy, reliability and performance capabilitiies which reflect the importance of isolating these penetrations. The proposed change extends the allowed outage time for an inoperable containment isolation valve in a penetration that contains a closed piping system and one containment isolations valve. During this extended period, the only affect on the containment isolation function is a loss of redundancy since the remaining isolation barrier ensures that leakage from the containment atmosphere is within the limits assumed in the safety analysis. As such, there is no increase in the amount of radioactive material released to the environment. Therefore, the proposed change does not involve a significant reduction in a margin of safety .

  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-12 DOC M.2 JFD 8 CTS 3.6.3 CTS 4.5.3.d ITS SR 3.6.3.2 and SR 3.6.3.3 and Associated Bases CTS 3.6.3 specifies that prior to the reactor going critical after a refueling outage an administrative check will be made to confirm that all "locked closed" manual CIVs are closed and locked. A TS Amendment Request dated March 26, 1997, proposes to modify CTS 3.6.3 to change the "Administrative check" to a "visual check" and renumber the specification as CTS 4.5.3.d. This Amendment is under review by the staff. The corresponding ITS SRs are ITS SR 3.6.3.2 and SR 3.6.3.3. The ITS SRs verify only that the "not-locked closed" valves are closed, shown in the ITS markup by the implementation of TSTF-45 (JFD 8). The CTS markup does not show this Less Restrictive (L) change nor is it as part of the overall More Restrictive changes (DOC M.2) made to this CTS SR. Comment: Revise the CTS markup and provide the appropriate discussion and justification for this Less Restrictive (L) change. Consumers Energy RffPonse:

A new DOC (DOC L.4) and NSHC (NSHC L.4) have been provided to justify the less restrictive . requirement that only valves which are not locked, sealed, or otherwise secured in position need to be verified in accordance with ITS SR 3.6.3.2 and ITS SR 3.6.3.3 .. Affected Submittal Pages: Att 3, CTS page 4-21 (ITS 3.6.3, page 4 of 5) Att 3, DOC 3.6.3, page 10of10 Att 4, NSHC 3.6.3, page 2 of 2 13

  • 4.S 4.S.2 ' -(continued)

(1) Individual onta1nment be leak r e testect.at a f equency of at 1 st every refuelin , not exceedin s ecifi in a and b shall The containment equipment h and the fuel transfer t e shall be tested at eac refueling outage or after ch time used if that be sooner. (b) A full air 1 k penetration test shall be p1rforme at six-month i ervals. During the period between t e six-month ests when CONTAINMENT INTEGRITY is re uired, a reduced p ssure test for the door seals or a f 1 air lock pen ration test shall be performed with 72 hours after e ther each air lock door opening or t first of a series of openings

  • I Se.c. 3.'4.7..>
  • ..1./..:..4

_____________ \ l 5

  • h.e isolation valves shall be demonstrated OPERABLE by performance I © of a cycling test and verification of isolation time for auto ._ '-isolation valves prior to declaring the valve to be OPERABLE after maintenance, repair, or replacement work 1s performed on the valve 3-b*3-30 3.Co.1.lo l ..... b:l'a ., o..... c.c.+J .... s;-\ ... t-.cd "3. 1..1. Ll or its associated actuator, control, or power circuit. . 4-21 \ I * / 3 -a-* Amendment No. -I-Ci, -t-14, "-------I C:"'2:PiictJ ""-

C.\Vs. . /" * \!::.:::J

  • *
  • LA.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES LA.5 TS 4.5.3b states " LE by erifying that on ea containment isolation right channel left channel test signal, pplicable isolatio valves actuate to their required positi during COLD f<.Pr I HUTDOWN or t least once per refueling cycle." Th requirement to test the valves*. 3." "during COLD HUTDOWN" is not included in the oposed ITS. The proposed ITS ill test the v ves every 18 months which is the equ* alent to the "once per refueling ycle" in the roposed ITS. The requirement totes the valves during "COLD SHUTDO " will be addressed by plant proced es. Changes to plant procedur_;§ are made* accordance with the plant procedure ge process.

This change is consiste with NUREG-1432. LESS RESTRICTIVE CHANGES (L) L.1 CTS Table 4.2.2 item 13.a specifies that the Containment Purge and Ventilation Isolation Valves shall be determined closed "at least once per 24 hours." The proposed. ITS SR 3.6.3.1 changes the frequency for verifying the 8 inch containment purge valve and 12 inch air room supply valves are closed from 24 hours to 31 days. This change is reasonable because these valves are not allowed to be open in proposed MODES 1-4. These valves are required to be locked closed, either electrically or by some other means to de-activate the valve. Therefore, the likelihood of the valve being opened is very remote. The 31 day frequency is consistent with the surveillance requirements for other safety related systems which require valve position verification. This change is consistent with NUREG-1432. L2.. 'I:tJ.St.n. r RA\ /...; {{Al !.(,,;-9 L. L{ 'C.f°'" R/11 j.f#.'3""/I-- Palisades Nuclear Plant Page 10of10 01/20/98 L 1 5 Se... ItJS a; 1M1 /3-b L.*h tl IH 5 Jo. 3-3 I

  • *
  • 3.6.3 DOC L.4 CTS 4.5.3d requires that prior to the reactor going critical after a refueling outage, a visual check will be made to confirm that all "locked-closed" manual containment isolation valves are closed and locked except for valves that are open under administrative controls.

In the ITS, manual valve position verification is required by proposed SR 3.6.3.2 and SR 3.6.3.3. The requirements of SR 3.6.3.2 and SR 3.6.3.3 are less restrictive than the CTS since they only apply to valves that are "not locked, sealed, or otherwise secured in position." This proposed change is acceptable since these valves are verified closed when they are locked, sealed, or otherwise secured in position. Administrative programs provide the appropriate controls to assure valves that are normally locked, sealed, or otherwise secured in position are in their correct position. This change is consistent with NUREG-1432 as modified by TSTF-45 Rev.l . J3-c_

  • ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
3. L.?.. L. LIL.I l.5 L.b The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation.

The proposed change will continue to verify that the purge exhaust and air room supply yalves are still closed. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? The proposed change increases the surveillance interval from 24 hours to 31 days. The margin of safety afforded by the purge exhaust and air room supply valves are that they are closed and thereby isolating a direct path from the containment atmosphere. Since the purge exhaust and air room supply valves are required to be closed in MODES 1, 2, 3 and 4, the likelihood of these valves being open is very small. Therefore, this change does not involve a significant reduction in a margin of safety . Se.t -1 / S<-t. I:MS C#-/(.1+ I 3 . -q Se.t. l 3 *bo 3 ... , 7-Sec. IrJSu.1 f?!+ I 3*&.3 -1'1/IS /Jt/17 Sec:.. Lu.su 7 Ill 1 3 . b* 3 -5 / Palisades Nuclear Plant Page 2of2 01/20/98 )3-d

  • 3.6.3 NSHC L.4 CTS 4.5.3d requires that prior to the reactor going critical after a refueling outage, a visual check will be made to confirm that all "locked-closed" manual containment isolation valves are closed and locked except for valves that are open under administrative controls.

In the ITS, manual valve position verification is required by proposed SR 3.6.3.2 and SR 3.6.3.3. The requirements of SR 3.6.3.2 and SR 3.6.3.3 are less restrictive than the CTS since they only apply to valves that are "not locked, sealed, or otherwise secured in position." This proposed change is acceptable since these valves are verified closed* when they are locked, sealed, or otherwise secured in position. Administrative programs provide the appropriate controls to assure valves that are normally locked, sealed, or otherwise secured in position are in their correct position. This change is consistent with NUREG-1432 as m,odified by TSTF-45 Rev. I. -1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated? Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change limits the periodic verification of manual containment isolation valves to only those valves that are not locked, sealed, or otherwise secured in position. The proposed change does not involve a change to any accident initiators or precursors. Therefore, the proposed change does not involve a significant increase in the probability of an accident. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change does not alter the assumed capability of the containment isolation function since the required valves are still assured to be in their correct position. Thus, the consequence of an accident remain unchanged. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change still ensures that an OPERABLE containment isolation barrier is available whenever the plant is in a condition that requires containment isolation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated . 13-e_

  • 3 . *
  • Does this change involve a significant reduction in a margin of safety? The margin of safety is a function of the overall containment leakage. The proposed change limits the periodic verification of manual containment isolation valves to only those valves that are not locked, sealed, or otherwise secured in position.

This change does not relax the requirement to maintain containment integrity, but recognizes the misalignment of valves which have been verified closed when they were locked, sealed, or otherwise secured in position is small. As such, this change does not result in any activity that would result in an increase in the amount of radioactive material released to the environment. Therefore, the proposed change does not involve a significant reduction in a margin of safety . 13-6

  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUE$T: 3.6.3-13 DOC M.2 DOC M.3 JFD2 CTS 4.2 Table 4.2.2, Item 13b CTS 4.5.3.d STS SR 3.6.3.4 and SR 3.6.3.6 ITS SR 3.6.3.3, SR 3.6.3.5 and Associated Bases The frequencies specified for CTS 4.2 Table 4.2.2, Item 13.b (ITS SR 3.6.3.5) and CTS-4.5.3.d (ITS SR 3.6.3.3) have been modified to "Once prior to entering .. ;92 days" and "184 days and once within 92 days after each opening of the valve," respectively.

These frequencies are different that the frequencies specified for the corresponding STS SRs. The justification used for these changes (JFD 2) is for clarity grammatical preference and to establish consistency. The staffs position is that the change does not fall into any of these categories and is considered to be a generic change which is beyond the scope of review for this conversion. Comment: Delete this generic change. Consumers Energy Remonse: The word "once" has been deleted from the Frequency of ITS SR 3.6.3.2. For other changes made to ITS SR 3.6.3.5, see RAI. 3.6.3-29. Affected Submittal pages: Att 1, ITS 3.6.3, page 3.6.3-6 Att 5, NUREG 3.6.3, page 3.6-13 " 14 Containment Isolation Valves . 3.6.3 .

  • SURVEILLANCE REQUIREMENTS SURVEILLANCE FRC:QUENCY SR 3.6.3.3 -------------------NOTE--------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative means. ------------------------------------------- RP.I Verify each manual containment isolation 9ftee-fo'i or to . valve and blind flange that is located entering MODE 4 inside containment and not locked, sealed, from MODE 5 if or otherwise secured, and required to be not performed closed during accident conditions, is within the closed, except for containment isolation previous valves that are open under administrative 92 days controls. SR 3.6.3.4 Verify the isolation time of each automatic In accordance power operated containment isolation valve with the

  • is within limits. Inservice Testing Program V..r 1+i 184 days RAI SR 3.6.3.5 Perform le rate fer 3,(,.,3-containment 8 inch purge exhaust and 12 inch air room supply valve1-w+!fl il i iRt SH:l sx is

'oy ('11111c< 0 a.. ra. fc. -kst. 92 da s afte each openin of the valve Z.( SR 3.6.3.6 Verify each automatic containment isolation 18 months valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal . ** Palisades Nuclear Plant 3.6.3-6 Amendment No. 01/20/98 )"-/-o-x Containment Valves li:Aimefspherii and Ou!ll] @'f .*. 3. 6. 3

  • SURVEILLANCE RE UIREMENTS continued SR 0."J. r.dr-\oc..k, I e..IU, o..-
c. TS 't. S'. 5. c.. SR 3.6.3, . .... * '" c...1S
  • l .. 'U . ."2. 0 3. SR 3.6.3i 3 .IJ.3-ZC\
  • SR g CEOG STS SURVEILLANCE

**************-NOTE-*******************

Valves and blind flanges in high radiation areas may be verified by use of administrative means. Verify valve and blind flange that 1s located inside containmentiiand required to bt

  • closed during accident conditions is closed, except for containment isolation that are open under administrative controls.

Verify the isolation time of leath po'llJ!rl ppef!ted indl each automatic ontainment iso ation valve.is within limits. i_ 1 Perfe,. le age rate te9tf ng far COntainmt(t.f>U91 valvew *Uh f:Ui] hA* 2 2<** e,.I...< -J. 11-""' "" .... -s .... u' {re; C.l11x.d 'of pcr.forma.nc.' of. Q. * -c. ro.-k -t<.s1 (j) FREQUENCY rior to entering HOOE 4 from MODE S i f not perfonned within the previous 92 days In accordance with the Inservice Testing ProMamloft 192 lS 184 days (J) (f) Ytrify 1ach automatic containment tfstlmonths valvt that is not locked, sealed, or oth1rwis1 secured in position, actuates to the isolation position on an actual or simulated actuation signal. I \ 3.6-13 } Ll-b (continued) Rev 1, 04/07 /95

  • * ---------CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.3-14 DOC LA.2 CTS 4.5.3.a CTS 4.5.3.a requires that the CIVs be demonstrated OPERABLE after maintenance, repair or replacement work is performed on the valve or its associated components.

This information, according to DOC LA.2 is being relocated to plant procedures. It is unclear from the discussion in DOC LA.2 if the procedure change control process is covered by 1 O CFR 50.59 or some other non-regulatory control process. If the procedure change control process is not covered by 1 O CFR 50.59 then the change is a Less Restrictive (L) change deletion of material rather than a Less Restrictive (LA) change. Less Restrictive (LA) changes are limited to those items which are relocated to licensee controlled documents covered by a 10 CFR 50.59 change control process. Comment: Provide additional discussion and justification on the plant procedure change control process. Consumers Energy Response: A new DOC (DOC L.5) and NSHC (NSHC L.5) have been provided to justify the deletion of extraneous details contained in CTS 4.5.3a. DOC L.5 supersedes the justification previously provided in DOC LA.2. As such, DOC LA.2 has been deleted in its entirety. Affected Submittal f'ages: NOTE: These pages are applicable to RAJ 3.6.3-14, 3.6.3-15, 3.6.3-16 and 3.6.3-17. Att 3, CTS, page 4-13 (ITS 3.6.3, page 5 of 5) Att 3, CTS_, page 4-21 (ITS 3.6.3, page 4 of 5) Att 3, DOC 3.6.3, page 9of10 Att 3, DOC 3.6.3, page 10 of 10 Att 4, NSHC 3.6.3, page 2 of 2 \ \ 15

  • 4.Z lZ. Verify tht Io nt Rtt10v1l Syst .. TSP baskets ar OPERABLE by tht surveillance Verify tht TSP baskets contain b wttn 8,300 ind 11,00 TSP tlCh 18 llOnths. Verify that a sa.plt froa t TSP baskets provide idequitt pH adJust .. nt of borated watt tach 18 months. 113.

PU1"'9t and Isolati6n Valves I 3.

  • 3.t...3.5
  • Tht Conta1n .. nt Purge and Ytnt11at1on Isolation Valves shall bt dttt,..intd-lostd: A.\!.\ 14. Shutdown Coo11nt 15. To .. ,t tht sh down cooling r1qu1r ... nts of b. Tht r uirtd r11ctor coohnt pump(s) if not in operation should dttt intd to bt OPERABLE onct per days by ver1fy1ng correct br ktr align81nts and indicated wer av111abi11ty.

ht rtqu1rtd sttl8 g1n1r1tor(s shall bt d1t1n11ned OPERAS by v1rif11nt the s1condar1 water 1v1l to bt at least ct per 12 hours. At least one coolant loop or train shall be verified o bt 1n operation and c1rculat1n reactor coolant at least net per lZ h a. Yer 1 that th* Matn F11dwat1r Regulating alvt and the issoc1 b ass valve close on an actual or simul ted Containment High ressure (CHP) signal once each 18 mon s. Verify that th* Main Feedw1t1rlR1gul ing valve and tht a sociattd bypass valve close on an actual or 1mulated Stt .. Gener. tor Low Pressure (SGLP) stgnal :onct each 110nths

  • Amtndmtnt No. &4-, HI, 165 May 19, 1995 4.13 I 5-.o.. 3.1)
  • 4.5 4.5.2 -(continued)

@ S pe.c :4=: c. c:.+: o.._ 4 (1) Individual ontainment iso tion valves shall be leak r e tested'. at a f equency of at 1 refuelin , not exceedin s ec1fi in a and b (b) A full air 1 k penetration test shall be perfonne at six-month i ervals. During the period between t 1 six-month ests when CONTAINMENT INTEGRITY is re uired, a reduced p ssur1 test for the door seals or a f 1 air '-... 1 ock pen ration test shall be performed with 72 hours Se.c. 3.1....7-/ after e ther each air lock door opening or t first of a series of openings.

14. S.3 I containment Isolltioli/Yalyes 1
  • Ci} h.e isolation valves shall be demonstrated OPERABLE by performance I .IJ of a cycling test and verification of isolation time for auto ")'ll isolation valves prior to declaring the valve to bt OPERABLE after 3.G,,3-30 maintenance, repair, or replace1111nt work is performed on the valve \ or its associated actuator, control, or power circuit. S"\2. 3.C...1 lo ..J. ,Alj :1:3:' *. Q .... 0.C:.+J OD" s;-\ ... t'<<d ..:__ __ ........____, 0 '3. 1..1. '-' 3.1., .1.'2. Co .. h:da s;R

.. e ,.: .... j "'\ ,..,._ Mooe S ot .... -e.J. . "";'""""' ..\"\..L rreo1:0 ... s "\'Z,. dA,s * ,, 4-21

  • Amendment No. ai, 28, -H4, .._ ____ < Abb c'iQ:.PT!eN ec,, Prx.t:tA,

"> $(!..u r'-11 C.1 I .SR ! . fl. 3 ./o / \.!:::.:..::::) I 5 -6 Ahl L.5 *

  • *
  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES LESS RESTRICTIVE CHANGES-REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) LA.1 LA.2 LA.3 In CTS 1.0, Containment Integrity is defined in parts (a) through (e). As stated in Administrative Change A.3, the CTS requirements of "Containment Integrity" are split into three LCOs in the proposed ITS. Parts (a) and (d) of the definition of Containment Integrity relate to containment isolation valves and effectively require that the containment isolation valves are OPERABLE.

In the proposed ITS LCO 3.6.3, the LCO statement is reduced to simply "Containment isolation valves are OPERABLE." Therefore in part (a) the reference to "nonautomatic" containment isolation valves and the requirement that "blind flanges are closed" is not needed. In part (d) the reference to "automatic" containment isolation valves and the fact that they are " .. .locked closed" is also not needed. The Bases for proposed ITS LCO 3.6.3 will discuss the different types of containment isolation valves and their associated requirements. Therefore, the additional information can be deleted from the LCO and will be addressed in the Bases. This change is maintains consistency with NUREG-1432. r.J .. a spec1 ies equrrements for demonstratmg contamment elation valve PERABILITY afte "maintenance, repair, or replacement wor is performed on the alves or its associ ed actuator, control, or power circuit." se actions are dictated y post maintena e and post-modification testing requiremen which are under icensee control. Changes to plant_procedures are made in a ordance with the plant rocedure chan e process. Therefore, this information is ot included in the ro osed ITS. This chan e maintains consistenc with N G-14 Palisades Nuclear Plant Page 9of10 01/20/98 ) 5-f_

  • *
  • LA.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES LA.5 LE by erifying that on ea left channel test signal, pplicable isolatio valves actuate to their required positi during COLD f<Pr I HUTDOWN or t least once per refueling cycle." Th requirement to test the valves 3. *lr "during COLD HUTDOWN" is not included in the oposed ITS. The proposed ITS ill test the v ves every 18 months which is the equ* alent to the "once per refueling ycle" in the roposed ITS. The requirement totes the valves during "COLD SHUTDO "will be addressed by plant proced es. Changes to plant procedur;§ are made
  • accordance with the plant procedure ge process. This change is . consiste with NUREG-1432.

LESS RESTRICTIVE CHANGES (L) L.1 CTS Table 4.2.2 item 13.a specifies that the Containment Purge and Ventilation Isolation Valves shall be determined closed "at least once per 24 hours." The proposed. ITS SR 3.6.3.1 changes the frequency for verifying the 8 inch containment purge valve and 12 inch air room supply valves are closed from 24 hours to 31 days. This change is reasonable because these valves are not allowed to be open in proposed MODES 1-4. These valves are required to be locked closed, either electrically or by some other means to de-activate the valve. Therefore, the likelihood of the valve being opened is very remote. The 31 day frequency is consistent with the surveillance requirements for other safety related systems which require valve position verification. This change is consistent with NUREG-1432. L2. st.' :t'tV.S uz. r /(.Al {{A\ ?i.C,.;-9 \ \ L. L/ s._ ;t I\)\ "C.(";-/Uri j.'1.'3""/')... Palisades Nuclear Plant Page 10of10 01/20/98 L.5 s c. '1).JS ta,; /l/11 N/ts/u,/17 !5-d L*b Se<, 5.<o. 3-31

  • *
  • 3.6.3 DOC L.5 CTS 4.5.3a, CTS 4.5.3b, and CTS 4.2 Table 4.2.2, Items 13.a and 13.b contain details that are not necessary to describe, or *are not pertinent to, any actual regulatory requirement.

As such, these details are proposed for deletion. Specifically: CTS 4.5.3a describes the testing necessary for CIVs prior to declaring the valves Operable after maintenance, repairs, or replacement work is performed on the valve or its associated actuator, control, or power circuit. Explicitly stating these tests as they relate to maintenance activities is unnecessary since the technical specifications stipulate the level of performance that must be met for an Operable CIV in the associated Surveillance Requirements. CTS 4.5.3b states that each CIV shall be demonstrated Operable by verifying ... valves actuate to their required position "during Cold Shutdown or at least once per refueling cycle". The phrase "during Cold Shutdown" is intended to describe the plant condition which best facilitates testing the applicable (automatic) CIVs. The phrase "at least once per refueling cycle" establishes the frequency for the test. Specifying the plant condition at which CIV testing is performed (i.e., Cold Shutdown) is a detail which is not pertinent to the actual requirement for testing CIVs. The LCO Applicability for CIVs stipulates the plant conditions when CIVs are required to be Operable. Testing within the Applicability is governed by valve Operability, testing in plant conditions outside the Applicability has no impact on safety . CTS 4.2 Table 4.2.2, Item 13a states the containment purge and ventilation isolation valves are determined closed "by checking the valve position indicator in the control room". The intent of verifying valve position is to ensure that the valve is in its correct position. Specifying that valve position be verified "by checking the valve position indicator in the control room" does not constitute a requirement assumed in the safety analyses. Rather, it simply provides a method for assuring the valve is in the correct position. Since the valves may be locked closed electrically, mechanically , or by other physical means, stipulating "by checking the valve position indicator in the control room" is an inappropriate detail not pertinent to the actual requirement. CTS 4.2 Table 4.2.2, Item 13b states the containment purge and ventilation isolation valves are determined closed by performing a leak rate test "between the valves." Specifying that a leakage rate test be performed "between the valves" does not constitute a requirement assumed in the safety analyses. Rather, it simply provides a method for conducting the leakage rate test. Since the above details are not necessary to describe, or are not pertinent to, any actual regulatory requirement, they can be deleted without an impact to public health and safety. These changes are consistent with . I '5 -

  • *
  • 2. 3. .. L. L. t..I L.5 L.b ATTACH1\.1ENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue to verify that the purge exhaust and air room supply yalves are still closed. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? The proposed change increases the surveillance interval from 24 hours to 31 days. The margin of safety afforded by the purge exhaust and air room supply valves are that they are closed and thereby isolating a direct path from the containment atmosphere. Since the purge exhaust and air room supply valves are required to be closed in MODES l, 2, 3 and 4, the likelihood of these valves being open is very small. Therefore, this change does not involve a significant reduction in a margin of safety. Se.t t\Al -1 I S<-c. X:rVS cµ. I 3 Sec. If.JSi:.rt l 3.b. 3 ... ,7-.. INS&-1 f<ff I 3.&.3-1'1/IS/l,/17 Sec,... IN..su 1 f.,q l 3 . f,. .3 -I Palisades Nuclear Plant Page 2of2 01/20/98 15-f

  • *
  • 3.6.3 NSHC L.5 CTS 4.5.3a, CTS 4.5.3b, and CTS 4.2 Table 4.2.2, Items 13.a and 13.b contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement.

As such, these details are proposed for deletion. Specifically: CTS 4.5.3a describes the testing necessary for CIVs prior to declaring the valves Operable after maintenance, repairs, or replacement work is performed on the valve or its associated actuator, control, or power circuit. Explicitly stating these tests as they relate to maintenance activities is unnecessary since the technical specifications stipulate the level of performance that must be met for an Operable CIV in the associated Surveillance Requirements. CTS 4.5.3b states that each CIV shall be demonstrated Operable by verifying ... valves actuate to their required position "during Cold Shutdown or at least once per refueling cycle". The phrase "during Cold Shutdown" is intended to describe the plant condition which best facilitates testing the applicable (automatic) CIVs. The phrase "at least once per refueling cycle" establishes the frequency for the test. Specifying the plant condition at which CIV testing is performed (i.e., Cold Shutdown) is,a detail which is not pertinent to the actual requirement for testing CIVs. The LCO Applicability for CIVs stipulates the plant conditions when CIVs are required to be Operable. Testing within the Applicability is governed by valve Operability, testing in plant conditions outside the Applicability has no impact on safety . CTS 4.2 Table 4.2.2, Item 13a states the containment purge and ventilation isolation valves are determined closed "by checking the valve position indicator in the control room". The intent of verifying valve position is to ensure that the valve is in its correct position. Specifying that valve position be verified "by checking the valve position indicator in the control room" does not constitute a requirement assumed in the safety analyses. Rather, it simply provides a method for assuring the valve is in the correct position. Since the valves may be locked closed electrically, mechanically , or by other physical means, stipulating "by checking the valve position indicator in the control room" is an inappropriate detail not pertinent to the actual requirement. . CTS 4.2 Table 4.2.2, Item 13b states the containment purge and ventilation isolation valves are determined closed by performing a leak rate test "between the valves." Specifying that a leakage rate test be performed "between the valves" does not constitute a requirement assumed in the safety analyses. Rather, it simply provides a method for conducting the leakage rate test. Since the above details are not necessary to describe, or are not pertinent to, any actual regulatory requirement, they can be deleted without an impact to public health and safety. These changes are consistent with NUREG-1432. \

  • 1. 2.
  • 3.
  • Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed changes delete details from the Technical Specifications that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. The deletion of details from the Technical Specifications is not assumed to be an initiator of any analyzed event. The proposed changes do not reduce the functional requirement or alter the.intent of any specification. As such, the.consequences of an accident remain unchanged. Therefore, the proposed.<:hanges do not involve a significant increase in the probability or consequences of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change deletes detail from the Technical Specifications that are nQt necessary to describe, or are not pertinent to, any actual regulatory requirement. The* changes will not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. The changes will not impose different requirements, and adequate control of activities will be maintained. The changes will not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes will not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? Margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. The proposed changes delete details from the Technical Specifications. Removal of these details is acceptable since this information is not directly pertinent to the actual requirement and does not alter the intent of the requirement. Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to lice.nsee controlled document without a significant impact on safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. \ \ /5-h_

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.3-15 DOC LA.3 CTS 4.2 Table 4.2.2 Item 13.a CTS 4.2 table 4.2.2 Item 13.a requires determining that the containment purge and ventilation isolation valves is closed by checking the valve position indicator in the control room. The "checking the valve position indicator in the control room" according to DOC LA.3 is being relocated to plant procedures.

It is unclear from the discussion in DOC LA.3 if the procedure change control process is covered by 10 CFR 50.59 or some other non-regulatory control process. If the procedure change control process is not covered by 10 CFR 50.59 then the change is a Less Restrictive (L) change deletion of material rather than a Less Restrictive (LA) change. Less Restrictive (LA) changes are limited to those items which are relocated to licensee controlled documents covered by a 10 CFR 50.59 change control process. Comment: Provide additional discussion and justification on the plant procedure change control process. Consumers Energy Response: A new DOC (DOC L.5) and NSHC (NSHC L.5) have been provided to justify the deletion of extraneous details contained in CTS 4.2. Table 4.2.2, Item 13a. DOC L.5 supersedes the justification previously provided in DOC LA.2. As such, DOC LA.2 has been deleted in its entirety. Affected Submittal Pages: See RAI 3.6.3-14. " 16

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  • CONVERSION TO IMPROVED .TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.3-16 DOC LA.4 CTS 4.2 Table 4.2.2 Item 13.b CTS 4.2 Table 4.2.2 Item 13.b requires performing a leak rate test of the containment purge and ventilation isolation valves by performing a leak rate test between the valves. The details of how to perform the test (between the valves) according to DOC LA.4 is being relocated to plant procedures.

It is unclear from the discussion in DOC LA.4 if the procedure change control process is covered by 10 CFR 50.59 or some other non-regulatory control process. If the procedure change control process is not covered by 1 O CFR 50.59 then the change is a Less Restrictive (L) change deletion of material rather than a Less Restrictive (LA) change. Less Restrictive (LA) changes are limited to those items which are relocated to licensee controlled documents covered by a 1 O CFR 50.59 change control process. Comment: Provide additional discussion and justification on the plant procedure change control process. Consumers Energy Response: A new DOC (DOC L.5) and NSHC (NSHC L.5) have been provided to justify the deletion of extraneous details contained in CTS 4.2. Table 4.2.2, Item 13b. DOC L.5 supersedes the justification previously provided in DOC LA.2. As such, DOC LA.2 has been deleted in its entirety. Affected Submittal Pages: See RAI 3.6.3-14. \ \ 17

  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.3-17 DOC LA.5 CTS 4.5.3.b CTS 4.5.3.b. requires that each CIV activates to ITS isolation position during COLD SHUTDOWN and at least once per refueling cycle. The "during COLD SHUTDOWN" according to DOC LA.5 is being relocated to plant procedures.

It is unclear from the discussion in DOC LA.5 if the procedure change control process is covered by 1 O CFR 50.59 or some other regulatory control process. If the procedure change control process is not covered by 1 O CFR 50.59 then the change is a Less Restrictive (L) change deletion of material rather than a Less Restrictive (LA) change. Less Restrictive (LA) changes are limited to those items which are refloated to licensee controlled documents covered by a 10 CFR 50.59 change control process. See Comment Number 3.6.3-18. Comment: Provide additional discussion and justification on the plant procedure change control process. See Comment Number 3.6.3-18. Consumers Energy Response: A new DOC (DOC L.5) and NSHC (NSHC L.5) have been provided to justify the deletion of extraneous details contained in CTS 4.5.3.b. DOC L.5 supersedes the justification previously

  • provided in DOC LA.2. As such, DOC LA.2 has been deleted in its entirety.

Affected Submjttal pagf§". See RAI 3.6.3-14. \I

  • 18
  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.3-18 DOC LA.5 CTS 4.5.3.b ITS SR 3.6.3.6 and Associated Bases CTS 4.5.3.b requires that each CIV be demonstrated OPERABLE by verifying on an actuation signal that the valve actuates to its required position.

The corresponding ITS SR is ITS SR 3.6.3.6. The CTS frequency for this SR is "during COLD SHUTDOWN or at least once per refueling cycle." The ITS frequency is 18 months which is the equivalent to "At least once per refueling cycle." The frequency of "during COLD SHUTDOWN" is relocated to plant procedures. It is unclear from the discussion in DOC LA.5 if there are certain CIVs because of plant specific design characteristics, operational constraints or current licensing basis which are required to be tested at each COLD SHUTDOWN, i.e., more than once per refueling cycle. If this is the case then the change is all!lO a Less Restrictive (L) change and could be considered as a beyond scope of review item for'this conversion. See Comment Number 3.6.3-17. Comment: Provide additional discussion and justification on this Less Restrictive change. Consumers Energv Response: See response to RAI 3.6.3-17 Affected Submittal Pages: No page changes . 19 CONVERSION TO IMPROVED TECHt-.!ICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION

  • SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.3-19 JFD 1 STS 3.6.3 RA E.1 and Associated Bases ITS 3.6.3 RA E.1 and Associated Bases STS/ITS 3.6.3 RA E.1 states that the affected penetration flow path shall be isolated "by use of at least one ... closed manual valve with resilient seals ... " STS B 3.6.3 Bases -E.1, E.2 and E.3 states in the first paragraph that the "Isolation barrier that meet this criterion are ... a closed manual valve with resilient seals". ITS B 3.6.3 Bases -E.1, E.2 and E.3 deletes the phrase "A closed manual valve with resilient seals" based on JFD 1 which implies that Palisades does not have manual valves in the containment purge system." See Comment Number 3.6.3-29.

Comment: Revise the ITS markup to make ITS 3.6.3 RA E.1 and its Associated Bases consistent. See Comment Number 3.6.3-29. Consumers Energy Remonse: See response to RAI 3.6.3-29. Affected Submittal Pages:

  • No page changes. ,, 20
  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-20 JFD 4 STS B 3.6.3 Bases BACKGROUND ITS 3.6.3 ACTION Note 1 ITS SR 3.6.3.2 and SR 3.6.3.3 ITS B 3.6.3 Bases -BACKGROUND STS B 3.6.3 Bases -BACKGROUND first paragraph third sentence states the following: "Manual valves, de-activated automatic

... passive devices." ITS B 3.6.3 Bases -BACKGROUND modifies this sentence by adding the word "Closed" prior to "Manual valves." Based on ITS 3.6.3 ACTION Note 1, ITS SR 3.6.3.2 and ITS SR 3.6.3.3 not all manual valves are required to be closed, only those required to be closed or locked closed for accident conditions. Thus, the adding of the word "closed" to the Bases sentence changes the intent of the sentence and excludes open manual valves. Comment: Delete this change. Consumers Energ.v Response: The word "closed" has been deleted from the first paragraph, third sentence, in the Bases Background section . Affected Submittal Pages: Att 2, ITS 3.6.3, page B 3.6.3-1 Att 5, NUREG 3.6:3, page B 3.6-19 ,, 21

  • *
  • Containment Isolation Valves B 3.6.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.3 Containment Valves BASES BACKGROUND The containment isolation valves and devices form part of the containment pressure boundary and provide a means for isolating penetration flow paths. These isolation);fevices RAI are either passive or active (automatic). .H,J-,2..0 valves, de-activated automatic valves secured in their closed position (including check with flow through the valve secured), blind flanges, and closed systems are considered passive devices. Check valves, or other automatic valves designed to close without operator action following an accident, are considered active devices. Two barriers in series are provided for each penetration so that no single failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the safety analysis.

One of these barriers may be a closed system

  • Containment isolation occurs upon receipt of a Containment High Pressure (CHP) signal or a Containment High Radiation (CHR) signal. However, not all containment isolatio.n valves are actuated by both signals. The signals close automatic containment isolation valves in fluid penetrations not required for*operation of Engineered Safety Feature systems in order to prevent leakage of radioactive material.

Other penetrations are isolated by the use of valves or check valves in the closed position, or blind flanges. As a result, the containment isolation valves (and blind flanges) help ensure that the containment atmosphere will be isolated in the event of a release of radioactive material to containment atmosphere from the Primary Coolant System (PCS) following a Design Basis Accident (OBA). The OPERABILITY requirements for containment isolation valves and devices help ensure that containment is isolated within the time limits assumed in the safety analysis. Therefore, the OPERABILITY requirements provide assurance that the containment leakage limits assumed in the accident analysis will be not exceeded in a OBA

  • Palisades Nuclear Plant B 3.6.3-1 01/20/98 c2/-o-
  • *
  • B 3.6 CONTAINMENT SYSTEMS Containment Isolation Valves j(Ati1)6spherfc and DAal}i-@ B 3:6 . B 3.6.3 Containment Isolation Valves ICAtmfspheric/and Duil) t--@ BASES . Jc..:c.C'j)

The containment isolation valve onn part of the **"'.\ I :::::,; SAC KG ROUND i s .,w; 0 .. ... .- containment pressure boundary and provide a means for fluid I {q'\ pene ra ns no serv ng e cons uence 1m1 1n \....;/ systems to be provided w h two 1sola on barrier. that are close on an automatic olation sf 1. These

  • dev ces. ire e t er pass ve or act 1Ye automatic)..

nua . 1 valves, de-activated automatic valves secured in the1r c

  • closed position (including check valves with flow througn the valve secured), blind flanges, and closed systems are considered passive devices. Check valves, or other .automatic valves designed to close without operator action following an accident, are considered active devices. Two barriers in series are provided for* each penetration so that no single credible failure or malfunction of an active component can result 1n a loss of isolation or leakage that exceeds limits assumed in the safety analysis.

One of these barriers may be a closed system. Tht OPERABILITY requirements for containnient isolation ensur1 that containment is isolated within the (Y -ti1111 limits assumed in the safety analysis. Therefore, the OPERABILITY re uirements provide assurance that the containment u c assumed in the accident analysis will CEOG STS bt I I B 3.6-19 Jl-b (continued) Rev 1, 04/07/95

  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-21 JFD 4 STS B 3.6.3 Bases -C.1 and C.2 ITS B 3.6.3 Bases -C.1 and C.2 STS B 3.6.3 Bases -C.1 and C.2 states in the last paragraph, last sentence, that "The probability of misalignment of these valves, once they have been verified ... is small." ITS B 3.6.3 Bases -C.1 and C.2 modifies this sentence by changing "valves" to "devices" . Blind flanges cannot be misaligned.

Thus, the STS wording is correct. Comment: Delete this change. Consumers Energy Response: In the ITS Bases for Actions C.1 and C.2, the last sentence in the last paragraph has been revised from " ... the probability of misalignment of these devices, once they have been verified to be in the proper position, is small" to ...... the probability of mispositioning these devices, once they have been verified to be in the proper position, is small". This change provides symmetry in the sentence structure by providing a correlation to the phrase "proper position" using the word "misposition".

  • Affected Submittal Pages: Att 2, ITS 3.6.3, page B 3.6.3-8 Att 5, NUREG 3.6.3, page B 3.6-26
  • 22
  • *
  • BASES ACTIONS (contin11ed)

C.l and C.2 Containment Isolation Valves B 3.6.3 With one or more penetration flow paths with one containment isolation valve inoperable, the inoperable valve must be restored to OPERABLE status or the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration. Required Action C.l must be completed within the 72 hour Completion Time. The specified time period is reasonable, considering the relative stability of the closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of supporting containment OPERABILITY during MODES 1, 2, 3, and 4. In the event the affected penetration is isolated in accordance with Required Action C.l, the affected penetration flow path must be verified to be isolated on a periodic basis. This is necessary to assure leak tightness of containment and that containment penetrations requiring isolation following an accident are isolated. The Completion Time of once per 31 days for verifying that each affected penetration flow path is isolated is appropriate considering the valves are operated under administrative controls and the probability of their misalignment is low. Condition C is modified by a Note indicating that this Condition is only applicable to those penetration flow paths with only one containment isolation valve and a closed system. This Note is necessary since this Condition is written to specifically address those penetration flow paths in a closed system. Required Action C.2 is modified by a Note that applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of these devices, once they have been be in the proper position, is small. 11ll$f6S,t1bn11)d /?.Al 3.<P.3-'Z-/ Palisades Nuclear Plant B 3.6.3-8 01/20/98 }_d.-o-

  • *
  • Containment Isolation Valves ((Atl1}6spherij?

and o)?a1i1-({3)

  • B 3.6.3 BASES ACTIONS C.l and C.2 (continued) path must be isolated.

The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate Required Action C.1 must be completed @ \!.!;;J Completion Time. The specified time I \1 period is reasonable, considering the relative stability of the closed syste* (hence, reliability) to act as a penetration isolation boundary and the relative importance of supporting containment OPERABILITY during HODES 1, 2, 3, and 4. In the event the affected penetration is isolated f n accordance with Required Action C.l, the affected penetration flow path must be verified to be isolated on a ?>>,, 5 periodic bash. This is necessary to assure leak tightness of containment and that containment penetrations requiring 0 rtn,rc. isolation following an accident are isolated. The Completion Tfmt of once per 31 days for verifying that each affected penetration flow path is isolated is appropriate considering the valves art operated under administrative controls and the probability of their misalignment is low. I).\ t I . -f "°'"C.. C.* ,-f00-S"'fP'" \o..-h*o.._ ->CL\A.\ ... +o c.\ose ca. 1..oc.A a.--..l ... .... .j.... 1.e. \-.1c.ul L+° .,..-. .+ .,J..\..<.Sf!, ..,,\..,...\

s .. --.l ,..+ \*c.'c..cJ Condition C is modified by a Nott indicating that this Condition is only applicable to those penetration flow paths with only one containment isolation valve and a closed system. This Nott is necessary since this Condition is written to specifically address those penetration flow paths in a closed system.
  • I l .f1...L ... t.,*...l r Required Action C.2 ts modified by a Note that applies to

..ic.\..re. I.e.. ; .... ... +&, valves and blind flanges. located. in high radiation areas and ... \1., * '** *, '.1.:i* allows these devices to be verified closed by use of *. , . I'> al' cnd*.lcs adminhtrativt means. Allowing verification by ld***e '"c -=' ' 1 ada:t.nhtrativ1 means 1s considered acceptable, since access i , * 'G \i * , i; . W * ** *-* t-o the st areas ts typ 1ca11 y restricted. Therefore, the JC\ "' ,,1. ... , I'"" d \oe: . probab11 ity thasa once they have\(:!; X w . ; +-1 .. i. al 1eJ Le. .been verified \.o be in the proper POlit o is small. within 1 it, the assumption of the safety anal sis are not (continued) CEOG STS \ B 3.6-26 Of((. howr 1s provid.tr) +o Joc..t CW'-d the a. ftc.c.lcd i/a.IW* Rev 1, /he I hovr C..o"fk+*1r -t (f'r.. prou1dey Q... f't.r1 cd o+ +1tiit rf1 C.oac+ Gmi'Y1ei,..,Svro.i1't +nc:.. !'Ma..111+o-1ri1nJ C.loSe.d,. +hf. f robl<.H\

  • Zl-b 04/07/95
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REOUE$T: 3.6.3-22 JFD 6 STS SR 3.6.3.1, SR 3.6.3.2 and Associated Bases ITS SR 3.6.3.1 and Associated Bases STS SR 3.6.3.1 specifies that the purge valves are verified sealed closed except for one purge valve in a penetration flow path while in Condition E. The ITS markup shows that STS SR 3.6.3.1 is not used and that STS SR 3.6.3.2 is modified in ITS SR 3.6.3.1 state the same intent as STS SR 3.6.3.1 except that the exception for one purge valve in a penetration not being sealed while in Condition Eis not included.

JFD 6 does not provide sufficient information as to why STS SR 3.6.3.2 and its associated Bases was modified to reflect plant specific design when STS SR 3.6.3.1 and its associated Bases would have been the appropriate SR to use. In addition, the deletion of the exception could result in the unnecessary shutdown of the unit if the purge valve is not within leakage limits and cannot be locked closed. Comment: Provide additional discussion and justification with regards to the changes made to STS SR 3.6.3.1, STS SR 3.6.3.2/ITS SR 3.6.3.1 and their associated Bases. Consumers Energy Response: JFD 6 has been revised to provide additional discussion for the deviation between ITS SR 3.6.3.1 and, ISTS SR 3.6.3.1 and SR 3.6.3.2. Affected Submittal Pages: Att 6, JFD 3.6.3, page 1 of 5 I \ 23

  • ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Change Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS. " The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has been provided.
2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of intent. 3. 4. 5. 6. The requirement/statement has been deleted since it is not applicable to this facility. The following requirements have been renumbered, where applicable, to reflect this deletion. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description. This change reflects the current licensing basis/technical specification. The Palisades purge and exhaust system consists of a supply through a 12 inch air RA-I 3.(D .') z..9 room supply path and two 8 inch exhaust paths. Previously the Palisades Nuclear Plant design contained "large" purge valves but they have been removed from the plant design. The NUREG-1432 portions dealing with the [42] inch purge valves is not included in the proposed ITS. Each path of the existing purge exhaust and air room supply contains two isolation valves with resilient seals. The primary difference between the use of the purge and exhaust path at Palisades and NUREG-1432 portions dealing with the "mini-purge" valves is that at Palisades the purge exhaust valves and air room supply valves are locked closed in MODES 1-4 and are not allowed to be opened. Therefore, for Palisades the valves are inoperable if the valves are open, or closed but not locked, since they have not been qualified to be able to close in the event of a Design Basis Accident. valv;s ae \:lRQctrge a leak test every 184 t:i&}'S.

  • TusaT A \ . . . .

Palisades Nuclear Plant Page 1 of 5 01/20/98 d-3 -o-

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  • 3.6.3 JFD 6 INSERT A Verification that the containment purge exhaust air room supply valves are closed is accomplished by performing a leakage rate test every 6 months. NUREG-1432 also specifies that a leakage rate test be performed within 92 days after opening the valves. That requirement has not been adopted in the ITS since the containment purge exhaust and air room supply valves are not allowed to be opened in MODES 1 -4. INSERT B ISTS SR 3. 6. 3 .1 requires a verification that each [ 42] inch purge valve is sealed closed except for one purge valve in a penetration flow path while in Condition E of the LCO. That SR was not adopted in the ITS since the Palisades plant design does not include 42" valves. Rather, ISTS SR 3.6.3.2 was revised to agree with LCO 3.6.3 (CTS 3.6.5) which requires the purge exhaust valves and air room supply valves to be closed in Modes 1, 2, 3, and 4. If one of these valves were open, the valve would be declared inoperable and the Required Actions of Condition D taken (See JFD 12). Proposed ITS Condition D, unlike ISTS Condition E, does not allow unrestricted operation in Modes 1, 2, 3, or 4. Thus, the exception contained in ISTS SR 3.6.3.1 related to Condition E does not apply. Since the purge valves are not used in MODES 1, 2, 3, or 4 for pressure control, ALARA, surveillance testing, or air quality considerations, that aspect of ISTS SR 3.6.3.2 has been deleted from the SR .
  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-23 JFD 6 JFD 15 STS B 3.6.3 Bases -BACKGROUND ITS B 3.6.3 Bases -BACKGROUND STS B 3.6.3 Bases -BACKGROUND in the fifth paragraph, second sentence, states that "Motor operated isolation valves are provided inside containment, and air operated isolation valves are provided outside containment." ITS B 3.6.3 Bases -BACKGROUND modifies this sentence by deleting "Motor operated ... inside containment, and", deleting the word "are" between "valves" and "provided" in the second part of the sentence, and adding "The purge exhaust and air room supply valves are." This change is designated and justified by JFD 6. This is incorrect.

JFD 6 does not describe the location of the containment purge valves, whereas JFD 15 does. In addition, based on the modifications made and JFD 15 the sentence would be more correct if "provided" were changed to located. Comment: Revise the ITS markup. Consumers Energy Response: The fifth paragraph, second sentence in the ISTS Bases Background section has been re-annotated to show that the appropriate NUREG deviation discussion is JFD 15 and not JFD 6 as previously indicated. In addition, the word "provided" has been replaced by the word "located" to correct that sentence. Affected Submittal Pages: Att 2, ITS 3.6.3, page B 3.6.3-2 Att 5, NUREG 3.6.3, page B 3.6-20 \ I 24

  • BACKGROUND (continued)

Containment Isolation Valves B 3.6.3 The 8 inch purge exhaust valves are designed for purging the containment atmosphere to the stack while introducing filtered makeup, through the 12 inch air room supply valves from the outside, when the plant is shut down during refueling operations and maintenance. The purge exhaust valves and air room supply valves are air operated isolation outside the containment. These valves are opera ed manually from the control room. These valves will close automatically upon receipt of a CHP or CHR signal. The air operated valves fail closed upon a loss of air. These valves are not qualified for automatic closure from their open position under OBA conditions. Therefore, these valves are locked closed in MODES 1, 2, 3, and 4 to ensure the containment boundary is maintained. Open purge exhaust or air room supply valves, following an accident that releases contamination to the containment atmosphere, would cause a significant increase in the containment leakage rate. APPLICABLE The containment isolation valve LCO was derived from the SAFETY ANALYSES assumptions related to minimizing the loss of reactor coolant inventory and establishing the containment boundary during major accidents. As part of the containment boundary, containment isolation valve OPERABILITY supports leak tightness of the containment. Therefore, the safety analysis of any event requiring isolation of containment is applicable to this LCO. The DBAs that result in a release of radioactive material within containment are a Loss of Coolant Accident (LOCA), a Main Steam Line Break (MSLB), and a control rod ejection accident. In the analysis for each of these accidents, it is assumed that containment isolation valves are either closed or function to close within the required isolation time following event initiation. This ensures that potential paths to the environment through containment isolation valves (including containment purge valves) are minimized. The safety analysis assumes that the purge exhaust and air room supply valves are closed at event initiation. Palisades Nuclear Plant \ \ 01/20/98

  • *
  • BASES BACKGROUND (continued)

Containment I so lat ion Valves /(Atm9Spheri? and Dua 1 a 3.5 . ..... I '" S "'-pp 1-, V<i. \.AS ,-....... i' ... ex!..*-';- .,__J .... .... -roo-a..l'e. * (10:WI (@ 1@ APPLICABLE The containment isolation valve lCO was derived from the SAFETY ANALYSES assumptions related to minimtzing the *loss of reactor coolant inventory and establishing the containment boundary during major accidents. As part of the containment boundary, containment isolation supports leak tightness of the containment. Therefore, the safety analysis of any event requiring isolation of containment is applicable to this LCO. CEOG STS The OBAs that result in a release of radioactive material within containment art a loss of coolant accident (LOCA), a \ \ B 3.6-20 (continued) Rev 1, 04/07/95

  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-24 JFD 9 CTS 4.5.3.c STS SR 3.6.3.5 ITS SR 3.6.3.4 and Associated Bases ITS SR 3.6.3.4 and its associated Bases modifies STS SR 3.6.3.5 and its associated Bases in accordance with TSTF-46 which verifies the isolation time of each automatic power operated CIV. CTS 4.5.3.c verifies the isolation time of "each power operated or automatic" CIV. The CTS and ITS are not in agreement, in that the CTS requires all power operated valves to be tested which would include all non-automatic power operated valves and all automatic valves to be tested which would include check valves. The ITS does not include all these valves. Comment: Revise the CTS markup and provide a discussion and justification for this Less Restrictive (L) change. Consumers Energy Response:

CTS 4.5.3c states that "the isolation time of each power operated or automatic valve shall be verified in accordance with Section XI of the ASME Boiler and Pressure Vessel Code." Prior to Amendment 184 to the Facility Operating License for the Palisades Plant, CTS 4.5.3c stated "the isolation time of each power operated or automatic valve shall be determined to be within its limit as specified in Table 3.6.1 when tested in accordance with Section XI of the ASME Boiler and Pressure Vessel Code." The isolation time limits specified in CTS Table 3.6.1 applied only to "Auto Isolation Valves" as denoted in the "REMARKS" column of Table 3.6.1. In the staffs Safety Evaluation for Amendment 184, it was stated "the staff has reviewed the licensee's proposed deletion of Table 3. 6. 1 and its associated TS changes and determined that the changes are in accordance with the guidance of GL 91-08 .. Deleting the list of containment isolation valves does not alter the existing TS requirement or the components they apply to. Lists of containment isolation valves are provided in the Final Safety Analysis Report and in the plant procedures for performing penetration leak testing and isolation valve closure time testing. The set of valves subject to the requirements of TS 3. 6 and 4. 5 will not change due to the proposed change. The staff, therefore, find the proposed change acceptable." Based on the contents of former CTS Table 3.6.1 and the intent of the change made in Amendment 184, the requirements of ITS SR 3.6.3.4 remain unchanged from CTS 4.5.3c. Affected Submittal Pages: No page changes. 25

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-25 JFD 10 CTS 3.6.5 ITS 3.6.3 ACTION D, F, and Associated Bases CTS 3.6.5 requires that with one containment purge exhaust or air room supply isolation valve not locked closed, the valve shall be locked closed within 1 hour or the plant shall be shutdown within 36 hours. The corresponding ITS ACTIONS are ITS 3.6.3 ACTION D and F respectively.

The Bases discussion of ITS 3.6.3 ACTION D does not provide an adequate justification for requiring the unlocked valve to be locked closed within 1 hour. The STS would allow via STS 3.6.3 ACTION A 4 hours to lock close the valve. Comment: Revise the ITS markup of ITS B 3.6.3 Bases -D.1 to provide a reasonable justification for the 1 hour Completion Time. Consumers Energy Response: The Bases discussion for ITS Condition D has been revised to provide a reasonable justification for the 1 hour Completion Time. Affected Submjttal Pages: Att 2, ITS 3.6.3, page B 3.6.3-9 Att 5, NUREG 3.6.3, page B 3.6-26 26 RA! -2..5' ACTIONS (continued) D"" IS Pmud-tJ fo ia<X C.l t&Cd +he. o. tfc:.c-kd !Ail V r. *S t(" The purge exhaust and air room valves have not been qualified to close following LOCA and are required to be locked closed. If one f these valves is found not locked closed, the potential exists for the valves to be inadvertently opened. Wbile tbe otRer iselatie" oalve R tRe patl:1 prgvi d&& i sgl iti or:i froiii tbi '9Rtai RRleRt tbe valve Wii Rot elesed be I resterea tl=li R E>Re-R0UF tg Ali r:ii za J:&-1--i aRse eR a s i val ven. In the event one or more ntainment purge valves in on or ore penetration flow pat s are not within the purge va ve leakage limits, purge va ve leakage must be restored t ithin limits, or the a ected penetration must be is ated. he method of isolatio must be by the use of at lea one isolation barrier that cannot be adversely affected y a single active failure Isolation barriers that mee this riterion are a clos and de-activated automatic lve with resilient seals, or blind flange. A purge valv with resilient seals uti ized to satisfy Required Acti n E.1 must ave been demonstr ted to meet the leakage requi ements of R 3.6.3.5. The ecified Completion Time is r asonable, onsidering that ne containment purge valve r ains closed o that a gross reach of containment does no exist. In accordance "th Required Action E.2, this penetration flow path mus be verified to be isolated o a periodic basis. The P, riodic verification is neces ry to ensure that contain ent penetrations required to e isolated following a accident, which are no long capable of being automatica y isolated, will be in the i elation position should an vent occur. This Required A does not require y testing or valve manipulat'on. Rather, it involves verification, through a syst walkdown, that those isolati n devices outside containmen capable of being mispos* ioned are in the correct po tion. For the isola on devices inside containmen , the time period speci ied as "prior to entering MO E 4 from MODE 5 if not perf rmed within the previous 92 ays" is based on I eng neering judgment and is reasonable in view of th inaccessibility of the isol ion devices and other a inistrative controls that wi 1 ensure that isolation 'd vice misalignment is an unli ely possibility

  • Palisades Nuclear Plant B 3.6.3-9 01/20/98
  • *
  • BASES ACTIONS Containment Isolation Valves i(Atl1)?spherij?

and o)la1)}-dj)

  • B 376.3 *c.1 and C.2 (continued) path must be isolated.

The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate Required Action C.l must be completed @ Completion Time. The specified time I \1 period is reasonable, considering the relative stability of the closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of supporting containment OPERABILITY during HODES 1, 2, 3, and 4. In the event the affected penetration is isolated in accordance with Required Action C.1, the affected penetration flow path must be verified to be isolated on a ?> .. 5 periodic bash. This is necesury to assure leak tightness of containment and that containment penetrations requiring 0 rrnc.r(. isolation following an accident are isolated. The Completion Time of once per 31 days for verifying that each affected penetration flow path is isolated is appropriate considering the valves art operated under administrative controls and the probability of their is low

  • Condition C 1s modified by a Nott indicating that this Condition is only applicable to those penetration flow paths with only one containment isolation valve and a closed syste11. This Nott is necessary since th1s Condition 1s written to specifically address those penetration flow paths in a closed system. Required Action C.2 is modified by a Nott that applies to valves and blind flanges.located.in high radiation areas and allows these devices to be verified closed by use of adm1n1strat1vt means. Allowing verification by means 1s considered acceptable, since access i' 1 r\;;, i; t*o these areas 1s typically restricted.

Therefore, the IC\ c\ *1e>-***, d . probability. ef th.ese once they have\ & x * ; += \ *'-,J *' r*ir 11e. .been verified \.o be in the proper poli a is small. *1 OK& .. -the condary contat ... n bypass .leakage ra not J c within 1 it, the assumption of the safety anal sis are not {continued) CEOG STS 11 B 3.6-26 Rev 1, 04/07 /95 hour IS f1rJtJ1d.rJ fo loc-t the a. f.fc.(;,icd l/a./IJcJ, the I hovr /( pro<.>1d'.> o.. (Jr.,;, c4 o+ rfG +nt f thc... 1'Ma..1"+o-m1nJ fA-f-1><.. lM '"": C./ o Sc d. '?., {o-b

  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST:
  • JFD 11 STS B 3.6.3 Bases -LCO ITS B 3.6.3 Bases -LCO STS B 3.6.3 Bases -LCO, second paragraph, first sentence, states that "The automatic power operated ... to actuate on an automatic isolation signal." ITS B 3.6.3 Bases -LCO modifies the end of the sentence to read as follows: "To actuate on upon receipt of a CHP or CHR signal as appropriate signals." The change does not make sense. Comment: Revise the ITS markup to correct this discrepancy.

Consumers Energ.v Response: The second paragraph, first sentence of the ISTS LCO Bases has been revised to state: "To actuate upon receipt of a CHP or CHR signal as appropriate." Affected Submittal Pages: Att 5, NUREG 3.6.3, page B 3.6-22 \ I 27

  • * *
  • Containment Isolation Valves [4t:lfiosphe/1c ant11ua1¥-{5)

.6 . BASES {continued) LCO , APPLICABILITY CEOG STS ,,.;ii 'l:ltc. r-e vi. ,..111! s ':,.O WC 1:'1 ....... @ The normally closed i so la

  • es are cons ered OPERABLE when manual valves are closed, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact. These passive isolation valves or devices are those listed in Reference 1 This LCO provides assurance that the containment isolation valves and purge valves will p1rfon11 their designed safety \,C"\ fu11ctions ta minimize the loss of o coolant inventory

'G..I and establish the containment oundary during accidents. In MODES 1, 2, *3, and 4, a OBA could cause a release of radioactive material to containment. In HODES 5 and 6, the probability and consequences of these events are reduced due to tht pressure and temperature limitations of these HODES. Therefore, the containment isolation valves are not required to be OPERABLE in MODE 5. The requirements for containment isolation valves during MOOE 6 art addressed in LCO 3.9.3, *contain1n1nt Penetrations.* \ \ B 3.6-22 d7-o-(continued) Rev 1, 04/07/95

  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-27 JFD 13 JFD 13 states in the second paragraph "TSTF-30, Rev. 2, also include a change to the Bases for SR 3.6.2.2 which stated: "The closed system must meet the requirements of SRP 6.2.4." This statement is incorrect.

TSTF-30, Rev 2 did not make a change to the Bases for SR 3.6.2.2. It did, however, add a similar statement to STS B 3.6.3 Bases -C.1 and C.2. Also, the quoted statement ends with the words "Requirements of Reference 3," not "requirements of SRP 6.2.4." Reference 3 in the REFERENCE Section of STS B 3.6.3 is SRP 6.2.4. Comment: Revise J FD 13. Consumers Energy Response: JFD 13 has been revised to correctly reflect the changes made by TSTF 30, Rev. 2. Affected Submittal Pages: Att 6, JFD 3.6.3, page 4 of 5 28

  • *
  • Ml 3.fi,.J-l.1a W1fl6i"'I

£] I U rc..h f()ft!J.' eiiaus+ a.rd o.ir rwrn PvfPly i.b.ti.tJ 1na1ea.1 s tnc. Lail.Jr. rro.v in . o 12. CTS 3.6.5 requires that the containment purge exliaust anM air room supply isotati6't lt.bd 13. valves shall be locked closed whenever the plant is above COLD SHUTDOWN. This"" Y C.

  • requirement is an implicit requirement in the proposed ITS 3.6.3 since it states "Each containment isolation valve shall be OPERABLE" and the proposed Bases state that the purge exhaust and air room supply valves must be locked closed. Therefore, an additional Action must also be provided in the proposed ITS if this requirement is not met. Proposed ITS Action D addresses the situation where "One purge exhaust or air room supply valve not locked closed" and requires that within 1 hour, the affected valve must be locked closed. This Action is added to the proposed ITS to reflect a plant specific change based on the Palisades Nuclear Plant CTS. NUREG-1432 Condition Chas a "bracketed" 4 hours as a Completion Time for an inoperable containment isolation valve in a penetration flow path with only one containment isolation valve and a closed system. TSTF-30, Rev. 2, revises this 4 hours to 72 hours. A Completion Time of 72 hours is appropriate since it involves a closed system which minimizes the potential of a leakage pathway. *since the Palisades Nuclear Plant CTS did not provide an explicit Completion Time for this condition, the 72 hours is included in the proposed ITS. . * .

Ac.tt&ns C,.1 C.. 'Z.. TSTF-30, Rev. 2, also included a change to the Bases fot"sft 3.6.2.2 which stated: "The closed system must meet the requirements o SRP 6.2. _)( Such a Bases statement is not consistent with the current licensing as1s an was not incorporate SRP 6.2.4 does not provide "requirements," but rather provides review guidance for acceptability of the design for containment isolation The containment penetrations which utilize closed systems have previously beln determined to be

  • acceptable.

The acceptability may have been based on compliance with SRP 6.2.4 or may have been justified on some other basis. Whatever the basis, the system design has been determined acceptable for use as a containment isolation feature, and it 'is inappropriate to now prevent such use unless it complies with SRP 6.2.4. Therefore, this change is consistent with TSTF-30, Rev. 2, except where the generic change is inconsistent with the current licensing basis. re. n ( IS ,, Palisades Nuclear Plant Page 4 of S 01/20/98 -0..

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.3-28 JFD 15 ITS 3.6.3 Required Action E.2 and Associated Bases ITS B 3.6.3 Bases -BACKGROUND ITS 3.6.3 RA E.2 requires the verification that the affected purge valve penetration flow path is isolated with a Completion Time of "Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devises inside containment." Based on the discussion in JFD 15 and ITS B 3.6.3 Bases -BACKGROUND, the containment purge valves are located outside containment.

Thus, the Completion Time of "Prior to ... inside containment" is not applicable to Palisades. See Comment Number 3.6.3-9 and 3.6.3-29. Comment: Revise the ITS markup to delete this Completion Time from ITS 3.6.3 RA E.2 and its associated Bases. Provide a discussion and justification for this change. See Comment Numbers 3.6.3-9 and 3.6.3-29. Consumers Energy Response: See response to RAI 3.6.3-29 Affected Submittal pages: No page changes. \ ., 29

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.3-29 CTS 3.0.3 CTS 3.6.1 ACTIONS CTS 4.2 Table 4.2.2, Item 13.b CTS 4.5.2.c ITS 3.6.3 ACTION E and Associated Bases ITS SR 3.6.3.5 and Associated Bases CTS 4.2 Table 4.2.2 Item 13.b requires a leak rate test of the containment purge and ventilation isolation valves. No action is provided in the CTS if this surveillance is failed. Thus, as a minimum entry into CTS 3.0.3 is required.

The corresponding ITS SR is ITS SR 3.6.3.5. Failure of ITS SR 3.6.3.5 would require entry into ITS 3.6.3 ACTION E. The remedial measures provided in ITS 3.6.3 ACTION E are Less Restrictive than requirements of CTS 3.0.3. The CTS markup of CTS 3.6.1 ACTIONS; CTS 4.2 Table 4.2.2, Item 13.b; or CTS 4.5.2.c do not show the addition of ITS 3.6.3 ACTION E. See Comment Numbers 3.6.3-9, 3.6.3-19 and 3.6.3-28. Comment: Revise the CTS markup and provide the appropriate discussions and justifications for the addition of this Less Restrictive (L) change. See Comment Numbers 3.6.3-9, 3.6.3-19 3.6.3-28. Consumers Enerav Response: As correctly noted by the NRC staff, the CTS markup failed to identify the addition of ITS 3.6.3 Condition E, and no justification was provided in the Discussion of Changes (DOCs) related to ITS 3.6.3 Condition E. This oversight has been corrected based on the following discussion. The purge exhaust valves and the air room supply valves have not been qualified to close following a LOCA. Thus, these valves must be locked closed whenever the plant is above Cold Shutdown. CTS 4.2, Table 4.2.2 Item 13 requires the containment purge and ventilation valves be determined "Closed" by: (a) checking the valve position indicator in the control room, and (b) performing a leak rate test between the valves. If either of these tests fail, the affected valve would be considered "Not Closed"and the actions of CTS 3.6.5 would be taken. The intent of the leakage rate test performed in accordance with CTS 4.2, Table 4.2.2 Item 13b is to verify the containment purge and ventilation valves are closed and that the resilient seal material in these valves has not deteriorated to the point where the valves could no longer be considered closed. This position is supported in the Basis of CTS 3.6.1 which states "to ensure that the (containment purge and ventilation) valves are closed and that the seals have not degraded, a between the valves leak rate test is periodically performed." The Basis for CTS 3.6.1 was previously accepted by the NRC staff in their approval of Amendment

  1. 90 to the Palisades plant Provisional Operating License. " (continued) 30
  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.3-29 (continued)

Consumers Energy Response: Based on the requirement of the CTS, ITS 3.6.3 has been revised to delete Condition E in its entirety. If one or more containment exhaust valves and air room supply valves are not locked closed as determined by proposed SR 3.6.3.1 and SR 3.6.3.5, the Required Actions of ITS Condition D must be taken. In addition, the wording of SR 3.6.3.5 has been revised to clearly indicate the purpose of that SR, and to delete the 92 day Frequency interval since this Frequency was chosen (as stated in the ISTS) "recognizing that cycling the valve could introduce additional seal degradation beyond that occurring to a valve that has not been opened." The Bases of ITS 3.6.3 has been revised to reflect the changes to the Actions and SR 3.6.3.5, and to clarify that the containment exhaust valves and the air room supply valves are not required to meet additional leakage rate testing requirements. Lastly, conforming changes have been made to the DOCs and JFDs. Affected Submittal Pages: Att 1, ITS 3.6.3, page 3.6.3-1 Att 1, ITS 3.6.3, page 3.6.3-2 Att 1, ITS 3.6.3, page 3.6.3-4 Att 1, ITS 3.6.3, page 3.6.3-5 Att 1, ITS 3.6.3, page 3.6.3-6 Att 2, ITS 3.6.3, page B 3.6.3-4 Att 2, ITS 3.6.3, page B 3.6.3-5 Att 2, ITS 3.6.3, page B 3.6.3-7 Att 2, ITS 3.6.3, page B 3.6.3-9 Att 2, ITS 3.6.3, page B 3.6.3-10 Att 2, ITS 3.6.3, page B 3.6.3-13 Att 3, CTS, page 3-40 (ITS 3.6.3, page 1 of 5) Att 3, CTS, page 4-13 (ITS 3.6.3, page 5 of 5) Att 3, DOC 3.6.3, page 2of10 Att 3, DOC 3.6.3, page 8 of 10 Att 5, NUREG 3.6.3, page 3.6-8 Att 5, NUREG 3.6.3, page 3.6-9 Att 5, NU REG 3.6.3, page 3.6-10 Att 5, NUREG 3.6.3, page 3.6-11 Att 5, NUREG 3.6.3, page 3.6-13 Att 5, NUREG 3.6.3, page B 3.6-22 . Att 5, NUREG 3.6.3, page B 3.6-23 Att 5, NUREG 3.6.3, page B 3.6-25 Att 5, NUREG 3.6.3, page B 3.6-27 Att 5, NUREG 3.6.3, page B 3.6-28 Att 5, NUREG 3.6.3, page B 3.6-31 Att 6, JFD 3.6.3, page 1 of 5 Att 6, JFD 3.6.3, page 3 of 5 Att 6, JFD 3.6.3, page 4 of 5 31

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  • 3.6 CONTAINMENT SYSTEMS Containment Isolation Valves 3.6.3 3.6.3 Containment Isolation Valves LCO 3.6.3 Each containment isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS -------------------------------------NOTES------------------------------------

1. Penetration flow paths, except for 8 inch purge exhaust valves and 12 inch air room supply valves penetration flow paths, may be unisolated intermittently under administrative controls.
2. Separate Condition entry is allowed for each penetration flow path. 3. Enter applicable Conditions and Required Actions for made inoperable by containment isolation valves
  • 4. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," when leakage results in exceeding the overall containment leakage rate acceptance criteria.

A. CONDITION

NOTE---------

On l y applicable to penetration flow paths with two containment isolation valves. One or more penetration flow paths with one containment isolation valve inoperable (except for purge exhaust valve or air room supply valve not locked closed -&F l eakage flat *nfffR n HmH!) . Palisades Nuclear Plant A.1 REQUIRED ACTION Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, bl ind flange, or check valve with flow through the valve secured. COMPLETION TIME 4 hours (continued) 3.6.3-1 21 Amendment No. 01/20/98

  • ACTIONS CONDITION A. (continued)

A.2

  • B. ---------NOTE---------

-B.1 Only applicable to penetration flow paths with two containment tsolation valves. ---------------------- One or more penetration flow paths with two containment isolation valves inoperable (except for purge exhaust valve or air room supply valve not locked closed -et" l'lOt w;ti,;n +i ffli t-) Palisades Nuclear Plant Containment Isolation Valves 3.6.3 REQUIRED ACTION COMPLETION TIME --------NOTE--------- Isolation devices in high radiation areas may be verified by use of administrative means. --------------------- Verify the affected Once per 31 days penetration. flow path for isolation is isolated. devices outside containment film. Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment Isolate the affected 1 hour penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. ' ' 3.6.3-2 Amendment No. 01/20/98 x

  • *
  • ACTIONS CONDITION One or more p netration flow paths wi one or more cont inment purge valve not within purge v lve leakage limits. Palisades Nuclear Plant E.1 AND E.2 Mm E.3 Containment Isolation Valves 3.6.3 REQUIRED ACTION Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve with resilient seals, or bl ind flange.

Iso late devices in high radiation areas/ may be verified1' b use of administra ve means. / ______________ if ____ _ Verify the fected enetratio flow path isolat c.t. Perform SR 3.6.3.5 (leakage test) for the resilient seal purge valves closed to comply with Required Action E .1. COMPLr:TION TIME 24 hours I I I I Once per 31 days for isolation devices outside containment Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment days 3.6.3-4 31-C-Amendment No. 01/20/98

  • *
  • Containment Isolation Valves 3.6.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. f .1 PAI © Required Action and Be in MODE 3. 6 hours .3 C\ associated Completion Time not met. A.fill f 2 Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS
===========================r========

SR 3.6.3.1 SR 3.6.3.2 SURVEILLANCE FREQUENCY Verify each 8 inch purge valve and 12 inch 31 days air room supply valve is locked closed

  • Va l ves and blind flanges in high radiation areas may be verified by use of administrative*

means. Verify each manual containment isolation 31 days valve and blind flange that is outside containment and not locked, sealed, or otherwise secured, and is required to be closed during accident conditions, is closed, except for containment isolation valves that are open under administrative controls . Palisades Nuclear Plant 3.6.3-5 Amendment No. 01/20/98 31--d X) '/..

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  • Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.3 -------------------NOTE--------------------

Valves and blind flanges in high radiation

areas may be verified by use of administrative means. -------------------------------------------

/\Al Verify each manual containment isolation 9nee-fo'i or to valve and blind flange that is located entering MODE 4 inside containment and not locked, sealed, from MODE 5 if or otherwise secured, and required to be not performed closed during accident conditions. is within the closed, except for containment isolation previous valves that are open under administrative 92 days controls. SR 3.6.3.4 Verify the isolation time of each automatic In accordance power operated containment isolation valve with the is within limits. Inservice Testing Program

  • rate testiAg fer 184 days RAI SR 3.6.3.5 3,(, . .3-containment 8 inch purge exhaust and -12 inch air room supply

..,e;il i iRt &eal&x is by fYllllet 0 -ks+. 92 da s afte each openin of the valve SR 3.6.3.6 Verify each automatic containment isolation 18 months valve that is not locked, sealed, or otherwise secured in position. actuates the isolation position on an actual or to simulated actuation signal

  • Palisades Nuclear Plant 3.6.3-6 Amendment No. 01/20/98 2-'f x
  • *
  • BASES LCO (continued)

APPLICABILITY ACTIONS* Containment Isolation Valves B 3.6.3 This. LCO provides assurance that the containment isolation valves and purge valves will perform their designed safety functions to minimize the loss of primary coolant inventory and establish the containment boundary during accidents. In MODES 1. 2, 3, and 4, a OBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are*reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment isolation valves are not required to be OPERABLE in MODE 5. The requirements for containment isolation valves during MODE 6 are addressed in LCO 3.9.3, "Containment Penetrations." The ACTIONS are modified by four notes. Note one allows isolated penetration flow paths, except for 8 inch exhaust and 12 inch air room supply purge valve penetration flow paths, to be unisolated intermittently under administrative controls. These administrative controls consist of stationing a dedicated operator at the valve controls, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for containment isolation is indicated. Due to the fact that the 8 inch purge exhaust valves and the 12 inch air room supply valves may be unable to close in the environment following a LOCA and the fact that those penetrations exhaust directly from the containment atmosphere to the environment, these valves may not be opened under administrative

  • Palisades Nuclear Plant B 3.6.3-4 01/20/98
  • *
  • BASES ACTIONS (continued)

Containment Isolation Valves B 3.6.3 A second.Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable containment isolation valve. Complying with the Required Actions may allow for continued operation, and subsequent inoperable containment isolation valves are governed by subsequent Condition entry and application of associated Required Actions. The ACTIONS are further modified by a third Note, which ensures that appropriate remedial actions are taken, if necessary, if the affected systems are rendered inoperable by an inoperable isolation valve. A fourth Note .has been added that reqµires entry into the applicable Conditions and Required Actions of LCO when leakage results in exceeding the overall containment leakage l i mi t . A.1 and A.2 In the one containment isolation valve in one or more penetration flow paths is inoperable (except for purge ' exhaust or air room supply valves i& Ret or leakage R9t the affected penetration flow. path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de:..activated automatic containment isolation valve, a closed manual valve, a blind flange, and a check valve with flow through the valve secured. For penetrations isolated in accordance with Required Action A.1, the device used to isolate the penetration be the closest available one to containment. Required Action A.1 must be completed within the 4 hour Completion Time. The 4 hour Completion Time is reasonable, considering the time required to isolate the penetration and the relative importance of supporting containment OPERABILITY during MODES 1, 2, 3, and 4 . Palisades Nuclear Plant B 3.6.3-5 01/20/98

  • *
  • BASES ACTIONS (continued)

Containment Isolation Valves B 3.6.3 With two containment isolation valves in one or more RPt I penetration flow paths inoperable (except for purge exhaust valve or-air room supply valve not locked closed OF '1 Rot the affected penetration flow path must be isolated within 1 hour. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve; and a blind flange. The 1 hour Completion Time is consistent with the ACrIONS of LCO 3.6.1. . In the event the affected penetration is isolated in accordance with Required Action B.l, the affected . . penetration must be verified to be isolated on a periodic basis per Required Action A.2, which remains in effect. This periodic verification is necessary to assure leak tightness of containment and that penetrations requiring isolation following an accident are isolated. The Completion Time of once per 31 days for verifying each affected penetration flow path is isolated is appropriate considering the fact that the valves are operated under administrative controls and the probability of their misalignment is low.

  • Condition B is modified by a Note indicating this Condition is only applicable to penetration flow paths with two containment isolation valves. Condition A of this LCO addresses the condition of one containment isolation valve inoperable in this type of penetration flow path . Palisades Nuclear Plant B 3.6.3-7 01/20/98 1-l.
  • BASES ACTIONS RA! (continued)
  • Z . .5' *
  • o ... c. bur IS Pmvd-tJ 1'b iocx: CJ +he. a. H'c:.c+c.d

!Ail V<S, the. I hwr CmPk1,0,, -rim<.. frov1dcs.

a. pw-1 oJ of -h'mt f6 Lu 1"Th th1. I o..\ fr\tl.11"11:1.\nln?l_

fn-c.o<-\la.\ \kS CJ OSeJ Containment Isolation Valves B 3.6.3 t\1\-1 Jl....l '.!> le !i-5 lNI C'Q. The purge exhaust and air room valves have not been qualified to close following LOCA and are required to be locked closed. If one f these valves is found not locked closed, the potential exists for the valves to be inadvertently opened. Wl:lile tl:le otReF iselatie" *alve 4R patR provides tbi -a.tmospbere, tbe va,]ve wbicb Wii ROt loekeel elesed be eRe RQUr to eR a , .. al"& i T

  • In the event one or more ntainment purge valves in on or ore penetration flow pat s are not within the purge va ve leakage limits, purge va ve leakage must be restored t ithin limits, or the an ected penetration must be is ated. he method of isolatio must be by the use of .at lea one isolation barrier that cannot be adversely affected y a single active failure Isolation barriers that mee this riterion are a clos and automatic lve with resilient seals, or blind flange. A purge valv with resilient seals uti ized to satisfy Required Acti n E.l must ave been demonstr ted to meet the leakage requi ements of R 3.6.3.5. The ecified Completion Time is r asonable, onsidering that ne containment purge valve r ains closed so that a gross reach of containment does no exist. In accordance "th Required Action E.2, this penetration flow path mus be verified to be isolated o a periodic basis. The P. riodic verification is neces ry to ensure that contain ent penetrations required to e isolated following a accident, which are no long capable of being automatica y isolated, will be in the i elation position should an vent occur. This Required A tion does not require y testing or valve manipulat"on.

Rather, it involves verification, through a syst walkdown, that those isolati n devices outside containmen capable of being mispos* ioned are in the correct pos tion. For the isola on devices inside containmen , the time period speci ied as 11 prior to entering MO E 4 from MODE 5 if not perf rmed within the previous 92 ays" is based on l eng*neering judgment and is reasonable in view of th inaccessibility Qf the isol ion devices and other a inistrative controJs that wi 1 ensure that isolation l d vice misalignment is an unli ely possibility . Palisades Nuclear Plant B 3.6.3-9 01/20/98 L

  • *
  • BASES ACTIONS SURVEILLANCE REQUIREMENTS E Containment Isolation Valves B 3.6.3 (continued)

For the contai ment purge valve with resilient seal that is isolated in cordance with Required Action SR 3.6.3.5 must be perf rmed at least once every 92 days., This assures that degrad,!ti on of the resilient seal is and confirms t}fat the 1 eakage rate of the con ta i mnent purge valve doe,t not increase during the time the;penetration is isolated/ The normal Frequency for SR 3.613.5, 184 days, is based an NRC initiative, Generic Issue a-20 (Ref. 2). 1 Since more reliance is placed on a while in ! this Cbndition, it is prudent to perfornj'°the SR more often. I a Frequency of once per 92 qays was chosen and I shown to be acceptable based .on operati'ng er1ence. I t' E-vel1 andl(f). 2 If the Required Actions and associated Completion Times are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SR 3.6.3.1 ilili.tt ma.y ioc..t<uJ c.lcSd e. fec.+r1coilf )l . or o+r<<.r-fk)Si(.(>L mOll\S. )" This SR ensures that the 8 inch purge exhaust and 12 inch air room supply valves are locked closed as required. If a valve is open, or closed but not locked, in violation of this SR, the is considered inoperable. but i

e. 'lal *1e is Ret etReF'1Ji se te A ave eKse&&ive leikige, it Reed RQt be CORiidered te leaka§e ef 1;mit.. These valves may be unable to close in the environment following a LOCA. Therefore, each of the valves is required to remain closed during MODES 1, 2, 3, and 4. The 31 day Frequency is consistent with other containment isolation valve requirements discussed in SR 3.6.3.2
  • Palisades Nuclear Plant B 3.6.3-10 01/20/98 31-'
  • *
  • BASES SURVEILLANCE REQUIREMENTS (continued)

QS 1 n fh(.; Sa. t(, C va. I of\ ..for ArntrJMi.,...t /tJ, D +o fu fao/1tj 6 u c,,11 te.. Containment Isolation Valves B 3.6.3 ,i-c \b.l \JCS a.rt Y c_/as,ed l SR 3*5*3*5 (SR 3 (d.I 1.JJ.liJ:. t.S c.Jo;J).J For containment 8 inch purge exhaust and 12 inch air room supply valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J (Ref. 3), is required to ensure Operating experience has demonstrated that this type of seal

  • has the potential to degrade in a shorter time period than do other seal types. Based on this observation and the importance of maintaining this penetration leak tight (due to the direct path between containment and the environment), a Frequency of 184 days was established as part of the NRC resolution of Generic Issue B-20, "Containment Leakage Due _to Seal Deterioration" (Ref.
  • y, t is mu e per orme wit in 92 day after opening the valve. Th 92 day Frequency was chosen recognizing that eye ng the could introduce additional seal de adation (beyond that occurrin to a valve that has no been opened). Thus, decreasi the interval (from 4 days) is a prudent measure a er a valve has been o en
  • SR 3.6.3.6 Automatic containment isolation valves close on a containment signal to prevent leakage of radioactive material from containment following a OBA. Thi.s SR ensures each automatic containment isolation valve will actuate to its isolation position on an actual or simulated actuation signal, i.e., CHP or CHR. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.

The 18 month Frequency was developed considering 1t is prudent that this SR be performed only during a plant outage, since isolation of penetrations would eliminate cooling water flow and disrupt normal operation of many critical components. Operating experience has shown that these components usually pass this SR when performed on the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

  • Palisades Nuclear Plant B 3.6.3-13 01/20/98 31-Av
  • (...C.O I :,.1g.3 . 3.6 J.6.3 3.6.4 (1tt1b /JaC> ACTION: and 1.0 ps1g when in POWER OPE With containment internal pres re above the limit, r tore pressure t within the limit within 1 hou , or be in at least HO SHUTDOWN within the next 6 hours and in COLD HUTDOWN within the fo owing 30 hours. Two ndependent containment hydroge recombiners shall be O ERABLE when th plant ts in POWER OPERATION or. OT STANDBY. With one ecombiner 1 operable, restore the inoperab recombiner to OPERABL status within or be in 1 least HOT S TDOWN within the next l..GoJ.i..J * .:(AOD <ADD Ad'I Or-l c. 0
  • 4.Z lZ. Vtrify tht Io nt Rtt10v1l Syst .. TSP baskets ar OPERABLE by tht survtillanee
  • I Verify tht TSP baskets contain b wttn 8,300 and 11,00 TSP each 18 llOnths. . b. Verify that 1 SlllPlt froa t TSP baskets provide adequate pH adjust .. nt of borated w1t1 each 18 110nths. I 13. Cont!/ru11nt Purgt and Y'nt11ation Isol1tf6n Valves I
  • 3.t...3.5
15.
  • 1. b. Tht r uirtd reactor coolant pump(s) if not in operation should e dtte intd to bt OPERABLE onct ptr days by verifying correct
  • br tr al1gn111nts and 1nd1cattd w1r availability.

ee. ht rtqu1rtd st1aa g1n1r1tor(s shall bt d1ttn1ined OPERAS '3,'I verifying the secondary water 1v1l to bt at least 12 hours. At least on* coolant loop or tr11n shall bt vtr1f1ed o bt tn optrat1on and c1rculat1n r11ctor coolant at least net ptr 12 h a. Vtr 1 that tht Main F11dw1t1r R1gul1t1ng alvt and tht 1ssoei b ass valvt clost on an actual or simul ttd Containmtnt High ressurt (CHP) signal onet each 18 mon s

  • Verify that tht Main fttdw1t1r,R1gul 1n9 valve and tht a sociattd bypass valvt close on an actual or 1mulattd St1111 Gener. tor Low Pressure (SGLP) signal _once 11ch 10nths. Amtndmtnt No. &h 1-il, 1-*, -I-ii, 165 Hay 19. 1995 ,_,, ...-, I
  • *
  • A.4 A.5 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES CTS 3.6. l ACTION c states " ... the plant shall be placed in at least HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 30 hours." With respect to temperature when comparing the CTS and ITS, the proposed ITS MODE 3 is specified by being greater than 300°F while the CTS HOT SHUTDOWN is greater than 525°F. While the ITS covers a broader range, for a shutdown there is no effective difference for achieving either temperature in 6 hours *since they are specified as "greater than." For the CTS COLD SHUTDOWN versus the ITS MODE 5, the temperature requirement is being less than 210°F versus being* less than 200°F in the ITS. This difference of 10 degrees is negligible and has no significant impact on operations.

The other parameter which is common between the CTS terms HOT SHUTDOWN and COLD SHUTDOWN and the corresponding ITS MODES 3 and 5 is the reactivity condition. The ITS MODE 3 and 5 are defined, as a reference point, by a reactivity condition of Keff < . 99. However, in ITS Section 3 .1, the equivalent amount of SHUTDOWN MARGIN is required as that specified in the CTS definitions of HOT SHUTDOWN and COLD SHUTDOWN. Therefore, .the amount of SHUTDOWN MARGIN is considered to be same when the requirements of proposed ITS 3 .1 are considered. The time to reach the CTS COLD SHUTDOWN is specified as " ... within the following 30 hours" while the ITS allows a total of 36 hours to reach MODE 5. Therefore, even if the plant is already in MODE 3 the full 36 hours are allowed. This change reflects the usage rules as specified in NUREG-1432. These changes are all considered to be administrative in that there is no significant impact to operation of the plant and they reflect the terminology and usage rules of NUREG-1432.

  • The phrase "(except for purge exhal.1-St valve or air room supply valve not locked closed er leakage net witlttn limits)" is added to CTS 3.6.1 ACTION statement.

This statement is added to tell the user that another Condition applies for this situation even though it could be considered that this Condition is applicable. Purge. exhaust valve and air room supply valve closure status is addressed by proposed ITS Condition D. Pl.lrge exl:lau&i Valve and air J;QQM snppl:f valve leakase limit& are aSQHSSl!S ey prepgsea ITS Csndition E. This is considered to be an administrative change to clarify the proper usage and application of the proposed ITS. This change maintains consistency with the usage requirements of NUREG-1432 . Palisades Nuclear Plant Page 2of10 01/20/98 -,v x

  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES M.2 (continued)

M.3 v Sc.-c. Il\JSmT These changes are considered to be Restrictive changes since the CTS only requires-*the verification to be performed prior to going critical, does not require verification between refueling outages, and does not address blind flanges. This change is consistent with NUREG-1432. requires that the containment purge and v ntilation isolation valves be leak tested every 6 months to verify adequate closure. The proposed ITS SR 3.6.3.5 re ires the six month frequency (184 days) p sit requires that they be leak tested wi 92 days after opening the valves. The 9 day requirement exists becaus the valves have resilient seals that may inc additional degradation beyond that hich would occur if it was not opened. se valves should not normally be opened MODES 1-4, but this additional require ent is added if for . . some reason they mus opened. Since the Palisades CTS does ot have this requirement the addif n of the 92 day test after the valves have een opened is considered a More estrictive chan e. This chan e is consiste wi G-1432. I:>roposed ITS includes Required Action A.2 which states "Verify the affected penetration flow path is isolated." This verification is not required in CTS 3.6.1 Actions. This verification is necessary to ensure that the flow path with the inoperable valve remains isolated until the inoperable valve is Completion Time for A.'l. l!ftts ttetieB is "Once per 31 days for isolation devices outside containment AND Prior to entering MODE 4 from MODE S if not performed within the previous 92 days for isolation devices inside containment." This verification is not included in the CTS and therefore this is considered to be a more restrictive change. This change is consistent with NUREG-1432 .

  • Palisades Nuclear Plant Page 8of10 01/20/98
  • CiS l.C..I f\ck:o ..... * . -Containment Isolation Valves i<AtmofPheric and . 6 . 3.6 CONTAINMENT SYSTEMS 3.6.3 Conhin111ent Isolation Valves /(Atmo;s'pheric and CR!al)f.--@

LCO 3.6.3 Each containment isolation valve shall be OPERABLE. APPLICABILITY: HODES 1, 2, 3, and 4. ACTIONS . I 2.. ; """"' c-.:, t'OO -i: *

  • i *;;;;*;;th;
  • path%f may be uniso1atec1'intennittently under administrative controls.
  • 2. Separate Condition entry is allowed for each penetration flow path.
  • 3. Enter applicable Conditions and Required Actions for system(s) made inoperable by containment isolation valves. 4. inter applicable Conditions and Required Actions of LCO 3.6.1, *containment,*

when leakage results tn exceeding the overall containment leakage rate acceptance criteria * ------------------------------------------------------------------------------ CONDITION REQUIRED ACTION COMPLETION TIME C.TS ? '-.I. A ... +i*, .... b A. ---------NOTE*****-*** A. l Isolate the affected 4 hours Only applicable to penetration flow path penetration flow paths by use of at least with two containment one closed and isolation valves. de-activated G--------------------- automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured. !fil2 (continued) CEOG STS 3.6-8 Rev 1, 04/07 /95 j/-f

  • *
  • Containment Isolatio11 Valves fil¥'spher1c/and I@ . .3 ACTIONS CONDITION A. {continued)

A.Z B. ---------NOTE********* B.1 Only applicabl1 to penetration flow paths with two containment isolation valv11. ---*------------------ CEOG STS REQUIRED ACTION --------NOTE--------- Isol ation devices in high radiation areas may be verified by use of administrative means. Verify the affected penetration flow path 1s isolated.

  • COMPLETION.

TIME Once per 31 days for isolation devices outside containment Prior to en'teri ng MOOE 4

  • from MODE 5 if not perf armed. . within the previous 92 days for isolation devices inside containment Isolat1 the affected 1 hour penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. (continued) 3.6-9 Rev 1, 04/07 /95
  • * * .::rs J.&;..s
  • ACTIONS continued CONDITION C.

Only applicable to penetration flow paths with only one containment 1so1at1on valve and a closed syste*. ----------------------, On* or lllOrt penetrat1on flow paths with one containment isolation valve inoperable

  • L Seco dary containment byp ss leakage not w1 in 1 i111t. E. Ont or 110r1 perietratiC.

f1ov*p* w1 th ou or 110n c:ont.ai-t*pu valvet*IOt wit n purge valve, 1 akage 1 i*its. CEOG STS t>. e,."-... s+ o .... a.: ... roo-'"'eP'1 v .. l.-t. ,..+ c..\4s,.d, Containment Isolation 3.6.3 0.1 _ , l Isolate tht 1ffecte *penitrat1on flow th by use of at lea one* [closed an de-activated automatic va e with* resilient s als, closed" man al valve w1th *resi ient seals, or blin flange]. 3.6-10 \),\ L 11c.k.. c.\oHIA ... c3)-]\_ 24 hours I {continued) Rev 1, 04/07/95 I he,. ......

  • *
  • ACTIONS CONDITION
t. (continued)

@ e.. J.t..\ AJ-;o,..t 0 Rlquhwl Actt* and issectll&lll C...letton Ti* "°' *t. CEOG STS

  • Containment Isolation ValvesJ(Atrnf1'c;phericfind REQUIRED ACTION --------NOTE---------

Isolation devices in high radiation areas may be verified by use of administrativ* means. Verify the affected penetration flow pith is isolated. Perform-SR for the resilien seal purge valv closed to compl Required Actio Ba in MODE 3. 81 in MODE S. 3.6-11 3 /--s COMPLETION TIME Once per 1 days for 1sol t ion devices outside contai ent Pri r to en er1 ng HOOE 4

  • f 111 MODE S if t perfonned*

i thin the previous 92 days for 1so ht devices inside containment

  • Once per j CJ @ 6 hours 36 hours Rev 1, 04/07 /95
  • Containment Isolation Valves .ITA!mefspheri?

and ouill] (I:£j 3.6.3 SURVEILLANCE RE UIREMENTS continued c..rs Y. s. 3.ct SR 3.6.3.£ °'"""" r.dr-Id I 'ita.lt4, o..- H!(.._,.cJ. c,TS 't. S'. 3. '-SR 3.6.34} * .. r c..1S * ."2.."2. ?,. b) SR AAI SR '1 CEOG STS SURVEILLANCE

NOTE--------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative means. Verify valve and blind flange that is located inside containment"1nd required to be

  • closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

Verify the isolation time of leath power! ppe\i!ted indJ each automatic ontainment iso ation valve is within lim ts. \- 41., I Vtr1W Perfe,. le age rate teiting fer conta t nmecitu:r

  • 11 II hM 2

._,.i,..... 14 "...,_ s .... p,1 bf pcrt'ormanc.c. of Q, . -c. ro.-k -t'-51 (j) FREQUENCY rior to entering HOOE 4 from HOOE S 1 f not performed within the previous 92 days In accordance with the Inservice Testing Pro,am17f 192 ,lS 184 days CD © Verify each automatic containment isolatjon j{"(stlmonths v1lv1 that is not locked, sealed, or othirwis1 secured in position, actuates to the isolation position on an actual or simulated actuation signal. (continued) 3.6-13 Rev 1, 04/07/95 K X'< I

  • *
  • Containment Isolation BASES (continued)
  • LCD r....p ..... f" ec:.&:,+ o-c d,.. c. s: ... ...Q td APPLICABILITY CEOG STS """'/ 'f:ltc.. vi.,..e r,,.,.,..

@ The normally closed isola

  • es are cons1 ered OPERABLE when manual valves are automatic valves are de-activated and secured 1n their closed position, blind flangts*are in place, and closed systems are intact. These passive isolation valves or devices are those listed in Reference 1 as Type C Th1s LCO provides assurance that the conta1n.ment isolation valves and purge valves will perfom their designed safety l,G\ fu11ctions to 11ini111z1 the loss of to coolant inventory

'IJ.J and establish th1 containment oundary during accidents. In MODES 1 1 2, 3, and 4, a OBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment isolation valves are not required to bt OPERABLE in MODE 5. The requirements for containment isolation valves during MOOE 6 art addressed in LCO 3.9.3, *containment Penetrations.* B 3.6-22 3) -Lt (continued) Rev 1, 04/07/95

  • BASES (continued)

ACTIONS <>.-.J. I '2. ; Ac.J,.. c..:.rroo-. s....,erl-, Containment Isolation Valves\<Aimosphefic and B .6.3 (9 fc-c..t s p IM"ff" u.l..it.S *

  • 1-z._ *, ... ,.:.,. s"'"rr'i

,-o.'"' be. """" o.bte. c..l.:..S.IZ.. i"' ........ , A second Note has been added to provide clarification that, for this LCO, sepc ite Condition entry is allowed for each penetration flow path. This 1s acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable containment isolation valve. Complying with the Required Actions may allow for continued operation, and.subsequent inoperable containment isolation valves are governed by subsequent., Condition entry and application of associated Required Actions

  • The ACTIONS are further modified by a third Note, which ensures that appropriate remedial actions are taken, if necessary, if the affected systems are rendered inoperable by an inoperable containment isolation valve. A fourth Note has been added that requires entry into the applicable Conditions and Required Actions of LCO 3.6.1 when leaka91 results in exceeding the overall containment leakage l111it. ___ .... @ ... .. A.1 and In the event one containment isolation valve in one or more penetration flow paths is inoperable xce t for purge alvevl or °"'",,. **

an 1 n ass ea a e fta1i within A. 'fO\ ..,,.\.,.._ l111;t" the affected penetration flow path must be isolated. -.,J-i...-,u...-,.-".'5 no'° \-.k<cl The method of isolation must include the use of at least one \/ "---......... isolation barrier that cannot be adversely affected by a I\ single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic containment isolation valve, a closed manual valve, a blind (continued) . CEOG STS B 3.6-23 Rev 1, 04/07/95 ,J}-( * .-----------------------

* *

  • BASES ACTIONS Containment Isolation ICAtmfspheric al){j B 3.6.3 A.l and A.2 (continued) means. Allowing verification by administrative means is* considered acceptable, since access to these areas ts typically restricted.

Therefore, the probability of misalignment* of these devices, once they have been verified to bt in the proper position, ts small. , /UH With two containment isolation valves in one or mor C\ penetration flow paths inoperable 'fixcept for purge alve ... as 1 ea It age not *Ith iii limit§< the a ecte penetrat on ow path must be isolated s ..... u:., ,_,a.\A' within 1 hour. The method of isolation must include the use

s of at least one 1solat1on barrier that cannot be adversely 1 1 affected by a single active failure. Isolation barriers I that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. CEOG STS
  • The 1 hourmmpletion Time is consistent with the LCO 3.6.l n the event the affected penetration is isolated i accordance with Required Act1on B.1, the affected penetration must be verified to be isolated on a periodic basis per Required Action A.2, which remains in effect. This periodic verification 1s necessary to assure leak tightness of containment and that penetrations requiring following an accident are isolated.

The Completion T1me of once per 31 days for verifying each affected penetration flow path is isolated is appropriate the fact that the valves are operated ur.der administrative controls and the probability of their misalignment is low. Condition B is modified by a Note indicating this Condition is only applicable to penetration flow paths with two containment isolation valves. Condition A of this LCO address1s the condition of one containment valve inoperable in this type of penetration flow path. C.1 and C.2 With one or more penetration flow paths with one containment isolation valve inoperable, th* inoperable valve must be restored to OPERABLE status or the affected penetration flow (continued) B 3.6-25 Rev 1, 04/07/95 3*1-w

  • BASES ACTIONS *
  • CEOG STS
  • Containment Isolation Therefo , the 1 eakage must 1 restored to w1 th*
  • limit within hours. Restorat1o can be accompli.sh by isolating t penetrat1on(s) tha caused the limit o be exceeded b use of one closed d de-activated aut tic c* valve, cl ed manual valve, or blind flange. Whe a penetrat n is isolated, the eakaga rate for th isolated penetra on is assumed to b the actual pathw&Y. leakage throug the isolation devi
  • If two isolatio devices are *used. isolate the penet tion, the leakage ate is assumed to the lesser actual thway leakage of e two devices. The hour.Completion T is reasonable c sidering the ti required to resto the leakage by 1 lat1ng the p etration(s) and th relitive importan of secondaJ'y
  • ntainraent bypass*l kige to the over& containment unction. In the event one or more co ainment purge* valves in one or more penetration flow path are not within the purge valve leakage limits, purge val leakage must be restored to within limits, or the af cttd penetration must be isolat
  • The 11111thod of isolation st be by the use of at least o isolation barrier that annot be adversely affected by single active failure Isolation barriers that meet th s criterion are a fClo **and de-activated automatic val e I li\l with resilient seal , a c manu va ve w res ent l:.-' fiii!i, or a blind angef':' A purge va v1 with res11 ent seals ut11 ized to satisfy Required Act1oR.-E'. l must ave demonstrated ta et the leakage requirements of S The specified C pletion Time is reasonable, cons ering that one conta n1111nt purge valve remains closed that a gross br11ch f containment does not exist. In accorda 1* with Required Action E.2,. this netration flow path ust b1 verified to be isolated on periodic basis. t e periodic verification is necess y to ensure that co ainment penetrations required to isolated follow g an accident, which are no longer, capable of being auto ically isolated, w111 be in th* is lation position shou an event occur. This Required Ac ion does not req re any testing or valve manipulat1
n. Rather, it in lves verification, through a syste walkdown, that those (continued)

B 3.6-27 Rev 1, 04/07/95 j -x

  • *
  • BASES ACTlONS ...__ SURVEILLANCE REQUIREMENTS CEOG STS Containment Isolation Valves UCM:mos@ic "4

.6. E".1. E.2. and E.3 (c ntinued) isolation devices o side containment capable of being mispositioned are the correct position. For the isolation devices nside containment, the time period specified as *pri r to entering HOOE 4 MOOE 5 if n t performed withi the previous 92 days* is based on engineering jud ment and is reasonable in of th* inaccessib ity of the isolation devices and oth administrativ controls that will ensure* that isola on device mi£a1 n11Gnt 1s an unlikely possibility. For the co ainment purge valve with resilient s 1 that is isolated accordance with Required Action E.l, SR must be p rformed at least once eve,*y 'l92f"days This -......... i© assures at degradation of the resilient seal is detected and con rms that the leakage rate of the con ainment purge valve es not increase during the time the is isola d. Th* normal Frequency for SR 3.6 .. days, is bas on an NRC Generic* Issue 20 {Ref. 3). Sin more reliance is placed on a single alve while in thi Condition, it is prudent to perfor111 he SR more often

  • Th refore, a Frequency of once per ays wu chosen and IU) h s been shown to b* acceptable based operating . eriance. . I If the Required Actions and associated Completion Times are not met, the plant must be brought to a MOOE which the LCO does not apply. To achieve this status, the plant must be brought to at least MOOE 3 within 6 hours and to HOOE 5 within 36 hours. The allowed Completion Times are reasonable, basld on operating experience, to reach the required plant conditions frOll full power conditions in an ordtrly manner and without challenging plant systems
  • Each 42] inch containmen purg1 valve is re ired to be veri ied sealed closed a 31 day intervals.

This Su eillanc1 is designe to ensure that a oss breach of ainment is not cau d by an inadverte or spurious * {continued) B 3.6-28 Rev 1, 04/07/95

  • BASES SURVEILLANCE REQUIREMENTS
    • the. \b.I V<.S ac.M.1 I'/ C.t65-tcl (SR 3.61.3.1 llo..lvc.

Is \n cwlU<Ltitin "'t>.,. mc.nt NO* C\o to -Tr.-. fuc .. .li +. Y uinsi.I * * \ CEOG STS Conta;nment Isolation Valves ICAtli\6'spheriqand Dual)}-@ B 3.6.3 SR administrative means is considered acceptable, since access to these areas 1s typically restricted during HODES 1, z, and 3 for ALARA reasons. Therefore, the probab;11ty of misalignment of these containment fsolatfon valves, once they have been verified to be 1n their proper position, is small. SR 3.6.3.it \(D Verifying that the 1solat1on time of each fRDwer operi\ed and 1 I /Zf\ automatic;..containment isolation valve is within limits is CV required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analysis * .9f,'The l isolation time and Frequency of this SR are fn accordance with the Inservice Testing Program pr )2 da;f"S\.}'-' SR f ....... \"Z.;,.cJ... .. , ... f' __ ,. .. fl'1'11 For with resilient seals, . additional leakag1 rat1 the test requirements of 10 CFR SO, Appendix J (Ref. '1}J11s required to ensure \ OPERASILH\'. Operating experience has demonstrated that '{ this typ1 of seal has th1 potential to degrade fn a shorter I\ time period than do oth1r seal types. Based on this observation and th1 importanc1 of maintaining this penetration leak tight (du1 to th1 direct path between containment and th1 environment), a Frequency of 184 days was established as part of the NRC resolution of Generic Issue B-20, *containment Leakag1 Due to Seal Deterioration* (i) @ Additionally, thi SR must be performed within 92 day after opening the valv

  • The 92 day Frequency was chosen recognizing tha cycling the valve could introduce additional sea degradation (beyond that occurring o a valve that ha not been opened).

decreasin the interval (fr 184 days) 1s a prudent measure aft r a valve has been op Id. (continued) B 3.6-31 Rev 1, 04/07/95 .

  • *
  • ATTAC1ŽENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Chanu Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG . MARKUPS. " The first five justifications were used generically throughout the markup of the NUREG. Not all generic justµications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has been provided.
2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and* do not involve technical changes or changes of intent. 3 . The requirement/statement has been deleted since it is not applicable to this facility. The following requirements have been renumbered, where applicable, to reflect this deletion.

  • 4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
5. This change reflects the cilrrent licensing basis/technical specification.
  • 6. The.Palisades purge and exhaust system consists of a supply through a 12 inch air room supply path and two 8 inch exhaust paths. Previously the Palisades Nuclear Plant design contained "large" purge valves but they have been removed from the plant design. The NUREG.:1432 portions dealing with the [42] inch purge valves is not included in the proposed ITS. Each path of the existing purge exhaust and air room supply contains two isolation valves with resilient seals. The priuiary difference between the use of the purge and exhaust path Palisades and NUREG-1432 portions dealing with the "mini-purge" valves is that at Palisades the purge exhaust valves and air room supply valves are locked closed in MODES 1-4 and are not allowed to be opened. Therefore, for Palisades the valves are inoperable if the valves are open, or closed but not locked, since they have not been qualified to be able to close in the event of a Design Basis Accident.

These \1alves ae aa'1erge a leak test every 184 ae:ys. (aifferesee.

R.Al 13 . . . n Palisades Nuclear Plant Page 1of5 01/20/98

  • 3.6.3 JFD 6* INSERT A *
  • Verification that t4e containment purge exhaust and air room supply valves are closed is accomplished by performing a leakage rate test every 6 months. NUREG-1432 also specifies that a leakage rate test be performed within 92 days after opening the valves. That requirement has not been adopted in the ITS since the containment purge exhaust and air room supply valves are not allowed to be opened in MODES 1 -4. INSERT B --ISTS SR 3.6.3.1 requires a verification that each [42] inch purge valve is sealed closed' except for one purge valve in a penetration flow path while in Condition E of the LCO. That SR was not adopted in the ITS since the Palisades plant design does not include 42" valves. Rather, ISTS SR 3.6.3.2 was revised to agree with LCO 3.6.3 (CTS 3.6.5) which requires the purge exhaust valves and air room supply valves to be closed in Modes 1, 2, 3, and 4. If one of these valves were open, the valve would be declared inoperable and the Required Actions of Condition D taken (See JFD 12). Proposed ITS Condition D, unlike ISTS Condition E, does not allow unrestricted operation in Modes 1, 2, 3, or 4. Thus, the exception contained in ISTS SR 3.6.3.1 related to Condition E does not apply; Since the purge valves are not used in MODES 1, 2, 3, or 4 for pressure control, ALARA, surveillance testing, or air quality considerations, that aspect of ISTS SR 3.6.3.2 has been deleted from the SR .
  • *
  • ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Change Discussion
10. CTS 3.6.5 requires "The containment purge exhaust and air room supply isolation valves shall be locked closed whenever the plant is above COLD SHUTDOWN." This portion of the CTS is addressed by the proposed ITS LCO 3.6.3 for the containment
11.
  • isolation valves being OPERABLE and the ITS MODES 1-4 are substituted for the "above COLD SHUTDOWN." If one of these valves were not locked shut the CTS requires "With one containment purge exhaust or air room supply isolation valve not locked closed, lock the valve closed within 1 hour." This becomes ACTION D in the proposed ITS 3.6.3. This r:equirement is included in the proposed ITS because the purge exhaust and air room supply isolation valves are have not been qualified to close following a LOCA. The proposed Conditions A and B have been modified to reflect this change with wording which states "(except for purge exhaust valve or air room
  • supply valve not locked closed ef le!tlEage net 'Yvithin NUREG-1432..affcad:y . include.(&exclusion for the purge valve leakage limits to instruct the user that another condition should be entered and so similar wording is added to address the case where a purge exhaust or air room supply valve is not locked closed. These changes represent a plant specific to the Palisades Plant CTS and _ analysis.

uili.t a.re. f)S+ O.tfl1CAhl< +o fo l/{tutS 1i\ Conci . ._W A o.,J B la<c:." dt. h-tul. The Palisades Nuclear Plant design includes actuating the containment isolation valves by a Containment Pressure (CHP) signal, or a Containment High Radiation (CHR) signal, or both. Mosfcontainment isolation valves receive both signals, however there are some .valves which only receive one signal. For example, the component cooling water is only closed on a CHP signal. Therefore, in the proposed ITS, .the Bases are revised to repeat the actual SR wording and then identify the applicable signals for clarity. This is a plant specific change to the Palisades Nuclear Plant design . Palisades Nuclear Plant Page 3 of 5 01/20/98 3 I --o.-r:__.,

  • *
  • ATTACH1\.1ENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Change Djscussjon Da LSTS W1f1ar-i [J
  • u rc..-k. fDr

---" . <?ihM+ a.nd " 1' f'f:Dl'n PvfPlf \.CJi..tJ I nd1c.a.4d

the. a.t{c:. Ll2JllC. 'I 'nRMu rl.J. (} 12. CTS 3.6.5 requires that the contamment purge exliaust and all' llbd 13. valves shall be locked closed whenever the plant is above COLD SHUTDOWN.

This"' t C.

  • requirement is an implicit requirement in the proposed ITS 3.6.3 since it states* "Each containment isolation valve shall be OPERABLE" and the proposed Bases state that the purge exhaust and air room supply valves must be locked closed. Therefore, an additional Action must also be provided in the proposed ITS if this requirement is not met. Proposed ITS Action D addresses the situation where "One purge exhaust or air room supply valve not locked closed" and requires that within 1 hour, the affected valve must be locked closed. This Action is added to the proposed ITS to reflect a plant specific change based on the Palisades Nuclear Plant CTS. NUREG-1432 Condition Chas a "bracketed" 4 hours as a Completion Time for an inoperable containment isolation valve in a penetration flow path with only one
  • containment isolation valve and a closed system. TSTF-30, Rev. 2, revises this 4 hours to 72 hours. A Completion Time of 72 hours is appropriate since it involves a closed system which minimizes the potential of a leakage pathway. Since the Palisades Nuclear Plant CTS did not provide an explicit Completion Time for this condition, the 72 hours is included in the proposed ITS. RI1rc4 Ac.tfOl\S C. ( Q.f\.& C.. 'Z. TSTF-30, Rev. 2, also included-a change to the Bases for"s 3.6.2.2 which stated: "The closed system must meet the requirements o SRP 6.2 .. >< Such a Bases statement is not consistent with the current licensing asis an was not incorporate SRP 6.2.4 does not provide "requirements," but rather provides review guidance for acceptability of the design for containment isolation capabili!}'.

The containment penetrations which utilize closed systems have previously determined to be . acceptable. The acceptability may have been based on compliance with SRP 6.2.4 or may have been justified on some other basis. Whatever the basis, the system design has been determined acceptable for use as a containment isolation feature, and it is inappropriate to now prevent such use unless it complies with SRP 6.2.4. Therefore, this change is consistent with TSTF-30, Rev. 2, except where the generic change is inconsistent with the current licensing basis . Palisades Nuclear Plant Page 4 of S 01/20/98 SI -a.d

  • * * *CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT

________ _ NBC REQUEST: 3.6.3-30 CTS 4.5.3.b ITS SR 3.6.3.6 and SR 3.6.3.7 The CTS markup of CTS 4.5.3.b shows that this CTS SR is ITS SR 3.6.3.6 and SR 3.6.3.7. ITS 3.6.3 does not contain an SR numbered SR 3.6.3. 7.

  • Comment: Revise the CTS/ITS markup accordingly.

Consumers Energy Response: The CTS markup has been revised to delete the cross reference to ITS SR 3.6.3. 7 since that SR does not appear in the ITS. Affected Submittal Pages: Att 3, CTS, page 4-21 (ITS 3.6.3, page 4.of 5) 32

  • 4.5 4.5.Z
  • 4.5.3 Cl}

3-"*3-3() \&' 'ti:!:} (1) *(continued) 1l.. ' (b) A air 1 k penetration test shall b1 performe it six-month 1 ervals. During th1 period between t 1 six-month 1sts when CONTAINMENT 1NTEGRIT1 is re u1rtd, a reduced p ssur1 test for th1 door seals or 1 f 1 air lock pen ration test shall b1 performed with 72 hours , after e ther each air lock door opening or t first of a series of openings

  • 4-21 *
  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQVEST: 3.6.3-31 CTS 4.5.3.b ITS SR 3.6.3.6 and Associated Bases CTS 4.5.3.b verifies that each isolation actuates to its correct or required position upon receipt of an actuation signal. ITS SR 3.6.3.6 verifies the same function, however, it exempts those automatic CIVs that are locked, sealed or otherwise secured in position.

Thus the ITS does not encompass as many valves as the CTS. Comment: Revise the CTS markup to correctly reflect ITS SR 3.6.3.6 and provide a discussion and justification for this Less Restrictive (L) change. Consumers Enerqv Resoonse: A new DOC (DOC L.6) and NSHC (NSHC L.6) have been provided to justify the deletion of actuation testing for CIVs that are locked, sealed or otherwise secured in position. Affected Submittal Pages: Att 3, CTS page 4-21 (ITS 3.6.3, page 4 of 5) Att 3, DOC 3.6.3, page 10 of 10 Att 4, NSHC 3.6.3, page 2 of 2 33

  • 4.5 4.5.2 14.S.3 (!)

3.b,3-3l> 

.. ; C&n.

... -7 s;-\ .. b:d (::'/(i' * ....__ __ Si: '3.t..l.Ll -(continued) S

c. (.+; o"" S ' (b) A full air 1 k penetration test shall be p1rforme at s1x*month 1 1rvals. Dur1ng tht period between t 1 six-month 1sts when CONTAIMMENT INTEGRITY is re u1red, a reduced p ssurt test for tht door seals or a f 1 air lock pen ration test shall bt performed w1th 72 hours ' after e th1r each air lock door opening or t first of a series of openings. '3.(,.'3.4(1

.. sR e ,,.: .. 1 t"'Q-MoOE 5' ot .... -e.J. -tt.a. rreo1:o"S "\?.. ..i ... ,s

  • 4-21 *
  • **
  • LA.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES TS Table . -tern . con ms a requrremen e contamme purge and entilation isolation Ives be determined closed at least once every months "by erforming a leak r e test between the The overall requ* ment is that the alves be demons ated to be closed by performance of a leak test The method of ccomplishing s testing can be addressed by plant procedures hich are under the ontrol of the 1censee. Therefore, the method of testing is not eluded in the roposed IT and will be by plant procedures.

C ges to these procedures are made i accordance with the plant change control proces . This change is G-32. LA.5 TS 4.5 .3b states " LE by erifying that on ea containment isolation right channel left channel test signal, pplicable isolatio valves actuate to their required positi during COLD f< Pr I HUTDOWN or t least once per refueling cycle." Th requirement-to test the valves 3. "during COLD HUTDOWN" is not included in the oposed ITS. The proposed ITS ill test the v ves every 18 months which is the equ* alent to the "once per refueling ycle" in the roposed ITS. The requirement totes the valves during "COLD SHUTDO " will be addressed by plant proced es. Changes to plant are made* accordance with the plant procedure ge process. This change 1s with NUREG-1432. LESS RESTRICTIVE CHANGES (L) L.1 CTS Table 4.2.2 item 13.a specifies that the Containment Purge and Ventilation Isolation Valves shall be determined closed "at least once per 24 hours." The proposed. ITS SR 3.6.3.1 changes the frequency for verifying the 8 inch containment purge valve and 12 inch air room supply valves are closed from 24 hours to 31 days. This change is reasonable because these valves are not allowed to be open in proposed MODES 1-4. These valves are required to be locked closed, either electrically or by some other means to de-activate the valve. Therefore, the likelihood of the valve being opened is very remote. The 31 day frequency is consistent with the surveillance requirements for other safety related systems which require valve position verification. This change is consistent with NUREG-1432. L 2.. "I:tJ.SUL 1" RA\ 3i.{g.!>*ll /{Al ?i.(,.;-9 L. L/ /?ft I j.(t,.?,-/')_ Palisades Nuclear Plant Page 10of10 01/20/98 L*5 Sc... ":Qt.JS t11, 1 !?Al l... h "'!NS'&1 5.41. 3-3 / 33-h

  • *
  • 3.6.3 DOC L.6 CTS 4.5.3b requires that each containment isolation valve be demonstrated Operable by verifying it actuates to its required position.

In the ITS, an equivalent test is required by SR 3.6.3.6. However, ITS SR 3.6.3.6 does not require containment isolation valves under administrative controls that are locked, sealed, or otherwise secured in position to be tested. This is because these valves are already in the position necessary to perform the containment isolation function. Thus, there is no need to verify these valve can reposition on an actual or simulated actuation signal. The allowance not to test containment isolation valves that are locked, sealed, or otherwise secured in position is a relaxation from the requirements of the CTS. This change is consistent with NUREG-1432 . 33-c...

  • *
  • 2. 3. L.i L.-, L., L.I L.5 L.17 ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue to verify that the purge exhaust and air room supply yalves are still closed. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? The proposed change increases the surveillance interval from 24 hours to 31 days. The margin of safety afforded by the purge exhaust and air room supply valves are that they are closed and thereby isolating a direct path from the containment atmosphere. Since the purge exhaust and air room supply valves are required to be closed in MODES 1, 2, 3 and 4, the likelihood of these valves being open is very small. Therefore, this change does not involve a significant reduction in a margin of safety. Se.c. INSfA T £\Ai -1 I s Lt. I:MS C# I 3 Se.c.. l 3 .b-3 .. XNSu.1 RI+ I j.{,.3-1'1/IS/Jt./17 S LJ.Su1 f<.AI 3:b.3-51 Palisades Nuclear Plant Page 2 of 2 01/20/98 33-d

  • *
  • 3.6.3 NSHC L.6 CTS 4.5.3b requires that each containment isolation valve be demonstrated Operable by verifying it actuates to its required position.

In the ITS, an equivalent test is required by SR 3.6.3.6. However, ITS. SR 3.6.3.6 does not require containment isolation valves under administrative controls that are locked, sealed, or otherwise secured in position to be tested. This is because these valves are already in the position necessary to perform the containment isolation function. Thus, there is no need to verify these valve can reposition on an actual or simulated actuation signal. The allowance not to test containment isolation valves that are locked, sealed, or otherwise secured in position is a relaxation from the requirements of the CTS. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change deletes the requirement to perform an actuation test of containment isolation valves that are locked, sealed, or otherwise secured in position since these valves are already in the position necessary to perform their containment isolation function. The proposed change does not involve a change to any accident initiators or precursors. Therefore, the proposed change does not involve a significant increase in the probability of an accident. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change does not alter the assumed capability of the containment isolation function. Thus, the consequence of an accident remain unchanged. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change still ensures that an OPERABLE containment isolation barrier is available whenever the plant is in a condition that requires containment isolation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated .

  • 3. *
  • Does this change involve a significant reduction in a margin of safety? The margin of safety is a function of the overall containment leakage. The proposed change deletes the requirement to perform an actuation test of containment isolation valves that are locked, sealed, or otherwise secured in position.

This change does not relax the requirement to maintain containment integrity, but recognizes that valves that are locked, sealed, or otherwise secured in position are already capable of performing their isolation function. As such, this change does not result in any activity that would result in an increase in the amount of radioactive material released to the environment. Therefore, the proposed change does not involve a significant reduction in a margin of safety . 33-f

  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6 Containment Cooling Systems 3.6.6-1 DOC A.1 DOC M.1
  • DOCM.7 DOC LA.1 DOCL1 JFD2 JFD4 JFD 6 JFD 8 JFD 11 CTS 3.4 ITS 3.6.6 and Associated Bases ITS B 3.6.6 Base -BACKGROUND and JFD 6 state that the original design of the system . considered the containment spray pumps and the containment air coolers to be redundant systems. Subsequent analysis required that one containment spray pump be used when the three containment air coolers are being relied on for heat removal. Thus the ITS is structured accordingly.

It is unclear from the information provided as to whether the staff reviewed and approved the subsequent analysis and whether the CTS reflects this analysis. If the staff has reviewed and approved the subsequent analysis and the CTS reflects this analysis, then the .proposed ITS ACTIONS seem to allow more components to be inoperable than is currently allowed. Insufficient information is provided by the DOCs and JFDs to evaluate this change. If, however, the staff has not reviewed a*nd approved the subsequent analysis and the CTS reflects the original design and analysis, then the proposed changes to the ITS ACTIONS constitute a beyond scope of review item for this conversion. The responses to Comment Numbers 3.6.6-2, 3.6.6-3, 3.6.6-5, 3.6.6-6, 3.6.6-14, 3.6.6-15 and 3.6.6-22 will depend on which situation applies.

  • Comment: Revise the DOCs and JFDs to 9larify which situation applies and provide sufficient information to evaluate the changes. See Comment Numbers 3.6.6-2, 3.6.6-3, 3.6.6-5, 3.6.6.-6, 3.6.6-14, 3.6.6-15, and 3.6.6-22.

Consumers Energy Re§Qonse: , . The re-analysis of the containment cooling systems was presented to the NRC staff in the request for TS amendment 104, by letters dated October 20, 1986 and November 21, 1986. The amendment was approved and issued by the NRC on March 24, 1987, and is accompanied by a Safety Evaluation Report. The CTS Bases issued with this amendment include discussion of the system response to postulated accidents. In addition, FSAR sections 6.2 and 6.3 contain descriptions of the systems design and operation during postulated emergency conditions. Response to specific comments on DOCs is presented in the responses to Comment Numbers 3.6.6-2, -3, -5, -6, -14, -15, and -22. '

  • Affected Submittal Pages: No page changes. 34
  • * * *CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT . NRC REQUEST: 3.6.6-2 DOC A.1 DOC M.1 DOCM.7 DOC LA.1 DOC L.1 JFD2 JFD4 JFD6 JFD 8 JFD 11 CTS 3.4 STS 3.6.6.A and Associated Bases STS 3.6.6B and Associated Bases STS 3.6. 7 and Associated Bases ITS 3.5.5 and Associated Bases ITS 3.6.6 and Associated .Bases ITS 8 3.6.6 Bases -BACKGROUND "Containment Spray System" states the following: "In addition, the Containment Spray System in conjunction with the use of trisodium phosphate (LCO 3.5.5 "Trisodium Phosphate")

serve to remove iodine which may be released following an accident." Based on this statement the ITS markup used the wrong specifications, STS 3.6.6 A "Containment Spray and Cooling Systems (Atmospheric and Dual) (credit taken for iodine removal by the Containment Spray System)" should have been used rather than STS 3.6.6.B which does not take credit for iodine removal by the Containment Spray System. The Trisodium Phosphate system and associated LCO take the place of the Spray Additive System description in STS B 3.6.6.A Bases -BACKGROUND and STS 3.6. 7. See Comment Numbers specified in Comment Number 3.6.6.-1, as well as 3.6.6-16, 3.6.6-18 and 3.6.6-19. Comment: Revise the ITS markup to use the STS 3.6.6.A rather than STS 3.6.6.B or provide justification to show that STS 3.6.6.B is the appropriate specification to use. See Comment Numbers specified in Comment Number 3.6.6-1, as well as 3.6.6-16, 3.6.6-18 and 3.6.6-19. Consumers Energy Response: Amendment Nos. 158 and 165 to the Palisades TS, dated September 17, 1993 and May 19, 1995 respectively, removed the requirement for hydrazine and sodium hydroxide addition capability from the Technical Specifications. As described in the Safety Evaluation Report issued with amendment number 158, the hydrazine injection system performance did not have a substantive effect on the iodine removal function and was not needed. Hence it was deleted . (continued) 35

  • * * . CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS . RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-2 Consumers Energy Response: (continued)

Amendment 165 and the associated SER approved the removal of the sodium hydroxide addition function with the addition of hydrated Tri-Sodium Phosphate (TSP) in baskets to the lower level of the containment. This change assured that the post-accident containment sump solution would be maintained in a slightly caustic condition (pH between 7 and 8). Maintaining the sump solution slightly caustic assists in the retention of iodine in the solution. The performance of post-accident iodine control (retention in solution) is now dependent on only the TSP and any water (postulated break flow, ECCS, containment spray, or other) that accumulates in the containment. TSP requirements are addressed in proposed specification 3.5.5. In addition, the Bases for NUREG-1432 specification 3.6.6A, hydrazine and sodium hydroxide addition are described as a function of the containment spray and cooling system. This does not reflect the Palisades design. Therefore, the ISTS markup was base.d on the 3.6.68 specification of NUREG-1432 . The ultimate factor in determining which specification to use as the model for the Palisades ITS was that which most closely reflects the plant design and licensing basis. It was determined that the 3.6.68 version did so. This minimized the amount of deviation from an existing ISTS, consistent with guidance from the NRC staff to minimize deviation to the extent possible, consistent with the facility design. Affected Submittal Pages: No page changes . 36

  • * * .CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-3 DOC A.1 DOC M.1 DOC M.7 DOC LA.1 DOC L.1 JFD 2 JFD4 JFD6 JFD 8 JFD 11 CTS 3.4 ITS 3.6.6 and Associated Bases CTS 3.4.1 provides an LCO that is based upon component OPERABILITY.

CTS 3.4.2, 3.4.3, 3.4.4;, and 3.4.5 lists the permissible component inoperabilities which limit the number of concurrent components inoperable that are associated with diesel generator 1-2 and 1-1 divisions. ITS 3.6.6 provides an LCO that is based upon maintaining the OPERABILITY of two trains. The CTS markup indicates that DOC A.1 is the justification used for converting from a component based LCO to a train based LCO. This is not acceptable. A new specific DOC which explains how the two trains of Containment Cooling System are derived from the CTS 3.4.1 requirements, should be provided. Also, in order to understand the system design and the proposed ITS 3.6.6 ACTIONS, simplified schematics of the Containment Cooling Systems to describe what specific components comprise each train should be provided. An in-depth technical explanation of how the non-safety-related containment air cooler VHX-4 associated with containment spray pumps 548 and 54C is determined to be equivalent to a 100% cooling capacity cooling train also needs to be provided. In addition, the following information will be needed to evaluate the changes and the conversion from CTS to ITS: What are the cooling capacities of these two trains without any component being inoperable? How is the cooling capacity of each train changed, as one or more components become inoperable? Tabulate these resulting reduced train cooling capacities. Illustrate what favorable combinations of remaining OPERABLE train components can still yield the redundant capacity required to comply with this ITS ACTION A Condition. Explain how the "alternate" trains configurations are created. Identify the cross-over valves between trains that wili permit inoperable components to be removed from service and explain how the inoperable components are replaced with redundant OPERABLE components. Are all these re-alignment of train configurations made from the control room. See Comment Numbers specified in Comment Number 3.6.6-1. Comment: Provide the requested information. Revise the CTS/ITS markup as specified above and provide the appropriate discussions and justifications for the above changes. See Comment Numbers specified in Comment Number 3.6.6-1 . (continued) 37

  • * ---------*CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUE$T: 3.6.6-3 (continued)

Consumers Energy Response: DOC LA.1, in combination with A.1, provide a description of how the details of the CTS requirements relocated to the associated Bases, and the requirements are reformatted consistent with NUREG-1432. A summary description of the way the containment cooling systems are arranged is provided in the Palisades FSAR, at sections:

  • 6.2, "Containment Spray System,"
  • 6.3, "Containment Air Coolers,"
  • 9.1, "Service Water System," .and
  • 9.3, "Component Cooling System." The CTS component lists are grouped by associated Diesel Generator (DG), i.e., by train. That
  • is, components associated with DG 1-1 include one train of containment cooling consisting of three containment air coolers that are provided cooling water by the service wat.er system, and one containment spray pump whose flow is cooled by the component cooling water, and subsequently removed by the service water system . Similarly, components associated with DG 1-2 include two containment spray pumps whose flow is cooled by component cooling water, and subsequently removed by the service water system. The number of pumps assigned to each.train (i.e., associated with one or the other OG) reflects the relative magnitude of the heat removal load via the service water and component cooling water systems. However these cooling systems are designed so that they can be connected and flow provided by opposite train components, given power and component availability.

Extensive detail regarding normal alignment and train cross-connection capability is provided in the above referenced FSAR sections. FSAR Tables referenced in the above listed FSAR sections describe flow rates, heat transfer capabilities, and associated parameters for components listed in the CTS and credited in the licensing basis. These sections, and the associated tables, provide the licensing basis for the plant design. Heat Exchanger VHX-4 does not provide any post-accident cooling function, and its heat removing service water flow is automatically isolated. Fan V-4A continues to operate after a postulated accident to provide mixing of the containment atmosphere only. Therefore the requirements for the heat exchange were removed from the TS in amendment 104, issued March 24, 1987. '

  • Affected Submittal pages: No page changes. 38
  • . CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: . 3.6.6-4 DOC A.5 CTS 4.6.3 ITS SR 3.6.6.5 CTS 4.6.3 specifies that the containment spray pumps be tested at intervals "not to exceed three months." The corresponding ITS SR is ITS SR 3.6.6.5 which has a frequency of "In accordance with the lnservice Testing (IST) Program." This change is described in DOC A.5. Insufficient information is provided in DOC A.5 to determine if the change is an Administrative change. If the current IST Program specifies a frequency of 3 months for pump testing, then the change can be considered as Administrative (change in terminology). However, if the current IST Programs specifies a frequency other than 3 months, then the change would be a Less Restrictive (LA) change (movement of detail) plus a More or Less Restrictive (L) change depending on whether the current IST frequency is greater than or less than 3 months respectively.

Comment: Provide additional discussion and justification to show that this change in frequency nomenclature is an Administrative change.

  • Consumers Energy Response:
  • DOC A.5 has been revised to indicate that the current IST program frequency for pump testing is once every 3 months. Affected Submittal Pages: Att 3, DOC 3.6.6, page 2 of 8 39
  • *
  • A.4 A.5 RA-1:

-'t fe> A.'S'. ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS CTS 3 .4.4 specifies that valves, interlocks and piping that are directly associated with the "above" (CTS 3.4.1) components shall meet the same requirements as listed for that component CTS :3.4.5 specifies that valves, interlocks and piping which is associated with the containment cooling system and not covered by Specification 3 .4.4 may be inoperable for no more than hours if it is required to function during an accident. These requirements are addressed by the definition of OPERABILITY which requires that all associated equipment be OPERABLE. In the proposed ITS all equipment in a particular train which is required to function during an accident must be OPERABLE and all equipment in the train will have the same Completion Time. This is an administrative change since the requirement remains that all equipment in a train of containment cooling must be OPERABLE. This change is consistent with NUREG-1432. CTS 4.6.3 specifies that the containment spray pumPrs be tested at intervals "not to . exceed three months." In the proposed ITS 3.6.6.5 this is changed to be "In accordance with the Inservice Test Program." The lnservice Test Program will specify the frequency for pump testing. This is considered to be an administrative change since the three month frequency will now be contained in the lnservice Test Program. This change is consistent with NUREG-1432. A.6 CTS 4.6.2a specifies in part that for the Containment Spray System Test: "Operation of the system is initiated by tripping* the normal actuation instrumentation. " The proposed ITS add the following phrase "by an actual or simulated actuation signal." The allowance to use an actual or simulated actuation signal is based on the fact that the channel being tested cannot differentiate between an actual or simulated signal and either initiation can demonstrate system OPERABILITY. This is considered to be *an administrative change since the system testing requirements have not changed. This change is consistent withNUREG-1432 . Palisades Nuclear Plant Page 2 of 8 01/20/98

  • *
  • DOC 3.6.6, A.5 [add the bold italici.Zed words to the A.5 DOC as shown] A.5 CTS 4.6.3 specifies that the containment spray pumps be tested at intervals "not to exceed three months." In the proposed ITS SR 3.6.6.5 this is changed to be "In accordance with the Inservice Test Program." The Inservice Test Program will specify the frequency for pump testing*.

The lnservice Test Program requires testing these pumps once every 3 months. This is considered to be an administrative change since the three month frequency will now be contained in the Inservice Test Program. This change is consistent with NUREG-1432 . J9-b -* *

  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-5 DOC M.1 DOC L.2 JFD 8 CTS 3.4.1 CTS 3.4.2 CTS 3.4.3 ITS 3.6.6 APPLICABILITY and Associated Bases ITS 3.6.6 ACTION B and Associated Bases CTS 3.4.1 specifies that "The reactor shall not be made critical.

.. unless the following conditions are.met." CTS 3.4.2 and 3.4.3 require a shutdown to COLD SHUTDOWN (MODE 5) if system/co.mponent inoperabilities cannot be restored to OPERABLE status. ITS 3.6.6 . APPLICABILITY specifies the applicability of the Containment Cooling Systems as MODES 1, 2 and 3 with shutdown requirement of MODE 4 in ITS 3.6.6 ACTION B. The staffs position is that the APPLICABILITY should be in MODES 1, 2, 3, and 4. The staff bases its position on a number of factors. ITS B 3.6.6 Bases -APPLICABLE SAFETY ANALYSES states that the Containment Cooling Systems limit the temperature and pressure that could be experienced following a OBA. Furthermore, it refers to ITS 3.6.4 "Containment Pressure" and ITS 3.6.5 "Containment Air Temperature" for a more detailed discussion on the intent of the design basis and the accident analysis and evaluation. The APPLICABILITY of these two specifications is MODES 1, 3, 3 and 4 because a OBA could in these MODES cause a release of radioactive material into containment. Therefore, the Containment Cooling Systems need to be consistent with the analyses, and thus would be needed in MODE 4. Another factor that was considered was that the CTS in CTS 3.4.2 and 3.4.3 require a shutdown to COLD SHUTDOWN not HOT SHUTDOWN if the inoperable system/component cannot be restored to OPERABLE status. The intent of the CTS from the staffs perspective is that the APPLICABILITY would be equivalent of ITS MODES 1, 2, 3, and 4. Finally, the actuation instrumentation for the Containment Cooling System -CHP -is applicable or required to be OPERABLE in MODES 1, 2, 3, and 4. There is no mention in the submittal of system interlocks, cutouts or other switches which would prevent the system from actuating/aligning in MODE 4 regardless of the shutdown cooling alignment (See Comment Number 3.6.6-6). Comment: Revise the CTS/ITS markup and provide the appropriate discussion and justification to show the APPLICABILITY as MODES 1, 2, 3, and 4. Consumers Energy Res.goose: As described in DOC M.1, the applicability of CTS 3.4.1 (which corresponds to the LCO) is provided as "the reactor shall not be made critical... unless all of the following conditions are met:" CTS 3.4.2 and 3.4.3 constitute the equivalent to ITS Condition A and B and specify that they only apply "During power operation .... " Based on literal compliance with these CTS statements, CTS 3.4.2 and 3.4.3 do not apply when the reactor is not in "power operation." (continued) 40 *. I

  • * ** . CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-5 Consumers Energ.,v Response: (continued)

If the condition occurs while in power operation, CTS 3.4.2 or 3.4.3 would be entered. In amendment 85 to the CTS, Section 3.0, Applicability was added which explicitly indicates that LCOs must be met when in the applicable mode or other specified condition. Therefore, CTS 3.4.2 and 3.4.3 would have to be met until the reactor is made subcritical, then the associated LCO, CTS 3.4.1, would no longer apply. In summary, the CTS do not specify any required actions below 2% power, and CTS 3.0.3 would apply. If operating at greater than 2% power, the Actions would no longer be required when the reactor is subcritical. This is a legacy of the historically inconsistent Technical Specifications that resulted in the need for, and conversion to, the ITS. Although CTS only require the containment cooling systems to be operable when the reactor is critical, and the actions to address inoperabilities only apply in "power operation," the Palisades design basis is more accurately reflected by the proposed ITS that will require _the system to be OPERABLE in MODES 1, 2, and 3. Therefore this is accurately described as a more restrictive change to the CTS. Additional discussion related to the facility design is provided in JFD 8 for ISTS 3.6.6, which notes that the bracketed MODE 4 Applicability in the ISTS does not apply and some Required Action changes are made because the design and licensing basis for Palisades is to align the plant for shutdown cooling in MODE 4. This prevents the containment spray pump flow path from being aligned to the containment spray headers as required to support the containment cooling system function. The Containment High Pressure (CHP) signal is required to be operable in MODES 1, 2, 3, and 4 because it also provides an actuation signal for the containment isolation function. The requirement for CHP OPERABILITY does not reflect any requirement to support the containment cooling system in MODES other than 1, 2, and 3. Affected Submittal Pages: No page changes . 41

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.6-6 DOC M.1 DOC L.2 JFD 8 CTS 3.4.1 ITS 3.6.6 ACTION Band Associated Bases In light of the discussion in Comment Number 3.6.6-5, ITS 3.6.6 ACTION B needs to be modified.

The issue discussed in JFD 8 with regards in the Shutdown Cooling System alignment in MODE 4 using the containment spray pumps needs to be resolved. One possibility is the adding of a note to the ITS APPLICABILITY which states "A containment spray pump flowpath may be considered OPERABLE to the SOC heat exchanger during alignment and operation for shutdown cooling, if capable of being manually realigned to the containment spray mode of operation." This note will allow ITS 3.6.6 to be applicable in MODE 4 and not interfere with shutdown cooling operations, provided the concern about interlocks in Comment Number 3.6.6-5 can be addressed. Comment: Revise the CTS markup, DOCs and JFDs, ITS markup and Bases as necessary to resolve this issue . Consumers Energy ReS.Qonse: As described in the response to Comment 3.6.6-5, no change is required. The proposed specifications accurately reflect the licensing and design bases. Affected Submittal Pages: No page changes . 42

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.6-7 DOC M.4 CTS 4.6.5.b ITS SR 3.6.6.2 and Associated Bases CTS 4.6.5.b specifies that each fan and valve required to function during accident conditions will be exercised at intervals not to exceed three months. The corresponding ITS SR for the fans *is ITS SR 3.6.6.2 which requires that the fan be operated for :::: 15 minutes, every 31 days. As currently written, the CTS does not require the fan to operate for a specific period of time, thus, the addition of the ":::: 15 minutes" is a More Restrictive change. The CTS does not show this change. Comment: Revise the CTS markup and provide the appropriate discussion and justification for this More Restrictive change. Consumers Energy Response:

A new DOC, M.9, has been prepared to describe the more restrictive change. Affected Submittal pages: Att 3, CTS 3.6.6, page 4-25 Att 3, DOC 3.6.6, page 5 of 8 " 43

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  • 14: 6 SAfUY INJECTION ANO CONJA[Mt1ENT SPRAY SYSTEMS Syrye1111nc1 Begy1rerntnts (cont1nu1d)

I 14.6.4/ yaJv/s (conti)u1d) !--@ Th safety tnjectton rectrc atton path shall bt v tf11d OPERABLE thin 7 d1ys prtor to H reactor st&rtup by v f'ytng valves . See J.5> CV-3027 and 3056 ar1 op yid thttr swttch1s HS 027A, HS-30278, HS*3056A, and HS-3056 art optn. d. Each C nt1tn .. nt Spray V v1 control all bt v1rtft1d to be OPE LE at 11ast once ch r1fu1ltng by CY, 1ng 1ach valve from tht control roOll whtl1 observing valve op atton at l1ast each 1 110nths. {(B.lgyConh1nwnt A1r Cs)gl 1na Sutn <i) .sR .. c..l. <,4/JO SR J. b.(.. I c..r f"eu,,.-1-cJ

A.

< .4 1.(..(., 3 c.r f'"cu,.-1-e.J ,._ Z:TI)@ < 110/) Sf J. '*'* '/ *d f'f!uAeJ ; ... rrr-) @ \ U IV I ?t_,t, ,zs> :W RA-C 3.r...t..- \0 ., I; .* ' ., 4*25 AMnd.ent No. 61, ;;, W. 174 J ( .l. October 31, 1996 / ...J y .,.} L{ ?c..er

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  • A_,,.-r: c..t,-10 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS M.8 The propose TS SR 3.6.6.7 req
  • es that each conta' ent spray pumps actual ors* ulated actuation sig every 18 months. e CTS does note require a eriodic verification f containµlent spray tarting, although th umps are verifie to start as part of th CHP Logic Channe unctional Test req ed by CTS .17.3. la. This pro des an explicit requ* ement to verify pu starting.

This ch ge is considered to ea more restrictive ge since this ver* cation is not ntained in the exist' g CTS. Thi chang is consistent with N G-1432. LESS RESTRICTIVE CHANGES -REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) LA. l In the CTS 3 .4.1-a-and b, the major components (i.e., Containment Air Coolers and Containment Spray Pumps) of the containment cooling trains are listed with an associated diesel generator and required to be OPERABLE. In CTS 3A.lc it also requires that "All heat exchangers, valves, piping, and interlocks associated with the above components and required to function during accident conditions are OPERABLE." In the proposed ITS the LCO will simply read: "Two containment cooling trains shall be OPERABLE." In the proposed ITS, the information regarding what will comprise a train of containment cooling is not included in the TS LCO but will be addressed in the Bases. Changes to the Bases are made under licensee control in accordance with the Bases Change Control Program which is addressed in Chapter 5 of the ITS. This change is consistent with NUREG-1432. '3*C,.f.,-\\ LA.2 CTS 4.6.2a specifies t for the Containmen pray System test "The st shall be performed with th isolation valves in the ray supply lines at the ntainment blocked closed." CTS .6.2c specifies that for e Containment Spray S em test: "The test will be cons* ered satisfactory if vis observations indicate components have operated tisfactorily." This lev of detail with respect t ow the test is perfo a and w t the testing acceptanc riteria are is more app priately addressed in nt test' g procedures. The re rrement to perform the ntainment spray test i ot anged but the testing thodology and accep e criteria will be not ocated in the proposed ITS but ill be addressed in pl rocedures. Changes t plant procedures are ma in accordance with th lant procedure change rocess. These changes are co "stent with NUREG-14 . Palisades Nuclear Plant Page S of 8 01/20/98 13-b

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  • Insert to 3.6,6, DOC M.9 Proposed SR 3.6.6.2 requires the containment cooler fans to be operated for 15 minutes every 31 days. CTS 4.6.5.b only requires that each fan be exercised, with no minimum operating time. The addition of a required duration to this surveillance requirement is a more restrictive change, consistent with NUREG-1432 .
  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUE$T: 3.6.6-8 . DOC M.6 JFD4 JFD 14 CTS 4.6.5.b STS SR 3.6.6.3 and Associated Bases ITS SR 3.6.6.4 and Associated Bases ITS SR 3.6.6.4 requires that the total service water flow rate is > 3935.1 gpm, to Containment Air Coolers (CACs) VHX-1, VHX-2, and VHX-3 when aligned for accident conditions.

The frequency for this SR is every 18 months. The justification used to justify the 18 months JFD 14 states that the flow rate can only be verified during shutdown conditions. This is insufficient justification for changing the corresponding STS SR 3.6.6.3 frequency from 31 days to 18 months. Based on the requirement in CTS 4.6.5.b to exercise/stroke each valve in the system every three months and the Bases statement in ITS B 3.6.6 Bases -BACKGROUND "Containment Air Recirculation and Cooling" insert that the CACs are used during normal operation that is there is service water flow through the coolers, the staff cannot justify a frequency of 18 months for this SR. In addition, the wording of the Bases for STS SR 3.6.6.3 would allow any flow rate and system alignment to be used as long as it can be shown that it would be the equivalent to the design flow rate assumed in the accident analyses with the system in the accident alignment. Thus the requirement in ITS SR 3.6.6.4 to have the system aligned for accident conditions becomes unnecessary. See Comment Numbers 3.6.6-9 and 3.6.6-20. Comment: Provide additional discussion and justification to support the 18 month frequency and the accident alignment requirement for ITS SR 3.6.6.4 based on plant specific design, operational constraints or current licensing basis, or revise the ITS markup to reflect the STS or current licensing Basis (CTS 4.6.5.b). See 3.6.6-9 and 3.6.6-20. Consumers Energy Response: The Palisades service water supply to the containment air coolers is arranged as shown in FSAR Figure 9-1, Sheet 1. As shown on this drawing, the air coolers inside the containment are supplied via a common header. No installed instrumentation is available to indicate the required flow is supplied to the individual coolers credited in the associated safety analyses, i.e. VHX-1, VHX-2, and VHX-3. Therefore flow measurements must be made using local instrumentation _that *can be aligned and calibrated during plant outages: As described in FSAR Section 6.3.2.2, the normal operating configuration for the system is with service water supplied to all four coolers, with two fans in operation associated with each. The service water flowpath is via, temperature control lines on the three safety related coolers (VHX-1, -2, and -3)', and through the high-capacity discharge line of the VHX-4 cooler. Upon actuation by a safety injection signal, the high-capficity discharge valves on the safety related coolers open, and the high-capacity inlet valve on VHX-4 closes. This directs the service water flow through the three safety related cool_ers. r . . (continued) 44

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-8 (continued)

Consumers Energy Response: In addition, the flow characteristics of the service water system are significantly different after an emergency actuation due to the realignment of portions that provide non-safety related functions during normal operations. The combination of a lack of available instrumentation, significant realignment of flow associated with the containment air coolers inside the containment, and realignment of flow from non-safety related components and functions, prevents the measurement of service water flow while in operation. Therefore ITS SR 3.6.6.4 is proposed with an 18 month frequency requirement that will facilitate its performance during refueling outages. As discussed in DOC M.7, there is no existing SR that requires the performance of a corresponding flow test. Based on this, the addition of SR 3.6.6.4 is a more restrictive change, proposed with the 18 month frequency that is consistent with the design and operation of the Palisades plant. A new J FD 16 is provided to describe the deviation from the ISTS because of the physical configuration of the flow system in the containment and lack of available flow instruments capable of measuring the flow rate through each containment air cooler individually, and the changes in service water flow that occur when it is called upon to provide post design basis accident function. Affected Submittal Pages: Att 6, JFD 3.6.6, page 5 of 5 ,, 45

  • *
  • Change ATTACHl\'IENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING SYSTEMS Discussion
14. NUREG-1432 requires a verification of the cooling water flow rate to the containment air coolers every 31 days. For the Palisades Nuclear Plant design, this flow rate can only be verified during a condition due to the valve alignments involved.

Therefore, the NUREG-1432 31 day Frequency is replaced with an 18 month . Frequency in the proposed ITS to reflect the test being performed during shutdown conditions. This change is a plant specific change to reflect the Palisades Nuclear Plant design. 15. The Palisades plant was designed prior to issuance of the General Design Criteria (GDC) in 10 CFR 50. Therefore, reference to the GDCs is omitted and appropriately replaced by reference to "Palisades Nuclear Plant design criteria . " The Palisades Nuclear Plant design was compared to the GDCs as they appeared in 10 CFR 50 Appendix A on July 7, 1971. It was this updated discussion, including the identified . exemptions, which formed the original plant Licensing Basis for future compliance with the GDCs . "31=t> \ <,.. '3.<...t. -f Mc.) D \1 Ml: <..

  • 1 f"fDlg {<.A I '7.",. ,g-... . *-. -Palisades Nuclear Plant Page 5 of 5 01/20/98
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  • Insert to 3.6:6, JFD 16. ISTS SR 3.6.6.3 requires the verification that the required flow rate through each containment cooling train *is greater than the specified limit. The proposed Palisades ITS modifies this requirement to verify that the total flow rate through the containment air coolers is greater than the specified limit through the specified coolers when the system is aligned for accident conditions.

This change is required because the Palisades design does not include flow instrumentation on each cooler individually and significant changes to the flow distribution occur as the result of automatic realignment of service water system valves. The flow distribution through the containment air coolers when in the accident alignment is governed by the piping configuration of the system, i.e. valves are placed in their wide-open state, flow is isolated to one heat exchanger, and non-safety related portions of the system are automatically realigned to support the system safety function. CTS does not contain a surveillance requirement that corresponds to ISTS SR 3.6.6.3. ISTS. 3.6.6.3 has been modified into ITS 3.6.6.4 that requires flow testing of the containment air coolers in a manner that is consistent with the Palisades design, installed .. instrumentation, and operating practice .

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.6-9 DOC M.6 ITS SR 3.6.6.4 and Associated Bases ITS SR 3.6.6.4 verifies the service water flow rate to the safety related CACs. DOC M.6 states this fact in adding ITS SR 3.6.6.4. However, the third sentence in DOC M.6 states "This requirement to verify service water flow to the CACs is not included in the proposed ITS." This seems to be a contradiction.

Comment: Correct this discrepancy. Consumers Enerqv Remonse: The sentence should say "This requirement to verify service water flow to the CACs is not

  • included in the CTS." DOC M.6 has been corrected.

Affected Submittal pages: Att 3, DOC 3.6.6, page 4 of 8 46

  • **
  • M.4 M.5 M.6 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS CTS 4.6.5b specifies that for the Containment Air Cooling System "Each fan and valve required to function during accident conditions will be exercised at intervals not to exceed three months. " The proposed ITS in SR 3. 6. 6. 2 changes this interval to every 31 days. The change from every 3 months to 31 days is a more restrictive change. This change is consistent with NUREG-1432.
  • The proposed ITS SR 3. 6. 6 .1 requires that each containment spray manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position be verified to be in the correct position every 31 days. The CTS does not require a periodic verification of containment spray valve position.

This surveillance is added to give increased confidence that the valves are in their appropriate position. This change is considered to be a more restrictive change since this valve position verification is not contained in the existing CTS. This change is consistent with NUREG-1432.

  • The proposed ITS SR 3.6.6.4 requires that the total service water flow rate is > 3935.l gpm to the Containment Air Coolers (CACs) VHX-1, VHX-2, and VHX-3 . This requirement is provided to ensure that the CACs are receiving the necessary flow as specified in the safety analysis to ensure that sufficient heat is removed following a design basis accident to ensure that the containment pressure and temperature is maintained within the analyzed values. This requirement to verify service water flow to the CACs is not included in the e S. Therefore, including this requirement in the proposed ITS is considere to be a more restrictive change. This change is consistent with NUREG-1432.

M.7 CTS 3.4.2 and 3.4.3 specifies that following the reactor being placed in a hot shutdown condition due to an inoperability, that if the inoperability is not restored "within an additional 48 hours," the reactor shall be placed in a cold shutdown condition within 24 hours. The allowance for the additional 48 hours before a further shutdown is required is not included in the proposed ITS. In the proposed ITS Condition B specifies the shutdown requirements as discussed in other DOCS and does not include this intermediate completion time to restore an inoperable component. The Completion Time of Required Action B.2 to be in MODE 4 in a total of 30 hours provides a reasonable time to restore an inoperable containment cooling train to OPERABLE status. The deletion of this 48 hour allowance is considered to be a more restrictive change since it decreases the amount of time to restore an inoperable containment cooling train to OPERABLE status. Th,is change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 4of8 01/20/98 90 -tJ-

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6..:1 O DOC M.8 CTS 4.6.2.a CTS 4.17.3.1.a ITS SR 3.6.6. 7 ITS SR 3.6.6. 7 requires that each containment spray pump starts on an actual or simulated actuation signal every 18 months. DOC M.8 justifies the adding of this SR as a More Restrictive change since the verification (pump starting) is not contained in the CTS. This is incorrect.

DOC M.8 states that the pumps are started as a result of the performance of CTS 4.17.3.1.a but is not explicitly verified to start based on that SR. The staff disagrees. The performance of CTS 4.17.3.1.a in and of itself may not explicitly require verification of pump start but it does require a pump start. However, CTS 4.6.2.a does require that verification, whether it is performed in conjunction with or independent of CTS 4.17.3.1.a, CTS 4.6.2.a requires that system operation be verified at each refueling outage by tripping the actuation instrumentation. Since the pump is part of the system, this CTS would verify pump start. Therefore, the staff considers this change. to be an Administrative change rather than a More Restrictive change. Comment: Revise the CTS markup and provide additional discussion and justification for this Administrative change. Also, include the appropriate markup of CTS 4.17.3.1.a in the CTS markup of 3.6 to reflect this change. Consumers Energy ReS.Ponse: DOC A.7 has been added to describe the reformatting of the CTS 4.6.2.a requirement to initiate the Containment Spray pumps to proposed SR 3.6.6.7. CTS 4.6.2.a requires the performance of a system test, including a requirement to initiate the test by tripping the normal actuation instrumentation. The CS pumps are an integral part of the system, and this CTS surveillance requirement includes the start of the pumps. DOC M.8 has been deleted in lieu of the new DOCA.7. CTS 4.17.3.1.a, requires the performance of a Channel Functional Test (CFT) of the Containment High Pressure (CHP) logic trains. This requirement has become ITS SR 3.3.4.3. This requirement directs the testing of the logic to the " ... trip initiating function." For clarity in the distinction between the logic testing of SR 3.3.4.3 and the required pump start of SR 3.6.6.7, CTS 4.17.3.1.a was not included in the markup pages of CTS 4.6.2.a. Affected Submittal pages: Att 3, CTS 3.6.6, page 4-24 Att 3, CTS 3.6.6, page 4-25 Att 3, DOC 3.6.6, pages 2 and 8 47

  • *
  • 4. 6 isAfiri INJEtJ'.{ON AHO jcDNTAI"1ENT SPMY sysr01s TESTS I 4. 6 .1 a. System tests shall bt perfot"!llld teach reactor refu1lin9 int val. A test saf1ty tnj1ctton signal tll bt applted to initiate operatton of the syst ... Tht safety 1nj1ctton and shutdow cooling '-system pulq) motors HY bed 1n1rgtztd for this test. Th system S'ee 3.S/ will bt considtrtd s*t1sfa ory tf control board indicat*on and visual tnd1 tt that all have r. c1ived the safety tnjection signal n the proper s1quenc1 and ti ng (ie, the appropriate pump break s shall havt opened and clos d, and all valves shall have c 11ttd their travel). Cont11n11nt Spray Syst11 c. At least 1v1ry tin years tht spray nozzlts shall bt verified to be open. l4J.ll flmlR1

(!> 'Yro re---. 6.4 b. Acceptablt 11v1ls of ptrfoC91nct shall bt that tht Du.Os start, n:ch tht1r rated h11ds on rec1rculat1on. flow, land/operate for atl 111 ii f1ftnn j1nut1j * (lAl: 13 Each Saf1t1 InJtet1on Tank f1 path shall be vtr1fitd OP. RABLE w1th1n 1 days prior to each r ctor startup by ver1fy1n 1ach motor optrattd 1soht1on valvt is pen by observing valvt po tion 3 s"'-1nd1cat1on and valvt 1tsel , and locking open tht us httd S"e.e.

  • Y c1rcu1t br1ak1rs.

\ Iht Low Pressure Safet Injection flow path shall v1rift1d OPERABLE w1th1n 7 dax prior to each reactor star. up by verifying flow control valvt * *3001 ts open, and 1ts air upply is isolated. 4-24 !Wtndllint No. Q, ;.i, 9i, 174 October 31. 1996 f'.,_"I .._ 'J ot_ 4

  • *
  • 14. 6 SAFETY INJECTION ANO CONTAINMENT SPRAY SYSTEMS TESTS f-.@ Syrye111ance Begyirements (continued)

I f4.6.4/ l/alyis {contt'4iued) 1--@ Th safety injection recirc ation path shall be v ified OPERABLE thin 7 days prior to u reactor startup by ve fyi ng va 1 ves See. 5. )> CV-3027 and 3056 are op and their switches HS 027A, HS-30278, HS*3056A, and HS-3056 art open. d. Each C ntain .. nt Spray V ve control OPE LE at least once ch rtfue11ng by CY, ing each valve from tht control roOll .t\111 obs1rvi09 valvt op ation at least each 1 110nths. Air Cgo11nq Syst11 (j'.) E .. MJency llOd1 .sR for OPERABILITY during 11c ct ue sh J.7) Each fan (and valv* required to function duri w111 be not f 0 ,.. < ,4/JO s R J. {:.._(g. I S!-, 1., .. G., 3 o.. r '"ere ... *hd .... > B c..s ,-...._ Z:T!) @> RA-r: 3.f...l.-\0 4*25 AmtndMnt No. W, ;.a, ;;., W, 174 1 1 October 31, 1996 'i l-b LLi4 i'c..cr

  • *
  • A.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS CTS 3 .4.4 specifies that valves, interlocks and piping that are directly associated with the "above" (CTS 3.4.1) components shall meet the same requirements as listed for that component CTS 3.4.5 specifies that valves, interlocks and piping which is associated with the containment cooling system and not covered by Specification 3 .4.4 may be inoperable for no more than 24 hours if it is required to function during an accident.

These requirements are addressed by the definition of OPERABILITY which requires that all associated equipment be OPERABLE. In the proposed ITS all equipment in a particular train which is required to function during an accident must be OPERABLE and all equipment in the train will have the same Completion Time. This is an administrative change since the requirement remains that all equipment in a train of containment cooling must be OPERABLE. This change is consistent with NUREG-1432. A.5 CTS 4.6.3 specifies that the containment spray pumps be tested at intervals "not to exceed three months." In the proposed ITS SR 3.6.6.5 this is changed to be "In accordance with the Inservice Test Program." The Inservice Test Program will specify the frequency for pump testing. This is considered to be an administrative change since the three month frequency will now be contained in the Inservice Test Program. This change is consistent with NUREG-1432. A.6 CTS 4.6.2a specifies in part that for the Containment Spray System Test: "Operation of the system is initiated by tripping the normal actuation instrumentation. " The proposed ITS add the following phrase "by an actual or simulated actuation signal." The allowance to use an actual or simulated actuation signal is based on the fact that the channel being tested cannot differentiate between an actual or simulated signal and either initiation can demonstrate system OPERABILITY. This is considered to be an administrative change since the system testing requirements have not changed. This change is consistent with NUREG-1432. A .'1-s RA/ 3 &. (;.-;D Palisades Nuclear Plant Page 2of8 01/20/98

  • Insert to DOC A.7 CTS 4.6.2.a requires a system test of the containment spray system that is initiated by tripping the normal actuation instrumentation.

Since the containment spray pumps are an integral part of the system this CTS requirement verifies that the pumps are capable of being started by an actuation signal. ITS SR 3 .6.6. 7 requires verification of automatic containment spray pump start upon receipt Of an actuation signal. This reformatting of 'the CTS 4.6.2.a requirement into ITS SR 3.6.6.7 is an administrative change that does not result in addition or removal of any requirement. This change is consistent with NUREG-1432 . L/1-d

  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQVEST: 3.6.6-11 DOC LA.2 CTS 4.6.2.a CTS 4.6.2.c CTS 4.6.2.a specifies for the Containment Spray System test that "The test shall be performed with the isolation valves in the spray lines at the containment blocked closed." CTS 4.6.2.c specifies for the Containment Spray System test that 'The test will be considered satisfactory if visual observations indicate all components have operated satisfactorily." This information according to DOC LA.2 is being relocated to plant procedures.

It is unclear from the discussion in DOC LA.2 if the procedure change control process is covered by 10 CFR 50.59 or some other non-regulatory control process. If the procedure change control process is not covered by 10 CFR 50.59 then the change is a Less Restrictive (L) change deletion of material rather than a Less Restrictive (LA) change. Less Restrictive (LA) changes are limited to those items which are relocated to licensee controlled documents covered by a 10 CFR 50.59 change control process. See Comment Number 3.6.6-12. Comment: Provide additional discussion and justification on the plant procedure change control process. See Comment Number 3.6.6*12. Consumers Energy Response: DOC LA.2 has been deleted and a new L.3 DOC has been prepared to describe the removal of these details from the proposed specifications. Affected Submittal Pages: Att 3, CTS , page 4-24 Att 3, DOC 3.6.6, page 5 of 8 Att 3, DOC 3.6.6, page 8 of 8 Att 4, NSHC 3.6.6, page 5 of 5 \ 48

  • ':::; .r. I ....
r. c.o.r\ o -4 '6 [sA¢ij INJECJAON AHO jcOHTAI"'ENT SPRAY sysn;11s TESTS I 4. 6 .1 l4J.3 I System tests shall be p1rfoM11d teach *reactor refueling int val. A test safety injection signal 111 be applttd to initiate operation of th* syst*. *Tht safety injection and shutdow cooling ) syste* 110tors 1ay bt d energized for this test. Th system See 3.5 will be considtrld s&ti\fa ory tf control board 1ndtcat*on and visual observations indt te that 111 cQ11Pon1nts have r. c1iv1d tne safety tnJectton signal n the proper s1qu1nc1 and ti ng (11, th* appropriate PUllP break s shall have opened and clos d, and valves shall have c lettd their travel). Contatn19nt

$pray Syst11 c. At least 1ver1 tin years th* spray nozzles shall b9 vtr1f11d to bt op*n. 5"S J. C...t..S' (!) -+--------- .6.4

  • b. R.Al: Acc1ptabl*

ltv*ls of p1rfot'9lnc1 shall b9 that th* DU!PS start, r::ch th1tr rattd heads on r1ctrculat1on flow, land/Operate fOt atj 11 ii f1ftnn jtnutli* Each Saf1t1 Injection Tank fl path shall b9 ver1f1td OP. RABLE w1th1n 7 d111 prior to each r ctor startup b1 ver1fy1n 1ach motor operated tsolatton valve ts pen b 1 1 okbs1rv1ng vahlve po 1 tt1odn 3.S'\. 1nd1cat1on and valv1 1tse1 , and oc 1nt open t e ass a

  • ctrcutt breakers.

\ The Low Pressure Saftt flow path shall

  • v1r1f11d OPERABLE within 7 dax prtor to each reactor star. up by verifying flow control valve -3001 ts open, and tts air upply is isolated.

4*24 Amtndllint No. ii-, 1-i, 9i, w. -t*, -tii. l 74 October 31, 1996 r ... " e. J o-l= L(

  • *
  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS M.8 The propose TS SR 3.6.6.7 req
  • es that each contai on an actual or s
  • ulated actuation sig l every 18 months. he CTS does not e licitly require a eriodic verification f containment spray tarting, although th umps are verifie to start as part of th CHP Logic Channe unctional Test req red by CTS .17 .3. la. This pro des an explicit requ* ement to verify pu starting.

This ch ge is considered to ea more restrictive ange since this ver* ntained in the exist* g CTS. This chang

  • is consistent with N LESS RESTRICTIVE CHANGES -REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) LA. l In the CTS 3 .4. la and b, the major components (i.e., Containment Air Coolers and *Containment Spray Pumps) of the containment cooling trains are listed with an associated diesel generator and required to be OPERABLE.

In CTS .3.4. lc it also requires that "All heat exchangers, valves, piping, and interlocks associated with the . above components and required to function during accident conditions are OPERABLE." In the proposed ITS the LCO will simply read: "Two containment cooling trains shall be OPERABLE." In the proposed ITS, the information regarding , what will comprise a train of containment cooling is not included in the TS LCO but will be addressed in the Bases. Changes to the Bases are made under licensee control in accordance with the Bases Change Control Program which is addressed in Chapter 5 of the ITS. This change is comistent with NUREG-1432. . fVo} ((_,,,.._.( RA!:. 3 *(,. (.. -\\ LA.2 CTS 4.6.2a specifies t for the Containmen pray System test "The st shall be performed with th isolation valves in the ray supply lines at the ntainment blocked closed." CTS .6.2c specifies that for e Containment Spray S tern test: "The test will be cons* ered satisfactory if vis observations indicate components have operated tisfactorily." This lev of detail with respect t ow the test is perfo and w t the testing acceptanc riteria are is more app priately addressed in nt test' g procedures. The re rrement to perform the ntainment spray test i ot anged but the testing thodology and accep e criteria will be not b ocated in the proposed ITS but 111 be addressed in pla rocedures. Changes plant procedures are ma

  • in accordance with th lant procedure change rocess. These changes are co 'stent with NUREG-14 Palisades Nuclear Plant Page 5of8 01/20/98
  • *
  • L.1 (Continued)

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS The support role and requirements of the service water pumps and component cooling water pumps to the containment air coolers and containment spray pumps are addressed by LCO 3.0.6 and the Safety Function Determination Program as well as the definition of OPERABILITY. This helps to ensure that the containment cooling trains have adequate support equipment to perform their function. This change maintains consistency with the intent of NUREG-1432. L.2 CTS 3.4.2 and 3.4.3 require that "If the inoperable component (train in ITS) has not been restored to operability within an additional 48 hours the reactor shall be placed in a cold shutdown condition within 24 hours." The proposed ITS requires that for Required Actions and Associated Completion Times not met the plant is eventually required to be placed in MODE 4 within a total of 30 hours. The "additional 48 hours" allowance of the CTS for restoring inoperable components before taking actions is not included in the proposed ITS. A shutdown to MODE 4 required in the proposed ITS places the plant out of the MODE of Applicability. CTS 3 .4.1 only requires that the equipment associated with Containment Cooling Systems be OPERABLE prior to making the reactor critical. As discussed in "more restrictive" change M.1, the proposed ITS will require that the containment cooling equipment be OPERABLE in MODES 1, 2, and 3 which is the MODES for which the equipment is required to function in an accident. The CTS requirement to place the plant in a cold shutdown condition (i.e, < 210 °F in the CTS) if the required actions and associated completion times are not met is overly restrictive since the equipment is not needed below 300°F. Therefore, only requiring that the plant be placed in MODE 4 is considered to be a less restrictive change. This change maintains consistency with NUREG-1432. C..."3 .:rf\.Oef"+ f\-ev.J '-( Palisades Nuclear Plant "3. (... -I I RA-I 3.V.L,-13 Page 8 of 8 01/20/98

  • *
  • DOC L.3 to 3.6.6 CTS 4.6.2a specifies that for the Containment Spray System test, "The test shall be performed with the isolation valves in the spray supply lines at the containment blocked closed." CTS 4.6.2c specifies that for the Containment Spray System test, "The test will be considered satisfactory if visual observations indicate all components have operated satisfactorily." This level of detail, with respect to how the test is performed and what the testing acceptance criteria are, is more appropriately addressed in plant procedures.

Removal of these details from the Technical Specifications will not affect the requirement to perform the containment spray test, however the methodology and acceptance criteria will not be included in the proposed specifications. This change is consistent with NUREG-1432 . \

  • *-* 2. 3. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
  • The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation.

The proposed change will continue require that the plant be placed in a MODE which does not require the equipment to be OPERABLE if the Required Actions and Associated Completion Times are not met. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? The proposed change allows the plant to be placed in MODE 4 ( < 300 °F) as opposed to the CTS "cold shutdown condition" ( < 210 °F) is the Required Actions and Associated Completion Times have not been met. The margin of safety afforded by the containment cooling systems is to ensure that containment pressure and temperature are maintained within the analytical values. Since the change still requires that the plant be placed in a MODE where the equipment is not required to operate if the Required Actions and Associated Completion Times have not been met no margin of safety is reduced. Therefore, this change does not involve a significant reduction in a margin of . safety. /Q_A-1 5.'4.(, -13 \ Palisades Nuclear Plant Page 5 of 5 01/20/98 '-18 e_

  • NSHC for L.3 to 3.6.6 *
  • LESS RESTRICTIVE CHANGE L.3 CTS 4.6.2a specifies that for the Containment Spray System test, "The test shall be performed with the isolation valves in the spray .supply lines at the containment blocked closed." CTS 4.6.2c specifies that for the Containment Spray System test, "The test will be considered satisfactory if visual observations indicate all components have operated satisfactorily." This level of detail, with respect to how the test is perfqrmed and what the testing acceptance criteria are, is more appropriately addressed in plant procedures.

Removal of these details from the

  • Technical Specifications will not affect the requirement to perform the containment spray however the methodology and acceptance criteria will not be included in the proposed specifications.

This change is consistent with NUREG-1432.

1. 2. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

The proposed change would remove details of how the containment system is aligned and regarding the how test acceptance criteria are to be met.* These directions do not provide direction or limits that must be verified. The details of how the system should be aligned during testing are merely a prerequisite designed to prevent inadvertent actuation of the containment spray system duriiig testing. The acceptance criteria details merely indicate that "visilal observation" is an acceptable way to verify system function. Since the details provide no substantive limits, removing this detail from the surveillance testing requirements associated with the containment spray system testing does not effect the testing or operability of the system in anyway. Since no effect on the system will occur, the probability of any accident previously evaluated is not effected. The consequences of a previously analyzed event are dependent on the initial conditions ' ' assumed for.the arialysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The presence or lack of the test details *being removed from the specifications does not effect the ability of the containment.spray system to perform its safety function. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The propa,sed change will merely remove the detail of test alignments and acceptance criteria from the Technical Specifications. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

  • 3 . *
  • Does this change involve a significant reduction in a margin of safety? The proposed change will remove details regarding how the containment spray system is aligned during testing, and a statement that indicates visual observation is an acceptable way of confirming acceptable results. The margin of safety afforded by the containment spray system is to ensure that the containment pressure and temperature are maintained within the analytical values. The removal of this unnecessary detail from the Technical Specifications will not effect the way testing is performed or the availability of the containment spray system. Therefore, this change does not involve a significant reduction in a margin of safety . f8-j
    • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-12 DOC LA.2 CTS 4.6.2.a ITS SR 3.6.6.6 CTS 4.6.2.a specifies that for the Containment Spray System test "The test shall be performed with the isolation valves in the spray lines at the containment blocked closed." Insufficient information is provided to justify the relocation of this detail to plant procedures.

CTS 4.6.2.a tests in part that the automatic containment spray valves actuate to their correct position on an actuation signal. The corresponding ITS SR is ITS SR 3.6.6.6. ITS SR 3.6.6.6 test all automatic valves except those locked, sealed or otherwise secured in the required position under Administrative control. Thus if the isolation valves at the containment are manual valves, then the relocation of the statement in CTS 4.6.2.a is acceptable. However, if the isolation valve is an automatic valve which actuates, then the statement cannot be relocated to procedures, but must be specified in the Bases. The ITS SR 3.6.6.6 exception for locked valves, which is not justified or indicated in the CTS markup, applies to those automatic valves that are normally locked in their correct position during operation and does not apply to valves that are locked for testing purposed. Comment: Revise the CTS markup to correctly reflect ITS SR 3.6.6.6 and provide the appropriate discussion and justification for the changes associated with ITS SR 3.6.6.6 discussed above. In addition, if the locked closed isolation valves are automatic valves describe how these valves will be tested to verify that they will actuate to their correct position on an actuation. Consumers Energy Response: As described in the response to RAI Comment 3.6.6-11, DOC LA.2 has been deleted and a new DOC L.3 prepared to described the removal of this requirement to align the containment spray isolation valves. This alignment does not support system OPERABILITY, it merely provides a cautionary allowance to prevent inadvertent spray actuation into the containment. The containment spray system isolation valves are normally closed valves that automatically open if a containment high pressure signal is received .. The valves are clearly subject to the requirements of SR 3.6.6.6 because they are not "locked, sealed or otherwise secured in the required position" as described in the Bases for this SR. The valves automatic actuation is a required part of the overall system response to an accident. SR 3.6.6.6 and SR 3.6.6.7 together constitute testing that provides the same degree of testing as presented in CTS 4.S.2a. The CTS markup pages were modified as a result of the changes required to address RAI Comment 3.6.6-10 and-11 to reflect this correspondence. For additional information, see the responses to these comments. Affected Submittal Pages: No page changes. \ \ 49

  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SEC"l'ION 3.6, CONTAINMENT NRC REQUEST: 3.6.6-13 DOC LA.3 CTS 4.6.3 CTS 4.6.3.a states for the Containment Spray Pump that "Alternate manual starting between control room console and the local breaker shall be practiced in the test program." CTS 4.6.3.b goes on to require that the pumps operate for a specific period of time 15 minutes).

This information according to DOC LA.3 is being relocated to plant procedures. It is unclear from the discussion in DOC LA.3 if the procedure change control process is covered by 1 O CFR 50.59 or some other non-regulatory control process. If the procedure change control process is not covered by 1 O CFR 50.59, then the change is a Less Restrictive (L) change deletion of material rather than a Less Restrictive (LA) change. Less Restrictive (LA) changes are limited to those items which are relocated to licensee controlled documents covered by a 10 CFR 50.59 change control process. Comment: Provide additional discussion and justification on the plant procedure change control process. Consumers Energy Response: DOC LA.3 has been deleted and a new L.4 DOC has been prepared to describe the removal of

  • these details from the proposed specifications.
  • Affected Submjttal Pages: Att 3, CTS 3.6.6, page 4-24 Att 3, DOC 3.6.6, page 6 of 8 Att 3, DOC 3.6.6, page 8 of 8 Att 4, NSHC 3.6.6, page 5 of 5 ' \ 50
  • 4. 6 /sA¢TY INJEC}!Off AHO /cONTAIHMENT SPRAY SYSTE.MS TESTS
  • 4.5.1 Syste* tuts shall t Heh *ructor refuel 1ng int Vil. A test safety tnjection stgnal ill be applied to initiate operation of the syst ... -The safety tnjectton and shutdow cooling system pump 110tors 1111 bt d enef91Zld for th1s test. Th system 3.S':> will be considered Sit1ifa ory tf control board 1ndicat'on and v1sual observat1ons tndi t1 that all c011ponents have r, c1iv1d the safety injection stgnal n the proper sequence and ti ng (ie, the appropriate PUllP break s shall have op1ntd and clos d, and all valves shall have c 11tld their travel). Cont1tn11nt Spray Syst11 c. At least 1v1ry ten years the spr11 nozzles shall bt vertf1ed to be open. I 4J.3 I ElmR1 )>,.o . 6.4 b.
  • Acc1ptabl1 l1v1ls of p1rfon11nc1 shall be that the ouaps start, r::cb thetr ratld heads on rectrculatton flow, land/operate f0£ atj 11 ii ttttun jinut1j
  • Ml: 3.t.(.-1} Each SafttJ Injection Tank f1 path shall be v1rtfitd RABLE within 7 days prior to each r ctor startup by v1rifytn each motor optrattd tsolatton valve 1S pen by obstn1ng vaht po tion S' 3 5\. indication and vaht itstl , and locktnt open th* us httd
  • circuit br1ak1rs.

The Low Pressure Saftt InJiction flow path shall verified OPERABLE within 7 dax prior to each reactor star, up by verifying flow control valve *3001 is open, and tts atr upply is isolated. 4-24 AmtndMint No. ii-, 9', H-1-, 1*, 174 October 31, 1996 o J o-1= 4 r .... '\ c. 50-CL

  • *
  • LA.3 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS g *--*-----LA.4 CTS 4.6.4d specifies that "Each Containment Spray Valve manual control be verified to be OPERABLE at least once each refueling by cycling each valve from the control room while observing valve operation at least each 18 months." The requirement to cycle the containment spray valves by manual control are details addressed by the Inservice Testing Program which are not included in the proposed
  • ITS. The Inservice Testing Program is discussed in ITS Section 5.5. The details of the program are located in the plant procedures which implement the program. Therefore, the requirement to manually cycle t:P.e containment spray valves will not be included in the proposed ITS but will be located in the plant procedures which implement the Inservice Testing Program. Proposed SR 3.6.6;6 will demonstrate that the valves are OPERABLE and actuate to the correct position on receipt of an actuation signal. This change is consistent with NUREG-1432.

,, Palisades Nuclear Plant Page 6 of 8 01/20/98 5o-b

  • *
  • L.1 (Continued)

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS The support role and requirements of service water pumps and component cooling water pumps to the containment air coolers and containment spray pumps are addressed by LCO 3.0.6 and the Safety Function Determination Program as well as the definition of OPERABILITY. This helps to ensure that the containment cooling trains _have adequate support equipment to perf9rm their function. This change maintains . consistency with the intent of NUREG-1432. L.2 CTS 3 .4.2 and 3 .4.3 require that "If the inoperable component (train in ITS) has not been restored to operability within an additional 48 hours the reactor shall be placed in a cold shutdown condition within 24 hours." The proposed ITS requires that for Required Actions and Associated Completion Times not met the plant is eventually required to be placed in MODE 4 within a total of 30 hours. The "additional 48 hours" allowance of the CTS for restoring inoperable components before taking actions is not included in the proposed ITS. A shutdown to MODE 4*required in the proposed ITS places the plant out of the MODE of Applicability. CTS 3 .4.1 only requires that the equipment associated with Containment Cooling Systems be OPERABLE prior to making the reactor critical. As discussed in "more restrictive" change M. l, the proposed ITS will require that the containment cooling equipment be OPERABLE in MODES l, 2, and 3 which is the MODES for which the equipment is required to function in an accident. The CTS requirement to place the plant in a cold shutdown condition (i.e, < 210* °F in the CTS) if the required actions and associated completion times are not met is overly restrictive since the equipment is not needed below 300°F. Therefore, only requiring that the plant be placed in MODE 4 is considered to be a less restrictive change. This change maintains consistency with NUREG-1432. t...3 rl'\6et-+ n-ei.J k ._, Palisades Nuclear Plant "3. '-. (..

  • I I RA-I 3.l.l.lr 13 Page 8 of 8 SD-C-01/20/98
  • *
  • DOC L.4 to 3.6.6 CTS 4.6.3a and 4.6.3b contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirellient.

As such, these details are proposed for deletion. CTS 4.6.3a states that "alternate manual starting [of the containment spray pumps] between the control room console and the local breaker shall be practiced in the test program." The ability to demonstrate the manual starting capability of the pumps from various plant locations is not assumed in the safety analyses and is not relevant to demonstrating that the pumps are capable of meeting their intended safety function. CTS 4.6.3b requires that the containment spray pumps " ... operate for at least fifteen minutes. " The ability to demonstrate pump operation for an arbitrary period such as fifteen minutes is not representative of the ability of the pump to operate for an extended period following a postulated accident, as assumed in the safety analyses. Since the above details are not necessary to describe, or are not pertinent to, any actual regulatory requirement, they can be deleted without an impact of public health and safety. This change is consistent with NUREG-1432 . \ \ 50-d

  • *
  • 2. 3. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING Does the change create the possibility of a new or different kind of accident*

from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue require that the plant be placed in a MODE which does not require the equipment to be OPERABLE if the Required Actions and Associated Completion Times are not met. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a reduction in a margin of safety? The proposed change allows plant to be placed in MODE 4 ( < 300 °F) as opposed to the CTS "cold shutdown condition" ( <210 °F) is the Required Actions and Associated Completion Times have not been met. The margin of safety afforded by the containment cooling systems is to ensure that containment pressure and temperature are maintained within the analytical values. Since the change still requires that the plant be placed in a MODE where the equipment is not required to operate if the Required . Actions and Associated Completion Times have not been met no margin of safety is reduced. Therefore, this change does not involve a significant reduction in a margin of safety . Palisades. Nuclear Plant Page S of 5 01/20/98 50-e._

  • *
  • NSHC for LA to 3.6.6 LESS RESTRICTIVE CHANGE L.4 CTS 4.6.3a and 4.6.3b contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement.

As such, these details are proposed for deletion. CTS 4.6.3a states that "alternate manual starting (of the containment spray pumps) between the control room console and the local breaker shall be practiced in the *test program." The ability to demonstrate the manual starting capability of the pumps from various plant locations is not assumed in the safety analyses and is not relevant to demonstrating that the pumps are capable of meeting their intended safety function. CTS 4.6.3b requires that the containment spray pumps " ... operate for at least fifteen minutes." The ability to demonstrate pump operation for an arbitrary period such as fifteen minutes is not representative of the ability of the pump to operate for an extended period following a postulated accident, as assumed in the safety I analyses. Since the above details are not necessary to describe, or are not pertinent to, any actual regulatory requirement, they can be deleted without an impact of public health and safety. This change is consistent with NUREG-1432.

1. 2. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change deletes details from the Technical Specifications that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. The deletion of details from the Technical Specifications is not assumed to be an initiator of . any analyzed event. The proposed changes do not reduce the functional requirement or alter the intent of any specification. As such, the consequences of an accident remain unchanged. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

  • Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change deletes detail from the Technical Specifications that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. The changes will not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. The changes will not impose different requirements, and adequate control of information I . will be maintained. The changes will not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

  • *
  • 3 . Does this change involve a significant reduction in a margin of safety? Margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative are initiated.

There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. The proposed changes deletes details from the Technical Specifications. Removal of these details is acceptable since this information is not directly pertinent to the actual requirement and does not alter the intent of the requirement. Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to licensee controlled document without a s*ignificant impact on safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety .

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.6-14 DOC L.1 JFD6 CTS 3.4.2 CTS 3.4.3 . CTS 3.4.4 CTS 3.4.5 STS 3.5.2 ITS 3.6.6 ACTION A. and Associated Bases CTS 3.4.2 and CTS 3.4.3 provide a 7-day allowed outage time for when one component is inoperable and a 24 hour allowed outage time when two components are inoperable.

ITS 3.6.6 ACTION A permits one or more trains to be inoperable provided there is a least 100% cooling capacity equivalent to a single OPERABLE containment cooling train. This approach is based upon STS 3.5.2; however, acceptability of this potentially generic change will depend on demonstrating that the extra containment cooling capacity is available, (See Comment Number 3.6.6-3) and that the design of the Palisades Containment Spray and Cooling Systems are sufficiently different from the other PWR designs to warrant a unique set of ACTIONS. In. addition, the following issues assume that an alternate train configuration can be established. It is tentatively acceptable to have a 72 hour Completion Time, rather than 7 days, for one train of . containment cooling inoperable under ITS 3.6.6 ACTION A. However, DOC L.1 does not specifically address this change which appears to be a More-Restrictive change to CTS 3.4.2. The CTS 3.4.3 change from 24 hours to 72. hours is Less-Restrictive. Insufficient information is provided in DOC L.1 for these changes. In CTS 3.4.5, there are components associated with the Containment Cooling System train which only have an outage time limited to 24 hours. Should these certain containment train be handled in a new Condition? Are there components which should have a longer Completi.on Time? For example -Can either pump 54b or 54c being solely inoperable in one train only be justified as allowable for 7 days? Are there trains or components which should have a shorter Completion Time? For example -Should a Containment Cooling System train cross-over valve be limited to a 24 hour period of

  • inoperability?

Comment: Provide additional discussions and justifications for the above Restrictive or More Restrictive changes that are separate from the overall grouping provided in DOC L.1. JFD 6 may need to be revised based on the responses to the above.

  • Consumers Enemv Ruponse: . Palisades containment cooling system design utilizes four Containment Air Coolers (CACs) and three Containment Spray (CS) pumps. The system and associated specification are unique because of a combination of shared system piping on the containment spray, Component Cooling Water (CCW), and Service Water systems (SW), and the asymmetric diversity of credited heat removal paths. The design is shown in FSAR sections 6.2, 6.3, 9.1, and 9.3, as well as the associated figures. The CS, CCW, and SW systems are designed so that a common pump discharge header is used to ensure that any individual pump can supply the entire system. Train separation and single failure protection is provided through isolation valv.es that can be used to mitigate failures. (continued) 51
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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT f\'RC REQUEST: 3.6.6-14 Consumers Energy Response: (continued)

This common header design provides the flexibility to use a component from an opposite train, if available, with no re-alignment of components necessary. This, in tum, provides the added reliability of opposite train availability given the appropriate power supply. For example each CS pump supplies water through each shutdown cooling heat exchanger, to each spray header in the containment building. Similarly, each CCW pump provides cooling to each shutdown cooling heat exchanger, and each SW pump supplies cooling water to each CCW heat exchanger. The asymmetry of the credited containment cooling trains lies with the heat removal path associated with the right and left trains. The heat removal path credited for the left train is the CS pumps and the shutdown cooling heat exchangers. The heat removal paths for the right train are a combination of three CACs, and a CS pump and the shutdown cooling heat exchangers. The CACs transfer heat directly to the SW system. The CS system removes heat from the containment when operating in the recirculation mode via the shutdown cooling heat exchanger, the CCW system, the CCW heat exchangers, and subsequently to the SW system. The diversity in credited heat removal paths, and the fluid systems common header design, led to the component level requirements of the CTS. The primary 'train' specific aspects of these systems is their association with a specific safety related power source. With the common header I system design of the fluid systems trains, no specific actions are required to compensate for unavailable individual components. With the loss of an individual component, full capacity for containment cooling remains available with no specific actions required to align or compensate. The proposed ITS Actions are structured to capture the existing requirements, however with the additional flexibility described in DOC L.1, i.e. any number of components may be inoperable on either train, as long as 100% of the cooling capability equivalent to a single train, remains OPERABLE. If three CACs are inoperable, the CTS would require an immediate shutdown. However with the three CACs inoperable, and any combination of two CS pumps and one containment air cooling safety related fan OPERABLE, the ITS will allow continued operation for up to 72 hours. This is clearly a less restrictive change in that a forced shutdown would be avoided. A separate Condition with a 7 day Completion Time for one inoperable component was not sought because of the lack of total redundancy between the two trains of cooling systems. A more analogous specification to model for an inoperable train is that provided in ISTS Condition D, with its 72 hour Completion Time. Based on this, the Completion Time for proposed Action A was established. Palisades believes that the proposed specificatioq 1 appropriately reflects the design and licensing basis of the plant. As described above, the plant specific design is unique in its use of diverse systems in an asymmetric arrangement, supported by fluid systems that share flow headers and piping in a non-train specific.arrangement, to justify the proposed deviation from the NUREG-1432. Affected Sµbmjttal f'ages: No page changes. 52

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  • TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-15 DOC L1 JFD6 CTS 3.4.3 CTS 3.4.4 CTS 3.4.5 CTS 3.4.6 STS 3.5.2 ITS 3.6.6 ACTION A and Associated Bases CTS 3.4.2 and CTS 3.4.3 provide a 7-day allowed outage time for when one component is inoperable and a 24 hour allowed outage time when two components are inoperable.

ITS 3.6.6 ACTION A permits one more trains to be inoperable provided there is at least 100% cooling capacity equivalent to a single OPERABLE containment cooling train. This approach is based upon STS 3.5.2; however, acceptability of this potentially generic change will depend on demonstrating that the extra containment cooling capacity available (See Comment Number 3.6.6-3) and that the design of the Palisades containment spray pump and cooling systems are sufficiently different from the other PWR designs to warrant a unique set of ACTIONS. In addition, the following issues assume than an alternate train configuration can be established. DOC L.1 and JFD 6 define only two configurations of trains of Containment Cooling Systems that have. the capacity to mitigate the assumed accident based upon emergency power from diesel generator 1-1 and 1-2 divisions. The single failure or unavailability of either emergency power division leaves only the redundant train and there appears to be no extra capacity that can be utilized. The ITS 3.6.6 ACTION A phrase allowing "one or more trains to be inoperable has not been justified. With one train inoperable due to no emergency power from diesel generator 1-1 division and the second train inoperable due to Pump 54A, there is no cooling capacity available which is a loss of function requiring a shutdown per CTS 3.0.3. Therefore, the ITS 3.6.6 ACTION A phrase addition (of "or more") should be removed or any explanation of how two trains can be inoperable and yet 100% cooling capacity is still available needs to be provided. Comment: Provide *additional discussions and justifications for this less-restrictive change that is separate from the overall DOC. L.1. Consumers Energy Response: . As described in the response to comment 3.6.6.-14, the common header fluid systems design and the diverse, asymmetric cooling system trains result in unique combinations of equipment that can provide 100% of the cooling capacity provided by one train. Unlike most newer plants, Palisades does not have two independent containment spray piping trains. The pump motors, and electrical and pneumatic controls are arranged in separate, redundant trains, but the piping is not. The three spray pumps take suction from two separate \ headers from the SIRWT, but discharge into a single common discharge header. The flow from this header splits and goes through one of the two Shutdown Cooling Heat Exchangers (SDHX) whose outlet headers are intercoonected. The two containment spray headers come off this SDHX outlet header and flow through their own spray isolation valves and th_eir own sets of spray nozzles. (continued) . 53

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-15 Consumers Energy Response: (continued)

Therefore, spray flow from any spray pump is indistinguishable from the flow from any other; the operation of any one pump provides the same effects on the containment atmosphere as any other. From an analytical perspective, it does not matter which spray pumps are operating, or which train they are associated with. One hundred percent cooling capacity can be provided by: 1) any one fan and any two spray pumps, or by 2) the three fans powered by DG 1-1 with their associated coolers, and any one spray pump. There are numerous combinations of multiple equipment inoperabilities which would make each train by itself inoperable, but would still provide adequate containment cooling. The active components of the containment cooling train powered by DG 1-1 are: Containment Air Cooler Fan V1A Containment Air Cooler Fan V2A Containment Air Cooler Fan V3A Containment Spray Pump P54A The active components of the train powered by DG 1-2 are: Containment Spray Pump P54B Containment Spray Pump P54C Containment Air Cooler Fan V4A A similar configuration, in which individual components on opposite trains can provide the capacity of a single full capacity train is described in Specification 3.5.2, ECCS Operating. The use of a condition in which one or more trains may be inoperable is appropriate because the inoperable individual components may not effect the ability to provide the equivalent of one full train of OPERABLE components. The comment postulated a possible condition in which one train is out of service due to an inoperable diesel generator, and a pump is inoperable on the opposite train. An inoperable diesel generator does not cause associated components to be inoperable as discussed in Specification 3.8.1 and therefore the postulated example is not applicable. For additional information see the discussion of Required Actions for an inoperable diesel generator in the Bases for Specification 3.8.1. The unique flexibility of the Palisades systems provide numerous permutations and combinations of equipment that provide the required cooling system function. Therefore it is appropriate to include the "or more" addition within the description of Condition A of the proposed specification. Affected Submittal Pages: \ No page changes . 54

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-16 JFD 1 JFD 12 STS B 3.6.6B Bases APPLICABLE SAFETY ANALYSES ITS B 3.6.6 Bases APPLICABLE SAFETY ANALYSES The second paragraph, fifth sentence in STS B 3.6.6B Bases -APPLICABLE SAFETY ANALYSES states the following: "The analyses and evaluations assume ... conditions of [120]°F and [14.7] psia." ITS B 3.6.6 Bases -APPLICABLE SAFETY ANALYSES deletes this statement based on the justification (JFD 12) that it duplicates information in ITS 3.6.4 and ITS B 3.6.5. This is unacceptable.

The information provided by this statement provides useful information for the understanding of the intent of this specification, so that the operator/user does not have to go to other specifications to gain an understanding of the intent. Also, this change would be considered as a generic change which is a beyond scope of review item for this conversion. Comment: Delete this generic change. Consumers Energy Response: The Palisades safety analyses identify two limiting events for containment peak pressure and peak temperature respectively. This information is incorporated into an insert that replaces the STS Bases sentence previously deleted. JFD 12 is revised to discuss the addition of this plant specific information that parallels the information provided in the STS. Affected Submittal Pages: Att 2, ITS 3.6.6, page B 3.6.6-4 Att 5, NUREG, B 3.6-57 Att 6, JFD 3.6.6, page 4 of 5 55

  • BASES *
  • APPLICABLE SAFETY ANALYSES Containment Cooling Systems B 3.6.6 The Containment Spray System and Containment Air Recirculation System limit the temperature and pressure that could be experienced following a OBA. The limiting OBAs considered relative to containment temperature and pressure are the Loss of Coolant Accident (LOCA) and the Main Steam Line Break (MSLB). The OBA LOCA and MSLB are analyzed using computer codes designed to predict the resultant containmen+

pressure and temperature transients. No OBAs are assumed to occur simultaneously or consecutively. The postulated LOCA OBA is analyzed, in regard to containment ESF systems, assuming the loss of one ESF bus, which is the worst case single active failure, resulting in one train of Containment Cooling rendered inoperable (Ref. 6). The postulated MSLB OBA is analyzed, in regard to containment ESF systems, assuming the loss of two containment spray pumps, which is the worst case single active failure (Ref. 7). The analysis and evaluation show that under the worst case scenario, the highest peak containment pressure and the peak containment vapor temperature are within the intent of the design basis. (See the Bases for Specifications 3.6.4, "Containment Pressure," and 3.6.5, "Containment Air Temperature," for a detailed discussion.) +The analyses also . assume a response time delayed i ni ti ati on *in order to -r"'""".lc provide conservative peak calculated containment pressure and temperature responses. The external *design pressure of the containment shell is 3 psig. This value is approximately 0.5 psig greater than the maximum external pressure that could be developed if the containment were sealed during a period of low barometric pressure and high temperature and, subsequently, the containment atmosphere was cooled with a concurrent major rise in barometric pressure. The modeled Containment actuation from the containment analysis is based on a response time associated with exceeding the Containment High Pressure setpoint to achjeve full flow thrgyghethe containment spray nozzles. The spray lines within containment are maintained filled to the 735 ft elevation to provide for. rapid spray initiation. lhe System total response time of < 60 seconds' includes diesel generator startup (for loss of offsite ower , loadin of e ui ment containment spray pump s artup, an spray ine i ing. \ \ Palisades Nuclear Plant B 3.6.6-4 55-o.-01/20/98

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  • BASES BACKGROUND APPLICABLE SAFETY ANALYSES "'f""M. ro.st..\ .. h .l 1\1\SL& >>IA / S re., "".l +o c..,. ... b,;,...""" a"+ E \: F' s1Jtf'-S,

\05s ....... e .. t .. -hr.;\-c fJ. R.c\"I: 3.r..,-11'- l-3 l".J.\i 14.11.1.-l CEOG STS

  • Canta i nment 11 ng Systems ill_"i4her; c ,.d&j,Af-<lo)

B 3.6.&l.-@ Contajnment Cooling System ..(continued) " B 3.6-57 55-b 'M:r. 3.L..C.. _,'I Rev 1, 04/07/95

  • *
  • ITS Bases page B 3.6.6-4 and STS Page B 3.6-57 The analyses and evaluations considered a range of power levels and equipment configurations as described in Reference
2. The peak containment pressure case is the 0 % power MSLB with initial (pre-accident) conditions of 140°F and 16.2 psia. The peak temperature case is the 102% power MSLB with initial (pre-accident) conditions of 140°F and 15.7 psia . \1 55-c_
  • *
  • ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING SYSTEMS Change Discussion
11. NUREG-1432 includes LCO 3.6.6A for Containment Spray Systems which take credit for iodine removal and LCO 3.6.6B for Containment Spray Systems which do not take credit for iodine removal. LCO 3.6.6A, which takes credit for iodine removal, is based on the use of sodium hydroxide addition to the containment spray water as a means of iodine removal. The Palisades Nuclear Plant no longer utilizes the addition of sodium hydroxide to containment spray, but rather uses Trisodium Phosphate (TSP) in containment which is dissolved when fluid is released into containment and then recirculated by the Containment Spray System. Therefore, the proposed Palisades ITS is based on Nl,JREG-1432 3.6.6.B(Credit not taken for iodine removal by the 12. .. Containment Spray System) because the iodine removal function is facilitated by TSP as opposed to sodium hydroxide which is the model for NUREG-1432 3.6.6.A. The "B" designation is removed frqm the titles since it serves no purpose in the proposed ITS. The proposed Bases includes wording to reflect the iodine removal function provided by the Containment Spray System in conjunction with the TSP. !: .... ,,_ J.1..1.-11.

The proposed Par ades Nuclear Plan TS contains the di ussion of conta* nt pressure and te erature analysis i specification 3.6.4 d 3.6.5 respectiv

y. Therefore, to oid duplication an confusion, the de led discussion incl Cled in B 3.6.6 App cable Safety Analy sis not included*

the proposed ITS. s. 1s implied as e intent of NURE -1432 with the stat ent to see the Ba s of 3.6.4 and 3.6.5 for detailed discussio . However, N -1432 repeats so of this detailed informa on in the followin sentence which is t included in oposed Palisades ITS. 13. The Palisades Nuclear Plant analysis does not include an explicit evaluation of an inadvertent containment spray event. There is an analysis regarding the external design pressure of containment which is addressed in the proposed Bases for ITS specification 3.6.4. Therefore, the NUREG-1432 discussion of an inadvertent containment spray actuation is not included in the proposed ITS. Palisades Nuclear Plant ' ' Page 4 of S 55--d 01/20/98

  • JFD 3.6.6 -12 (replaces old version entirely)
  • 12. The Palisades safety analyses identify two limiting events for containment peak pressure and peak temperature respectively.

This information is incorporated into an insert to B 3.6.6, Applicable Safety Analyses, that replaces the STS sentence deleted. The inserted information reflects the intent of the STS bases in that the initial conditions assumed in the limiting events are described .

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  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-17 JFD T ITS B 3.5.4 Bases -ACTIONS ITS B 3.6.6 Bases -SR 3.6.6.6 and SR 3.6.6. 7 The proposed ITS B 3.6.6 Bases SR 3.6.6.6 and SR 3.6.6.7 are acceptable; however, JFD 7 presents a question pertaining to Insert #1 (page B 3.5-25) of ITS B 3.5.4 Bases -ACTION. There appears to be "operational restrictions" placed upon the containment spray pumps when the valves are aligned to the recirculation mode of ECcs*operation.

These same "operational restrictions" are not found to be discussed in ITS B 3.6.6 Bases which covers these pumps. Comment: Provide further explanation and technical justification of what "operational restrictions" are necessary for this mode of operation in this alignment. Also, provide a discussion to ensure that all the valves involved will be verified by which? identified SR to be in their correct position as required by the collective surveillance requirements of Section 3.5 and Section 3.6. Consumers Energy Response: The Containment Spray (CS) pumps at Palisades provide a required support function for the High Pressure Safety Injection (HPSI) pumps. Details of this required support function are provided in the Bases LCO section for ECCS -Operating, 3.5.2. The required alignments are directed by plant procedures, and are manually taken after the swap of HPSI pump suction from the SIRWT to the containment sump. The "operational restrictions" are merely plant procedures with descriptions of the appropriate variations of system alignment to ensure the CS and HPSI pump(s) can perform their assumed functions. A new insert to the Background of 3.6.6 is provided to describe this support function associated with the CS pump. New JFD 17 is provided to describe the addition of this plant specific information. Affected Sµbmittal f'ages: Att 2, ITS 3.6.6, page B 3.6.6-3 Att 5, NUREG, page B 3.6-56 Att 6, JFD 3.6:6, page 5 of 5 \ I 56 Containment Cooling Systems B 3.6.6 I

  • BASES 3 {p. /.; -;-7 *
  • BACKGROUND (continued)

Containment Air Recirculation and Cooling System The Containment Air Recirculation and Cooling System includes four air handling and cooling units, referred to as the Containment Air Coolers (CACs), which are located entirely within the containment building. Three of the CACs (VHX-1, VHX-2, and VHX-3) are safety related coolers and arc cooled by the critical service water. The fourth CAC (VHX-4) is not taken credit for in an OBA for maintaining containment temperature within limit, but is used during normal operation along with the three CACs to maintain containment temperature below design limits. The fan associated with VHX-4, V-4A, is in the safety analysis as assisting in the containment atmosphere mixing function. Each CAC has two vaneaxial fans with direct connected motors which draw air through the cooling coils. Both of these fans are normally in operation, but only one *fan and *motor for each CAC is rated for post OBA operation. The post-OBA rated "safety related" fan units, V-lA, V-2A, V-3A, and V-4A, serve not only to provide forced flow for the associated cooler, but also provide of the containment atmosphere. A single operating safety related fan unit will provide enough air flow to assure that there is adequate mixing of unsprayed containment areas to assure the assumed iodine removal by the containment spray. The fan units also support the functioning of the hydrogen recombiners, as discussed in the Bases for LCO 3.6.7, "Hydrogen In post accident operation following a SIS, all four Containment air coolers are designed to change automatically to the emergency mode. The CACs are automatically changed to the emergency mode by a Safety Injection Signal (SIS). This signal will trip the normal rated fan motor in each unit, open the high-capacity zJ service water discharge valve from VHX-1, , and VHX-3, - the high-capacity service wateria'charijel'valve r..; (2!2l!))VHX-4. The test to verify the service water ,valves actuate to their correct position upon receipt of an SIS

  • signal is included in the surveillance test performed as part of Specification 3.7.8, "Service Water System." The safety related fans are normally in operation and only receive an actuation signal through the OBA sequencers following an SIS in conjunction with a loss of offsite power. This actuatipn is tested by the surveillance which verifies the energizing of loads from the OBA sequencers in Specification 3.8.1, "AC Sources-Operating." Palisades Nuclear Plant B 3.6.6-3 01/20/98 \ SuPffj
  • *BASES BACKGROUND J:.,... ....

k c. ... +c.; __ S'1 s+e..:.. ; ..... '-'" .. ,;-c..+i* -.f .. _ (J .. C.o T,...:s .. ol:--i>'..*Sf-...\o& ")/ .. e +.. .-e-*.ie. \*l.:.-4. -*, *e 'f'c.l" .. .u:J ... : ... \ ........ FS f-.t.2..1 .., i./O c, *** J....;"' .... e ... + .s r'A'\ f'"'-f' s _; \\ t'l'\eet #-L pc.c.: 4-., .... ; .-e ...... e.,...ts i .... .. t' '-b?iA. cg . Contiinment Cooling Systems I (Atm.(i?pheri <;!and ouA)l-60) B 3. .containment Spray System (continued)

(5) (iAS°) Th f"c.c..;.re. ... \.,+cJ s .... -p ,., .. C.oolc.l b'1 tk .sl.. .... hl_.""' Cao\: .. \ he .. + e1'c.h. .... i (continued) CEOG STS B 3.6-56 Rev 1, 04/07/95

  • SlP.-b *
  • *
  • ITS Bases page B 3.6.6-3 and STS Page B 3.6-56 ' The containment spray pumps also provide a required support function for the High Pressure Safety Injection pumps as described in the Bases for specification 3.5.2. The High Pressure Safety Injection pumps alone may not have adequate NPSH after a postulated accident and the realignment of their suctions from the SIR WT to the containment sump. Provision is made to manually provide flow from the discharge of the containment spray pumps to the suction of the High Pressure Safety Injection pumps after the change to recirculation mode has occurred.

The additional suction pressure ensures that adequate NPSH is available for the High Pressure Safety Injection pumps .

  • *
  • Change ATTAC1ŽENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING SYSTEMS Discussion
14. NUREG-1432 requires a verification of. the cooling water flow rate to the containment air coolers every 31 days. For the Palisades Nuclear Plant design, this flow rate can only be verified during a shutdown condition due to the valve alignments involved.

Therefore, the NUREG-1432 31 day Frequency is replaced with an 18 month Frequency in the proposed ITS to Ieflect the test being performed during shutdown conditions. This change is a plant specific change to reflect the Palisades Nuclear Plant design. 15. The Palisades plant was designed prior to issuance of the General Design Criteria (GDC) in 10 CFR 50. Therefore, reference to the GDCs is omitted and appropriately replaced by reference to "Palisades Nuclear Plant design criteria . " The Palisades Nuclear Plant design was compared to the GDCs as they appeared in 10 CPR 50 Appendix A on July 7, 1971. It was this updated discussion, including the identified exemptions, which formed the original plant Licensing Basis for future compliance with the GDCs . 11oc c.,) ::S-ii= D 11 I NSE.A.1 tJC.'-41 Sf\) l ! 3,(,, <-. l 1 RAt 3.t.(,* t& Palisades Nuclear Plant Page 5 of 5 50-d 01/20/98

  • *
  • JFI>; 17 ', The Contairurient Spray pumps at Palisades provide a required support function for the High Pressure Safety Injection (HPSI) *pumps. A new insert to the Background of 3.6.6 is provided to describe the HPSI pump support function associated with the Containment Spray pumps . " -e...
  • *
  • CONVERSION TO IMPF<OVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUES"f FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-18 JFD 15 STS B 3.6.6. 8 Bases -APPLICABLE SAFETY ANALYSES ITS 8 3.6.6 Bases -APPLICABLE SAFETY ANALYSES The sixth paragraph of STS B 3.6.6B Bases APPLICABLE SAFETY ANALYSES is deleted in its entirety in ITS 8 3.6.6 Bases -APPLICABLE SAFETY ANALYSES.

The justification used for this deletion JFD 15 discusses the replacement of the General Design Criteria (GDC) in 1 O CFR 50 with reference to plant specific design critical. This paragraph has nothing to do with the GDC. Comment: Provide a discussion and justification for the deletion of this paragraph. Consumers Energy Response: A new JFD (JFD 18) has been provided to discuss the deviation from the sixth paragraph in the Applicable Safety Analyses Bases discussion for ISTS 3.6.6. Affected Submittal Pages: Att 2, ITS 3.6.6, page 8 3.6.6-4 Att 5, NUREG 3.6.6, page 8 3.6-58 Att 6, JFD 3.6.6, page 5 of 5 57

  • BASES *
  • APPLICABLE SAFETY ANALYSES Containment Cooling Systems B 3.6.6 The Containment Spray System and Containment Air Recirculation System limit the temperature and pressure that could be experienced following a OBA. The limiting DBAs considered relative to containment temperature and pressure are the Loss of Coolant Accident (LOCA} and the Main Steam Line Break (MSLB). The OBA LOCA and MSLB are analyzed using computer codes designed to predict the resuliant containmen+

pressure and temperature transients. No DBAs are assumed to occur simultaneously or consecutively. The postulated LOCA DBA is analyzed, in regard to containment ESF systems, assuming the loss of one ESF bus, which is.the worst case single active failure, resulting in one train of Containment Cooling rendered inoperable (Ref. 6). The postulated MSLB OBA is analyzed, in regard to containment ESF systems, assuming the loss of two containment spray pumps, which is the worst case single active failure (Ref. 7) . . The analysis and evaluation show that under the worst case scenario, the highest peak containment pressure and the peak containment vapor temperature are within the intent of the design basis. (See the Bases for Specifications 3.6.4, "Containment Pressure," and 3.6.5, "Containment Air Temperature," for a detailed discussion.) +The analyses also assume a response time delayed initiation

  • in order to -r .... provide conservative peak calculated containment pressure and temperature responses.
  • The external design pressure of the containment shell is 3 psig. This value is 0.5 psig greater than the maximum external pressure that could be developed if the containment were sealed during a period of low barometric pressure and high temperature and, subsequently, the containment atmosphere was cooled with a concurrent major rise in barometric pressure.

The modeled Containment actuation from the containment analysis is based on a response time associated with exceeding the Containment H1gh Pressure setpoint to achieye fyll flow thrqygh.,the containment spray nozzles. The spray lines within containment are maintained filled to the 735 ft elevation to provide for rapid spray initiation. lhe System total response time of < 60 seconds includes diesel generator startup (for loss of offsite ower , loadin of e ui ment containment spray pump s artup, an spray ine l 1ng. ' I Palisades Nuclear Plant B 3.6.6-4 01/20/98 -. ' 51-Q

  • *
  • i:::-<.Ae.

C... l. I *sASES APPLICABLE SAFETY ANALYSES (continued) .spre.1 c.. .... ... I cJ +o e I e. ve-J,* .. ..._ 71 S-f+ +r, f ........ +;,,.

... ""' . -n*a .. pu-+.. ... ,,_ ........ .h,;,.... ros-.\- .,.,,s. J.*-k 1 C.o-. *-** ; "' R ... u. 4

  • LCO CEOG STS
  • The model Containment Cool g System actuation from the containm t analysis is bas on the unit. speci ic response time as ciated with excee ng the CCAS. to ach eve full Contai ent Coolfog Syste air and safety gr e cooling water low. The Containment Spray System and the Containment Cooling System satisfy Criterion 3 of Policy/SJ;atementi. c.i::z. ?(p [c.,.)ft)

{continued)

  • , B 3.6-58 Rev 1, 04/07/95 51-b
  • *
  • ATIAC.Hl\.1ENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND .COOLING SYSTEMS Djscussjon
14. NUREG-1432 requires a verification of. the cooling water flow rate to the containment air coolers every 31 days. For the Palisades Nuclear Plant design, this flow rate can only be verified during a shutdown condition due to the valve alignments involved.

Therefore, the NUREG-1432 31 day Frequency is replaced with an 18 month Frequency in the proposed ITS to ieflect the test being performed during shutdown conditions. This change is a plant specific change to reflect the Palisades Nuclear Plant design. 15. The Palisades plant was designed prior to issuance of the General Design Criteria (GDC) in 10 CFR 50. Therefore, reference to the GDCs is omitted and appropriately replaced by reference to "Palisades Nuclear Plant design criteria . " The Palisades Nuclear Plant design was compared to the GDCs as they appeared in 10 CFR 50 Appendix A on July 7, 1971. It was this updated discussion, including the identified exemptions, which formed the original plant Licensing Basis for future compliance with the GDCs . 'I:"w..4'- Mc.) ::s-r: o n ,..,c.\,v l ca (V.'J: !. '-* c.

  • l 1 RA\

18 Palisades Nuclear Plant Page S of S 01/20/98

  • *
  • JED 18 The sixth paragraph in the Applicable Safety Analyses discussion has been deleted since this information has been adequately addressed in the fourth paragraph of the Applicable Safety Analyses discussion.

As stated in the Background discussion, the "Containment Air Recirculation and Cooling System" and the "Containment Spray System" are generally addressed as the "Containment Cooling Systems." Since the assumed response times in the containment analysis are the same for the containment air coolers as they are for the containment spray pumps (following a loss of offsite power), the discussion in the Applicable Safety Analyses has been revised to address the generic term "Containment Cooling Systems." For event without a loss of offsite power, no delay time is assumed for the containment air coolers since they are already in operation . ,, 51-d

  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-19 JFD 15 STS B 3.6.6B Bases -References ITS B 3.6.6 Bases -References In STS B 3.6.6B Bases -REFERENCES, Reference 1 lists the 10 CFR 50 Appendix A, GDC applicable to the Containment Spray and Cooling Systems. ITS B 3.6.6 Bases -REFERENCES, replaces the GOC with FSAR Section 5.1. This change has no J FD associated with it. The appropriate JFD should be JFD 15. Comment: Correct this discrepancy.

Consumers Enerqv Response: Page B 3.6-64 of the STS markup has been annotated to reflect JFD 15 as applicable to the change to Reference

1. Affected Submittal Pages: Att 5, NUREG, B 3.6-64 58
  • *
  • BASES SURVEILLANCE REQUIREMENTS REFERENCES

... S-.1 CEOG STS

  • Containment Cooling Systems/ (Atll)(S'spheric iid
3. 6. -:-\\) SR (continued) experience.

See SR and SR above, for* further discussion of .the basis for the Har month Frequency. Ii P SB 3. 6. 61.9 __,, . With the containment spray inlet valves closed and the spray header dJ11ined of any solution, low pressure air or smoke can be blown through test connections. Performance of this SR demonstrates that each spray nozzle is unobstructed and provides assurance that spray coverage of the containment during an accident is not degraded. Due to the assive design of the nozzle, a test at t e rst r ue 1n d at /f'1 10 year intervals fs considered a equa e o e ec \,.2) obstruction of the spray nozzles. 10 CFR 0, Appendix A, GD 38, GDC 39, GDC 4 , GDC 41, GDC and GDC 43. 2. FSAR, Section 1 3. FSAR, Sectiorf tf{..ty. l 4. / FSAR, Sec/tion C-* 1]. (g)@. FSAR, Section fQ.11; \ @© ASHE, Boiler and Pressure Vessel Code, Section XI. l... 1, 'l. FSA't, ltt.11,\-3 f= .. A-2., 'lc..\.l<. If. i-1 F'Siil?., l..l-\4 C\-\ '. B 3.6-64 5'8-0-I Bev 1, 04/07/95

  • * * ------------CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.6-20 CTS 4.6.5 ITS 3.6.6 and Associated Bases ITS 3.7.8.2 CTS 4.6.5.a requires emergency mode automatic valve operation to be checked for OPERABILITY during each refueling shutdown.

CTS 4.6.5.b requires the exercising of each Containment Air Cooling System/Service Water system valve at a frequency not to exceed three months. There is no justification for relocating these SRs to ITS 3.7.8, nor is there a ITS SR 3.7.8.2 or ITS 3.6.6 SR for exercising the valves every three months. See Comment Numbers 3.6.6-8 and 3.6.6-21. Comment: Revise the CTS markup and provide a discussion and justification for this change. See Comment Numbers 3.6.6-8 and 3.6.6-21. Consumers Energy Response: CTS 4.6.5.a, on CTS page 4-25 in Attachment 3 of the 3.6 submittal includes the words "and

  • valve" encircled and is noted with "see 3. 7" in carats. This is not a relocation, but a referential annotation to show the location in the overall ITS submittal where this requirement applicable lo the Service Water system is addressed

-at the CTS page markup for ITS 3.7.8 . Attachment 3 of the 3.7 submittal, also includes CTS page 4-25 in section 3.7.8 however with annotation that indicates that the substance of the requirement associated with the "and valve" portion of the SR is addressed by DOC A.8 to Specification

3. 7.8. DOC A.8 indicates that this testing requirement is established in ITS 5.5.7, "lnservice Testing Program." As described in DOC A.8, the frequency in the lnservice Testing Program applicable to these components is once per 92 days, i.e. equivalent to the CTS requirement of once every three months. Based on this, the CTS requirement is appropriately addressed in the submittal.
  • Affected Submittal Pages: No page changes . 59
  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.6-21 CTS 4.6.5.a ITS SR 3.6.6.8 CTS 4.6.5.a requires emergency mode automatic valve and fan operation will be checked for OPERABILITY during each refueling shutdown.

The corresponding SRs are ITS SR 3.6.6.8 for the fans and the appropriate SR in ITS 3.7.8 for the valves. ITS SR 3.6.6.8 requires verifying that the containment cooling fan starts on an actual or simulated actuation signal. The CTS markup does not indicate whether "emergency mode operation" is initiated by an actual or simulated actuation signal or just a simulated actuation signal. Comment: . Revise the CTS markup and provide the appropriate discussion and justification for this change. Consumers Energy Response: The CTS requirement to verify valve and fan emergency mode OPERABILITY does not provide

  • any detail of how the check for OPERABILITY is to be performed.

However, implicit in the CTS requirement is the need to cause their operation in the emergency mode. Any actuation of the emergency mode must either be initiated by an actual or simulated actuation signal, no other alternative exists. This is not a change from the CTS requirement to cause their operation in the emergency mode, without any specific direction of how to initiate the operation. Again the alternatives are either actual or simulated actuation signals. Obversely, if the ITS requirement directed eith 1 er an actual or simulated actuation be specifically used, the change would be a more restrictive change because it would limit the way the operation was to be initiated. Based on the fact that any actuation must be either actual or simulated, and the CTS does not restrict the existing actuation, the ITS use of actual or simulated is not a change from the CTS i;.. requirement. Therefore, no changes are required. Affected Submittal Pages:. No page changes. ' \ 60

  • **
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION . ' SECTION 3.6, CONTAINMENT NRC REQUEST: 3.6.6-22 CTS 3.4.2 CTS 3.4.3 ITS 3.6.6 ACTION A and Associated Bases CTS 3.4.2 and CTS 3.4.3 provide a 7-day allowed outage time for when one component is inoperable and a 24 hour allowed outage time when two components are inoperable.

ITS 3.6.6 ACTION A permits one or more trains to be inoperable concurrently with one inoperable Component Cooling Water (CCW) train and one inoperable Service Water System (SSW) train. ITS 3.6.6, ITS 3.7.7 and ITS 3.7.8 now permit seven components from CTS 3.4.1.a to be inoperable or five components from CTS 3.4.1.b to be inoperable. There is no DOC provided which identifies and describes the Less Restrictive (L) change that permits the concurrent degraded condition of CCW and SWS with ITS 3.6.6. Comment: Provide a discussion arid justification for this Less Restrictive (L) change. . Consumers Eneaz.y Re19onse:

  • -DOC L.1 to 3.6.6 specifically addresses this concern in the discussion provided at the top of the second page containing the L.1 DOC. The Service Water and Component Cooling Water systems provide a support function to the containment cooling systems. As described in the L.1 discussion, this support function is appropriately addressed in ITS format through the individual system specifications, and the application of the Safety Function Determination Program (SFDP). The SFDP will provide the controls over system and train interactions that may jeopardize the* ability to satisfy individual safety functions.

The L.1 discussion documents this change as a less restrictive change: Affected Submittal Pages: No page changes. \ \ 61

  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS*

RESPONSE ro JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-23 CTS 4.6.5.a ITS SR 3.6.6.8 and Associated Bases ITS SR 3.7.8.2 and Associated Bases CTS 4.6.5.a checks the Containment Air Cooler (CAC) emergency mode automatic valve and fan operation and function at refueling. ITS SR 3.6.6.8 verifies each "containment cooling" fan starts on an actual or simulated signal. The requirements of CTS 4.6.5.a are fulfilled by ITS SR 3.6.6.8 and ITS SR 3.7.8.2; however, wording changes should be made to ensure there is equivalency. In ITS SR 3.6.6.8, the actuation signal acts on the "containment air coolers trains or units" not just the "cooling fans". The action to be verified is the closing of the high.,.capacity discharge valve from the non-safety CAC VHX-4; opening of the high-capacity discharge valves from the safety related CAC VHX-1, 2, and 3; and tripping the normal rated fan motor in each unit. This action is not recognized by the ITS SR 3.6.6.8 or stated in the Bases to ITS SR 3.6.6.8. See Comment Number 3.6.7-2. Comment: Revise the ITS markup and provide additional discussion and justification for this change. See Comment Numbers 3.6.6-24 and 3.6.7-2. Consumers Energy Response: The emergency actuation of the containment air cooler fans at the Palisades plant only occurs upon receipt of a signal from the Design Basis Accident (OBA) sequencer; they are not directly started by either a Containment High Pressure or Safety Injection signal.\ This is discussed in the Bases Background and LCO sections for Specification 3.6.6. SR 3.6.6.8 (and SR 3.8.1.10) requires verification that the safety related fans are appropriately actuated by the OBA sequencer. The automatic trip of the normal rated containment air cooler fans occurs as a result of a Safety Injection signal. This actuation (trip) is a required part of the Safety Injection logic as described by Specification 3.3.4, "ESF Logic and Manual Initiation" and verified in accordance with SR 3.3.4.3. The alignment of the Service Water system valves to their appropriate emergency mode position is addressed explicitly in SR 3.7.8.2 that requires verification that each automatic valve "actuates to the correct position." Based on the above, the ITS ensures that all three sets of required automatic actuations occur; the safety related fans are automatically initiated by the OBA sequencer (SR 3.6.6.8 and SR 3.8.1.10), the normal rated fans are automatically tripped by a SIS (SR 3.3.4.3), and the Service Water valves are automatically aligned to the correct position (SR 3.7.8.2). Affected Submittal Pages:

  • No page changes. 62
  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION SECTION 3.6, CONTAINMENT NBC REQUEST: 3.6.6-24 CTS 4.6.5.b ITS SR 3.6.6.2 and Associated Bases CTS 4.6.5.b checks the Containment Air Cooler (CAC) fan operation every three months. ITS SR 3.6.6.2 operates each containment air cooler fan unit for 15 minutes every 31 days. It is unclear from the wording of ITS SR 3.6.6.2 and its associated Bases as to whether both fans in each CAC unit is verified OPERABLE for a total of eight or just the safety related fans. The CTS wording implies that only the safety related fans are tested. See Comment Number 3.6.7-2. Comment: Revise the ITS markup and provide a discussion and justification for this change. See Comment Number 3.6.7-2. Consumers Energy Response:

The words "safety related" have been inserted into the first sentence of the Bases for SR 3.6.6.2. The sentence now reads "Operating each safety related Containment Air Cooler fan unit for .... " The JFD applicable to the STS markup remains appropriate for the additional words . Affected Submittal Pages: Att 2, ITS 3.6.6, page 8 3.6.6-7 Att 5, NUREG page 8 3.6-62 ,_._; 63 .

  • BASES *
  • SURVEILLANCE REQUIREMENTS SR 3.6.6.1 Containment Cooling Systems B 3.6.6 Verifying the correct alignment for manual, power operated, and automatic valves, excluding check valves, in the Containment Spray System provides assurance that the proper flow path exists for Containment Spray System operation.

This SR also does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct positions prior to being secured. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation. Rather, it involve$ verification, through a system walkdown, that those valves outside containment and capable of potentially being mispositioned, are in the correct position. RAI: 3.C...t..-1..'-\ SR 3.6.6.2 . _L_f Operating Air Cooler fan unit for 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly. The 31 day Frequency was developed considering the known reliability of the fan units and controls, the two train redundancy available, and the low probability of a significant degradation of the containment cooling train occurring between surveillances. SR 3.6.6.3 Verifying the containment spray header is full of water to the 735 ft elevation minimizes the time required to fill the header. This ensures that spray flow will be admitted to the containment atmosphere within the time frame assumed in the containment analysis. The 31 day Frequency is based on the static nature of the fill header and the low probability of a significant degradation of the water level in the piping occurring between surveillances

  • Palisades Nuclear Plant B .3.6.6-7 63-0--01/20/98
  • I ceLr * * /£) Containment Cooling Systems l(Ati'll6soheric B

BASES SURVEILLANCE REQUIREMENTS SR 3.6.sj.I those valves outside containment and capable of potentially being mispositioned, are in the correct position. . SR 3.6.fi,2. I V li c ..... + ... :--e-+ /F,. c-Je.r . ==-tSperat i ng each 1Contjl1 omen Ve pp I 1ngltra1 n*: fan unit for 3.c,.1.-214 15 minutes ensures that all trains are OPERABLE and that all asso,iated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. The 31 day Frequency was developed considering the known reliability of the fan units and controls, the two train redundancy available, and the low probability of a significant degradation of the containment cooling train occurring between surveillances. -/oJ./e 1-I SR M,611,r4 @ (D /rYJ water flow rate to@ HcOG'l ingitJ provides assurance the design fl,gw rate assumed in the safety analyses will be achieved Also considered in selectin this Frequency were the known rel hb111ty of the

,tYste11, the <t'lii¥tri}ft CEOG STS
  • redundancy, and thM low probability of a significant degradation of flow occurring between surveillances.

SB 3.6.6f1.K'3'@ . © l 1 -.1.it. v""" the containment spray header is full of water to the ... ir:ft 111 n m zes the time required to fil 1 the header. This ensures that spray flow will be admitted to the containment atmosphere within the time frame assumed in the containaient analysis *. The 31 day Frequency is based on the static nature of the fill header and the low probability of a significant degradation of the water level in the piping occurring between surveillances. (continued) \ B 3.6-62 . Rev 1, 04/07/95

  • *
  • ENCLOSURE 2 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS PARTICAL REPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.6
  • *
  • CONVERSION TO IMPROVED TECHNICAL SPECIFICATIONS RESPONSE TO JANUARY 26, 1999 REQUEST FOR ADDITIONAL INFORMATION REVISED PAGES FOR SECTION 3.6 *Page Change Instructions Due to the number of ITS Section 3.6 pages affected by our response to the NRC staff comments, replacement pages have been provided for all of SECTION 3.6. REMOVE PAGES INSERT PAGES REV DATE NRC COMMENT# ATTACHMENT 1 TO ITS CONVERSION SUBMITTAL All pages of Attachment 1 are replaced.

05/31/99 ATTACHMENT 2 TO ITS CONVERSION SUBMITTAL All pages of Attachment 2 are replaced. 05/31/99 ATTACHMENT 3 TO ITS CONVERSION SUBMITTAL All pages of Attachment 3 are replaced. 05/31/99 ATTACHMENT 4 TO ITS CONVERSION SUBMITTAL All pages of Attachment 4 are replaced. 05/31/99 ATTACHMENT 5 TO ITS CONVERSION SUBMITTAL All pages of Attachment 5 are replaced. 05/31/99 ATTACHMENT 6 TO ITS CONVERSION SUBMITTAL All pages of Attachment 6 are replaced. 05/31/99 ,. 3.6 CONTAINMENT SYSTEMS *

  • 3.6.1 Containment LCO
  • 3. 6 .1 Conta.i nment sha 11 be OPERABLE.

APPLICABILITY: MODES l, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. Containment A.1 Restore containment inoperable. to OPERABLE status. B. Required Action and B.1 Be in MODE 3. associated Completion Time not met . AND B.2 . Be in MODE 5. SURVEILLANCE REQUIREMENTS SR 3. 6 .1.1 SURVEILLANCE Perform required visual examinations and Type A leakage rate testing in accordance with the Containment Leak Rate Testing Program. Containment 3.6.1 COMPLETION TIME 1 hour 6 hours 36 hours

  • FREQUENCY In accordance with the Containment Leak Rate Testing Program Palisades Nuclear Plant 3.6.1-1 Amendment No. 05/31/99
  • *
  • SURVEILLANCE REQUIREMENTS SR 3. 6 .1.2 SR 3. 6 .1. 3 SURVEILLANCE Verify containment structural integrity in accordance with the Containment Structural Integrity Surveillance Program.

Loca l leak rate tests shall be performed at 55 psi g.

  • Perform required Type B and C leakage rate testing, except for containment air lock testing, in accordance with 10 CFR 50, Appendix J, Option A, as by approved exemptions.

The leakage rate acceptance criterion is 1.0 La. However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakage rate acceptance criteria are < 0.6 La for the Type B and Type C tests

  • Containment 3.6.1 FREQUENCY In accordance with the Containment Structural Integrity Surveillance Program ------NOTE--'---*

SR 3.0.2 is not applicable. In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions Palisades Nuclear Plant 3.6.1-2 Amendment No. 05/31/99 I i

  • I
  • *
  • 3.6 CONTAINMENT SYSTEMS 3.6.2 Containment Air Locks Containment Air Locks 3.6.2 LCO 3.6.2 Two containment air locks shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS -------------------------------------NOTES------------------------------------

1. Entry and exit is permissible through a 11 locked 11 air lock door to perform repairs on the affected air l.ock components.
  • 2. Separate Condition entry is allowed for each. air lock. 3. Enter applicable Conditions and Reqtiired Actions of LCO 3.6.1, 11 Containment, 11 when leakage results in exceeding the overall containment leakage rate acceptance criteria.
  • CONDITION A. One or more containment air locks with one containment . air lock door inoperable.

Palisades Nuclear Plant REQUIRED ACTION

1. Required Actions A.1, A.2, and A.3 are not
  • app l i cable if both doors in the same air lock are inoperable and Condition C is entered. 2. Entry and exit is permissible for 7 days under administrative controls if both air locks are inoperable.

COMPLETION TIME (continued) 3.6.2-1 Amendment No. 05/31/99

  • ACTIONS CONDITION A. (continued)
  • B. One or more containment air locks with containment air lock interlock mechanism inoperable.
    • Palisades Nuclear Plant A.1 AND A.2 AND A.3 Containment Air Locks 3.6.2 REQUIRED ACTION COMPLETION TIME Verify the OPERABLE 1 hour door is closed in the affected air lock. Lock the OPERABLE 24 hours door closed in the affected air lock. --------NOTE---------

Air lock doors in high radiation areas may be verified locked closed by administrative means. --------------------- Verify the OPERABLE Once per 31 days door is locked closed in the affected air 1 ock. ------------NOTES------------

1. Required Actions B.l, B.2, and B.3 are not applicable if both doors in the same air lock are inoperable and Condition C is entered. 2. Entry and exit of containment is permissible under the control of a dedicated individual.

(continued) 3.6.2-2 Amendment No. 05/31/99

  • ACTIONS CONDITION B. (continued)

B.1 AND B.2 AND B.3

  • c. One or more C.1 containment air locks inoperable for reasons other than Condition A or B. AND C.2 AND C.3
  • Palisades Nuclear Plant Containment Air Locks 3.6.2 REQUIRED ACTION COMPLETION TIME Verify an OPERABLE 1 hour door is closed in the affected air lock. Lock an OPERABLE door 24 hours closed in the affected air lock. --------NOTE---------

Air lock doors in high radiation areas may be verified locked closed by administrative means. --------------------- Verify an OPERABLE Once per 31 days door is locked closed in the affected air lock. Initiate action to Immediately evaluate overall containment leakage rate per LCO 3.6.1. Verify a door is 1 hour closed in the affected air lock. Restore air lock to 24 hours OPERABLE status . 3.6.2-3 Amendment No. 05/31/99

  • ACTIONS CONDITION D. Required Action and associated Completion Time not met. *
  • Palisades Nuclear Plant REQUIRED ACTION D.1 Be in MODE 3. AND D.2 Be in MODE 5. 3.6.2-4 Containment Air Locks 3.6.2 COMPLETION TIME 6 hours 36 hours Amendment No. 05/31/99 Containment Air Locks 3.6.2
  • SURVEILLANCE REQUIREMENTS
  • SR 3.6.2.1 SURVEILLANCE

NOTES-------------------

1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. 2. Results shall be evaluated against acceptance criteria of SR 3.6.1.3 in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions.
3. A seal contact check shall be performed on the emergency escape air lock following each full pressure test. Emergency escape air lock door opening, solely for-the purpose of strongback removal and performance of the seal contact check, does not necessitate additional pressure testing . 4. Local leak rate tests, other than personnel air lock doors, shall be performed at 55 psig. Perform required air lock leakage rate testing in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.

The acceptance criteria for air lock testing are: a. Overall air lock leakage rate is 1. 0 La when tested at Pa and combined with all penetrations and valves subjected to Type B and C tests. However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakage rate acceptance criteria is < 0.6 La when combined with all penetrations and valves subjected to Type B and C tests. FREQUENCY SR 3.0.2 is not applicable In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions (continued) Palisades Nuclear Plant 3.6.2-5 Amendment No. 05/31/99

  • *
  • Containment Air Locks 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1 (continued)

SR 3.6.2.2 b. For each personnel air lock door, leakage rate is 0.023 La when tested at;::: 10.0 psig. c. An acceptable emergency escape air lock door seal contact check consists of a verification of continuous contact between the seals and the sealing surfaces. Verify only one door in the air lock can be 18 months opened at a time . Palisades Nuclear Plant 3.6.2-6 Amendment No. 05/31/99

  • *
  • 3.6 CONTAINMENT SYSTEMS 3.6.3 Containment Isolation Valves Containment Isolation Valves 3.6.3 LCO 3.6.3 Each containment isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS -------------------------------------NOTES------------------------------------

1. Penetration flow paths, except for 8 inch purge exhaust valves and 12 inch air room supply valves penetration flow paths, may be unisolated intermittently under administrative controls.
2. Separate Condition entry is allowed for each penetration flow path. 3. Enter applicable Conditions and Required Actions for system(s) made inoperable by containment isolation valves. 4. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," when leakage results in exceeding the overall containment leakage rate acceptance criteria . CONDITION A. ---------NOTE---------

A.1 Only applicable to penetration flow paths with two containment isolation valves. One or more penetration flow paths with one containment isolation valve inoperable (except for purge exhaust valve or AND air room supply valve not locked closed). Palisades Nuclear Plant REQUIRED ACTION COMPLETION TIME Isolate the affected 4 hours penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured. (continued) 3.6.3-1 Amendment No. 05/31/99

  • ACTIONS CONDITION A. (continued)

A.2 '

  • B. ---------NOTE---------

B.1 Only applicable to penetration flow paths with two containment isolation valves. ---------------------- One or more penetration flow paths with two containment isolation valves inoperable (except for purge exhaust valve or air room supply valve not locked closed) .

  • Palisades Nuclear Plant Containment Isolation Valves 3.6.3 REQUIRED ACTION COMPLETION TIME --------NOTE---------

Isolation devices in high radiation areas may be verified by use of administrative means. --------------------- Verify the affected Once per 31 days penetration flow path for isolation is isolated. devices outside containment "-AND Prior to entering* MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment Isolate the affected 1 hour penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. 3.6.3-2 Amendment No. 05/31/99

  • ACTIONS CONDITION
c. ---------NOTE---------

C.1 Only applicable to penetration flow paths with on 1 y one containment isolation valve and a closed system. ---------------------- One or more AND penetration flow paths with one containment C.2 isolation valve inoperable.

  • D. One or more purge D.l exhaust or air room supply valves not locked closed. E. Required Action and E.1 associated Completion Time not met. AND E.2
  • Palisades Nuclear Plant Containment Valves 3.6.3 REQUIRED ACTION COMPLETION TIME Isolate the affected 72 hours penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or bl ind flange. --------NOTE---------

Isolation devices in high radiation areas may be verified by use of administrative means. --------------------- Verify the affected Once per 31 days penetration flow path i s i so 1 a ted. Lock closed the 1 hour affected valves. Be in MODE 3. 6 hours ( I Be in MODE 5. 36 hours 3.6.3-3 Amendment No. 05/31/99

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  • Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SR 3.6.3.1 SR 3.6.3.2 SR 3.6.3.3 SURVEILLANCE FREQUENCY Verify each 8 inch purge valve and 12 inch 31 days air room supply valve is locked closed.

Va l ves and blind flanges in high radiation areas may be verified by use of administrative means. ) ------------------------------------------- Verify each manual containment isolation valve and blind flange that is outside containment and not locked, sealed, or otherwise secured in position, and is required to be closed during accident conditions, is closed, except for containment isolation valves that are open under administrative controls * -------------------NOTE-------------------- Va l ves and blind flanges in high radiation areas may be verified by use of administrative means. Verify each manual containment isolation valve and blind flange that is located inside containment and not locked, sealed, or otherwise secured in position, and required to be closed during accident conditions, is closed, except for containment isolation valves that are open under administrative controls . / 31 days Prior to entering MODE 4* from MODE 5 if not performed within the previous 92 days Palisades Nuclear Plant 3.6.3-4 Amendment No. 05/31/99

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  • Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SR 3.6.3.4 SR 3.6.3.5 SR 3.6.3.6 SURVEILLANCE Verify the isolation time of each automatic power operated containment iso'lation valve is within limits. Verify each containment 8 inch purge exhaust and 12 inch air room supply valve is closed by performance of a leakage rate test. Verify each automatic containment isolation valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal . FREQUENCY In accordance with the Inservice Testing Program 184 days 18 months .Palisades Nuclear Plant 3.6.3-5 Amendment No. 05/31/99
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  • 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure Containment Pressure 3.6.4 LCO 3.6.4 Containment pressure shall be 1.0 psig in MODES 1 and 2 and 1.5 psig in MODES 3 and 4. APPLICABILITY:

MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. Containment pressure A. l Restore containment not within limit. pressure to within limit. B. Required Action and B.1 Be in MODE 3. associated Completion Time not met . AND B.2 Be in MODE 5. SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.6.4.1 Verify containment pressure is within limit. COMPLETION TIME 1 hour 6 hours 36 hours FREQUENCY 12 hours Paltsades Nuclear Plant 3.6.4-1 Amendment No. 05/31/99 Containment Air Temperature 3.6.5

  • 3.6 CONTAINMENT SYSTEMS *
  • 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be 140°F. APPLICABILITY:

MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. Containment average A.1 Restore containment air temperature not average air within limit. temperature to within limit. B. Required Action ana B.1 Be in MODE 3. associated Completion Time not met. AND B.2 Be in MODE 5. SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.6.5.1 Verify containment average air temperature is within limit. COMPLETION TIME 8 hours 6 hours 36 hours FREQUENCY 24 hours Palisades Nuclear Plant 3.6.5-1 Amendment No. 05/31/99 Containment Cooling Systems 3.6.6

  • 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Cooling Systems LCD 3 .. 6 .6 Two containment cooling trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore containment 72 hours containment cooling cooling train to trains inoperable. OPERABLE status. AND At least 100% of the cooling capability

  • equivalent to a single OPERABLE containment cooling train. B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE A. 30 hours
  • Palisades Nuclear Plant 3.6.6-1 Amendment No. 05/31/99
  • *
  • Containment Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS SR 3.6.6.1 SR 3.6.6.2 SR 3.6.6.3 SR 3.6.6.4 SR 3.6.6.5 SR 3.6.6.6 SURVEILLANCE FREQUENCY Verify each containment spray manual, power 31 days operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

Operate each Containment Air Cooler Fan 31 days Unit for 15 minutes. Verify the containment spray piping is full 31 days of water to the 735 ft elevation in the containment spray header . Verify total service water flow rate, when aligned for accident conditions, is 3935.5 gpm to Containment Air Coolers VHX-1, VHX-2, and VHX-3. Verify each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head. Verify each automatic containment spray valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to its correct position on an actual or simulated actuation signal . 18 months In accordance with the Inservice Testing Program 18 months Palisades Nuclear Plant 3.6.6-2 Amendment No. 05/31/99

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  • Containment Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS SR 3.6.6.7 SR 3.6.6.8 SR 3.6.6.9 SURVEILLANCE Verify each containment spray pump starts automatically on an actual or simulated actuation signal. Verify each containment cooling fan starts automatically on an actual or simulated actuation signal. Verify each spray nozzle is unobstructed.

FREQUENCY 18 months 18 months 10 years Palisades Nuclear Plant 3.6.6-3 Amendment No. 05/31/99

  • _ 3.6 CONTAINMENT SYSTEMS 3.6.7 Hydrogen Recombiners Hydrogen Recombiners 3.6.7 LCO 3.6.7 Two hydrogen recombiners shall be OPERABLE.

APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One hydrogen A.1 --------NOTE--------- recombiner inoperable. LCO 3.0.4 is not applicable.

Restore hydrogen 30 days recombiner to OPERABLE status. -' B. Required_ Action and B.1 Be in MODE 3. 6 hours associated Completion Ti me not met .

  • Palisades Nuclear Plant 3.6.7-1 Amendment No. 05/31/99
  • SURVEILLANCE REQUIREMENTS SR 3.6.7.1 SR 3.6.7.2 SR 3.6.7.3 *
  • SURVEILLANCE Perform a system functional test for each hydrogen recombiner.

Visually examine each hydrogen recombiner enclosure and verify there is no evidence of abnormal conditinns. Perform continuity and a resistance to ground test for each heater phase . Hydrogen Recombiners 3.6.7 FREQUENCY 18 months 18 months 18 months Palisades Nuclear Plant 3.6.7-2 Amendment No. 05/31/99

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  • Containment B 3.6.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.l Containment BASES BACKGROUND The containment consists of a concrete structure lined with steel plate, and the penetrations through this structure.

The structure is designed to contain radioactive* material that may be released from the reactor core following a Loss of Coolant design basis accident. Additionally, this structure provides shielding from the fission products that may be present in the containment atmosphere following accident conditions. The containment is a reinforced concrete structure with a cylindrical wall, a flat foundation mat, and a shallow dome roof. The foundation slab is reinforced with conventional mild-steel reinforcing. The internal pressure loads on the base slab are resisted by both the external soil pressure and the strength of the reinforced concrete slab. The cylinder wall is prestressed with a post tensioning system in the vertical and horizontal directions. The dome roof is prestressed utilizing a three way post tensioning system . The inside surface of the containment is lined with a carbon steel liner to ensure a high degree of leak tightness during operating and accident conditions.

  • The concrete structure is required f9r structural integrity of the containment under Design Basis Accident (OBA) conditions.

The steel liner and its penetrations establish the leakage limiting boundary of the containment. Maintaining the containment OPERABLE limits the leakage of fission product radioactivity from the containment to the environment .. SR 3.6.1.1 and SR 3.6.1.3 leakage rate requirements comply with 10 CFR 50, Appendix J, Option B for Type A tests and Option A for Type B and C test.s, as modified by approved exemptions. The isolation devices for the penetrations in the containment boundary are a part of the containment leak tight barrier. To maintain this leak tight barrier: a. All penetrations required to be closed during accident conditions are either: 1. capable of being closed by an OPERABLE automatic containment isolation system, or Palisades Nuclear Plant B 3.6.1-1 05/31/99

  • *
  • BASES BACKGROUND (continued)

APPLICABLE SAFETY ANALYSES Containment B 3.6.1 2. closed by manual valves, blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in LCO 3.6.3, "Contairiment Isolation Valves"; b. Each air lock is OPERABLE, except as provided in LCO 3.6.2, "Containment Air Locks"; c. The equipment hatch is properly closed and sealed. The safety design basis for the containment is that the containment must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate. The DBAs that result in a release of radioactive material within containment are a Loss of Coolant Accident (LOCA), a Main Steam Line Break (MSLB), and a control rod ejection accident (Ref. 1). In the analysis of each of these . accidents, it is assumed that containment is OPERABLE such that release of fission products to the environment ts controlled by the rate of-containment leakage. The containment was designed with an allowable leakage rate of 0.10% of containment air weight per day (Ref. 3). This leakage rate is defined in 10 CFR 50, Appendix J, as La: the maximum allowable containment leakage rate at the ca1culated maximum peak containment pressure (Pa) of 53 psig, which. results from the design basis LOCA. (Ref. 2). For the Palisades Nuclear Plant, the calculated maximum peak containment pressure results from a MSLB accident. However, since the limiting accident from an offsite dose perspective is a LOCA, this pressure is used as Pa. The Pa value of 53 psig represents the value found in Reference 1, rounded up to the next whole number. Satisfactory leakage rate test results are a requirement for the establishment of containment OPERABILITY. The containment satisfies Criterion 3 of 10 CFR 50.36(c)(2) . Palisades Nuclear Plant B 3.6.1-2 05/31/99

  • *
  • BASES LCO APPLICABILITY ACTIONS Containment B 3.6.1 Containment OPERABILITY is maintained by limiting leakage to 1.0 La, except prior to the first startup after performing a required 10 CFR 50, Appendix J leakage test. At this time, the applicable leakage limits must be met. Compliance with this LCD will ensure a containment configuration, including the equipment hatch, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis.

Individual leakage rates specified for the containment air lock (LCO 3.6.2) and purge valves which have resilient seals (LCO 3.6.3) are not specifically part of the acceptance criteria of 10 CFR 50, Appendix J. Therefore, leakage rates exceeding these individual limits only result in the containment. being inoperable when the leakage results in exceeding the overall acceptance criteria of 1.0 La. In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material into containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, containment is not required to be OPERABLE in MODE 5 to prevent leakage of radioactive material from containment. The requirements for containment during MODE 6 are addressed in LCO 3.9.3, "Containment Penetrations." In the event containment is inoperable, containment must be restored to OPERABLE status within 1 hour. The 1 hour Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining containment OPERABILITY during MODES 1, 2, 3, and 4. This time period also ensures that the probability of an accident (requiring containment OPERABILITY) occurring, during periods when containment is inoperable, is minimal .. Palisades Nuclear Plant B 3.6.1-3 05/31/99 Containment B 3.6.1 *

  • ACTIONS (continued)

SURVEILLANCE REQUIREMENTS B.1 and B.2 If containment cannot be restored to OPERABLE status within . the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

  • SR 3. 6 .1.1 Maintaining the containment OPERABLE requires compliance with the visual examinations and Type A leakage rate test requirements of the Containment Leak Rate Testing Program. As left leakage prior to the first startup after performing a required leakage test is required to be 0.75 La for overall Type A leakage following an outage that included . Type A testing. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit 1.0 La.

1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by the Containment Leak Rate Testing Program. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis. SR 3.6.1.2 This SR ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Structural Integrity Surveillance Program . Palisades Nuclear Plant B 3.6.1-4 05/31/99

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  • BASES SURVEILLANCE REQUIREMENTS (continued)

REFERENCES SR 3.6.1.3 Containment B 3.6.1 Maintaining the containment OPERABLE requires compliance with the Type B and C leakage rate test requirements of 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. Testing is performed at pressures 55 psig. As left leakage prior to the first startup after performing

  • a required 10 CFR 50, Appendix J, Option A, leakage test is required to be < 0.6 La for combined Type B and C leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of 1.0 La. At 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis.

SR Frequencies are as required by Appendix J, Option A, as modified by approved exemptions. Thus, SR 3.0.2 (which allows Frequency extensions) does not apply. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis .. SR 3.6.1.3 is modified by a Note which states that local leak tests shall be performed at pressures 55 psig. This value corresponds to the design pressure of the containment and bounds the maximum expected internal pressure resulting from an MSLB or design basis LOCA. 1. FSAR, Chapter 14 2. FSAR, Section 14.18 3. FSAR, Section 5.8 Palisades Nuclear Plant B 3.6.1-5 05/31/99 Containment Air Locks B 3.6.2

  • B 3.6 CONTAINMENT SYSTEMS *
  • B 3.6.2 Containment Air Locks BASES BACKGROUND Containment air locks form part of the containment pressure boundary and provide a means for personnel access during all MODES of operation.

Two air locks provide access into the containment. Each air lock is nominally a right circular cylinder, with a door at each end. The personnel air lock doors are 3 foot, 6 inches by 6 foot, 8 inches. The emergency escape air lock doors are 30 inches in diameter. The doors are interlocked to prevent simultaneous opening. During periods when containment is not required to be OPERABLE, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary. Each air lock door has been designed and tested to certify its ability to withstand a pressure in excess of the maximum expected pressure following a Design Basis Accident (OBA) in containment. As such, closure of a single door supports containment OPERABILITY. Each of the doors contains double gasketed seals and local testing capability to ensure pressure integrity.* To effect a leak tight seal, the air lock design uses pressure seated doors (i.e., an increase in containment internal pressure results in increased sealing force on each door). The containment air locks form part of the containment pressure boundary. As such, air lock integrity and leak tightness is essential for maintaining the containment leakage rate within limit in the event of a OBA. Not maintaining air lock integrity or leak tightness may result *in a leakage rate in excess of that assumed in the plant safety analysis . Palisades Nuclear Plant B 3.6.2-1 05/31/99

  • BASES Containment Air Locks B 3.6.2 APPLICABLE The DBAs that result in a release of radioactive material SAFETY ANALYSES within containment are a Loss of Coolant Accident (LOCA), a Main Steam Line Break (MSLB) and a control rod ejection accident (Ref. 1). In the analysis of each of these accidents, it is assumed that containment is OPERABLE such that release of fission products to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate of 0.10% of containment air weight per day (Ref. 2). This leakage rate is defined in 10 CFR 50, Appendix J, Option A, as La : the maximum allowable containment leakage rate at the calculated maximum peak containment pressure (Pa). For a LOCA, the calculated maximum peak containment pressure is approximately 53 psig. For an MSLB, the calculated maximum peak containment pressure is approximately 54 psig. However, to ensure sufficient margin and to bound all DBAs, Type B leakage rate testing is performed at or above the containment design pressure of 55.0 psig. This allowable leakage rate.forms the basis for the acceptance criteria imposed on the SRs associated with the air lock. LCO The containment air locks satisfy Criterion 3 of 10 CFR 50.36(c)(2).

Each containment air lock forms part of the containment pressure boundary. As part of the containment pressure boundary, the air lock safety function is related to control of the containment leakage rate resulting from a oaA. Thus, each air lock's structural integrity and leak are essential to the successful mitigation of such an event. Each air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of containment does not exist when containment is required to be OPERABLE. Closure of a single OPERABLE door in each air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into or* exit from containment . Palisades Nuclear Plant B 3.6.2-2 05/31/99

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  • BASES APPLICABILITY ACTIONS Containment Air Locks B 3.6.2 In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment air locks are not required in MODE 5 to prevent leakage of radioactive material from containment. The requirements for the containment air locks during MODE 6 are addressed in LCD 3.9.3, "Containment Penetrations." The ACTIONS are modified by three notes. The first note *allows entry and exit to perform repairs on the .affected air lock component. If the outer door is inoperable, then it may be easily accessed for most repairs. It is preferred that the air lock be accessed from inside containment by entering through the other OPERABLE air lock. However, if this .is not* practfcable, or if repairs on either door must be performed from the barrel side of the door then it is permissible to enter the air lock through the OPERABLE door, even if this door has been locked to comply with ACTIONS. This there is a short time during the containment boundary is not intact (during access through* the OPERABLE door) .. The ability to open the OPERABLE door, even if it means the containment boundary is temporarily not intact, is acceptable because of the low probability of an event that could pressurize the containment during the short time in which the OPERABLE door is expected to be open. After each entry and exit, the OPERABLE door must be .. immediately closed. If ALARA conditions permit, entry and exit should be vfa an OPERABLE air lock. A second Note has been added to provide clarification that, for this LCD, separate Condition entry is allowed for each air lock. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable air Complying with the Required . Actions may allow for continued operation, and a subsequent inoperable air lock is governed by subsequent Condition entry and application of associated Required Actions. A third Note has been included that requires entry into the applicable Conditions and Required Actions of LCO 3.6.1, "Containment," when leakage results in exceeding the overall containment leakage limit. \ . Palisades Nuclear Plant B 3.6.2-3 05/31/99 Containment Air Locks B 3.6.2 *

  • ACTIONS (continued)

A.1. A.2. and A.3 With one air lock door inoperable in one or more containment air locks, the OPERABLE door must be verified closed (Required Action A.1) in each affected containment air lock. This ensures that a leak tight containment barrier is maintained by the use of an OPERABLE air lock door. This action must be completed within 1 hour. This specified time period is consistent with the ACTIONS of LCO 3.6.1, which requires containment be restored to OPERABLE status within 1 hour. In addition, the affected air lock penetration must be . isolated by locking closed an OPERABLE air lock door within the 24 hour. Completion Time. The 24 hour Completion Time is considered reasonable for locking the OPERABLE air lock door, considering the OPERABLE door of the affected air lock is being maintained closed. Required Action A.3 verifies that an air lock with an inoperable door has been isolated by the use of a locked and closed OPERABLE air lock door. This ensures that an acceptable containment leakage boundary is maintained. The Completion Time of once per 31 days is based on engineering judgment and is considered adequate in view of the low likelihood of a locked door being mispositioned and other administrative controls. Required Action A.3 is modified by a Note that applies to air lock doors located in high radiation areas and allows these doors to be verified locked' closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of the door, once it has been verified to be in the proper position, is small . Palisades Nuclear Plant B 3.6.2-4 05/31/99

  • * *
  • BASES ACTIONS A.2. and A.3 (continued)

Air Locks B 3.6.2 The Required Actions have been modified by two Notes. Note 1 ensures that unly the Required Attions and associated Completion Times of Condition Care required if both doors in the same air lock are inoperablew With both doors in the same lock inoperable, an OPERABLE door is not available to be closed. Required Actions C.1 and C.2 are the appropriate remedial actions. The exception provided by Note 1 does not affect tracking the Completion Time from the. initial entry into Condition A; only the requirement to comply with the Required Actions. Note 2 allows use of the air lock for entry and exit for 7 days under administrative controls if both air locks have an inoperable door. This 7 day restriction begins when the second air lock is discovered inoperable. Containment entry may be required to perform Technical Specifications (TS) Surveillances and Required Actions, as well as other on equipment inside containment that are required by TS or activities on equipment that support TS-reqlii.red . equipment. This Note is not intended t.o preclude performing other activities (i.e., non-TS-required activities) if the containment was entered, using the inoperable air lock, to perform an allowed activity listed above. This allowance is acceptable due to the low probabilfty of an event that could the containment during the short time that the OPERABLE door is expected to be open. B.1. B.2. and B;3 With an air lock interlock mechanism inoperable in one or more air locks, the Required.Actions and associated Completion Times are consistent with those specified in Condition A. The Required Acti-0ns have modified by two Notes. Note 1 ensures that only the Required Actions and associated Completion Times of Condition C are required if both doors in the same air lock are inoperable. With both doors in the same air lock inoperable, an OPERABLE door is not available to be closed. Required Actions C.1 and C.2 are the appropriate remedial actions. Note 2 allows entry into and exit from containment under the control of a dedicated individual stationed at the air lock to ensure only one door is opened at a tit11e (i.e., the individual performs the function of the interlock) . Palisades Nuclear Plant B 3.6.2-5 05/31/99

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  • BASES ACTIONS B.1. B.2. and B.3 (continued)

Containment Air Locks B 3.6.2

  • Required Action B.3 is modified by a Note that applies to air lock doors located in high radiation areas and allows these doors to be verified locked closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted.

Therefore, the probability of misalignment of the door, once.it has been verified to be in the proper position, is small. C.l. C.2. and C.3 With one or more air locks for reasons other than . those described in Condition A or B, Required Action C.1 requires action to be initiated immediately to evaluate previous combined leakage rates using current air lock test results. If the overall containment leakage rate exceeds the limits of LCO 3.6.l, the conditions of that LCO must be entered in accordance Actions Note 3. An evaluation is acceptable since it is overly conservative to immediately declare. the.containment inoperable if both doors in an air lock have failed a seal test or if the overall air lock

  • leakage is not within limits. In many instances (e.g., only one seal per door has failed), containment remains OPERABLE, yet only 1 hour (per LCO 3.6.1) would be provided to restore the air lock door to OPERABLE status prior to requiring a plant shutdown.

In addition, even with both doors failing the seal test, the overall containment leakage rate can still be within limits. Required Action C.2 requires that one door in the affected containment air lock must be verified to be closed. This action must be completed within the 1 hour Completion Time. This specified time period is consistent with the ACTIONS of LCO 3.6.1, which requires that containment be restored to OPERABLE status within 1 hour. Additionally, the affected air lock(s) must be restored to OPERABLE status within the 24 hour Completion Time. The specified time period is considered reasonable for restoring an inoperable air lock to OPERABLE status, assuming that at least one door is maintained closed in each affected air lock . Palisades Nuclear Plant B 3.6.2-6 05/31/99

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  • BASES ACTIONS (continued)

SURVEILLANCE REQUIREMENTS D.1 and D.2 Containment Air Locks B 3.6.2 If the inoperable containment air lock cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

  • SR 3.6.2.1 Maintaining containment air locks OPERABLE requires compliance with the leakage rate test requirements of 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.

This SR reflects the leakage rate testing. requirements with regard to air lock leakage (Type B leakage tests). The acceptance criteria, were established during initial air lock and containment Operability testing. Subsequent amendments to the Technical Specifications revised the acceptance criteria overall Type B and C leakage limits and provided new acceptance criteria for the personnel air lock doors and the emergency air lock doors (Ref. 2). The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall containmerit leakage rate. Leak rate tests,

  • other than the personnel air lock doors, are performed a_t pressure 55 psig. The Frequency is required by 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.

Thus, SR 3.0.2 (which allows Frequency extensions) does not apply. Two exemptions to the requirements of 10 CFR 50, Appendix J have been granted for the containment air locks. The exemption granted by letter dated December 6, 1989 provides partial relief from the requirement of Paragraph III.D.2.(b)(ii) to leak test, at or above the calculated design basis accident peak containment pressure (Pa), containment air locks which were opened during a period when containment integrity was not required. This exemption permits the substitution of a between-the-seal leak test at

  • a reduced pressure, but not less than 10 psig, provided that. no maintenance, or other activity has been performed which could affect the sealing capability of the air locks . Palisades Nuclear Plant B 3.6.2-7 05/31/99
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  • BASES SURVEILLANCE REQUIREMENTS SR 3.6.2.1 (continued)

Containment Air Locks B 3.6.2 The exemption granted by letter dated September 30, 1997 applies only to the emergency escape. air lock and provides partial relief from the requirement of Paragraph III.D.2.(b)(ii) and Paragraph III.D.2.(b)(iii). The requirement of Paragraph III.D.2.(b)(ii) is discussed above. Paragraph III.D.2.(b)(iii) requires air locks opened during periods when containment integrity is required to undergo a

  • full air lock pressure test within 3 days after being opened. This exemption permits the performance of a door seal contact verification check in lieu of the final pressure test fo 11 owing the opening of the emergency escape air lock doors for post-test restoration or seal adjustment.

This exemption does not affect compliance with the

  • requirement to perform a full pressure air lock test at 6 month intervals, or the requirement to perform a full pressure air lock test within 72 hours of opening either air lock door during periods when containment integrity is required.

The SR has been modified by four Notes. Note 1 states that an inoperable air lock door does not*invalidate the previous successful performance of the overall air leakage test

  • This is considered reasonable since either air lock door is 6apable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR requiring the results to be evaluated against the acceptance criteria of SR 3.6.1.3. This ensures that air lock leakage is properly accounted for in determining the overall containment leakage rate. Note 3 clarifies that iterative pressure testing of the emergency escape air lock is not required when the air lock doors are opened solely for the purpose of strongback removal and performance of the seal contact check. Note 4 ensures that air lock testing, other that door seal testing, is performed at a pressue 55 psig consistent with other Type B and C tests . Palisades Nuclear Plant ( B 3.6.2-8 Containment Air Locks B 3.6.2 *

.* -* SURVEILLANCE REQUIREMENTS (continued) REFEREN.CES ./ SR 3.6.2.2 The air lock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY. Thus, the door interlock feature supports containment OPERABILITY while the air lock is being used for personnel transit into and out of containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur. Due to the purely mechanical nature of this interlock, and given that the i nter-1 ock mechanism is not norma 11 y cha 11 enged when the airlock is used for entry and exit (procedures require strict adherence to single door opening), this test is only required to be performed every 18 months .. The 18 month frequency is based on the need to perform this Surveillance under the conditions that apply during plant outage, and the potential for loss of containment OPERABILITY if the Surveillance were performed with the reactor at power . The 18 month for is justified based on generic operating

  • The Frequency is based on engineering judgment and is considered adequate given that the interlock is not normally challenged during use of the airlock. *
  • 1. FSAR, Chapter 14 2. FSAR, Section 5.8 Palisades Nuclear Plant B 3.6.2-9 05/31/99 \'
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  • Containment Isolation Valves B 3.6.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.3 Containment Isolation Valves BASES* BACKGROUND The containment isolation valves and devices form part of the containment pressure boundary and provide a means for isolating penetration flow paths. These isolation devices are either passive or active (automatic).

Manual valves, de-activated automatic valves secured in their closed position (including check valves with flow through the valve secured), blind flanges, and closed systems are considered devices. Check valves, or other automatic valves designed to close without operator action following an accident, are considered active devices. Two barriers in series are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the safety analysis. One of these barriers may be a closed system. Containment isolation occurs upon receipt of a Containment High Pressure (CHP) signal or a Containment High Radiation (CHR) signal. However, not all containment isolation valves are actuated by both signals. The signals close automatic

  • containment isolation valves in fluid penetrations not required for operation of Engineered Safety Feature systems in order to prevent leakage of radioactive material.

Other penetrations are isolated by the use of valves or check valves in the closed position, or blind flanges. As a result, the containment isolation valves (and blind flanges) help ensure that the containment atmosphere will be isolated in the event of a release of radioactive material to containment atmosphere from the Primary Coolant System (PCS) following a Design Basis Accident (DBA). The OPERABILITY requirements for containment isolation valves and devices help ensure that containment is isolated within the time limits assumed in the safety analysis. Therefore, the OPERABILITY requirements provide assurance that the containment leakage limits assumed in the accident analysis will be not exceeded in a DBA . Palisades Nuclear Plant B 3.6.3-:1 05/31/99

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  • BASES BACKGROUND (continued)

Containment Isolation Valves B 3.6.3 The 8 inch purge exhaust valves are designed for purging the containment atmosphere to the stack while introducing filtered makeup, through the 12 inch air room supply valves from the outside, when the plant is shut down during refueling operations and maintenance. The purge exhaust valves and air room supply valves are air operated isolation valves located outside the containment. These valves are operated manually from the control room. These valves will close automatically upon receipt of a CHP or CHR signal. The air operated valves fail closed upon a loss of air. These valves are not qualified for automatic closure from their open position under OBA conditions. Therefore, these valves are locked closed in MODES 1, 2, 3, and 4 to ensure the containment boundary is maintained. Open purge exhaust or air room supply valves, following an accident that releases contamination to the containment atmosphere, would cause a significant increase in the containment leakage rate. APPLICABLE The containment isolation valve LCO was derived from the SAFETY ANALYSES assumptions related to minimizing the loss of reactor coolant inventory and establishing the containment boundary during major accidents. As part of the containment boundary, containment isolation valve OPERABILITY supports leak tightness of the containment. Therefore, the safety analysis of any event requiring isolation of containment is applicable to this LCO. I The DBAs that.result in a release of radioactive material within containment are a Loss of Coolant Accident (LOCA), a Main Steam Line Break (MSLB), and a control rod ejection accident. In the analysis for each of these accidents, it is assumed that containment isolation valves are either closed or function to close within the required isolation time following event initiation. This ensures that potential paths to the environment through containment isolation valves (including containment purge valves) are minimized. The safety analysis assumes that the purge exhaust and air room supply valves are closed at event initiation . Palisades Nuclear Plant B 3.6.3-2 05/31/99

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  • BASES APPLICABLE SAFETY ANALYSES (continued)

LCO Containment Isolation Valves B 3.6.3 The OBA analysis assumes that, within 25 seconds after receiving a CHP or CHR signal each automatic power operated valve is closed and containment leakage terminated except for the design leakage rate, La. The single failure criterion required to be imposed in the tonduct of plant safety analyses was considered in the design of the containment purge valves. Two valves in series on each line provide assurance that both the supply and exhaust lines could be isolated even if a single failure occurred. Both isolation valves on the 8 inch and 12 inch lines are pneumatically operated spring closed valves. The 8 inch purge exhaust and 12 inch air room supply valves may be unable to close in the environment following a LOCA. Therefore, each of the purge valves is required to remain locked closed during MODES 1, 2, 3, and 4. In this case, the single failure criterion remains applicable to the containment purge valves due to the potential for failure in the control circuit associated with each valve. Again, the purge system valve design precludes a single failure from compromising the containment boundary as long as the system is operated in accordance with the subject LCO . The containment isolation valves satisfy Criterion 3 of 10 CFR 50.36(c)(2). Containment isolation valves form a part of the containment boundary. The containment isolation valve safety function is related to minimizing the loss of primary coolant inventory and establishing the containment boundary during a OBA. The automatic power operated isolation valves are required to have isolation times within limits and to actuate upon receipt of a CHP or CHR signal as appropriate. The purge exhaust and air room valves must be locked closed. The. valves covered by this LCO are listed with their associated stroke times in the FSAR (Ref. 1). The normally closed isolation valves are considered OPERABLE when manual valves are closed, automatic valves are de-activated and secured in their closed position, blin4 flanges are in place, and closed systems are intact. These passive isolation or devices are those listed in Reference 1 . Palisades Nuclear Plant B 3.6.3-3 05/31/99

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  • BASES LCO (continued)

APPLICABILITY ACTIONS ' Containment Isolation Valves B 3.6.3 Containment isolation valve leakage rates are addressed by LCO 3.6.1, 11 Containment, 11 as Type C testing. This LCO provides assurance that the containment isolation valves and purge valves will perform their designed safety functions to minimize the loss of primary coolant inventory and establish the containment boundary during accidents. In MODES 1, 2, 3, and 4, a OBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment isolation valves are not required to be OPERABLE in MODE 5. The requirements for containment isolation valves during MODE 6 are addressed in LCO 3.9.3, 11 tontainment Penetrations. 11 The ACTIONS are modified by four notes. Note one allows isolated penetration flow paths, except for 8 inch exhaust and 12 inch air room supply purge valve penetration flow paths, to be unisolated intermittently under administrative controls. These administrative controls consist of

  • stationing a dedicated operator at the valve controls, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for containment isolation is indicated.

Due to the fact that the 8 inch purge exhaust valves and the 12 inch air room supply valves may be unable to close in the environment following a LOCA and the fact that those penetrations exhaust directly from the containment atmosphere to the environment, these valves may not be opened under administrative controls. A second Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable containment isolation valve. Complying with the Required Actions may allow for continued operation, and subsequent inoperable containment isolation valves are governed by subsequent Condition entry and application of associated Required Actions . Palisades Nuclear Plant B 3.6.3-4 05/31/99 Containment Isolation Valves B 3.6.3 *

  • ACTIONS (continued)

The ACTIONS are further modified by a third Note, which ensures that appropriate remedial actions are taken, if necessary, if the affected systems are rendered inoperable by an inoperable containment isolation valve. A fourth Note has been added that requires entry into the applicable Conditions and Required Actions of LCO 3.6.1 when leakage results in exceeding the overall containment leakage 1 i mit. A.1 and A.2 In the event one containment isolation valve in one or more penetration flow paths is inoperable (except for purge exhaust or air room supply valves), the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic containment isolation valve, a closed manual valve, a blind flange, and a check valve with flow

  • through the valve secured. For penetrations isolated in accordance with Required Action A.l, the device used to isolate the penetration should be the closest available one to containment.

Required Action A.1 must be completed within the 4 hour Completion Time. The 4 hour Completion Time is reasonable, considering the time required to isolate the penetration and the relative importance of supporting containment OPERABILITY during MODES 1, 2, 3, and 4 . Palisades Nuclear Plant B 3.6.3-5 05/31/99

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  • BASES ACTIONS A.1 and A.2 (continued)

Containment Isolation Valves B 3.6.3 For affected penetration flow paths that cannot be restored to OPERABLE status within the 4 hour Completion Time and that have been isolated in accordance with Required Action A.l, the affected penetration flow paths must be verified to be isolated on a periodic basis. This is necessary to ensure that containment penetrations required to be isolated following an accident and no longer capable of being automatically isolated will be in the isolation position should an event occur. This Required Action does not require any testing or device manipulation. Rather, it involves verification, through a system walkdown, that those isolation devices outside containment and capable of being mispositioned are tn the correct position. The Completion Time of "once per 31 days for isolation devices outside containment" is appropriate considering the fact that the devices are: operated under controls and the probability of their misalignment is low. For the isolation devices inside containment, the time period specified as "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 .days" is based on engineering judgment and is considered reasonable in view of the of the isolation devices and other administrative controls that will ensure that isolation misalignment is an unlikely possibility. Condition A been modified by a Note indicating that this Condition is only applicable to those penetration flow paths with two containment isolation valves. For penetration flow paths with only one containment isolation valve and a closed system, Condition C provides appropriate actions. Required Action A.2 is modified by a Note that applies to isolation devices located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is

  • typically restricted.

Therefore, the probability of misalignment of these devices, once they have been verified to be in the proper position, is small . Palisades Nuclear Plant B 3.6.3-6 05/31/99

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  • BASES Containment Isolation Valves B 3.6.3 ACTIONS B.1 (continued)

With two containment isolation valves in one or more penetration flow paths inoperable (except for purge exhaust valve or air room supply valve not closed), the affected penetration flow path must be isolated within 1 hour. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. The 1 hour Completion Time ;s consistent with the ACTIONS of LCD 3.6.1. In the event the affected penetration is isolated in accordance with Required Action B.l, the affected penetration must be verified to be isolated on a periodic basis per Required Action* A.2, which remains in effect. This periodic verification is necessary to assure leak tightness of containment and that penetrations requiring isolation following an accident are isolated. The Completion Time of once per 31 days for verifying each* affected penetration flow path is isolated is appropriate considering the fact that the valves are operated under administrative controls and the probability of their misalignment is low. Condition B is modified by a Note indicating this Condition is only applicable to penetration flow paths with two containment isolation valves. Condition A of this LCD addresses the condition of one containment isolation valve inoperable in this type of penetration flow path . Palisades Nuclear Plant B 3.6.3-7 05/31/99

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  • BASES ACTIONS (continued)

C.1 and C.2 Containment Isolation Valves B 3.6.3 With one or more penetration flow paths with one containment isolation valve inoperable, the inoperable valve must be restored to OPERABLE status or the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active Isolation barriers that meet this criterion are a closed and* de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration. Required Action C.1 must be completed within the 72 hour Completion Time. The specified time period is reasonable, considering the relative stability of the closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of supporting containment OPERABILITY during MODES 1, 2, 3, and 4. In the event the affected penetration is isolated in accordance with Required. Action C.l, the affected penetration flow path must be verified to be isolated on a periodic basis. This is necessary to assure leak tightness

  • of containment and that containment penetrations requiring isolation following an accident are i.solated.

The Completion Time of once per 31 days for verifying that each affected penetration flow path is isolated is appropriate

  • considering the valves are operated under administrative controls and the probability of their misalignment is low. Condition C is.modified by a Note indicating that this. Condition is only .applicable to those penetration flow paths with only one containment isolation valve and a closed system. This Note is necessary since this Condition is written to specifically address those penetration flow paths in a closed system. Required Action C.2 is modified by a Note that applies to valves and blind flanges located in high radiation areas and allows these devices to be verified-closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted.

Therefore, the probability of mispositioning these devices, once they have been verified to be in the proper position, is small . Palisades Nuclear Plant B 3.6.3-8 05/31/99

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  • BASES Containment Isolation Valves B 3.6.3 ACTIONS D.1 (continued)

SURVEILLANCE REQUIREMENTS The purge exhaust and air room supply isolation valves have not been qualified to close following a LOCA and are required to be locked closed. If one or more of these valves is found not locked closed, the potential exists for the valves to be inadvertently opened. One hour is provided to lock closed the affected valves. The 1 hour Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining these valves closed. E.1 and E.2 If the Required Actions and associated Completion Times are not met, the plant must be brought to a MODE' in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems . SR 3.6.3.1 This SR ensures that the 8 inch purge exhaust and 12 inch air room supply valves are locked closed as required. If a valve is open, or closed but not locked, in violation of this SR, the valve is considered inoperable. Valves may be locked closed electrically, mechanically, or by other physical means. These valves may be unable to close in the environment following a LOCA. Therefore, each of the valves is required to remain closed during MODES 1, 2, 3, and 4. The 31 day Frequency is consistent with other containment isolation valve requirements discussed in SR 3.6.3.2 . Palisades Nuclear Plant B 3.6.3-9 05/31/99 Containment Isolation Valves B 3.6.3 *

  • SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.3.2 This SR requires verification that each manual .containment isolation valve and blind flange located outside containment, and not locked, sealed, or otherwise secured in position, and required to be closed during accident conditions, is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown, that those containment isolation valves outside containment and capable of being mispositioned are in the correct position. Since verification of valve position for containment isolation valves outside containment is relatively easy, the 31 day Frequency is based on engineering judgment and was chosen to provide added assurance of the correct positions. Containment isolation valves that are open under administrative. controls are not required to meet the SR during the time the valves are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing . The Note applies to valves and blind flanges located in high radiation areas and allows these devices to.be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, is small . Palisades Nuclear Plant B 3.6.3-10 05/31/99_

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  • BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.3.3 Containment Isolation Valves B 3.6.3 This SR requires verification that each containment isolation manual valve and blind flange inside containment and not locked, sealed or otherwise secured in position, and required to be closed during accident conditions, is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the containment boundary is within design limits. For containment isolation valves inside containment, the Frequency of "prior to entering MODE 4 from MODE 5 if not -performed within the previous 92 days" is appropriate, since these containment isolation valves are operated under administrative controls and the probability of their misalignment is low. Containment isolation valves that are open under administrative controls are not required to meet the SR during the time that they are open. This SR does hot apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing. The Note allows valves and blind flanges located in high* radiation areas td be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in their proper position, is small. SR 3.6.3.4 Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analysis. The .isolation time and Frequency of this SR are in accordance the Inservice Testing Program

  • Palisades Nuclear Plant B l.6.3-11 05/31/99 Containment Isolation Valves B 3.6.3 *
  • SURVEILLANCE REQUIREMENTS (continued)

REFERENCES SR 3.6.3.5 For containment 8 inch purge exhaust and 12 inch air room supply valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J (Ref. 3), is required to ensure the valves are physically closed (SR 3.6.3.1 verifies the valves are locked closed). Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than do other seal types. Based on this observation and the importance of maintaining this penetration leak tight (due to the direct path between containment and the environment), a Frequency of 184 days was established as part of the NRC resolution of Generic Issue B-20, 11 Containment Leakage Due to Seal Deterioration 11 (Ref. 2) as specified in. the Safety Evaluation for Amendment No. 90 to the Facility Operating License. SR 3.6.3.6 Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a OBA. This SR ensures each automatic containment isolation valve will actuate to its isolation position on an actual or simulated actuation signal, i.e., CHP or CHR. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 18 month Frequency was developed considering it is prudent that this SR be performed only during a plant outage, since isolation of penetrations would eliminate cooling water flow and disrupt normal operation of many critical components. Operating experience has shown that these components usually pass this SR when performed on the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

1. FSAR, Section 5.8 2. Generic Issue B-20 3. 10 CFR 50, Appendix J Palisades Nuclear Plant B 3.6.3-12 05/31/99
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  • Containment Pressure B 3.6.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.4 Containment Pressure BASES BACKGROUND The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). . Containment pressure is a process variable that is monitored
  • and controlled.

The containment pressure limits are derived from the input conditions used in the containment functional analyses. Should operation occur outside these limits coincident with a Design Basis Accident (DBA), post accident containment pressures could exceed calculated values. APPLICABLE Containment pressure is an initial condition used SAFETY ANALYSES in the DBA analyses to establish the maximum peak

  • containment internal pressure.

The limiting DBAs considered for determining the maximum containment internal pressure are the LOCA and MSLB. An MSLB at 0% RTP with the MSIVs open results in the highest calculated internal containment pressure of 54 psig, which is below the internal design pressure of 55 psig. The value of 54 psig represents the analytical value presented in Reference 1, rounded up to the next whole number. The postulated DBAs are analyzed assuming degraded containment Engineered Safety Feature (ESF) systems (i.e., assuming the limiting single active failure, resulting in the loss of two Containment Spray pumps since a common mode failure disables two of the three pumps. While the maximum containment internal pressure results from an MSLB, the licensing basis dose limitations are based on the LOCA (see the Bases for 3.6.1, "Containment," for a discussion on containment pressures resulting from a LOCA) . Palisades Nuclear Plant B 3.6.4-1 05/31/99

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  • BASES APPLICABLE SAFETY ANALYSES (continued)

LCO APPLICABILITY Containment Pressure B 3.6.4 The initial pressure condition used in the containment analysis was 15.7 psia (1.0 psig) in MODES 1 and 2 and 16.2 psia (1.5 psig in MODES 3 and 4). The LCO limits of 1.0 psig in MODES 1 and 2, and 1.5 psig in MODES 3 and 4 ensures that, in the event of an accident, the maximum accident design pressure for containment, 55 psig, is not exceeded. A higher containment pressure limit is allowed in MODES 3 and 4 where the reactor is not critical and the resulting heat addition to containment in a DBA is lower. The external design pressure of the containment shell is 3 psig. This value is approximately 0.5 psig greater than the maximum external pressure that could be developed if the containment were sealed during a period of low barometric pressure and high temperature and, subsequently, the containment atmosphere were cooled with a concurrent major rise in barometric pressure. Vacuum breakers are, therefore, not provided and no minimum containment pressure specification is required. Containment pressure satisfies Criterion 2 of 10 CFR 50.36(c)(2) . Maintaining containment pressure less than or equal to the LCO upper pressure limit ensures that, in the event of a DBA, the resultant peak containment accident pressure will remain below the containment design pressure. Two limits for containment pressure are provided to reflect the analyses which.allow for a higher containment pressure when the reactor is not critical due to less heat input to containment in the event of a DBA. In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. Since maintaining containment pressure within limits is essential to ensure initial conditions assumed in the accident analysis are maintained, the LCO is applicable in MODES 1, 2, 3, and 4. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining containment pressure the limits of the LCO is not required in MODE 5 or 6 . Palisades Nuclear Plant B 3.6.4-2 05/31/99 I

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  • BASES ACTIONS SURVEILLANCE REQUIREMENTS REFERENCES Containment Pressure B 3.6.4 When containment pressure is not within the limits of the LCD, containment pressure must be restored to within these limits within 1 hour. The Required Action is necessary to return operation to within the bounds of the containment analysis.

The 1 hour Completion Time is consistent with the ACTIONS of LCD 3.6.1, "Containment," which requires that containment be restored to OPERABLE status within 1 hour. B.1 and B.2 If containment pressure cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCD does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to.reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems . SR 3.6.4.1 Verifying that containment pressure is within limits ensures that operation remains within the limits assumed in the aicident analyses. The 12 hour Frequency of this SR was developed after taking into consideration operating experience related to trending of containment pressure variations during the applicable MODES. Furthermore, the 12 hour Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal containment pressure condition. The limit of 1.0 psig for MODES 1 and 2, 1.5 psig for MODES 3 and 4 are the actual limits used in the accident analysis and do not account for instrument inaccuracies.

1. FSAR, Section 14.18 Palisades Nuclear Plant B 3.6.4-3 05/31/99 Containment Air Temperature B 3.6.5
  • B 3.6 CONTAINMENT SYSTEMS *
  • B 3.6.5 Containment Air Temperature BASES BACKGROUND APPLICABLE SAFETY ANALYSES I The containment structure serves to contain radioactive material that may be released from the reactor core following a Design Basis Accident (DBA). The containment average air temperature is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). Containment air temperature is a process variable that is monitored and controlled.

The containment average air temperature limit is derived from the input conditi*ons used in the containment accident analyses .. This LCO ensures that initial conditions assumed in the anilysis of containment response to a DBA are not violated during plant operations. The total amount of energy to be removed from containment by the Containment Spray and Cooling systems during post accident conditions is dependent on the energy released to the containment due to the event, as well as the initial containment temperature and pressure. The higher the .. initial temperature, the more energy that must be removed, resulting in a higher peak containment pressure and temperature. Exceeding containment design pressure may result in leakage greater than that assumed in the accident analysis (Ref. 1). Operation with containment average air temperature in excess-of the LCO limit violates an initial condition assumed in the accident analysis. Containment average air temperature is an initial condition used in the DBA analyses that establishes the containment environmental qualification operating envelope for both pressure and temperature. the limit for containment average air temperature ensures that operation is maintained within the assumptions used in the DBA analysis for containment. The accident analyses and evaluations considered both LOCAs and MSLBs for determinitig the maximum peak containment pressures and temperatures. The worst case MSLB generates larger mass and energy releases than the worst case LOCA. Thus, the MSLB event bounds the LOCA event from the containment peak pressure and temperature standpoint

  • Palisades Nuclear Plant B 3.6.5-1 05/31/99
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  • BASES APPLICABLE SAFETY ANALYSES (continued)

LCO APPLICABILITY ACTIONS Containment Air Temperature B 3.6.5 The initial pre-accident temperature inside containment was assumed to be 140°F (Ref. 2). The initial containment average air temperature condition of 140°F resulted in a maximum vapor temperature in containment of 399°F. This value represents the analytical value presented in Reference 3, rounded up to the next highest number. This exceeds the containment building design temperature of 283°F. The effect on the structure is negligible due to the short. period of time the temperature exceeds the design value. Containment average air temperature satisfies Criterion 2 of 10 CFR 50.36(c)(2). During a DBA, with an initial containment average air . temperature less than or equal to the LCO temperature limiti the resultant peak accident pressure is maintained below the containment design pressure. As a result, the ability of containment to perform its function is ensured . .In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining containment average air temperature within the limit is not required in MODE 5 or 6. When containment average air temperature is not within the limit of the LCO, it must be restored to within limit within 8 hours. This Required Action is necessary to return operation to within the bounds of the containment analysis. The 8 hour Completion Time is acceptable considering the sensitivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems . Palisades Nuclear Plant B 3.6.5-2 05/31/99

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  • BASES ACTIONS (continued)

SURVEILLANCE REQUIREMENTS REFERENCES B.1 and B.2 Containment Air Temperature B 3.6.5 If the containment average air temperature cannot be

  • restored to within its limit within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within*36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SR 3.6.5.1 Verifying that containment average air temperature is within the LCO limit ensures that containment operation remains within the limit assumed for the containment analyses.

The 140°F limit is the actual limit assumed for the accident analyses and does not account for instrument inaccuracies. The 24 hour Frequency of this SR is considered acceptable based on the observed slow rates of temperature increase within containment as a result of environmental heat sources (due to the large volume of containment).

1. FSAR, Section 5.8 2. FSAR, Section 14.18 3.
  • FSAR, Table 14.18.2-3 Palisades Nuclear Plant B 3.6.5-3 05/31/99
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  • Containment Cooling Systems B 3.6.6 B 3.6 CONTAINMENT SYSTEMS B 3.6.6 Cooling Systems BASES BACKGROUND The Containment Spray and Containment Air Recirculation and Cooling systems provide containment atmosphere cooling to limit post accident pressure and temperature in containment to less than the design values. Reduction of containment pressure reduces the release of fission product radioactivity from containment to the environment, in the eveni of a Design Basis Acctdent (OBA). The Containment Spray and Containment Air Recirculation and Cooling systems are designed to the requirements of the Palisades Nuclear Plant design criteria (Ref. 1). The Containment Air Recirculation and Cooling System and Containment Spray System are Engineered Safety Feature (ESF) systems. They are designed to ensure that the heat removal capability required during the post accident period can be attained.

The systems are arranged with two spray pumps and one air cooler fan powered from one diesel generator, and with one spray pump and three air cooler fans powered from the other diesel generator. The Containment Spray System was originally designed to be redundant to the Containment Air Coolers (CACs) and fans. These systems were originally

  • designed such that either two containment spray pumps or three CACs .tould limit containment pressure to less than design. However, the current safety analyses credit for one containment spray pump when evaluating cases with three CACs, and for one air cooler fan in cases with two spray pumps. To address this dependency between the Containment Spray System and the Containment Air Recirculation and Cooling System the title of this Specification is "Containment Cooling Systems," and includes both systems. The LCO is written in terms of trains of containment cooling. One train of containment cooling is associated with Diesel Generator 1-1 and inclutles Containment Spray Pumps P-548 and P-54C, and Air Cooler Fan V-4A. The other train of containment cooling is associated with Diesel 1-2 and includes Containment Spray Pump P-54A along with CACs VHX-1, VHX-2, and VHX-3 and their associated safety related fans, V-lA, V-2A, and V-3A . Additional details of the required equipment and its operation is discussed with the containment cooling system with which it is associated . Palisades Nuclear Plant B 3.6.6-1 05/31/99
  • BACKGROUND (continued)

Containment Spray System Containment Cooling Systems B 3.6.6 The Containment Spray System consists of three half-capacity (50%) motor driven pumps, spray headers, two full sets of full capacity (100%) nozzles, valves, and piping, two full capacity (100%) pump suction lines from the Safety Injection and Refueling Water Tank (SIRWT) and the containment sump with the associated piping, valves, power sources, instruments, and controls. The SIRWT supplies borated water to the containment spray during the injection phase of In the recirculation mode of operation, containment spray pump suction is transferred from the SIRWT to the containment sump. The Containment Spray System provides a spray of cold borated water into the upper regions of containment to reduce the containment pressure and temperature during a OBA. In addition, the Containment Spray System in conjunction with the use of trisodium phosphate (LCO 3.5.5, 11 Trisodium Phosphate,") serve to remove iodine which may be released following an accident. The SIRWT solution temperature is an important, factor in determining the heat removal capability of the Containment Spray System during the injection phase. In the recirculation mode of operation, heat is removed from the containment sump water by the shutdown cooling heat exchangers. Two containment spray pumps will meet the capacity requirements in the event of a OBA. * . The Containment Spray System is actuated either automatically by a Containment High Pressure (CHP) signal or manually. An automatic actuation opens the containment spray header isolation valves, starts the three containment spray pumps, and begins the injection phase. A actuation of the Containment Spray System is available on the main control board to begin the same sequence. The injection phase continues until an SIRWT Level Low signal is received. The Low Level signal for the SIRWT generates a Recirculation Actuation Signal (RAS) that aligns valves from the containment spray pump suction to the containment sump. The Containment Spray System in recirculation mode maintains an equilibrium temperature between the containment atmosphere and the recirculated sump water. *Operation of the Containment Spray System in the recirculation mode is controlled by the operator accordance with the emergency operating procedures. , Palisades Nuclear Plant B 3.6.6-2 05/31/99

  • BACKGROUND Containment Cooling Systems B 3.6.6 Containment Spray System (continued)

The containment spray pumps also provide a required support function for the High Pressure Safety Injection pumps as described in the Bases for specification. 3.5.2. The High Pressure Safety Injection pumps alone may not have adequate NPSH after a postulated accident and the realignment of their'suctions from the SIRWT to the containment sump. Provision is made to manually provide flow from the discharge of the containment spray pumps to the suction of the High Pressure Safety Injection pumps after the change to recirculation mode has occurred. The additional suction pressure ensures that adequate NPSH is available for the High Pressure Safety Injection pumps. Containment Air Recirculation and Cooling System The Containment Air Recirculation and Cooling System includes four air handling and cooling units, referred to as *the Contai.nment Air Coolers (CACs), which are located entirely within the containment building. Three of the CACs . (VHX-1, VHX-2, and VHX-3) are safety related coolers and are cooled by the critical service water: The fourth CAC (VHX-4) is not taken credit for in an DBA for maintaining containment temperature within limit, but is used during normal operation along with the three CACs to maintain containment temperature below design limits; The fan associated with VHX-4, is assumed in the safety analysis as assisting in the containment atmosphere mixing function.

  • Each CAC has two vaneaxial fans with direct connected motors which draw air through the cooling coils. Both of these* fans are normally in operation, but only one fan and motor for each CAC is rated for post DBA operatibn.

The post-DBA rated "safety related" fan 'units, V-lA, V-2A, V-3A, and V-4A, serve not only to provide forced flow for the associated cooler, but also provide mixing of the containment atmosphere. A single operating safety related fan unit will provide enough air flow to assure that there . is adequate mixing of unsprayed containment areas to assure the assumed iodine removal by the containment spray. The fan units also support the functioning of the hydrogen recombiners, as discussed in the Bases for LCO 3.6.7, "Hydrogen In post accident operation following a SIS, all four Containment air coolers are designed to cha_nge automatically to the emergency mode . Palisades Nuclear Plant B 3.6.6-3 05/31/99

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  • BASES BACKGROUND Containment Cooling Systems B 3.6.6 Containment Air Recirculation and Cooling System (continued)

The CACs are automatically changed to the emergency mode by a Safety Injection Signal (SIS). This signal will trip the normal rated fan motor in each unit, open the high-capacity service water discharge valve from VHX-1, VHX-2 , and VHX-3, and close the high-capacity service water supply valve to VHX-4. The test to verify the service water valves actuate to their correct position upon receipt of an SIS signal is included in the surveillance test performed as part of Specification 3.7.8, "Service Water System." The safety related fans are normally in operation and only receive an actuation signal through the DBA sequencers following an SIS in conjunction with a loss of offsite power. This actuation is tested by. the surveillance which verifies the energizing of loads from the DBA sequencers in Specification 3.8.1, "AC Sources-Operating." APPLICABLE The Containment Spray System and Containment Air . SAFETY ANALYSES Recirculation System limit the temperature and pressure that could be experienced following a DBA. The limiting DBAs considered relative to containment temperature and pressure are the Loss of Coolant Accident (LOCA) and the Main Steam Line Break (MSLB). The DBA LOCA and MSLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature tr.ans i en ts. . No DBAs are assumed to occur simultaneously or consecutively. The postulated LOCA DBA is analyzed, in regard to containment ESF systems, *assuming the loss of one ESF bus, which is the worst case single active failure, resulting in one train of Containment Cooling rendered inoperable (Ref. 6). The postulated MSLB DBA is analyzed, in regard to containment ESF systems, assuming the loss of two containment spray pumps, which is the worst case single active failure (Ref. 7).

  • Palisades Nuclear Plant 8 3.6.6-4 05/31/99
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  • BASES APPLICABLE SAFETY ANALYSES (continued)

Containment Cooling Systems B 3.6.6 The analysis and evaluation show that under the worst case scenario, the highest peak containment pressure and the peak containment vapor temperature are within the intent of the design basis. (See the Bases for Specifications 3.6.4, "Containment Pressure," and 3.6.5, "Containment Air Temperature," for a detailed The analyses and evaluations considered a range of power levels .and equipment configurations as described in Reference

2. The peak containment pressure case is the 0% power MSLB with initial (pre-accident) conditions of 140°F and 16.2 psia. The peak temperature case is the 102% power MSLB with initial (pre-accident) conditions of 140°F and 15.7 psia. The analyses also assume a response time delayed initiation in order to provide conservative peak calculated containment pressure and temperature responses.

The external design pressure of the containment shell is-3 psig. This value is approximately 0.5 psig greater than the maximum external pressure that could be developed. if the containment were sealed during a period of low barometric pressure and high temperature and, subsequently, the containment atmosphere was cooled with a concurrent major rise in barometric pressure . The modeled Contairiment Cooling System actuation from the containment analysis is based on a response time associated with exceeding the Containment High Pressure setpoint to achieve full flow through the CACs and containment spray nozzles. The spray lines within containment are maintained filled to the 735 ft elevation to provide for rapid spray. initfation. The Containment Cooling System total response time of< 60 seconds includes diesel generator startup (for loss of offsite power), loading of equipment, CAC and containment spray pump startup, and spray line filling. The performance of the Containment Spray System for post accident conditions is given in Reference

3. The performance of the Containment Air Coolers is given in Reference
4. The result of the analysis is that each train of containment cooling can provide 100% of the required peak cooling capacity during the post accident condition.

The Containment Spray System and the Containment Cooling System satisfy Criterion 3 of 10 CFR 50.36(c)(2) . Palisades Nuclear Plant B 3.6.6-5 05/31/99

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  • BASES LCO Containment Cooling Systems B 3.6.6 During a DBA, a minimum of one containment cooling train is required to maintain the containment peak pressure and temperature below the design limits (Ref. 2). One train of containment cooling is associated with Diesel Generator 1-1 and includes Containment Spray Pumps P-54B and P-54C, and air cooler fan V-4A. The other train of containment cooling is associated with Diesel Generator 1-2 and includes Containment Spray Pump P-54A along with CACs VHX-1, VHX-2, and VHX-3 and their associated safety related fans, V-lA, V-2A, and V-3A. To ensure that these requirements are met, two trains of containment cooling must be OPERABLE.

Therefore, in the event of an accident, the minimum requirements are met, assuming the worst case single active failure occurs.

  • The Containment Spray System portion of containment cooling train includes a spray pump, spray headers, nozzles, valves, piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the SIRWT upon an ESF actuation signal and automatically transferring suction to the containment sump
  • The Containment Air Recirculation and Cooling System portion of the containment cooling train which must be OPERABLE includes the three safety related air coolers which each consist of four cooling coil banks, the safety related fan which must be in operation to be OPERABLE, gravity-operated fan discharge dampers, instruments, and controls to ensure an OPERABLE flow path. . CAC fans V-lA, V-2A, V-3A, and V-4A must be in operation to be considered OPERABLE.

These fans only receive a start, signal from the DBA sequencer; they are assumed to be in operation, and are not started by either a CHP or an SIS signal

  • Palisades Nuclear Plant B 3.6.6-6 05/31/99
  • APPLICABILITY ACTIONS Containment Cooling Systems B 3.6.6 In MODES 1, 2, and 3, a OBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature requiring the operation of the containment spray trains and containment cooling trains. In MODES 4, 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Thus, the Containment Spray and Containment Cooling systems are not required to be OPERABLE in MODES 4, 5 and 6. If one or more trains of containment cooling are inoperable, but at least 100% of the containment cooling capability . delivered by a single OPERABLE containment cooling train is 1 available, the inoperable containment cooling train must be restored to OPERABLE status within 72 hours. The components in this degraded condition are capable of providing greater than 100% of the heat removal needs (for the condition of one containment cooling train inoperable) after an accident.

This condition allows for the loss of any two containment spray pumps, or the three required CACs and one containment spray pump, even if they are supplied from different electrical trains of power as long as two containment spray pumps and one cooling fan, or the three required CACs and one containment spray pump are available. The 72 hour Completion Time was developed taking into account the redundant heat removal capabilities afforded by the other Containment Cooling train and the low probability of a OBA occurring during this period. B.1 and B.2 If the Required Action and associated Completion Time of this LCO are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditi-0ns from full power conditions in an orderly manner and without challenging plant systems . Palisades Nuclear Plant B 3.6.6-7 05/31/99

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  • BASES SURVEILLANCE REQUIREMENTS SR 3.6.6;1 Containment Cooling Systems B 3.6.6 Verifying the correct alignment for manual, power operated, and automatic valves, excluding check valves, in the Containment Spray System provides assurance that the proper flow path exists for Containment Spray System operation.

This SR also does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct positions prior to being secured. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation. Rathei, it involves verification, through a system walkdown, that those valves outside containment and capable of potentially being mispositioned, are in the correct position. SR 3.6.6.2 Operating each safety related Containment Air Cooler fan unit 15 minutes ensures that all trains are OPERABLE and that all associated controls are*functioning properly . The 31 day Frequency was developed considering the known reliability of the fan units and controls, the two train redundancy available, and the low probability of a significant degradation of the containment cooling train occurring between surveillances. SR 3.6.6.3 Verifying the containment spray header is full of water to the 735 ft elevation minimizes the time required to fill the header. This ensures that spray flow will be admitted to the containment atmosphere within the frame assumed in the containment analysis. The 31 day Frequency is based on the static nature of the fill header and the low probability of a significant degradation of the water level in the piping occurring between surveillances

  • Palisades Nuclear Plant B 3.6.6-8 05/31/99
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  • BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.6;4 Containment Cooling Systems B 3.6.6 Verifying a total service water flow rate of 3935.5 gpm to CACs VHX-1, VHX-2, and VHX-3, when aligned for accident conditions, provides assurance the design flow rate assumed in the safety analyses will be achieved (Ref. 8). Also considered in selecting this Frequency were the known reliability of the cooling water system, the two train and the low probability of a significant degradation of flow occurring between surveillances. SR 3.6.6.5 Verifying th.at each containment spray pump 1 s deve 1 oped head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has.not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by Section XI of the ASME Code (Ref. 5). Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program. SR 3.6.6.6 and SR 3.6.6.7 SR 3.6.6.6 verifies each automatic containment spray valve actuates to its correct position upon receipt of an actual or simulated actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. SR 3.6.6.7 verifies each containment spray pump starts automatically on an actual'or simulated actuation signal. The 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillances were performed with the reactor at power . Palisades Nuclear Plant B 3.6.6-9 05/31/99

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  • BASES SURVEILLANCE REQUIREMENTS Containment Cooling Systems B 3.6.6 SR and SR 3.6.6.7 (continued)

Operating experience has shown that these components usually pass the Surveillances when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. Where the surveillance of containment sump isolation valves is also required by SR 3.5.2.5, a single surveillance may be used to satisfy both requirements.

  • SR 3.6.6.8 This SR verifies each containment cooling fan actuates upon receipt of an actual or simulated actuation signal. The 18 month Frequency is based on engineering judgement and has been shown to be acceptable through operating experience.
  • See SR 3.6.6.6 and SR 3.6.6.7, above, for further discussion of the basis for the 18 month Frequency . SR 3.6.6.9 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections.

Performance of this SR demonstrates that each spray nozzle is unobstructed and provides assurance that spray coverage of the containment during an accident is not degraded. Due to the passive design of the nozzle, a test at 10 year intervals is considered adequate to detect obstruction of the spray nozzles . Palisades Nuclear Plant B 3.6.6-10 05/31/99

  • BASES REFERENCES
1. 2. 3. 4. 5. 6. 7. 8 . *
  • Palisades Nuclear Plant Section 5.1 FSAR, Section 14.18 FSAR, Sections 6.2 FSAR, Section 6.3 Containment Cooling Systems B 3.6.6 ASME, Boiler and Pressure Vessel Code, Section XI FSAR, Table 14.18.1-3 FSAR, Table 14.18.2-1 FSAR, Table 9-1 B 05/31/99 Hydrogen Recombiners B 3.6.7
  • B 3.6 CONTAINMENT SYSTEMS *
  • B 3.6.7 Hydrogen Recombiners BASES BACKGROUND The function of the hydrogen Recombiners is to eliminate the potential breach of containment due to a sudden hydrogen oxygen reaction.

Per 10 CFR 50.44, "Standards for Combustible Gas Control Systems in Light-Water-Cooled Reactors" (Ref. 1), and the Palisades Nuclear Plant design criteria (Ref. 2), hydrogen recombiners are required to the hydrogen concentration in the containment following a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). The recombiners accomplish this by recombining hydrogen and oxygen in a controlled manner to form water vapor. The vapor remains in containment, thus eliminating any discharge to the environment. The hydrogen recombiners are manually initiated since flammability limits would not be reached. until several days after a Design Basis Accident (DBA).

  • Two 100% capacity independent hydrogen recombiners are provided.

Each consists of controls and a power supply located in the auxiliary building, and a recombiner located in containment. The recombiners have no moving parts. When a hydrogen recombiner is placed in operation, containment atmosphere is drawn through the unit by natural convection and the temperature of the air is raised to a level sufficient for recombination of the hydrogen and oxygen to occur (approximately 1150°F). A single recombiner is capable of maintaining the hydrogen concentration in containment below the 4.1 volume. percent (v/o) flammability limit. Two recombiners are provided to meet the requirement for redundancy and independence. Each recombiner is powered from a separate Engineered Safety Features bus and is provided with a separate power panel and control panel . Palisades Nuclear. Plant B 3.6.7-1 05/31/99

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  • BASES BACKGROUND (continued)

Hydrogen Recombiners B 3.6.7 In order for a Hydrogen Recombiner to be considered OPERABLE,*at least one containment cooling safety related fan powered from the same electrical train must be in operation or available for operation. Fan operation is necessary to ensure that the post-accident containment atmosphere is adequately mixed preventing local hydrogen buildups in excess of the flammability limit. The supporting fan must be powered from the same electrical train as the recombiner, to assure that it would be available in the event of an accident combined with a Loss of Offsite Power and failure of the opposite Diesel Generator (DG). The fan must be started or verified to be in operation when the recombiners are placed in operation because these fans do not receive an SIS start signal (although they do receive a DBA sequencer start signal). If there is no qualified fan available, the associated recombiner would not have all of its required support equipment and would have to be declared inoperable. Only fans V-lA, V-2A, V-3A, and V-4A (associated with VHX-1, VHX-2, VHX-3, and VHX-4 respectively) are qualified for the post-accident containment environment. Recombiner M-698, poweted from the Left Train (DG 1-1), is supported by V-4A, which is the only qualified fan powered from the Left Train. If V-4A was unavailable, Hydrogen Recombiner M-698 would have to be declared inoperable. I Recombiner M-69A, powered from the Right Train (DG 1-2), can *be supported by V-lA, V-2A, or V-3A, all of which are powered from the Right Train. If V-lA, V-2A, and V-3A were all unavailable, Hydrogen Recombiner M-69A would have to be declared inoperable. LCO 3.6.6, "Containment Cooling Systems," also contains *requirements for containment cooling fan OPERABILITY. The restoration time specified for Containment Air Coolers in LCO 3.6.6 is more restrictive than that specified for Hydrogen Recombiners in LCO 3.6.7 . Palisades Nuclear Plant B 3.6.7-2 05/31/99

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  • BASES Hydrogen Recombiners B 3.6.7 APPLICABLE The hydrogen recombiners provide for controlling the bulk SAFETY ANALYSES hydrogen concentration in containment to less than the lower flammable concentration of v/o following a OBA. This control would prevent a containment wide hydrogen burn, thus ensuring the pressure and temperature assumed in the analysis are not exceeded and minimizing damage to safety related equipment located in containment.

The limiting relative to hydrogen generation is a LOCA. LCO Hydrogen may accumulate within containment following a LOCA as a result of: A metal steam reaction between the zirconium fuel rod cladding and.the primary coolant; b. Radiolytic decomposition of water in the Primary Coolant System (PCS) and the containment sump; c. Hydrogen in the PCS at the time of the LOCA (i.e., hydrogen dissolved in the primary coolant and hydrogen gas in the pressurizer vapor space); or d. Corrosion of metals exposed to Containment Spray System and Emergency Core Cooling Systems solutions . To evaluate the potential for hydrogen accumulation in contaihment following a LOCA, hydrogen generation as a function of time following the initiation of the accident is calculated.*. Conservative assumptions discussed in

  • Referente 3 are used to maximize the amount of hydrogen calculated.

The hydrogen recombiners satisfy Criterion 3 of 10 CFR 50.36(c)(2). Two hydrogen recombiners must be OPERABLE. In addition, one *safety related cooling fan associated with each train must be in operation. These requirements ensure OPERABILITY of at l*east one hydrogen recombiners and . adequate mixing of the containment atmosphere in the event of a worst case single active failure. Operation with at least one hydrogen recombiner ensures that the post LOCA hydrogen concentration can be prevented from exceeding the flammabi,lity limit . Palisades Nuclear Plant B 3.6.7-3 05/31/99

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  • BASES APPLICABILITY ACTIONS Hydrogen Recombiners B 3.6.7 In MODES 1 and 2, two hydrogen recombiners are required to control the post LOCA hydrogen concentration within containment below its flammability limit of 4.1 v/o, assuming a worst case single failure.
  • In MODES 3 and 4, both the hydrogen production rate and the total hydrogen produced after a LOCA would be less than that calculated for the OBA LOCA. Also, because of the limited time in these MODES, the probability of an accident requiring the hydrogen recombiners is low. Therefore, the hydrogen recombiners are not required in MODE 3 or 4. In MODES 5 and 6, the probability and consequences of a LOCA are low, due to*the pressure and temperature limitations.

Therefore, hydrogen recombiners are not required in these MODES. With one coritainment hydrogen recombiner inoperable, the tnoperable recombiner .be restored to OPERABLE status within 30 days. In this condition, the remaining OPERABLE hydrogen recombiner is adequate to perform the hydrogen control function. The 30 day Completion Time is based on the availability of the other hydrogen recombi ner, the sma 11

  • probability of a LOCA or MSLB occurring (that would generate .an amount of hydrogen that e:xceeds the flammability limit), and the amount of time available after a LOCA or MSLB (should one occur) for operator action to prevent hydrogen accumulation from exceeding the flammability limit. Required Action A.1 has been modified by a Note stating that the provisions of LCO 3.0.4 are not applicable.

As a result, a MODE change is all owed when one hydrogen recombiner is inoperable. This allowance is based on the availability of the other hydrogen recombiner, the small probability of a LOCA or MSLB occurring (that would generate an amount of hydrogen that exceeds the flammability limit), and the amount of time available after a LOCA or MSLB (should one occur) for operator action to prevent hydrogen accumulation from exceeding the flammability limit . Palisades Nuclear Plant B 3.6.7-4 05/31/99

  • * ** BASES . Hydrogen Recombiners B 3.6.7 ACTIONS B.1 (continued)

SURVEILLANCE REQUIREMENTS If the inoperable hydrogen recombiner cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours.* The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

  • SR 3.6.7.1 Performance of a system functional test for each hydrogen recombiner ensures that the recombiners are operational and can attain and sustain the temperature necessary for hydrogen recombination.

In particular, this SR requires verification that the minimum heater sheath temperature increases to 700°F in 90 minutes. After reaching 700°F, the power is i.ncreased to maximum for approximately 2 minutes and verified to be 60 kW. The 18 month Frequency is based on past operating history of the recombiners and engineering judgement. SR 3.6.7.2 This SR ensures that there are no physical problems that could affect operation. Since the are mechanically passive, they are not subject to mechanical failure. The only credible failures involve loss of power, blockage of the internal flow path, missile impact, etc. A. visual inspection is sufficient to determine abnormal conditions that could cause such failures. The 18 month Frequency for this SR was developed considering that the . incidence of hydrogen recombiners failing the SR in the past is low . Palisades Nuclear Plant B 3.6.7-5 05/31/99

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  • BASES SURVEILLANCE REQUIREMENTS (continued)

REFERENCES SR 3.6.7.3 Hydrogen Recombiners B 3.6.7 This SR requires performance of a resistance to ground test for each heater phase to ensure that there *are no detectable grounds in any heater phase. This is accomplished by verifying that the resistance to ground for any heater phase is 10,000 ohms. This SR also requires the performance of a continuity test to ensure there is no single phase fault or any openings of a single heater bank. The continuity test measures the resistance of each phase to neutral. The 18 month Frequency for this SR was developed considering that the incidence of hydrogen recombiners failing the SR in the past is low. 1. 10 CFR 50.44 2. FSAR, Section 5.1 3. Regulatory Guide 1.7, Revision 1 Palisades Nuclear Plant B 3.6.7-6 05/31/99

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  • ATTAC1ŽENT 3 PALISADES NUCLEAR PLANT SECTION 3.6, CONTAINMENT SYSTEMS CTSMARKUP AND DISCUSSION OF CHANGES \ \
  • 3.6 Lc.o 3.6.1 * .3.6.2 3.6.4 3.6.5 ** . I. ...\_-, I \

t--.cr .... 'o-. .. I CONTAINMENT SYSTEM {f) . CONTAINMENT shall be I /@ @ When the pl ant is labole COLD/SHUTDNI, j i" Moi)E"S I, 2.,3 '°'".I "t I @ fSu.a.ls6 ) '( .2-/ b. When the reacto'r vessel head is r. moved (unless the CS boron concentratio,r( is at REFUELING B ON CONCENTRATION and 1/

c. When posit've reactivity ch nges are made boron dilution or CONTROL R D motion (except for testing one ONTROL ROD at a time ACTION: With one or more c tainment isolation valves in erable (including during performanc of valve testing), maintain t 'ktas-t-one isolation valve OPERABLE
  • each affected penetration t t is open and either: a. Restor the inoperable valves to OPE LE status within 4 hours; or b. Isol te each affected penetrati'on thin 4 hours by use of at least on closed and deactivated automa c valve, closed manual valve, or ind flange; or c. Be in at least HOT SHUTDOWN thin the next 6 hours and in COLD ______ SJ:i_UTDQWN within the follow* 30 hours.
  • The containment ternal pressure shall not exceed: a. 1.5 psig hen above COLD SHUTDOWN and below HOT b. 1.0 p g when in POWER OPERATION or HOT STAN Y. With co ainment internal pressure above the 1
  • it, restore pressure to withi the limit within 1 hour, or be in at l ast HOT SHUTDOWN within the ext 6 hours and in COLD SHUTDOWN withi the followin 30 hours. The containment a erage air temperature shall not exceed 14 F when the plant is above LO SHUTDOWN.

With containment average ai temperature above the limi , restore temperature to within the limit ithin 8 hours, or be in at l ast HOT SHUTDOWN within the next 6 hours d in COLD WN wi in the followin 30 hours.

  • wo independent containment ydrogen recombiners bPERABLE when the plant is in POWER OPE ION or HOT STANDBY. With on recombiner inoperable, restore the i operable recombiner to OPERAS status within 30 da s or be in at lea HOT SHUTDOWN the next 2.hQYJ'S.

Th.e con ainment purge exhaust and air oom supply isolation v lves shall be lo ed closed whenever the plant i above COLD SHUTDOWN. ith one cont nment purge exhaust or air ro supply isolation valv not locked cl ed, lock the valve closed with' 1 hour or be in at 1 st HOT s NOBY within the next 6 hours a a in COLD SHUTDOWN wit n the ollowin 30 hours. ', 3--40 Amendment No. -l-2-8, Revised 05/31/99 1.0 ** I se:u .. 1.l * {continued) A C NNEL FUNCTIONAL TEST shall be the inje ion of a simulated signal i o the channel to verify that it is OPE BLE, including any alarm and rip initiating function. COLP SHUTDOWN The COLD SHUTDOWN condition shall e when the primary coolant SHUTDOWN BORON CONCENTRATION and llY* is less than 21o*F. at I (,. J a. All nonautomatic containment 1solaticin valves and blind flanges are 1----. closed {OPERABLE). The equipment hatch is properly closed and sealed. r---------""" le. At, least one door in each air lock is properly closed

d. All automatic containment isolation valves are OPERABLE or are locked closed. (t3. The_ uncontrolled containment leakage satisfies Specificati.on [ .@ CONTROL RODS CONTROL RODS shall full-length shutdown and regulating rods. ument that provides cycle specific parameter im.its for the current load cycle. These cycle specific parameter imits shall be determ ed for each reload cycle in accordance with Specification

.6.5. Plant* operation within these limits i addressed in individua Specifications. DOSE E !VALENT I-131 shall be that concentration o I-131 {µCi/gm) whic alone would produce the same thyroid dose a the quantity and iso pie mixture of I-131, I-132, I-133, 1-134 a I-135 actually pr ent. The thyroid dose conversion u ed for this calculation s all be those listed in Table III of TID-1484 , "Calculation of istance Factors for Power and Test Reactor tes." 1-2 Amendment No. a+, 43, i4, i+, ii, ' Ž' 28, w, -t-74, Revised 05/31/99 *

  • *
  • S,8,B,'Z.:Z

.. 4.4 4. 5 . 4.5.l 4.5.2 3.t...\.,3. Inte rated Leak Tests The containment integrated leak rate testing shall be performed in accordance with the Containment Leak Rate Testing Program. QC a. Test 1 Local leak rate tests other than Eersonnel Airlock doors between e per orme at psig. (2} Local leak rate t sts for checking airlock door seals within 72 hours of each door opening shall be per armed as follows: '(a} A between e seals test shall be pe formed on the Personnel irlock at 10 psig. (b} A full pr ssure test shall be perfo med on the Emergency Escape Ai lock at 55 psig. A se contact check shall be perfo ed on the Emergency Esca e Airlock following each ful pressure test. Emergen Escape.Airlock door opening, solely for the purpose o strongback removal and perform nee of the*seal contact eek, does not necessi ate additional ressure estin . Acceptable methods of testing are halogen gas detection, soap ... bubble, pressure decay, or equivalent. The local leak rate shall be measured{for/each of/the) r;:I\

  • 10 ltPPj. -,.-{ D (a} Containment penetrations that employ resi ient seal {c) G-(d) (e) gaskets, sealant compounds, or bellows. Fuel transfer tube. Isolation valves on the testable fluid systems' lines penetrating the containment.

Other components which leak repair in order to meet the acceptance criterion for any integrated leak rate test. 4-19 ,, \ Amendment No . .J-2., -+/--3-&, 74, 177 Revised 05/31/99

  • * *

)£5.2 I fi..d'tl>f\ fl._(,t'1ofl I I l 3,t,,Z I CONTAINMENT c,,," t .. '" "'" ""T ,Wal Leak Detect.r6n con inued) b. Acceotance Criteria (b ( 1) and isolation valves ,-(2) The l akage for a Personnel airlock doc seal test shall not exce 0. 023 L *. (3) An cceptable Emergency Escape Airlo contact check co sists of a verification of conti between the @ © s als and the sealing surfaces. If at any ime it is determined that 0.6 L. is exceeded, . repairs hall be initiated immediately. If repairs are not compl ed and conformance to the acce ance criterion of 4.5 .. b(l) is not dem9nstrated withi 48 hours, the plant sh 1 be placed in at least HOT SH DOWN within the next hours and in COLD SHUTDOWN wit 'n the following 30 hours. § If at any time it is determined that total containment leakage exceeds L., within one ho r

  • shall be initiated to glace the
  • least HOT HUT within the next 6 hours and in OLD HUT within l owing 30 hours. (3) (4) ctie.5 f'AocE.3 If the Perso el airlock door seal leakage is grea 0.023 L., or if the Emergency Escape Lock door se contact check fail to meet its acceptance criterion, re irs shall be initiated nwned1ately to restore the door seal t the acceptan criteria of specification 4.5.2.b(2) or 4.5.2.b(3).

In the e ent repairs cannot be completed withi 7 days, the plant s all be placed in at least HOT SHUTDOW within the next 6 hour and in COLD SHUTDOWN within the foll ing 30 hours. lock door seal leakage results in on door causing tot containment leakage to exceed 0.60 L, the door shall be de ared inoperable and the remaining OPE BLE door shall be i d1ately locked closed* and tested wi in-4 hours. As long the remaining door is found to be OP BLE, the provisions f 4.5.2.c(2) do not apply. Repairs sh 11 be initiated irmied1ately to establish conformance w th specification 4.5.2.b(l). In the event conformance to this specification cannot be established within 48 hour the plant shall be placed in at least HOT SHUTDOWN wit n the next 6 hours and in COLD SHUTDOWN within the following 0 hours. Entry and exit is permissible throu a "locked" air lock door' to perform repairs on the affected air lock components. 4-20 CONTAINMENT TSCR REV 2 GtlaA§e 7, Amendment No. +74, -!++, \I Revised 05/31/99

  • 4.5 4.5.2
  • 4.5.3
  • S" pLc.: +:i-c..,__.ti-f i.,,.1 ' The cont nment equipment tch and the fuel ansfer tube sh 1 be tested at e h refueling outag or after each t' used, if that e sooner. b) A full air ock penetration test shall be perf rmed at six-mont intervals.

During the period betw n the tests when CONTAINMENT INTEGRITY . required, a reduc pressure test for the door seals o a full air lock enetration test shall be performed ithin 72 hours af r either each air lock door opening r the first of a s ies of openings.

a. The isolatio valves shall be demonstrated OPERABLE by performan of a cyclin test and verification of isolation time for auto isolation lves prior to declaring the valve to be OPERABLE a ter maintenan , repair, or replacement work is performed on the alve or its a ociated actuator, control, or power circuit. b. c. Each 1 lat1on valve shall be demonstrated OPERABLE by ve fying that each containment isolation right channel or left hannel test signal, applicable isolation valves actuate to the required pos tion during COLO SHUTDOWN or at least once per ref eling cycle. T e isolation time of each power operated or automat c valve shall e verified in accordance with Section XI of the A E Boiler and Pressure Vessel Code. Prior to the reactor going critical after a ref ling outage, a visual check will be made to confirm that all " ocked-closed" manual containment isolation valves are close and locked (except for valves that are open under administrativ control as permitted by LCO 3.6.1). Each three months the isolation valves m st be stroked to the position required to fulfill their safe y function unless it is established that such operation is no practical during plant
  • operation.

The latter valves shall full-stroked during each COLO SHUTDOWN

  • 4-21 Amendment No. -l-2&, Revised 05/31/99
  • * * ) i * ? I \
  • -IQ
  • S'R J.l..\.7-4.5 4.5.4 \ ! l I I I I CONTAINMENT TESTS a. Tendon tion shall be accomplished at five year intervals for the life of he plant. The scheduled inspectio dates for all subsequent
  • nspections may be varied by not mo than plus or minus one year fr m the base schedule.
b. The survei lance tendons shall be randomly bu representatively selected f om each of the following groups:. 1. A mi 1mUll of 4 dome tendons including o e from each dome tend n group. Z. A mini11U* of 4 vertical tendons. 3. A of 5 hoop tendons. For eac inspection, the tendons shall be lected on a random basis e cept that those .tendons whose rout ng has been modified to clear p netrations shall be excluded fro* he sample. c. During each tendon inspection, the follow ng field testing shall be perfo d: 1. s. ift-off readings shal 1 be taken fo 11ch of the surv_eil lance endons. The tests shall include 1 following actions: (b) On1 tendon, randoaly s1l1ctld from each group of tendons during each inspection, shall be subjected to essentially cQ11Plet1 detensioning to 1de tify broken or damaged wires. The simultaneous rne1sure11en of elongation and jacking fore* during retensioning all be l!!ld* at a minimum of three approxi*ately equall spaced of force between th* seating force nd zero. While the tendon is in the det nsionld state, each wire in the tendon will be checkld for con 1nu1ty. Three wires, one froa each of a vertical, a hoop and a dome tendon will be r1110ved and i ntified for inspection.

At each succ1ss1v1 surv1ill1nc1, thJwirts will be selected from diff1rent tendons. Each of ht inspection wires removed will be visually inspected for e idenc1 of corrosion or other deleterious effects and t1k1n for laboratory testing. Th* sheathing filler shall/be inspected visually for color and coverage and SIJIPles shall be obtained for laboratory testing. I I 1 . Tendon anchorage hardwartfsuch as bearing p ates, stressing washers, shias and shall be visually inspected for evidence of corrosion or deleterious effects. Revised 05/31/99 i 1741' October 31, 1996 ,,.; -Ii,.... #-<. .. Str"-ch.-o..f .I .. -te1rit'f 5 .. rvei \\("'Lot°

  • *
  • 4.5 CONTAINMENT TESTS .5.4 (continued)
d. Following the fiel testing of 4.5.4c, the follo ing laboratory testing shall bed ne: 1. Three tensile test specimens shill be cut om each of the three inspect on wires removed (one from e ch end and one from the middle). One additional specimen shal be cut from the wire deten11i ed by field visual inspection to have the greatest amo nt of corrosion.

Each of the wire samples sh al*. be tested fo ultimate strength, yield st ngth, and elon91tion.

2. The she1thi g filler SillPlts shall be tak n fro* each end of each tendon exaained.

Vertical tendon s les shall be taken frOtl the lo r end. SillPles shall be th oughly mixed and 1nalyzed fo reserve alk111nity, water c ntent, and concentrati n of water soluble chlorides nitrates, and sulfides. Analyses shall be perfonaed i accordance with the procedures ind within the acceptance li ts specified in ASHE Codi Secti n XI, Table IWL-2525-1

  • Procedure shall bt establisned to *1ni 1z1 voids and to assure th t the volu .. of sheathing fil er removed has been replaced pon completion of the inspec on and amounts document
  • e. Acceptance criteria shall be as follows: 1. The aver 91 of all measured tendon fo ces for each type of tendon s all bt equal to or gr11t1r t an the minimu* required prestr1s level, of 584 kips per tend n for dome tendons and, 615 kip per tendon for hoop and v1rt cal tendons. The 111asur force in each individual te on .shall not be less than 95 of the predicted force, or (l) .. asured force in not 110r1 han one tendon is between and 951 of the predicted f rc1, and (b) t .. asured forces in two ten ons located adj1cent to t t tendon in (a) above 1r1 no less than 951 of the ld1ct1d forces, and (c) hi .. asured forces in all th re111ining slllJ)lt tendons r1 not less than 951 of the redicted force. If asured force in any tendon i less than 90S of its pred cted force, the tendon shall bt CQllPletely detensioned and det1n1in1tion shill bt *1d1 as to the cause of such an occu rence and corrective action hall bt taken. In addition, all such tendons shall have thei forces measured as add tional tendons in the next s heduled inspection period. The Co1111ission shall be notified in accordance with Paraarach
4. 5 4f. Revised 4-2lb 05/31/99 Amendment No. -t-4, 1-i, 174 October 31. 1996
  • *I
  • I
  • 4.5 4.5.4 (continued) 4.5.5 I 4.5.6 f. ** b. c. 2. Inspe tion wires shall indicate no signif cant loss of section by co rosion or pitting. 3. 4. Tensi e test specimens cut fro* inspecti wires shall be teste for ultimate strength.

Failure a less than 11.78 k.ips of-an one of the test samples requires he Co111nission be notif 'ed in accordance with specificatio 4.5.4f. Tend anchorage hardware shall be fret f significant S, S $"> corr sion, pitting, cracks or other d11e erious effects. c.e.. If any el aent of the prestressing-syst .. fai s to meet the acceptanc criteria of 4.5.41., the reportin provisions of 10 CFR 50.73 Shi l apply. A VT-1 vi ual 1x111in1tion shall bt perfo on the end anchorage concrete surface at the surv11111nc1 tendon anchor points for signs of crack ng, popouts, spilling, or cerrosio

  • Concrete cracks having w dths greater than 0.010 shall bt e aluatld and documented.

The end nchor1g1 concrete surv1t111nc1 ins action interval shall be the s as tendon surveillance interval 1. k widths shall be .. asurtd by usi g' optical comparators ire feeler gauge. Mov ... nts shal b1 111e1sured by using untablt .. chanical ext1nsa111ters.

2. Co cret1 anchorage areas 1r1 accept& 1 if no concrete cracks ar wider than 0.010 inches and no si ns of new or progressive de eriorat1on since the previous ins 1ction are found. 3. Co crttt surface conditions excetdin those stated in 4.5.5c.2 a vt shall bt 1v1lu1tld for the tff t on tendon and . c tain111nt structural integrity.

T t results of evaluation s all bt included in the final surv 111nc1 report. If, as a re ult of 1 prtstressing syst .. insp tion under Section 4.5.4, corrective etensioning of five percent (8) o 110r1 of the total number of da111 ten ons is necessary to restore thei liftoff forces to within the li*its of Specification 4.5.4, a dOlle d1l111in1tion inspection shall be perfo within 90 days following such c rrtctivt retensioning. The results of this inspection shall be reported to the NRC. 4-2lc Revised 05/31/99 ndment No. -t-i, 174 October 31, 1996

  • 3. 6 CONTAINMENT SYSTEM© A,*.,.. Lol-f:.S oPEIVH3L£ I ..J..,_f.2_

N GR Y shall be maintained:* _

a. When the plant is /abote COLD j;" t-'\o!JCS l,Z.,s

'-\I r

  • 3.6.2 3.6.4 3.6.5 b. he reactor vessel hea is removed un ess t PCS boron ti on is at REFUEL G BORON CONCENTRATIO and 1----"".J See S. 'I c. ACTION: With one or e containment isolation valve inoperable (including during perfo ance of valve testing), maint n at least-one isolation valve OPE LE in each affected penetrati that is open and either: a. Res re the inoperable valves to 0 RABLE status within 4 hours; o b. I olate each affected penetrati within 4 hours by use of at ast ne and deactivated aut atic valve, manual val e, bl ind fl ange; or . The contain nt internal pressure shal not exceed: a. 1.5 p ig when above COLD SHUTDOW and below HOT STANDBY and b. 1.0 sig when in POWER OPERATION or HOT STANDBY. With cont inment internal pressure a ave the limit, restore pressure to within t e limit within l hour, orb in at least HOT SHUT WN within next 6 hours and in COLD SHUTDOW within the followin 30 hours. Two i containment hydro en recombiners shall b OPERABLE when the lant is in POWER OPERATION o HOT STANDBY. With on recombiner inop rable, restore the inoperabl recombiner to OPERABL status within 30 a s or be in at least HOT SH TDOWN within the next 1 hours. Th containment purge exhaust a air room supply isola ion valves shall be 1 ocked c 1 osed whenever the p ant is .above COLD SHUTD WN. With one . co tainment purge exhaust or ai room supply isolation . alve not locked c osed, lock the valve closed w'thin l hour or be in a least HOT SJANDBY within the next 6 hour and in COLD SHUTDOWN w thin the 0 5ee 3.1. .. 4) LADO Nl:>.Tt 1.. o..S < ROD AC.\\Cl'l iJOTE 1 Amendment No. Revised 05/31/99
  • *
  • 1. 0 RA A-.1 RA S.I e.A C.."L
  • 7r_.-, -.ipe.c..+-Ce.."":i*:,,._....

--L ""'"D .................... +=N=S (continued) UNCT ONA T T A ANNEL FUNCTIONAL TEST shall be the injecti n of a simulated signal

  • to the channel to verify that it is OPERAS , including any alarm and trip initiating function.

COLP SHUTDOWN The COLD SHUTDOWN condition shall be when the primary coolant is at SHUTDOWN BORON CONCENTRATION and T

  • 1s less tha.n 210"F. NT GR TY INTEGRITY All nonaut atic containment iso t1on valves and bl1 d flanges are I /see closed 0 RABLE . / Tl).i equipment hatch 1s /roperly closed and s,!led.

1."1.1> r.

  • At le.st one door in each air lock is properly

.(§ All automatic containment. i,Olation valves are OPERA E or are locked closed. .

  • The uncontro ed containment leakag satisfies Specif1ca on 4.5. CONTROL RODS CONTROL RODS shall be shutdown and regulating rod . ent that provides cycle specific parame er limits ad cycle. These cycle specific paramet shall be determine for each reload cycle in accordance wi Specification 6.6.5. Plant operation within these limits is addressed in individual SP. cifications.

ENT 1-131 shall be that concentration o 1-131 (µCi/gm) which ala would produce the same thyroid dose as he quantity and isotopic ixture of I-131, I-132, I-133, I-134 an I-135 actually present The thyroid dose conversion factors us for this calculation shall e those listed in Table III o'f TID-14844 "Calculation of Dist ce Factors for Power and Test Reactor Si *1-2 Amendment No. 3*, i4, i+, ii. Hi Revised 05/31/99 See 3.ta.5) See. J_!,._\/ 4.4

  • 5 ** SR3.<...2..I
  • Deleted CONTAI ENT TESTS ection Tests local leak rate tests, other than Personnel Airlock doors between the seals tests, shall be performed at SS psig. (2) local leak rate tests for checking airlock door seals within 72 hours of each door opening shall be performed as follows: SR-3.Co.?.../

[ (a) 1\ b." ( A between the seals test shall be performed on the Personnel Airlock at 10 psig. SR 3,1.1,2,1 3/'j (b) A full pressure test shall be performed tn the Emergency Escape Airlock at 55 psig. A seal contact check shall be performed on the Emergency Escape Airlock following each full pressure test. Emergency Escape Airlock door opening, solely for the purpose of strongback removal and performance of the seal contact check, does not necessitate additional pressure testing . (3) Acceptable methods of testing are halogen gas detection, soap bubble, pressure decay, or equivalent. ($U--;..f.t.' (a) I @' (c) (d) -U ; ...i. )Q . ' <s-..3'.(..I

> Air lock equrent door seals. I olation valves-on the te table fluid systems' netrating the containme
t. Other containment order to meet the leak rate test. 4-19 ents which require tance criterion for k repair in y integrated Amendment No. f.74, 177 3 erfs Revised 05/31/99 (fi.2
  • S'lt 3. ll. z..1 t S'R I 'I I IO SR. 3. ro.z./ **c. *

(_J:JJD 1{,A.-A.I *

  • CONTAINMENT

[fij'ij) .5YSIE.MS Local Leaj/Petection Tests /('continued)) b. Acceotance Criteria (1) The total leakage from all penetrations and isolation valves shall not exceed 0.60 L *. (2) The leakage for a Personnel airlock door seal test shall not exceed O. 023 L *. (3) An Emergency Escape Airlock door seal contact check consists of a verification of contact between the and the sealing surfaces. [C.-/correct i vi,\ctioii)

1) 1 1s e ermrne t at 0.6 L. i exceeded,
  • repairs shal be initiated immediately.

If r. pairs are not completed a conformance to the acceptance riterion of 4.5.2.b(l) is not demonstrated within 48 h rs, the plant shall be aced in at least HOT SHUTDOWN thin the next 6 hours nd in COLfr SHUTDOWN within the E llowing 30 hours. Entry and exit is permissible through a "locked" air lock door to perform repairs on the affected air lock components. 4-20 Revised 05/31/99 Gl:la1:u3e 7, Amendment No. i-74, -!++,

  • 4.5 CONTAINMENT TESTS 4.5.2 {continued)

@:? Test Frequency (1) Indivi al penetrations and containm t isolation valves be l k rate tested at a frequency at least every ref eling, not exceeding a two-yea interval, except as s cified in a and b below: (a) ;.: containment equipment atch and the be ,shall be tested at ch refueling outag or after each time used if that e sooner. s R. 3./o.2.1 w:+i-tDc.FQ Sb, AAl 1, 0 PT Ai O.."d a. Pfr11*I 4.5.3 a. The isolation be demonstrated OPERABLE by p of a cycling st and verification of isolation time fo isolation va es prior to declaring the valve to be OP maintenance repair, or replacement work is performed on or its ass ciated actuator, corttrol, or power circui . rformance auto BLE after the valve b. Each is ation valve shall be demonstrated OPERABL by verifying that o each containment isolation right channel left channel test gnal, applicable isolation valves actuate o their required posi ion during COLD SHUTDOWN or at least once r refueling cycle. c. isolation time of each power operated or a tomatic valve shall verified in accordance with Section XI of he ASHE Boiler and ressure Vessel Code. Prior to the reactor going critical after refueling outage, a visual check will be made to confirm that all "locked-closed" manual containment isolation valves are osed and locked {except for valves that are open under administr. tive control as permitted by LCO 3.6.1). e. Each three months the isolation valv must be stroked to the position required to fulfill therr s fety function unless it is established that such operation is ot practical during plant operation. The latter valves shal be full-stroked during each COLD SHUTDOWN. \I 4-21 < A'DD I -h::> s R 1 lo. i. 1 .:..;. ... J : ... Dr'/ Amendment No. Revised 05/31/99 m, -!-23, -H-4, '----.c<'. Au.D SR 3.{o.1..1 .h,_ .... J-\"': "J: T.S'> /All). *-----.._ ____ ""'-s 3.(Q. l.. "2.. cs +.:. ....... J :.... I.IS"') @ c:;ee 3.iP.I) See I See

  • u: .. o
  • 3.6 3.6.00) 3.6.2 3.6.4 .(ADD A CONTAINMENT N shall be a
  • a. When the plant is Jab04'e COLD SHJltDOWN/r1i

... MoDES 1,Z.,)c....J Lf I c. When 11t>sitive reactivity are made by boron di}Ution or CONTROL ROD motion (except testing one CONTROL R9£l at __ . ACTION: The not exceed: a 1.5 psig when above COLD SHUTD N and below HOT STAND and b. 1.. 0 psi g when in POWER OPE ION or HOT STANDBY. With containment internal pres re above the limit, r tore pressure to within the limit within 1 hou , or be ,in at least HO SHUTDOWN within the next 6 hours and in COLD HUTDOWN within the fo owing 30 hours. The c tainment average air temperature sh 11 not exceed 14Q°F the pla is above COLD SHUTDOWN. With contai ment average air temper ure ab e the limit, restore temperature tow thin the limit within B ours, o be in at least HOT SHUTDOWN within t next 6 hours and in CO UTDOWN within, the following 30 hours Two *ndependent containment hydroge recombiners shall be O ERABLE when th plant is in POWER OPERATION or. HOT STANDBY. With one ecombiner i operable, restore the inoperab recombiner to OPERABL status within O da s or be in at least HOT S TDOWN within the next hours Sc:.e 5.l..\)

  • i <ADD ,;c..**n Or.J c..

_ l-J" 5

  • 1.0 LC.o lf.J
  • 0 J.to.3 <::: *I* '* -'fee.."*

c.c.:ho (continued) NEL FUNCTIONAL TEST shall b the injection of a simulate signal in the channel to verify that 't is OPERABLE, including any larm and t ip initiating function. COLD SHUTDOWN The COLD SHUTDOWN condit on shall be when the primary . ..Coolant is at SHUTDOWN BORON CONCENT TION and T..,. is less than 219-'F. I 3_/g_ .5 All no autom tic containment isolation valves tllind flaf\6es arel I co (OPERABLE). I b. Th' equipment hatch i{properly closed at('d sealed.J--<See. le . At le'ast one door in each a;; lock is properly clo{ed and. sealed. 3.1..:1 .. ) lffi le. uncontrolled leakage satisfies sr-cification 4.5. 1------------------------------*** CONTROL ROOS CONTROL ROOS shall -length shutdown and regulating rods. The COLR is the docume that provides cycle specific parameter imits for the current reloa cycle. These cycle specific parameter mits shall be determined or each reload cycle in accordance with Specification 6.6.5 Plant operation within these limits is ddressed in individual Spe fications. DOSE EQUIVA NT 1-131 shall be that concentration of -131 (µCi/gm) which alon would produce the same thyroid dose as e quantity and isotopic xture of 1-131, 1-132, I-133, I-134 and -135 actually present. The thyroid dose conversion factors use for this calculation shall those listed in Table III of TI0-14844, Calculation of Oista e Factors for Power and Test Reactor Sits." 1-2. Amendment No. i4, i+, ii, Revised 05/31/99 *

  • 4.5 CONTAINMENT TESTS 4.5.2 Local Leak Detection Tests (continued)

I (!:j b. ccep ance r1 er1a (1) The total lea ge from all penetrat ons and isolation valv s shall not ex ed 0.60 L *. (2) The leakag for a Personnel air ck door seal test shal exceed 0. 3 L *. (3) An acce able Emergency Esca e Airlock door seal check consis of a verification f continuous contact the seals nd the sealing surf ces. @./ Correct iyi Act i o"J (1) If at any time it is d termined that 0.60 L. is exceeded, r pairs shall be init ated immediately. If repairs are ompleted and confer ance to the acceptance criterion o 4.5.2.b(l) is not monstrated within 48 hours, the pl nt shall be placed i at least HOT SHUTDOWN within the xt 6 hours and in C D SHUTDOWN within the following 3 hours. A c.'i1or. ).lDt c. 4- If the Pe sonnel airlock door seal leakage i greater than 0.023 L. or if the Emergency Escape Lock d r seal contact check f ils to meet its acceptance criteri n, repairs shall be inHia ed i11111ediately to restore the door seal to the accep ance criteria of specification 4.5 2.b(2) or 4.5.2.b(3). In t e event repairs cannot be complete within 7 days, the pla shall be placed in at least HOT HUTDOWN within the next 6 ours and in COLD SHUTDOWN within e following 30 hours. f air lock door seal leakage resul s in one door causing otal containment leakage to*exce 0.60 L., the door shall declared inoperable and the rema* ing OPERABLE door shall b i11111ediately locked closed* and sted within 4 hours. As ong as the remaining door is found o be OPERABLE, the provis* ns of 4.5.2.c{2) do not apply. pairs shall be initiated immediately to establish con nnance'with specification 4.5.2.b{l). In the event c formance to this specific cannot be established with" 48 hours the plant shall e placed in at least HOT SH DOWN within the next 6 hou s and in COLD SHUTDOWN within the following 30 hours.

  • ntry and exit is penni ible through a "locked" air ock door to perform repairs on the ffected air lock components CONTAINMENT TSCR REV 2 4-20 Gl=iaA!Je 7, . Amendment No. Hi, +74, -l++, -:Po.1c 5 b9 5: Revised 05/31/99
  • *
  • 4.s* 4.5.2 ,,.. r . -.:,.pec:+-u.+.-:i,.., 4 (continued)

(1) Individual enetrations and containment iso tion valves shall ) be leak r e tested-. at a f equency of at l st every 3.lo.I refuelin , not exceeding two-year inter 1, except as s ecifi in a and .('t> elow: * (a) The containment equipment hand the fuel transfer t e shall be tested at eac refueling outage or after ach time used if that be sooner. (b) A full air 1 ck penetration test shall be performe at six-month.i ervals. During the period between t e six-month ests when CONTAINMENT INTEGRITY is re uired, a reduced p essure test for the door seals or a f 1 air lock pen ration test shall be performed withi 72 hours after e'ther each air lock door opening or t first of a series of openings. 14.5.3 Containme.t{t Isolatiori/Valves 1 CL.-. c..c.fJ ow-5;,,__\.,.hJ s: _,___ __ _,.______. 0 S'R 3. <... '1 e .-:"') Mc'Df ,..,, .... MoOE 5 ot re.rro.-.... eJ. w;-\1..-:..., -\'1-Jl. ... s "17-. J .... ,s The isolation valves shall be demonstrated OPERABLE by performance of a cycling test and verification of isolation time for auto isolation valves prior to declaring the valve to be OPERABLE after maintenance, repair, or replacement work is performed on the valve or its associated actuator, control, or power circuit. 1 10 ll@ 4-21 m, m, 74, Revised 05/31/99 *

  • *
  • 4.2 12. Verify the Io ne Ret10val System TSP biskets ir OPERABLE by the fol surveillance
i. Verify the TSP biskets contain b ween 8,300 and 11,00 of TSP eich 18 months. b. Verify that-a sample fro11 t TSP bukets provide adequite pH idjust111nt of borated wate tich 18 months.
  • 113. Cont!/(nment Purge and Isolatj'.bn Vilves I SR 3.(:,.5.I 3.l *. 3.5 The Cont1in111nt Purge and Ventilation Isolation Valves shall be detel"llined losed:

[@ 14. Shutdown Cooling 15. To .eet the sh down cooling requiri111ents of The re uired reactor coolint pump(s) if not in operation should e dete ined to be OPERABLE once per days by verifying correct br ker alignaents and indicated wer iVlilability.

e. he required steia generator ( s sh111 be detenni ned OPERAS verifying the secondary water tvel to be it least b. 12 hours. c. At least one coolant loop or train shall be verified o be in operation and circulitin reictor coolant it leasti nee per 12 hours. /
  • 1. Ver y that the Main Feedwattr Regulating alve and the associ b ass valve close on an actual or simul ted Contiinment High ressurt (CHP) signal once each 18 mon s. Verify that tht Main Feedwater"Regul ing valve and the a sociated bypass valve close on an actual or imulated Ste111 Gener. tor Low Pressure (SGLP) signal.once each months
  • Amendment No. &to, 9Q, Hi, 165 May 19, 1995 4-13 Revised 05/31/99 3.1)
  • **
  • 3.6 3 .6. llD CONTAINMENT SYSTEM CONTAINMENT INTEGR Y shall be maintained:*
a. When the is above COLD SHUTOOW ACTION: See 3.l:i.1) !-.<.. z_ With one or mor containment isolation valves inopera e (including during perfo nee of valve testing), maintain at le st one isolation valve OPERAS in each affected penetration that is open and either: a. Rest e the inoperable valves to OPERABLE s tus within 4 hours; or b. Is ate each affected penetration within hours by use of at least e closed and deactivated automatic val e, closed manual valve, or 1 ind flange; or Be in at least HOT SHUTDOWN within SHUTDOWN within the followin 30 h' The containment internal pressure shall not and in COLO a. 1.5 psig hen ab e COLO SHUT OWN and bel ; and i
b. 1.0 psig when 1n IPOWEo/OPERATION or/BOT STANDBY. MooES 1 .. .J z..I ; With containment Internal pressure above the 11m1tR to within the limit within.I hour, or be in at least I oysHUTD Nlwithrn !Mo the next 6 hours and in wi the fol owing 30 hours. i . MOL>* * ! e con a nmen average a r e era ure s a no exceed 140"F hen the : plant is above COLO SHUTDOWN With containment average air t perature above the limit, restore te erature to within the limit wi in 8 hours, or be in at least HOT SHU WN within the next 6 hours and n COLD H TDOWN within the fol win 30 hours. 3.6.4 Two dependent containment hydrogen r ombiners shall be OPERA E when 3.6.5
  • the plant is in POWER OPERATION or HO STANDBY. With one i perable, restore the inoperable r, combiner to OPERABLE st us days or be in at least HOT SHUT WN within the next 12 h ntainment purge exhaust and air r, om supply isolation valv be eked closed whenever the plant i above COLD SHUTDOWN.

Wi c tainment purge exhaust or air ro supply isolation valve t locked osed, lock the valve closed with' 1 hour or be in at leas HOT TANDBY within the next 6 hours a in COLD SHUTDOWN within the following 30 hours. 3-40 Amendment No. < 4/JD SR ?.t.H. I &..t

....
:crs>@ Revised 05/31/99 *
  • L..C...O Ac;nc.J A
  • 3.6 3. 6. .6. 2 CONTAINMENT SYSTEM CONTAINMENT shall be maint 'ned:* a. When is above COLO HUTOOWN 3.(_.y 3.1,..3 b. Wh the reactor vessel hea is removed (unless t PCS boron ncentration is at REFUE NG BORON CONC£NTRATIO , and l e. Wh *positive reactivity made by boron dilu 1or1QY./ "!..<.:.-'/

C TROL ROD motion except fort testing one CONTROL RO ACTION: I With one or more c tainment isolation valves inoperable ( eluding during perfonnanc of valve testing), maintain at least o e isolation valve OPERABLE

  • each affected penetration that is ope and either: a. Restore he inoperable valves to OPERABLE statu within 4 hours; or b. lsol e each affected penetration within 4 ho rs by use of at least one losed and deactivated automatic valve, or bl nd flange; or c. Be in at least HOT SHUTDOWN within the SHUTDOWN within the following 30 hours; 6 hours and in COLD The internal pressure shall not exceed* a/1.5 ps1g when above COLD SHUTDOWN and below OT STANDBY; and b. 1.0 psig when in POWER OPERATION or HOT ANDBY. / . //With containment internal pressure above t limit, restore pressure t within the limit within 1 hour, or be in least HOT SHUTDOWN within the next 6 hours and in COLO SHUTDOWN wi in the following 30 hours. 3'.o.4/*Two independent gen recombiners shall be OP RABLE when the plant is in POWER OPERATI or HOT STANDBY. With one combiner inoperable, restore the ino rable recombiner to OPERABL status within /, 30 days or be in at least T SHUTDOWN within the next hours. . ee 3.6.5
  • be l ked closed whenever the plant is abo COLO SHUTDOWN.

With ne tainment purge exhaust and air s pply isolation hall con ainment purge exhaust or air room sup y isolation valve no locked cl sed, lock the valve closed within 1 ur or be in at

  • STANDBY within the next 6 hours and in OLD SHUTDOWN within ,Y'le /'following 30 hours. . . 3-40 <ADD !e J.to.5". I c..s

... +,J .-.... -::crs)<E]} 1 o erform . 5ee:!>. lel.J Amendment No. Revised 05/31/99 . Su 3.t-J/

  • e operating status o the containment cool in systems. To ass t operability of eq pMnt required to rellM> heat from the conta'nment in non11l aper ing and e1111r9ency sit tions. ' Soecifications lGO Co .... -ti.,.: ...... ."t."'+ c. ..... \: ")I,,..,_\""
  • S ho.lR._ be. . I A C,.d:.h .. "" E,
  • 9ener1to 1-1 is See J.l) 11 heat exchangers, valves, piping and int1rlocks associated with th1 above CQllPonents and r1quirtd to function during accident conditions are operable.

c.o.,\.*,_ J +..re.:.., Amendment Ho. -t-Q4, 17 2 Revised 05/31/99 September 26, 1996

  • .\-.-.. :,,.. 0 + -c: .... c.oo\: ".l
  • lhtl:J
  • 3.35 ,.-

7 .C,. ton 1 n .. n too ;ng 1nop1r1bl1 for 1 Revised 05/31/99 Amtnd*nt No. 1-Q4, 172 September 26, 1996 Po..\ e Z.. o-1= '-/

  • s pe.c:

0 ?. l,.fp 4.6 !sA¢fY INJECj40N AHDjCONTAINMENT SPRAY SYSTEMS TESTS ' . 4. 6 .1 SI\ 1.i.l,,."11"7 -System tests shall be performed teach reactor refueling int val. A test safety injection signal ill be applied to initiate operation of the system. The safety injection and shutdown cooling '-... system pump motors may bed energized for this test. Th system See 3.S/ will be considered ory if control board indicat*on and visual observations indi t1 that all C()llponents have r, ceived the safety injection signal n the proper sequence and ti ng (ie, the appropriate PUllP break s shall have opened and clos d, and all valves shall have c letlcl their travel). Containment Spray Syste1 At least 1v1ry ten years the spray nozzles shall b1 verified to be open. c. The test w 1 be considered indicate 1 components hav ......._ ___ _:_ 14 J. 3 I flllllll1 (!> 4.6.4 b.

  • Acceptable 11v1ls of p1rfor111nc1 shall be that the ountps start, reach thtir rated h11ds on rec1rculit1on flow, land/Operate fi)f at/ 11 uif f i ft11n /ii nut1j. ' I@ Each Saf1ty Injection Tank fl path shall be verified OP. RABLE within 7 days prior to tach r ctor startup by v1r1fyin each motor operated isolation valve is pin by observing valve'po tion 3 s"'-indication and valve itsel , and locking open the ass iated
  • Y circuit breakers.

The Low Pressure Safet Injlttion flow path shall e verified OPERABLE within 7 diY, prior to each reactor star. up by verifying flow control valve -3006 is open, and its air upply is isolated. 4-24 Amendment No. i-1-, +-3, 9i, W, ffi, 174 October 31, 1996 Revised 05/31/99 ra.ie. ] o+: 4

  • 14. 6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS Syrvejllance Regyirements (continued}

I /4.6.4/ Valv{s (conti/riued) f--@ Th safety injection recirc at1on path shall be v ified OPERABLE thin 7 days prior to ea reactor shrtup by ve ifying valves See J.'S> CV-3027 and 3056 are op and their switches HS 027A, HS-30278, HS-3056A, and HS-3056 are open. d. Each C ntain111nt Spray Ya ve manual control all be verified to be 6 OPE LE at 1 east once ch refu1 l1 ng by CY, 1 ng each va 1 ve from L-A L\ the control roOtl while observing valve op ation at least each 1 110nths. Ate ()'.) E .. rg1ncy llOd1 v1lv d fan o tr1tion will be checked for OPERABILITY during each refu1 sh i:lo * ----.,.---.;-<See 3.7°"'1 @'.) Each fanCand valv* required to function during accident conditions

  • ,.u_ w1ll bl not * <ADO SK. 3.b.la.I c..r rrere,,,.-1-cd
,..

<AD.P 1.(..1..,, 3 e:..r f".e;e .... eJ ,-...._ X:T!) § < flfJ/J 5f!.. J.t..f.. '-/ o.1

,. z
-rr-)@ 4-25 Amend1111nt No. 69, R, W, 174 October 31, 1996 Revised r y .J' L{ 05/31/99 c._c(--
  • * **-3.6 3.6.2 3.6.3 3.6.S -----------

CONTAINMENT SYSTEM CONTAINMENT IN GRITY shall be maint;{ned:*

a. plant is above COLo,sliuTOOWN, b. W n the reactor vessel h oncentration is at REFU 3.t...l) 21."'. z.

With one or more ntainment isolation valves 1n erable (including during performan of valve testing), maintain least one isolation valve OPERABLE n each affected penetration t tis. open and either: a.

  • Restor the inoperable valves to OPE LE status within 4 hoursi or b. Iso te each affected .Penetration 1thin 4 hours by use of at least o closed and deactivated autom ic valve, closed manual valve, or ind flange; or* Be 1n at least HOT SHUTDOWN thin the next 6 hours and in COLD HUTOOWN within the followi 30 hours. The con ainment.internal pressure not exceed: above COLO SHUT WN and below_ HOT STANDBY; b 1.0 psig when With containment internal pr sure above the limit, restor, pressure to within the limit within 1 h r, or be in at least HOT SH DOWN within the next 6 hours and in-co SHUTDOWN within the follo n 30 hours. The average air temperature hall not exceed 140*F whe the is above COLD SHUTDOWN.

With co ainment average -air tempe ature a ve the limit, restore temperature o within the limit within hours, r be 1n at least HOT SHUTDOWN with the next 6 hours and in OLD SHUTDOWN within the following 30 h

  • Two e containment purge exhau and air room supply isolation valves shall e locked closed whenever e plant 1s above COLD SHUTDOW . With one containment purge exhaus or air room supply isolation v ve not locked closed, the valve osed within 1 hour o*r be 1n a least HOT STANDBY within the nex 6 hours and in COLD SHUTDOWN thin the follow1n 30 hours. erform 3*40 ------LC.0 s .o.4 ; s l'\ot c..tpl:c.J.l<

Amendment No. Revised 05/31/99 3.r...s) J.1t>.z.) <..3> _ .. --.... 5 G.l

  • 4.2 Hydrogen Recombiners Each be demonstrated aperable:

s g 3.b.1. [_ erunit tLA, I) that CY

  • 3. \ \ *As meas red b installe or portable temR rature measuring nstruments .
  • 4*12 Amendment No. 811 99, 162 October 26, 1994 Revised 05/31/99
  • *
  • ADMINISTRATIVE CHANGES (A) ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.1, CONTAINMENT A. l All reformatting and renumbering are in accordance with NUREG-1432.

As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications.

  • Editorial rewording (either adding or deleting) is made consistent with NUREG-1432.

During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change. A.2 Not used. A.3 CTS 3.6.1 specifies that "CONTAINMENT INTEGRITY shall be maintained." In the proposed ITS 3.6.1, this is changed to "Containment shall be OPERABLE." The CTS defines "CONTAINMENT INTEGRITY" in the definitions in Chapter 1.0 in parts (a) through (e). The proposed ITS separates the contents of the definition for "Containment Integrity" into the applicable ITS LCOs. Specifically, parts (b) and (e) of the definition are addressed in proposed ITS LCO 3 .6.1. Parts (a) and (d) are addressed in proposed ITS LCO 3.6.3, "Containment Isolation Valves." Part (c) is addressed in proposed ITS LCO 3.6.2, "Containment Air Locks." The proposed ITS 3.6.1, Containment, encompasses the applicable part of CTS 3.6.1 and definition for "Containment Integrity" and therefore this is considered to be an administrative change to match the terminology and structure of NUREG-1432 . Palisades Nuclear Plant Page 1of7 05/31/99

  • A.4 A.5 Not used. Not used. ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.1, CONTAINMENT A.6 CTS 4.5.4, "Surveillances for Prestressing System," CTS 4.5.5, "End Anchorage Concrete Surveillance," and CTS 4.5.6, "Dome Delamination Surveillance," specify surveillance tests which are required to be performed.

The proposed ITS has a Containment Structural Integrity Surveillance Program in TS Chapter 5.0, Administrative Controls which contains the programmatic requirements related to Containment Structural Integrity testing. The information in CS 4.5.4, 4.5.5, and 4.5.6 will be addressed as part of the testing requirements for the Containment Structural Integrity Surveillance Program. In proposed ITS 3.6.1, SR 3.6.1.2 is added to provide a reference to this program by stating "Verify containment structural integrity in accordance with the Containment Structural Integrity Surveillance Program." The Frequency for this program states "In accordance with the Containment Structural Integrity Surveillance Program. " This is considered to be an administrative change to provide a reference to the Containment Structural Integrity

  • Surveillance Program. This change is consistent with NUREG-1432

.. Palisades Nuclear Plant Page 2of7 05/31/99

  • *
  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.1, CONTAINMENT A.7 CTS 4.5.2.c(2) states " ... the plant shall be placed in at least HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 30 hours." In the proposed ITS, Condition B contains the same requirement when Required Actions and Associated Completion Times are not met except that the plant must be placed in MODE 3 within 6 hours and MODE 5 within a total of 36 hours. With respect to temperature between the CTS and ITS, the proposed ITS MODE 3 is specified by being greater than 300°F while the CTS HOT SHUTDOWN is greater than 525°F. While the ITS covers a broader range, for a shutdown there is no effective difference for achieving either temperature in 6 hours since they are specified.

as "greater than.". For the CTS COLD SHUTDOWN versus the ITS MODE 5, the temperature requirement is being less than 210°F versus being less than 200°F in the ITS: This difference of 10 degrees is negligible and has no significant impact on operations. The other parameter which is common between the CTS terms HOT SHUTDOWN and COLD SHUTDOWN and the corresponding ITS MODES 3 and 5 is the reactivity condition. The ITS MODE 3 and 5 are defined, as a reference point, . by a reactivity condition of Keff < . 99. However, in ITS Section 3 .1, the equivalent amount of SHUTDOWN MARGIN is required as that specified in the CTS definitions of HOT SHUTDOWN and COLD SHUTDOWN. Therefore, the amount of SHUTDOWN MARGIN is considered to be same when the requirements of propos.ed

  • ITS 3 .1 are considered.

The time to reach the CTS COLD SHUTDOWN is specified .as " ... withiil the following 30 hours" while the ITS allows a total of 36 hours to reach MODE 5. Therefore, even if the plant is already in MODE 3 the full 36 hours are allowed. This change reflects the usage rules as specified in NUREG-1432. These changes are all considered to be administrative in that there is no significant . impact to operation of the plant and they reflect the terminology and usage rules of NUREG-1432. A.8 The Frequency of proposed ITS SR 3.6.1.3 is modified by a Note which that "SR 3.0.2, is not applicable." The inclusion of this Note is for clarification 'purposes only and is considered Administrative in nature since the CTS does not contain an explicit exemption from the testing frequency of 10 CFR 50, Appendix J. Thus, both the CTS and ITS preclude frequency extensions for Type B and C leakage rate tests. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 3of7 05/31199

  • *
  • A.9 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.1, CONTAINMENT CTS 3.6.la requires that Containment Integrity be maintained "when the plant is above Cold Shutdown.

In the ITS, this statement is equivalent to an Applicability of Modes 1, 2, 3, and 4 since the CTS definition of Cold Shutdown is essentially the same as the ITS of Mode 5. That is, each represents a plant condition at which the average temperature of the primary coolant is below its boiling point at one standard atmosphere. From an analytical perspective, there is no significant difference between the CTS value of 210°F and the ITS value of 200°F for the upper temperature limit in Mode 5. Therefore, replacing the CTS phrase "when the plant is above Cold . Shutdown" with an Applicability of "Mode 1, 2, 3' and 4 II is considered to be Administrative in nature. This change is consistent with NUREG-1432. MORE RESTRICTIVE CHANGES (M) M.1 CTS 4.5.4, CTS 4.5.5 and CTS 4.5.6 specify the tests, inspections, frequencies and acceptance criteria necessary to *ensure the structural integrity of the containment is maintained. Although not explicitly stated, failure to meet these surveillance requirements would result in the containment being declared inoperable (in accordance with CTS 3.6) since containment integrity could no longer be assured. Since the CTS does not contain an explicit Action Statement for an inoperable containment, the plant must be placed in at least Cold Shutdown within 37 hours. In the ITS, periodic verification of containment structural integrity is required by SR 3.6.1.2. Upon failure to meet the requirements of SR 3.6.1.2, Required Actions A and B of ITS 3.6.1 would require the plant to be placed in Mode 5 within 36 hours. The addition of explicit actions in the ITS to address the failure to meet the containment structural integrity surveillances has been characterized as more restrictive since it requires the plant to be placed in Mode 5 within 36 hours, versus the 37 hours allowed in the CTS to reach Cold Shutdown. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 4of7 05/31199

  • *
  • ATTACIWENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.1, CONTAINMENT LESS RESTRICTIVE-REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENT (LA) LA.1 In CTS 1.0, the definition of Containment Integrity addresses, in part, the status of: a) all nonautomatic containment isolation valves and blind flanges, b) the equipment hatch, c) containment air lock doors, and d) all automatic containment isolation valves. Proposed ITS 3.6.1 requires the containment be Operable.

To be Operable, the containment and its associated penetrations m1:1st limit leakage to an acceptable limit. The Bases of ITS 3. 6 .1 describes the isolation devices for the containment penetrations that form part of the leak tight barrier. This includes all automatic and nonautomatic containment isolation valves, blind flanges, air locks, and the equipment hatch. Since the details of what constitutes containment integrity are adequately described in the Bases of ITS 3.6.1, a separate definition is no longer required. Changes to the Bases will be made in accordance with the Bases Control Ptogram as discussed in TS Chapter 5.0, Administrative Controls. This change maintains consistency with NUREG-1432. LA.2 CTS 4.5;2d specifies the test frequency for individual penetrations and containment isolation valves as "at least every refueling outage, not exceeding a two year in_terval except as specified in (a) below". CTS 4.5.2d(l)(a) states that "the containment equipment hatch and the fuel transfer tube shall be tested at each refueling outage or after each time used, if that be sooner." In the ITS, Type Band C containinent leak rate testing is required by SR 3.6.1.3. The Frequency for SR 3.6.1.3 is stated as "in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions." The frequencies for Type Band C leak tests required by CTS 4.5.2d are equivalent to the frequencies stipulated in 10 CFR 50, Appendix J, Option A. Specifically, Appendix J Section III, paragraph D.2(a) states "Type B tests, except tests for air locks, shall be performed during reactor shutdown for refueling, or other convenient intervals, but in no case at intervals greater than 2 years. If opened following a Type A or B test, containment penetrations subject to Type B testing shall be Type B tested prior to returning the reactor to an operating mode requiring containment integrity". Appendix J, Section III, paragraph D.3 states, "Type C tests shall be performed during each reactor shutdown for refueling but in no case at intervals greater than 2 years." Since the testing frequencies of CTS 4.5.2d are equivalent to the testing frequendes in 10 CFR 50, Appendix J, Option A, these details can be deleted from CTS 4.5.2d without any affects on public health and safety. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 5of7 05/31199

  • LA.3 LA.4 ATTAC1ŽENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.1, CONTAINMENT CTS 4.5.2a(3) specifies that "Acceptable methods of testing are halogen gas detection, soap bubble, pressure decay, or equivalent." The acceptable methods of testing are contained in the FSAR Section 5.8.8.2.2.

These testing details are not required to be included in the proposed ITS and will be controlled by the FSAR. Changes to-the FSAR are made in accordance with the provisions of 10 CFR 50.59. This change maintains consistency with NUREG-1432. CTS 4.5.2a.(4) parts (a) through (e) specifies the components which are. required to have a local leak rate measured. Part (b) specifies in part that the air lock door seals are tested. This becomes part of SR 3.6.2.1 and will be addressed with ITS LCO 3.6.2. The remaining information is considered to be details which are already addressed in the FSAR Section 5.8.8.2.2 regarding what must have a local leak rate test performed on it.

  • The testing is performed in accordance with the requirements specified in 10 CFR 5.0. Appendix J, Option A. *The content of the details of the types of things which must have a local leak rate test performed will be controlled in the FSAR. Changes to the FSAR are made i.n accordance with the provisions of 10 CFR 50.59. This change is consistent with NUREG-1432.
  • . LESS RESTRICTIVE CHANGES (L)
  • L.1 The CTS *requirements for Type -B and C leak rate testing are being revised such that the leakage limit for Type Band C testing is < 0.60 4 only during the first plant startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A. After this, the new limit will now become 1. 0 4. This means that if the testing is performed during a refueling outage, the total Type B and C leakage must be . "as-left" at < 0.60 La prior resuming power operations.

Following this, the leakage limit for the remainder of time until the test is performed again becomes 1. 0 4 for the total containment leakage. Overall containment integrity is maintained because the results of Type B and C testing must be compared against the overall c.ontainment leakage limit to ensure that the leakage remains 1. 0 4 . (continued) Palisades Nuclear Plant Page 6of7 05/31/99

  • *
  • L.1 (continued)

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.1, CONTAINMENT CTS 4.5.2.c(l) specifies actions to be taken if .60 La is exceeded for Local Leak Detection Tests. The actions in 4.5.2.c(l) to initiate repairs immediately and shut down if the acceptance criteria of 4.5.2.b(l) is not met (ensuring total leakage from all penetrations and isolation valves shall not exceed .60 La) will no longer apply. The acceptance criteria will be that the overall containment leakage will not exceed 1. 0 4. This change is acceptable since the overall containment leakage requirements of ITS 3. 6 .1 remains valid at all times. Any increase in Type B and C leakage will be evaluated against this limit. The change is considered Less Restrictive since the acceptance criteria of < 0.60 La for Type B and C tests will now become 1.0 4 except during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J. This change is consistent with NUREG-1432. L.2 CTS 3.6. lc requires Containment Integrity be maintained "when positive reactivity changes are made by boron dilution or control rod motion (except for testing one control rod at a time)." In the ITS, containment integrity is ensured by operating the plant within the limits established in LCO 3.6.1, "Containment", LCO 3.6.2, "Containment Air Locks", and LCO 3.6.3, "Containment Isolation Valves." The Applicability for these specifications is MODES 1, 2, 3, and 4. The ITS contains no requirement equivalent to containment integrity in MODE 5, and only a less stringent requirement in MODE 6 referred to as containment closure (LCO 3.9.3, "Containment Penetrations") which is only applicable during Core Alteration or movement of irradiated fuel in the containment. As such, the requirement to maintain containment integrity in the ITS is less restrictive than the requirement in the CTS since the ITS would allow positive reactivity and control rod motion in MODES 5 or 6. This change is acceptable since in MODES 5 and 6, the probability and consequences of an event which would required containment integrity is reduced due to the pressure and temperature limitations in these MODES. Furthermore, by definition MODE 5 requires the reactor to be shutdown (subcritical) by > 1 % b. k/k. In addition ITS 3.1.1, "Shutdown Margin (SDM)" requires SDM be z 2%. In MODE 6 ITS 3. 9 .1, "Boron Concentration" requires sufficient boron concentration to maintain the reactor subcritical by z 5 % .6.p with all control rods withdrawn. These requirements ensure that reactivity changes from boron dilutions or control rod motion can be made without approaching a condition (criticality) in which containment integrity would be needed. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 7of7 05/31199

  • *
  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS ADMINISTRATIVE CHANGES (A) A.l All reformatting and renumbering are in accordance with NUREG-1432.

As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications.

  • Editorial rewording (either adding or deleting) is made consistent with NUREG-1432.

During Improved Technical Specification (ITS) development certain wording' preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is coru;istent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change. A.2 Not used. A.3 CTS 3.6.1.2 specifies that "CONTAINMENT INTEGRITY shall be maintained" with CONTAINMENT INTEGRITY being defined in Chapter 1.0, Definitions. The proposed ITS takes the elements of "CONTAINMENT INTEGRITY" and restructures them into three LCOs: Containment; 3.6.2, Containment Air Locks; 3.6.3; Containment Isolation Valves. Therefore, for proposed.ITS 3.6.2, CTS 3.6.1 is restructured to read "Two Containment Air Locks shall be maintained.".* This change is consistent with NUREG-1432. A.4 Not used . Palisades Nuclear Plant Page 1of10 05/31/99

  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS A.5 In the proposed ITS, ACTION Note 3 is added which states "Enter applicable Conditions and Required Actions of LCO 3.6.1 "Containment," when air lock leakage results in exceeding the overall containment leakage rate." This note provides guidance in accordance with LCO 3.0.6 to specify when other TS should be followed.

This note is considered to be administrative in nature in that it provides guidance on the use and application of the ITS where it wasn't explicitly addressed as part of the CTS. The I intent of ITS 3 .6.2 Action Note 3 is to provide guidance to the ITS user to take the I .appropriate Actions of Specification 3 6 .1 when the containment leakage rate I acceptance criteria is exceeded. Note 3 does not imply any additional requirements, but I . simply provides a cross-reference between two specifications. Thus, the addition of I Note 3 can only be characterized as an "Administrative" change. The corresponding I requirements in the CTS that would be cross-referenced are; CTS 4.5.2c(2) which I requires that corrective actions be initiated "if at any time it is determined that the total I containment leakage rate exceeds La", and CTS 4.5.2d(l)(b) which requires a full air I lock penetration test. The inoperability of air lock doors due to excessive leakage are I treated consistently between the ITS and CTS. That is, if one door is declared I inoperable due to excessive leakage, the remaining Operable door fulfills the I containment isolation function. If both doors are inoperable due to excessive leakage, I the total leakage must be evaluated against the overall containment leakage limit. I

  • These changes are consistent with NUREG-1432.
  • A.6 The CTS Section 1.0, Definitions, contains the definition for "CONTAINMENT INTEGRITY." As stated in Administrative Change A.3, proposed ITS 3.6.2 specifies the requirements for Containment Air Locks. The element of the CTS defined term CONTAINMENT INTEGRITY which relates to air locks is item (c) which states "At least one door in each air lock is properly closed and sealed." This statement forms the Required Actions A.1, B.1, and C.2 in the ACTIONS for Containment Air which ensure that the Containment remains isolated.

The CTS phrase "and sealed" is not included in the proposed ITS as it is implicit in the use of an "OPERABLE" door. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 2of10 05/31/99

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS A.7 CTS 4.5.2.c(3) and (4) state " ... the plant shall be placed in at leas(HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 30 hours." In the proposed ITS, Condition D contains the same requirement when Required Actions and Associated Completion Times are not met except that the plant must be placed in MODE 3 within 6 hours and MODE 5 within a total of 36 hours. With respect to temperature between the CTS and ITS, the proposed ITS MODE 3 is specified by being greater than 300°F while the CTS HOT SHUTDOWN is greater than 525°F. While the ITS covers a broader range, for a shutdown there is no effective difference for achieving either temperature in 6 hours since they are specified as "greater than." For the CTS COLD SHUTDOWN versus the ITS MODE 5, the temperature requirement is being less than 210°F versus being less than 200°F in the ITS. This difference of 10 degrees is negligible and has no significant impact on operations.

The other parameter which is common between the CTS terms HOT SHUTDOWN and COLD SHUTDOWN and the corresponding ITS MODES 3 and 5 is the reactivity condition. The ITS MODE 3 and 5 are defined, as a reference point, by a reactivity condition of Keff < . 99. However, in ITS Section 3 .1, the equivalent amount of SHUTDOWN MARGIN is required as that specified in the CTS definitions of HOT SHUTDOWN and COLD SHUTDOWN. Therefore, the amount of SHUTDOWN MARGIN is considered to be same when the requirements of proposed ITS 3 .1 are considered. The time to reach the CTS COLD SHUTDOWN is specified

  • as " ... within the following 30 hours" while the ITS allows a total of 36 hours to reach MODE 5. Therefore, even if the plant is already in MOPE 3 the full 36 hours are allowed. This change reflects the usage rules as specified in NUREG-1432.

These changes are all considered to be administrative in that there is no significant impact to operation of the plant and they reflect the terminology and usage rules of NUREG-1432. A.8 Not used. A.9 The proposed ITS includes Note 1 at the top of the Required Actions of ACTION B.

  • Note 1 specifies that the Required Actions B. l, B .2, and B. 3 are not applicable if both doors in the same air lock are inoperable and Condition C is entered. This note ensures that Required Actions B.1, B.2, and B.3 do not have to be performed since Condition C requires the appropriate remedial measures if both doors in the same air lock are inoperable.

This is considered to be an administrative change to clarify the usage rules of TS. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 3of10 05/31/99 ATTACHMENT 3 DISCUSSION OF CHANGES

  • SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS *
  • A.10 The proposed ITS includes two Notes in SR 3.6.2.1. The first Note states "An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage tests." This Note provides clarification on the usage rules of TS for this particular application because it is recognized that one closed OPERABLE door will provide containment integrity and should not invalidate the overall air lock . leakage test. This is considered to be an administrative change to clarify the application of the TS usage rules. This change is consistent with NUREG-1432.
  • A.11 The proposed ITS includes Note 1 to the Required Actions for Condition A. The Note states "Required Actions A.1, A.2, and A.3 are not applicable if both doors in the same air lock are inoperable and Condition C is entered." This note is provided to clarify the usage rules in that if both air lock doors are inoperable and Condition C is entered, the Required Actions of Condition A are not entered but rather the Required Actions of Condition Care followed.

This change is consistent with NUREG-1432. A.12 . The proposed ITS includes two Notes in SR 3.6.2.1. The second Note states "Results shall be evaluated against acceptance criteria of SR 3.6.1.3 in accordance with I . 10 CFR 50, Appendix J, as modified by approved exemptions." CTS 4.5.2c(2) starts out by stating "If at any time it is determined that total containment leakage exceeds La** .. " This implies that the air lock leakage, as well as any other containment leakage, is always compared against the overall containment leakage requirements. Therefore, the addition of this note is an administrative change to provide a reminder to compare the test results against the overall containment leakage limit. This change is consistent with NUREG-1432. A.13 Not used. A.14 The Frequency of proposed ITS SR 3.6.2.1 is modified by a Note which states that I "SR 3.0.2 is not applicable." The inclusion of this Note is for clarification purposes -I only and is considered Administrative in nature since the CTS does not contain an I explicit exemption from the testing frequency of 10 CFR 50, Appendix J. Thus, both I the CTS and ITS preclude frequency extensions for Type B and C leakage rate tests. I This change is c;onsistent with NUREG-1432. I Palisades Nuclear Plant Page 4of10 05/31/99 ATTACHMENT 3 DISCUSSION OF CHANGES

  • SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS *
  • A.15 CTS 3.6.la requires that Containment Integrity be maintained "when the plant is above Cold Shutdown.

In the ITS, this statement is equivalent to an Applicability of Modes 1, 2, 3, and 4 since the CTS definition of Cold Shutdown is essentially the same as the ITS definition of Mode 5. That is, each represents a plant condition at which the average temperature of the primary coolant is below its boiling point at one standard atmosphere. From an analytical perspective, there is no significant difference between the CTS value of 210°F and the ITS value of 200°F for the upper temperature limit in Mode 5. Therefore, replacing the CTS phrase "when the plant is above Cold Shutdown" with an Applicability of "Mode 1, 2, 3, and 4" is considered to be Administrative in nature. This change is consistent with NUREG-1432. MORE RESTRICTIVE CHANGES (M) M.1 In ITS 3.6.2, for a containment air lock to be considered Operable, the air lock (door) I interlock mechanism must be Operable. In the CTS, the containment air lock door I interlock mechanism is not required for containment air .lock Operability. As such, I inclusion of the door interlock mechanism in the definitfon of an Operable containment I air lock is considered to be an additional restriction on plant operations. The addition I of this requirement is appropriate since the door interlock mechanism functions to I ensure that a gross breach of containment does not exist when the containment is I required to be Operable by only allowing one air lock door to be opened at a time. I Inclusive with this change is a new Condition (Condition B) which addresses the I inoperability of one or more interlockmechanism, and the corresponding Required I Actions and Completion Times (B.l, B.2, B.3) which provide the appropriate I compensatory measures. The Required Actions are modified by three Notes. Note 1 I states that Required Actions B.1, B.2, and B.3 are not applicable if both doors in the

  • I same air lock are inoperable and Condition C is entered, Note 2 allows containment I ingress and egress under the controls of a dedicated individual, and Note 3 allows lock I closed air lock door in high radiation areas to be verified by administrative means. The I. addition of Notes 2 and 3 are appropriate since they continue to ensure a leak tight I containment barrier is provided.

Note l eliminates conflicts in the ITS usage rules as I described in DOC A.9. Lastly, a new Surveillance Requirement (SR 3.6.2.2) has been I added to perform an interlock test every 18 months. That SR ensures the interlock I feature will functioned as designed. The addition of these More Restrictive changes are I consistent with NUREG-1432. I . \ Palisades Nuclear Plant Page 5of10 05/31/99

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  • M.2 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS Proposed ITS 3.6.2 adds a Condition C which addresses the cases where one or more containment air locks are inoperable for reasons other than Condition A or B. This would include situations such as both air lock doors inoperable at the same time. The* CTS does not specify requirements for air locks other than the door seals must be within leakage limits and the CTS definition of "Containment Integrity" requires that at least one door in each air lock is properly closed and sealed. If proposed Condition C is entered, Required Action (R.A.) A.1 requires action to be initiated immediately to evaluate overall containment leakage rate per LCO 3.6.1, and a door must be verified closed in the affected air lock within one hour. If the limits of LCO 3.6.1 is entered, then the R.A.s and Completion Times are consistent with the shutdown required by CTS LCO 3.0.3. However, R.A. C.3 requires that the air lock be restored to OPERABLE status within 24 hours. These actions are equal to or more restrictive than any actions specified in the CTS for air lock inoperability.

In particular, the requirement to restore the air lock to OPERABLE status within 24 hours is more restrictive than the CTS restoration times. Therefore, this is considered to be a more restrictive change. This change is consistent with NUREG-1432. I M.3 CTS 4.5.2c(3) provides the actions "if the personnel air lock door seal I leakage is > 0.023 La, or if the emergency escape (air) lock door seal contact check I fails to meet its acceptance criterion." CTS 4.5.2c(3) allows both personnel air lock

  • 1 doors to exceed their specified leakage limit, or both emergency escape air lock doors I to exceed their acceptance criterion for up to 7 days provided that leakage from one I door does not cause the total containment leakage to exceed 0.60 La. CTS 4.5.2c(4)

I provides the corrective actions if door seal leakage results in one door causing the total I containment leakage to exceed 0.60 La. This specification does not limit the magrutude I of the leakage rate provided one Operable door in the affected air lock is locked closed. I CTS 4.5.2c(4) requires that repairs be initiated immediately and conformance with the I specification established in 48 hours. If at any time it is determined that containment I leakage exceeds 0.60 La (e.g., from two inoperable doors in one air lock) but is I < 1.0 La, CTS 4.5.2c(l) requires that repairs be initiated immediately and I conformance with the specification established in 48 hours. In ITS 3.6.2, if both doors I in an air lock are declared inoperable for failure to meet their surveillance requirement I

  • acceptance criteria (i.e., leakage for the personnel air lock, and seal contact for the I emergency escape air lock), Condition C is entered. Provided the overall containment I leakage limit does not exceed 1. 0 La, 24 hours are provided to return one door to an
  • I Operable status. The Actions of the ITS are more restrictive than the CTS since they I limit the time that two air lock doors can be inoperable due to excessive leakage. This I additional restriction is appropriate since it limits the duration two air lock doors can be I inoperable to a reasonable time for restoring one door to Operable status. This change I is consistent with NUREG-1432.

I I Palisades Nuclear Plant Page 6of10 05/31/99

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS M.4 CTS 4.5.2c(3) provides the corrective actions if one personnel air lock door exceeds its leakage limit, or one emergency air lock door fails to meet its acceptance criterion for seal contact. In ITS 3.6.2, these same occurrences are addressed by Condition A. The CTS corrective actions require that repairs be initiated to restore the leak tightness of the door seal immediately, and to complete the repairs within 7 days or place the plant in cold shutdown.

The Required Actions of Condition A dictate the Operable air lock door be verified closed within 1 hour (RA A.1), the Operable air lock door be locked closed within 24 hours (RA A.2), and a verification that the Operable door is lock closed once per 31 days (RA A.3). The CTS corrective actions to initiate repairs immediately and to complete the repairs in 7 days has been deleted in the ITS as justified in DOC L.1. The addition of ITS Required Actions A.1, A.2, and A.3 represent additional restrictions on plant operations. These restrictions are appropriate since the remaining Operable door fulfills the function of the containment isolation barrier and, by locking the Operable door and performing periodic verifications, assurance is provided the isolation barrier is maintained. Required Action A.3 is also modified by a Note which allows air lock doors in high radiation areas that are locked closed in accordance with Required Action A.3 to be verified by admillistrative means. This allowance is appropriate since access to these areas is typically restricted* and thereby, the probability of a door being misaligned once it has been verified to be in the proper position is small. This change is consistent with NUREG-1432. LESS RESTRICTIVE CHANGES-REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) LA.1 CTS 4.5.2a(3) specifies that "Acceptable methods of testing are halogen gas detection, . soap bubble, pressure decay, or equivalent." The acceptable methods of testing are contained in the FSAR Section 5.8.8.2.2. These testing details are not required to be included in the proposed ITS and will be controlled by the FSAR. Changes to the FSAR are made in accordance with the provisions of 10 CFR 50.59. This change maintains consistency with NUREG-1432 . Palisades Nuclear Plant Page 7of10 05/31/99

  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS LA.2 CTS 4.5.2d.(l)(b) specifies that a full air lock penetration test is performed at six month intervals, and between this test period, either a reduced pressure test for the door seals, or a full air lock penetration test will be performed with 72 hours after either, each air lock door opening or the first of a series of openings.

Proposed SR 3.2.6.1 required the performance of testing "in accordance with 10 CFR 50, Appendix J, I Option A, as modified by approved exemptions." The frequencies for air lock testing I required by CTS 4.5.2d are equivalent to the frequencies stipulated in 10 CFR 50, I Appendix J, Option A. Specifically, Appendix J Section ill, paragraph D.2(b)(i) states I "air locks shall be tested prior to initial fuel loading and at 6 month intervals thereafter

  • I at an internal pressure not less than Pa." Paragraph D.2(b)(i) states "air locks opened I during periods when containment integrity is required by the plant's Technical
  • I Specifications shall be tested within 3 days after being opened" .... "For .air locks doors I having testable seals, testing the seals fulfills the 3 day test requirements." Since the I testing frequencies of CTS 4.5.2d(l)(b) are equivalent to the testing frequencies in I 10 CFR 50, Appendix J, Option A, these details can be deleted from CTS 4.5.2d(l)(b)

I without any* affects on public health and safety. This change is consistent with I NUREG-1432. I LESS RESTRICTIVE CHANGES (L) L.1 *' CTS 4.5.2c(3) requires that repairs be initiated.immediately for excessive air lock door seal leakage and repairs be completed within a specified time. CTS 4.5.2c(3) states "If air lock door seal leakage is greater than 0.023L.i, repairs shall be initiated immediately to restore the door to less than specification 4.5.2.b(2). In the event repairs cannot be completed within 7 days .. ;." The proposed ITS Required Actions (R.A.) for Condition A do not require repairs to be initiated immediately since the OPERABLE. door can be verified closed in one hour (R.A. A.1) which ensures that the safety function for containment is met. In addition, there is no requirement to complete the repairs in within a specified time since the safety function of the air lock is met as long as the OPERABLE door is closed. These changes are considered to be less restrictive since the specific repair actions are not specified in the proposed ITS. However, there is no impact on safety since an OPERABLE door is required to be verified closed within one hour (R.A. A. l), locked within 24 hours (R.A. A.2) and periodically verified every 31 days to be closed (R.A. A.3) which ensures the containment integrity function is maintained. This change is consistent with NUREG-1432. Palisades Nuclear Plant Page 8of10 05/31/99

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  • L.2 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS CTS 4.5.2c(4) specifies requirements for air lock door seal leakage causing the total containment leakage to exceed . 60 4. These requirements are not included in the proposed ITS 3.6.2 relative to exceeding

.60 4, as suffiCient actions are provided to ensure containment OPERABILITY is maintained. Proposed ITS 3.6.2 Condition A and C provide ACTIONS for air lock doors being inoperable. In accordance with the acceptance criteria in CTS 4.5.2b(2) which is consistent with proposed SR 3.6.2.1, the acceptance criteria for containment air lock door seal leakage is :=:;

  • 023 4 (See . DOC M.3). There is no impact on containment leakage if one air lock door seal leakage is > .60 La as long as there is a closed OPERABLE door in the air lock. The CTS requirement to test the remaining OPERABLE door, which was required to be locked closed, within 4 hours is also not required to be performed included in the proposed ITS. In accordance with the usage rules of NUREG-1432, equipment is assumed to be OPERABLE between surveillance periods unless there is some .reason to believe otherwise.

Therefore*, there is no reason to perform a confirmatory test for

  • OPERABILITY if there is no indication that the OPERABLE door seals are d.egraded.

The proposed ITS Required Actions and Frequency in ACTION A of verifying the OPERABLE door is closed in the affected air lock within 1 hour AND locking the OPERABLE door closed in the affected air lock within 24 hours AND verifying the OPERABLE door is locked closed in the affected air lock once per 31 days are sufficient to ensure that containment OPERABILITY is maintained. As long as. an OPERABLE door exists, the leakage through the other door is irrelevant. Therefore, the acceptance criteria of .60 4 and additional associated actions are no longer required. This change is considered to be less restrictive since additional actions beyond those addressed by proposed ACTION A for air lock door seal leakage being > .023 La are not needed to maintain containment OPERABILITY. This change is consistent with NUREG-1432. L.3 The actions of CTS 4.5.2c have been revised to allow entry and exiting_ of the containment when one door in both containment air locks are inoperable. Proposed ITS 3.6.2 Condition A, Required Action Note 2 states that "entry and exit is permissible for 7 days under admlnistrative controls if both air locks are inoperable." . I This purpose of this Note is to allow ingress and egress through an Operable containment air lock door, that is required to be closed and locked to isolated the air lock penetration, to perform technical specification related and non-technical specification related activities on equipment inside containment. The allowance is permitted for up to 7 days provided appropriate administrative controls are instituted. This allowance is acceptable due to the lQw probability of an event that could pressurize the containment during the short time that the Operable door is expected to be opened. , This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 9of10 05/31/99 ATTACHMENT 3 DISCUSSION OF CHANGES

  • SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS *
  • L.4 CTS 4.5.2c(3) provides the corrective actions if the personnel air lock doors exceed their leakage limit, or the emergency air lock fails to meet its acceptance criterion for seal contact. In ITS 3.6.2, these same occurrences are addressed by Condition A which states "one or more containment air focks with one containment air lock door inoperable." ITS Condition A represents a relaxation from CTS 4.5.2c(3) since the ITS allows one inoperable door in the personnel air lock to exist concurrently with one inoperable door in the emergency air lock. The CTS does not address multiple air locks with inoperable doors. Thus, if multiple inoperabilities occurred, entry into LCO 3.0.3 would be required.

In addition to allowing multiple inoperabilities, the ITS Actions for containment air locks are also modified by a Note which states " Separate Condition entry is allowed for each air lock." The addition of this Note .allows separate . Completion Time tracking for each Condition starting from the time of discovery of the situation that required entry into the specified Condition. The allowance to have one or . more containment air locks with one containment door inoperable, and to permit separate conditfon entry for each inoperability is acceptable since the Operable door in each air lock ensures that a leak tight containment barrier is established and that the appropriate compensatory measures are taken to ensure the barrier (door) is maintained Operable. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 10of10 05/31/99

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  • ATTACHMENT 3 DISCUSSION OF SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES ADMINISTRATIVE CHANGES (A) A.1 All reformatting and renumbering are in accordance with NUREG-1432.

As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications .. Editorial rewording (either adding or deleting) is made consistent with NUREG-1432. During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with r NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change. A.2 Proposed ITS Condition A contains a note which states "Only applicable to penetration flow paths with two containment isolation valves." This note is added to clarify which condition is to be entered based on the containment penetration configuration with respect to isolation valves. This is an administrative change to assist in TS usage. This change is consistent with NUREG-1432. A.3 CTS 3.6.1 specifies that "CONTAINMENT INTEGRITY shall be maintained" with CONTAINMENT INTEGRITY being defined in Chapter 1.0, Definitions. The proposed ITS takes the elements of "CONTAINMENT INTEGRITY" and restructures them into three LCOs : 3.6.1, Containment; 3.6.2, Containment Air Locks; 3.6.3, Containment Isolation Valves. Therefore, for proposed ITS 3.6.3, CTS 3.6.1 is restructured to read "Containment Isolation.Valves shall be OPERABLE." This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 1of11 05/31/99

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES A.4 CTS 3.6.1 ACTION c states " ... the plant shall be placed in at least HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 30 hours." With respect to temperature when comparing the CTS and ITS, the proposed ITS MODE 3 is specified by being greater than 300°F while the CTS HOT. SHUTDOWN is greater than 525°F. While the ITS covers a broader range, for a shutdown there is no effective difference for achieving either temperature in 6 hours since they are specified as "greater than." For the CTS COLD SHUTDOWN versus the ITS MODE 5, the temperature requirement is being less than 210°F versus being less than 200°F in the ITS. This difference of 10 degrees is negligible and has no significant impact on operations.

The other parameter which is common between the CTS terms HOT SHUTDOWN and COLD SHUTDOWN and the corresponding ITS MODES 3 and 5 is the reactivity condition. The ITS MODE 3 and 5 are defined, as a . reference point, by a reactivity condition of Keff < . 99. However, in ITS Section 3 .1, the equivalent amount of SHUTDOWN MARGIN is required as that specified in the CTS definitions of HOT SHUTDOWN and COLD SHUTDOWN. Therefore, the amount of SHUTDOWN MARGIN is considered to be same when the requirements of proposed ITS 3 .1 are considered. The time to reach the CTS COLD SHUTDOWN is specified as " ... within the following 30 hours" while the ITS allows a total of 36 hours to reach MODE 5. Therefore, even if the plant is already in MODE 3 the full 36 hours are allowed. This change reflects the usage rules as specified in NUREG-1432. These changes are all considered to be administrative in that there is no significant impact to operation of the plant and they reflect the terminology and usage rules of NUREG-1432. A.5 The phrase "(except for purge exhaust valve or air room supply valve not locked closed)" is added to CTS 3.6.1 ACTION statement. This statement is added to tell the user that another Condition applies for this situation even though it could be considered that this Condition is applicable. Purge exhaust valve and air room supply valve closure status is addressed by proposed ITS Condition D. This is considered to .be an administrative change to clarify the proper usage and application of the proposed ITS. This change maintains consistency with the usage requirements of NUREG-1432 . Palisades Nuclear Plant Page 2of11 05/31/99

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  • A.6 A.7 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES CTS 3. 6 .1 Action B provides means to isolate a penetration in the event that one containment isolation valve is inoperable in a flow path with two containment isolation valves by stating "Isolate each affected penetration within 4 hour by use of at least one closed and deactivated automatic valve, closed manual valve, or blind flange." NUREG-1432 and the proposed ITS include the provision for a check valve with the flow secured to be used as a mechanism to ensure that the flow path is isolated.

This_ same provision is included in the proposed ITS since it represents a viable method of isolation and a secured check valve could be considered a "deactivated automatic valve." This is considered an administrative change for completeness to explicitly allow a containment isolation valve with flow secured to be used to isolate the flow path where previously it was performed under the CTS term "deactivated automatic valve." This change is consistent with NUREG-1432. \ The proposed ITS has a Condition B for the case where the only two containment isolation valves in the same flow path are inoperable. In the CTS, this situation would result in entering LCO 3.0.3 which allows for one hour to initiate a shutdown and then HOT SHUTDOWN must be achieved within 6 hours and COLD SHUTDOWN must be . achieved within the following 30 hours. The Required Action in proposed ITS Condition B is to isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange within one hour. If this time is not achieved then proposed ITS Condition F is entered which requires that MODE 3 be entered within 6 hours and MODE 5 be entered within 36 hours. Since the CTS and proposed ITS actions result in equivalent Completion Times, this is considered to be an administrative change to maintain consistency with NUREG-1432. A.8 The CTS 3.6.1 has a footnote at the bottom of the page allowing penetration flow paths to be unisolated intermittently under administrative control. *This footnote becomes Actions Note 1 in the proposed ITS. The following exception is added after "Penetration flow paths" to clarify the applicability to purge exhaust valves and air room supply valves since they cannot be opened in MODES 1, 2, 3 or 4: " .... except for 8 inch purge exhaust valves and 12 inch air room supply valves penetration flow paths .... " This is an administrative change since it does not change any requirements but rather specifies what can be opened intermittently. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 3of11 05/31199

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES A.9 In the proposed ITS, ACTION Note 4 is added which is not included in the CTS. Note 4 states "Enter applicable Conditions and Required Actions of LCO 3. 6 .1 "Containment," when leakage results in exceeding the overall containment leakage rate acceptance criteria." This note provides guidance in accordance with LCO 3.0.6 to specify when other TS should be followed.

This note is considered to be administrative in nature in that it provides guidance on the use and application of the ITS where it wasn't explicitly addressed as part of the CTS. These changes are consistent with NUREG-1432. The intent of ITS 3.6.3 Action Note 4 is to provide guidance to the ITS user to take the appropriate Actions of Specification 3 .6.1 when the containment leakage rate acceptance criteria is exceeded. Note 4 does not imply any additional requirements, but simply provides a cross-reference between two specifications. Thus, the addition of Note 4 can only be characterized as an "Administrative" change. The corresponding requirements in the CTS that would be cross-referenced are; CTS 4.5 .2c(2) which requires that corrective actions be initiated "if at any time it is determined that the total containment leakage rate exceeds La", and CTS 4.5.2d(l) which requires individual penetrations and containment isolation valves to be leak tested. As such, containment isolation valves with excessive leakage are treated consistently between the ITS and CTS. That is, total leakage must be evaluated against the overall containment leakage limit . A.10 In proposed ITS, ACTION Notes 2 and 3 are added which are not included in the CTS. These notes involve the application of TS usage rules. Note 2 states "Separate Condition entry is allowed for each penetration flow path." This allows a separate "clock" for each penetration flow path in accordance with the usage rules of NUREG-1432. This would allow more than one flow path to have an inoperable valve. However, since it is only for a limited period of time before each flow path must be isolated, this allowance is acceptable. The application of Note 2 is consistent with the approach used in the CTS to address inoperable CIVs in multiple penetrations. That is, corrective actions are taken on a per penetration basis with the allowed outage time for each inoperability beginning at the time of discovery of the inoperable valve. Note 3 states "Enter applicable Conditions and Required Actions for systems made inoperable by containment isolation valves." Following the philosophy of ITS LCO 3.0.6, "cascading" is not _required unless specified in the individual specifications. If a containment isolation valve is closed, it most likely is going to have an impact on the system it is isolating. Therefore, the note requires that if a system is made inoperable by a containment isolation valve the applicable conditions and required actions should be entered. This philosophy is consistent with the CTS as specified in LCO 3.0.1 (i.e., take the corrective actions for the inoperable component). As such, Note 3 does not impose any additional requirement toi or relaxation from, the CTS but simply clarifies that support system Conditions and Required Actions must be entered and that *reliance exclusively on the support system Required Actions is not permissible. These changes are administrative in nature. The changes made are consistent with NUREG-1432. Palisades Nuclear Plant Page 4of11 05/31/99

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  • I '-A.11 A.12 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES CTS 3.6.1 ACTION a states "Restore the inoperable valve to OPERABLEstatus within 4 hours. " This statement is not included in the proposed ITS because it is not necessary.

The option to restore a component to OPERABLE status is already provided by LCO 3.0.2 and is redundant to specify it in the ACTIONS. This is an administrative change to reflect the usage rules of NUREG-1432. This change is consistent with NUREG-1432. CTS 4.5.3b references " ... each containment isolation right channel or left channel test signal" when discussing testing requirements for containment isolation valves. This phrase is replaced in the proposed ITS with "an actual or simulated actuation signal." The CHP and CHR are the actuation signals which isolation the containment isolation valves and they consist of a right channel and left channel signal. Therefore, the proposed wording envelopes the CTS discussion while providing the reference to the actuation signals. The proposed ITS allows an actual or simulated signal since either signal could function as a "test" and would have no effect on the results of the test. This is considered an administrative change to more accurately describe the test performed. This change maintains consistency with NUREG-1432. A.13 CTS 4.5.3e specifies "Each three months the isolation valves must be stroked to the position required to fulfill their safety funetion unless it is established that such operation is not practical during plant operation. The latter valves shall be full-stroked during each COLD SHUTDOWN." This statement is duplicative of the regulations contained in 10 CPR 50.55a, which incorporates by reference the requirements and frequencies of ASME Section XI, and is not needed to be included in the TS. This is considered an administrative change since the requirements of 10 CPR 50.55a, and by reference, the requirements of ASME Section XI, must be met. This change maintains consistency with NUREG-1432. A.14 CTS Table 4.2-2 Item 13 specifies minimum frequencies for Equipment Tests for Containment Purge and Ventilation Isolation Valves. The requirement states "The Containment Purge and Ventilation Isolation Valves shall be determined closed." The word "locked" is added to form the proposed ITS 3. 6. 3 .1. Adding the term "locked" is considered to be administrative change since CTS 3.6.5 already requires that "The containment purge exhaust and air room supply isolation valves shall be locked closed .... " A.15 Notused . Palisades Nuclear Plant Page 5of11 05/31/99

  • ** A.16 A.17 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VAL YES CTS 3.6.la requires that Containment Integrity be maintained "when the plant is above Cold Shutdown.

In the ITS, this statement is equivalent to an Applicability of Modes l, 2, 3, and 4 since the CTS definition of Cold Shutdown is essentially the same as the ITS definition of Mode 5. That is, each represents a plant condition at which the average temperature of the primary coolant is below its boiling point at one standard atmosphere. From an analytical perspective, there is no significant difference between the CTS value of 210°F and the ITS value of 200°F for the upper temperature limit in Mode 5. Therefore, replacing the CTS phrase "when the plant is above Cold Shutdown" with an Applicability of "Mode 1, 2, 3, and 4" is considered to be

  • Administrative in nature. This change is consistent with NUREG-1432.

The Actions of CTS 3.6.l state, in part, "with one or more containment isolation valves inoperable (including during performance of valve testing) maintain at least one ... " In ITS 3 . 6. 3, it is not necessary to include the parenthetical phrase "including during performance of valve testing" since LCO 3.0.2 stipulates that the Required Actions associated with a Condition be taken whenever the LCO is not met. Thus, excluding this phrase in the ITS is considered an Administrative change since the actual requirements of the CTS remains unchanged . A.18 CTS 3.6.5 provides the corrective actions when one containment purge exhaust or air room supply isolation valve is not locked closed. CTS 3.6.5 requires the valve be locked closed within 1 hour, or be in at least Hot Standby within the next 6 hours and in Cold Shutdown within the following 30 hours. CTS 3.6.5 does not provide corrective actions if two or more containment purge exhaust or air room supply isolation valves are not locked closed. Thus, the actions of LCO 3.0.3 must be invoked. The actions of LCO 3.0.3 are equivalent to the actions of CTS 3.6.5. That is, both specifications require that compliance with the LCO be established within ' 1 hour, or the plant must be placed in Cold Shutdown with the next 36 hours. In ITS 3.6.3, containment purge exhaust or air room supply isolation valve inoperability is addressed by Condition D. Condition D has been further modified from the requirements of CTS 3.6.5 to address "one or more" containment purge exhaust or air room supply isolation valves not locked closed. This change has been characterized*as an Administrative change on the basis that the corrective actions of CTS 3.6.5 for one valve not locked closed are equivalent to the actions of LCO 3.0.3 for multiple valves that are not locked closed. As such, the requirements of the CTS are equivalent to the requirements of the ITS for inoperable containment purge exhaust or air room supply isolation valves. Palisades Nuclear Plant Page 6of11 05/31/99

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES MORE RESTRICTIVE CHANGES (M) M.1 Not used. M.2 CTS 4.5.3d specifies requirements for ensuring that all locked-closed manual containment isolation valves are closed and locked except those open under administrative control. The proposed ITS addresses manual containment isolation valves and blind flanges for both outside and inside containment in proposed SR 3.6.3.2 and SR 3.6.3.3 respectively.

Blind flanges are specified in the proposed ITS sinc.e they are also iso_lation devices which must be in the closed position to function properly. The C_TS requires that this verification be performed prior to "the reactor goii:ig critical after a refueling outage. " The proposed ITS SR 3. 6. 3. 3 revises this to state prior to "entering MODE 4 from MODE 5 if not performed within the previous 92 days." The proposed revision is needed to match the Applicability of Specification 3.6.3 which is MODES 1-4. This verification is modified by the statement "if not performed within the previous 92 days" to ensure that for shutdowns after extended times in MODES 1-4, that the valve status is reverified. The proposed ITS SR 3.6.3.2 addresses the manual containment isolation valves and blind flanges which are outside containment and requires that they be verified closed every 31 days. The addition of the 31 day frequency is to address the fact that since these valves are outside containment, they are in a location which would be more susceptible to inadvertent mispositioning than the devices inside containment and therefore need to be verified more frequently. These changes are considered to. be More Restrictive changes since the CTS only requires the verification to be performed prior to going critical, does not require verification between refueling outages, and does not address blind flanges. This change is consistent with NUREG-1432. M.3 Not used. M.4 Proposed ITS includes Required Action A.2 which states "Verify the affected penetration flow path is isolated." This verification is not required in CTS 3. 6 .1 Actions. This verification is necessary to ensure that the flow path with the inoperable

  • valve remains isolated until the inoperable valve is restored.

Required Action A.2 is also modified by a Note which states "isolation devices in high radiation areas may be verified by use of administrative means." This allowance minimizes personnel' exposure and recognizes that high radiation areas are usually restricted such that the probability of a valve misalignment is small. The Completion Time for Required Action A.2 is "Once per 31 days for isolation devices outside containment AND Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment." This verification is not included in the CTS and therefore this is considered to be a more restrictive change. This change is consistent with NUREG-1432. Palisades Nuclear Plant Page 7of11 05/31/99

  • -* ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES LESS RESTRICTIVE CHANGES-REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) LA.1 In CTS 1.0, Containment Integrity is defined in parts (a) through (e). As stated in Administrative Change A.3, the CTS requirements of "Containment Integrity" are split into three LCOs in the proposed ITS. Parts (a) and (d) of the definition of Containment Integrity relate to containment isolation valves and effectively require that the containment isolation valves are OPERABLE.

In the proposed ITS LCO 3.6.3, the LCO statement is reduced to simply "Containment isolation valves are OPERABLE." Therefore in part (a) the reference to "nonautomatic" containment isolation valves and the requirement that "blind flanges are closed" is not needed. In part (d) the reference ', to "automatic" containment isolation valves and the fact that they are " .. .locked closed" is also not needed. The Bases for proposed ITS LCO 3.6.3 will discuss the different types of containment isolation valves and their associated requirements. Therefore, the additional information can be deleted from the LCO and will be addressed in the Bases. This change is maintains consistency with NUREG-1432. LA.2 Not used.

  • LA.3 Not used.
  • LA.4 Not used. LA.5 Not used. LESS RESTRICTIVE

-CHANGES (L) L.1 CTS Table 4.2.2 item 13.a specifies that the Containment Purge and Ventilation Isolation Valves shall be determined closed "at least once per 24 hours." The proposed ITS SR 3.6.3.1 changes the frequency for verifying the 8 inch containment purge valve and 12 inch air room supply valves are closed from 24 hours to 31 days. This change is reasonable because these valves are not allowed to be open in proposed MODES 1-4. These valves are required to be locked closed, either electrically or by some other means to de-activate the valve. Therefore, the likelihood of the valve being opened is very remote. The 31 day frequency is consistent with the surveillance requirements for other safety related systems which require valve position verification. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 8of11 05/31/99

  • * * -------------ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES L.2 CTS 3.6.1 contains the corrective actions for one or more inoperable containment isolation valves. For penetrations associated with a closed piping system and one CIV, CTS 3.6.1 action "c" would require the plant to be placed in at least Hot Shutdown within 6 hours and in Cold Shutdown within the following 30 hours. In ITS 3.6.3, this same inoperability would be addressed by Condition C which would require the affected penetration be isolated in 72 hours, and verified isolated once per 31 days. The Required Actions of the ITS are less restrictive than. the CTS since they allow 72 hours to isolate the affected penetration versus a forced shutdown.

The 72 hour period provides the necessary time to perform repairs on a failed CIV when relying on an intact closed system. A Completion Time of 72 hours is considered appropriate given the location of certain valves, the reliability of the closed system, and that 72 hours is typically provided for losing one train of redundancy throughout the ITS. If the closed system and associated CIV were both inoperable (as a containment boundary), the plant would be in LCO 3.0.3 since there is no specific Condition specified. Although ITS Required Action C.2 is an additional restriction on plant operations since it requires a verification that the affected penetration is isolated once I per 31 days, the overall change related to the addition of ITS Condition C is : j

  • characterized as Less Restrictive.

Lastly, Required Action C.2 is modified by a Note I which states "isolation devices in high radiation areas may be verified by use of I administrative means." This allowance minimizes personnel exposure and recognizes I that access to high radiation areas is restricted such that the probability of a valve I misalignment is small. This change is consistent with NUREG-1432 as modified by I TSTF-30. . I L.3 CTS 4.5.3d requires that prior to the reactor going critical a refueling outage, a visual check will be made to confirm that all locked-closed manual containment isolation valves are closed and locked except for valves that are open u11der administrative controls. In the ITS, manual valve position verification is required by proposed SR 3.6.3.2 and SR 3.6.3.3. SR 3.6.3.2 and SR 3.6.3.3 are modified by a Note which states "valves and blind flanges in high radiation areas may be verified by use of administrative means. " The allowance to verify valve position by use of "administrative means" is a relaxation from the CTS requirement to perform a "visual check. " The proposed change allows special provisions for high radiation areas to minimize _personnel exposure while still keeping track of containment isolation valve status. This change is considered acceptable since a verification requirement still exists, and because high radiation areas are restricted such that the probability of valve misalignment is small. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 9of11 05/31/99

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES L.4 CTS 4.5.3d requires that prior to the reactor going critical after a refueling outage, a I visual check will be made to confirm that all "locked-closed" manual containment I isolation valves are closed and locked except for valves that are open under I administrative controls.

In the ITS, manual valve position verification is required by I proposed SR 3.6.3.2 and SR 3.6.3.3. The requirements of SR 3.6.3.2 and SR 3.6.3.3 I are less restrictive than the CTS since they only apply to valves that are "not locked, I sealed, or otherwise secured in position. " This proposed change is acceptable since I / these valves are verified closed when they are locked, sealed, or otherwise secured in I position. Administrative programs provide the appropriate controls to assure valves I that are normally locked, sealed, or otherwise secured in position are in the correct I *position. This change is consistent with NUREG-1432 as modified by TSTF-45 Rev .1. I L.5 CTS 4.5.3a, CTS 4.5.3b, and CTS 4.2 Table 4.2.2, Items 13.a and 13.b contain details I that are not necessary to describe, or are not pertinent to, any actual regulatory I requirement. As such, these details are proposed for deletion. Specifically:

  • I . CTS 4.5.3a describes the testing necessary for CIVs prior to declaring .the valves I Operable after maintenance, repairs, or replacement work is performed on the valve or I its associated actuator, control, or power circuit. Explicitly stating these tests as they I relate to maintenance activities is unnecessary since the technical specifications stipulate I the level of performance that must be met for an Operable CIV in the associated I Surveillance Requirements. . I CTS 4.5.3b states that.each CIV shall be demonstrated Operable by verifying

... valves /. actuate to their required position*" during Cold Shutdown. or at least once per refueling cycle" .. The.phrase "during Cold Shutdown" is intended to describe the plant condition which best facilitates testing the applicable (automatic) CIV s. The phrase "at least once per refueling cycle" establishes the frequency for the test. Specifying the plant condition at which CIV testing is performed (i.e., Cold Shutdown) is a detail which is not pertinent to the actual requirement for testing CIV s. The LCO Applicability for CIVs stipulates the plant conditions when CIVs are required to be Operable. Testing within the Applicability is governed by valve Operability, testing in plant conditions outside the Applicability has no impact oil safety. CTS 4.2 Table 4.2.2, Item 13a states tlie containment purge and ventilation isolation valves are determined closed "by checking the valve position indicator in the control room". The intent of verifying valve position is to ensure that the valve is in its correct position. Specifying that valve position be verified "by checking the valve position indicator in the control room" does not constitute a requirement assumed in the safety

  • analyses.

Rather, it simply provides a for assuring the valve is in the correct position. Since the valves may be locked closed electrically, mechanically, or by other

  • physical means, stipulating "by checking the valve position indicator in the control room" is an inappropriate detail not pertinent to the actual requirement. (continued)

Palisades Nuclear Plant Page 10of11 05/31/99

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  • L. 5 (continued)

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES CTS 4.2 Table 4.2.2, Item 13b states th(f containment purge and ventilation isolation valves are determined closed by performing a leak rate test "between the valves." Specifying that a leakage rate test be performed "between the valves" does not constitute a requirement assumed in the safety analyses. Rather, it simply provides a for conducting the leakage rate test. Since the above details are not necessary to describe, or ate not pertinent to, any actual regulatory requirement, they can be deleted without an impact to public health and safety. These changes are consistent with NUREG-1432. L.6 CTS 4.5.3b requires that each containment isolation valve be demonstrated Operable by I verifying it actuates to its required position. In the ITS, an equivalent test is required I by SR 3.6.3.6. However, ITS SR 3.6.3.6 does not require containment isolation valves I under administrative controls that are locked, sealed, or otherwise secured in position *1 to be tested. This is because these valves are already in the position necessary to I perform the containment isolation function. Thus, there is no need to verify these valve I can reposition on an actual or simulated actuation signal. The allowance not to test I containment isolation valves that are locked, sealed, or otherwise secured in position is I a relaxation from the requirements of the CTS. This change is consistent with I NUREG-1432. I Palisades Nuclear Plant Page 11of11 05/31/99

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.4, CONTAINMENT PRESSURE ADMINISTRATIVE CHANGES (A) A. l All reformatting and renumbering are in accordance with NUREG-1432.

As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is made consistent with NUREG-1432. During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432 .. Since the design is already approved by the NRC, adding more details

  • does not result in a technical change.
  • A.2 CTS 3 .6.2a specifies a limit of 1.5 psig for containment pressure "when above COLD SHUTDOWN and below HOT STANDBY." The proposed ITS changes this to "in MODES 3 and 4." THE CTS COLD SHUTDOWN is defined with respect to temperature as Tave being less than 210°F. Therefore, for the CTS, above COLn SHUTDOWN would be 210°F or above. The ITS MODE 4 is >200°F. This temperature difference has very little impact operationally and is not considered a significant change. For the CTS HOT ST AND BY, Tave is > 525 °F. Therefore, "below" HOT STANDBY would be "below" 52S°F. In the CTS, there is no defined
  • plant operating conditions from a PCS temperature perspective for the region above* COLD SHUTDOWN and below HOT STANDBY. Proposed ITS MODE 3 is defined ** as being :::o: 300°F. Therefore, from a plant temperature perspective the region "above" COLD SHUTDOWN and "below" HOT STANDBY is considered to be the equivalent to the proposed ITS MODE 3 and 4. Therefore, specifying the condition for which the -1.5 psig limit applies as MODE 3 and 4 in the proposed ITS is considered to be the same as the CTS "above COLD SHUTDOWN and below HOT STANDBY." Based on this discussion, this is considered to be an administrative change. This change is consistent with NUREG-1432 . Palisades Nuclear J;>lant Page 1of2 05/31/99
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  • A.3 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.4, CONTAINMENT PRESSURE CTS 3.6.2b specifies a limit of 1.0 psig for containment pressure when in "POWER OPERATION or HOT STANDBY." The proposed ITS changes this to "MODES 1 and 2." The CTS term POWER OPERATION is defined to be when the reactor is critical and power is > 2 % of RA TED POWER. The proposed ITS definition of MODE 1 is with Keff 0.99 and power> 5% power. The CTS HOT STANDBY condition is when Tave > 525 °F, any of the CONTROL RODS are withdrawn and power is < 2 % of RATED POWER.* The proposed ITS definition of MODE 2 is 5 % power and Keff is 0. 99. From an operational standpoint there is minimal difference between the conditions described by the CTS HOT STANDBY and POWER OPERATION and the ITS MODES 1 and 2 because the combination of MODES 1 and 2 define conditiol1S similar to the combined CTS limits. Therefore, this is considered to be an administrative change. This change is consistent with NUREG-1432.

A.4 CTS 3.6.2 requires that if containment pressure is not restored to within limit, that the plant be placed in HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 30 liours. In the proposed ITS the CTS term HOT SHUTDOWN is replaced by MODE 3 and COLD SHUTDOWN is replaced by MODE 5. This is considered to be an administrative change since the effect on operations is similar. For a discussion of the change from the CTS operating condition definitions to the proposed ITS MODES refer to the information associated with Chapter 1.0. This change is consistent with NUREG-1432 .. MORE RESTRICTIVE CHANGES (M) M.1 CTS 3.6.2 specifies limits for containment internal pressure and actions to be taken if . the limits are not met. However, no surveillances are provided to verify that containment internal pressure is within limits. The proposed ITS requires that the containment pressure be verified within limits every 12 hours. This change is a more restrictive change since the CTS did not contain an explicit verification of containment pressure on a periodic basis. This change is consistent with NUREG-1432. LESS RESTRICTIVE CHANGES-REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) There were no "Removal of Detail" changes made to this specification. LESS RESTRICTIVE CHANGES (L) There were no "Less Restrictive Changes" made to this specification. Palisades Nuclear Plant Page 2of2 05/31/99 j _________________________________________ _

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.5, CONTAINMENT AIR TEMPERATURE ADMINISTRATIVE CHANGES (A) A.1 All reformatting and renumbering are in accordance with NUREG-1432.

As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is made consistent with NUREG-1432. During Improved Technical Specification (ITS) development certain wording preferences or* English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection., This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details* . does not result in a technical change: . * ...__ A.2 The Palisades CTS 3.6.3 requires that contailwlent average air temperature cannot exceed 140°F when the plant is above COLD SHUTDOWN. In the proposed ITS, COLD SHUTDOWN is replaced with MODE 5. The operational impact of the terminology difference is minimal and therefore this is considered to be an adrµinistrative change. See Section 1.0 for a more detailed discussion on the comparison of the CTS defined operation conditions and the ITS MODES. This change is consistent with NUREG-1432. A.3 The Palisades CTS 3.6.3 requires that if containinent average air temperature is not restored then the plant must be placed in HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. In the proposed ITS the CTS term HOT SHUTDOWN is replaced by MODE 3 .and COLD SHUTDOWN is replaceq by MODE 5. This is considered to be an administrative change since effect on operations is similar. For a discussion of the change from the. CTS operating condition definitions to the proposed ITS MODES refer to the information associated with Chapter 1.0. This change is consistent with NUREG-1432. TECHNICAL CHANGES -MORE RESTRICTIVE (M) M .1 CTS 3 6. 3 specifies a limit for containment average air temperature and actions to be taken if the limits are not met. However, no surveillances are provided to verify that

  • containment temperature is within limit. ,The proposed ITS requires that the containment pressure be verified within limit every 12 hours. This change is a more *restrictive change since the CTS did not contain an explicit verification of containment temperature on a periodic basis. This change is consistent with NUREG-1432.

Palisades Nuclear Plant Page 1of2 01/20/98

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.5, CONTAINMENT AIR TEMPERATURE LESS RESTRICTIVE CHANGES-REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENT (LA) There were no "Removal of Details" changes associated with this specification LESS RESTRICTIVE CHANGES (L) There were no "Less Restrictive" changes associated with this specification . Palisades Nuclear Plant Page 2of2 01/20/98
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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS ADMINISTRATIVE CHANGES (A) A.1 All reformatting and renumbering are in accordance with NUREG-1432.

As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing *Technical Specifications. Editorial rewording (either adding or deleting) is made consistent with NUREG-1432. During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change. A.2 CTS 3.4.3 states "Continued power operation with one component out of service shall be as specified in Section 3.4.2, with the permissible period in inoperability starting at the time that the first of the two components became inoperable." This explanatory information on the usage rules of technical specifications is addressed in the proposed ITS Section 1.3, Completion Times and does not need to be addressed in the Actions of proposed ITS 3.6.6. This is considered to be an administrative change since the requirements on complying with the completion times is addressed in the proposed ITS. This change is consistent with NUREG-1432. A. 3 . CTS 3 .4 1 specifies that the reactor shall not be made critical "except for low-temperature physics tests." Palisades no longer performs physics tests at low temperature conditions as were performed as part of initial criticality activities. The Special Tests Exceptions which are used are addressed in proposed ITS Section 3 .1. This is considered to be. an administrative change since Palisades no longer performs the tests referred to in CTS 3.4.1. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 1of8 05/31/99

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  • A.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS CTS 3.4.4 specifies that valves, interlocks and piping that are directly associated with the "above" (CTS 3.4.1) components shall meet the same requirements as listed for that component CTS 3.4.5 specifies that valves, interlocks and piping which is associated with the containment cooling system and not covered by Specification 3.4.4 may be inoperable for no more than 24 hours if it is required to function during an accident.

These requirements are addressed by the definition of OPERABILITY which requires . that all associated equipment be OPERABLE. In the proposed ITS all equipment in a particular train which is required to function during an accident must be OPERABLE and all equipment in the train will have the same Completion Time. This is an administrative change since the requirement remains that all equipment in a train of

  • containment*cooling must be OPERABLE.

This change is consistent with NUREG-1432. A.5 CTS 4.6.3 specifies that the containment spray pumps be tested at intervals "not to exceed three months." In the proposed ITS SR 3.6.6.5 this is changed to be "In accordance with the lnservice Test Program." The Inservice Test Program will specify the frequency for pump testing. CTS 4.6.3 specifies that the containment spray pumps be tested at intervals "not to exceed three months." In the proposed ITS SR 3.6.6.5 this is changed to be "In accordance with the lnservice Test Program." The lnservice Test Program will specify the frequency for pump testing. The Inservice Test Program requires testing these pumps once every 3 months. This is considered to be .an administrative change since the three month frequency will now be contained in the lnservice Test Program. This change is consistent with NUREG-1432. This is considered to be an administrative change since the three month frequency will now be contained in the Inservice Test Program. This change is consistent with NUREG-1432 .. A.6 CTS 4.6.2a specifies in part that for the Containment Spray System Test: "Operation of the system is initiated by tripping the normal actuation instrumentation." The proposed ITS add the following phrase "by an actual or simulated actuation signal." The allowance to use an actual or simulated actuation signal is based on the fact that the channel being tested cannot differentiate between an actual or simulated signal and either initiation can demonstrate system OPERABILITY. This is considered to be an administrative change since the system testing requirements have not changed. This change is consistent with NUREG-1432. A. 7 CTS 4.6.2.a requires a system test of the containment spray system that is initiated by tripping the normal actuation instrumentation. Since the containment spray pumps are

  • an intergral part of the system this CTS requirement verifies that the pumps are capable of being started by an actuation signal. ITS SR 3. 6. 6. 7 requires verification of automatic containment spray pump start upon receipt of an actuation signal. This reformatting of the CTS 4.6.2.a requirement into ITS SR 3.6.6.7 is an administrative change that does not result in addition or removal of any requirement.

This change is consistent with NUREG-1432. Palisades Nuclear Plant Page 2of8 05/31199

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS MORE RESTRICTIVE CHANGES (M) M.1 CTS 3 .4.1 for Containment Cooling Systems specifies that "The reactor shall not be made critical.. . unless all of the following conditions are met: " CTS 3 .4. 2 states in part "During power operation, one of the components listed in Specification 3 .4 .1 above ... " CTS 3.4.3 states in part "During power operation, the requirements of Specification 3 .4.1 may be modified .... " These statements represent the "Applicability" for the CTS. The proposed ITS has an Applicability of MODES 1, 2, and 3. This is a more restrictive change since the CTS only requires equipment to be OPERABLE before the reactor is critical and discusses the allowed inoperabilities "During power operation." In the proposed ITS, MODE 2 is where the reactor is first critical during a startup and CTS power operations would extend up through MODE 1 . *Extending the proposed . ITS Applicability down to MODE 3 is appropriate since there is sufficient heat and energy in the primary coolant system to require containment cooling systems to be OPERABLE.

This change is considered to be a more restrictive change since the , equipment is required to be OPERABLE over a larger range of operating conditions. This change maintains consistency with NUREG-1432. M.2 CTS 3.4.2 and 3.4.3 require in part that if inoperable components are not restored to OPERABLE status then the reactor shall be placed in a hot shutdown condition within 12 hours. In the proposed ITS, the requirement is to place the plant in MODE 3 within 6 hours. The CTS HOT SHUTDOWN is nominally equivalent to the proposed ITS MODE 3 as discussed in Section 1.1 and therefore, there is little effect in adopting the proposed ITS MODE 3 parameters. However, while the CTS allows 12 hours to reach HOT SHUTDOWN, the .Proposed ITS only allows 6 hours. Therefore, the proposed ITS is considered to be more restrictive. This change is consistent with NUREG-1432. M.3 The proposed ITS includes SR 3.6.6.3 which verifies that the containment spray header is filled to the 735 ft elevation every 31 days. The containment spray headers are filled to the 735 ft elevation to ensure a rapid response time for containment spray to be initiated following a design basis event which requires containment spray. By having the spray header column filled with water the containment spray system is able to begin its function of maintaining containment pressure to within acceptable levels sooner than if that volume of water had to be pumped up to that point. This practice of filling the containment spray header to elevation 735 ft is currently being used. However, since the requirement to verify this level every 31 days is not in the Palisades TS, this is considered to be a more restrictive change. This change is consistent with NUREG-i432 . Palisades Nuclear Plant Page 3of8 05/31/99

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  • M.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS CTS 4.6.5b specifies that for the Containment Air Cooling System "Each fan and valve required to function during accident conditions will be exercised at intervals not to exceed three months." The proposed ITS in SR 3.6.6.2 changes this interval to every 31 days. The change from every 3 months to 31 days is a more restrictive change. This change is consistent with NUREG-1432.

M.5 The proposed ITS SR 3.6.6.1 requires that each containment spray manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position be verified to be in the correct position every 31 days. The CTS does not require a periodic verification of containment spray valve position. This surveillance is added to give increased confidence that the valves are in their appropriate position. This change is considered to be a more restrictive change since this valve position verification is not contained in the existing CTS. This change is consistent with NUREG-1432. M.6 The proposed ITS SR 3.6.6.4 requires that the total service water flow rate is > 3935.1 gpm to the Containment Air Coolers (CACs) VHX-1, VHX-2, and This requirement is provided to ensure that the CACs are receiving the necessary flow . as specified in the safety analysis to ensure that sufficient heat is removed following a . design basis accident to ensure that the containment pressure and temperature is maintained within the analyzed values. This requirement to verify service water flow to the CACs is not included in the CTS. Therefore, including this requirement in the proposed ITS is considered to be a more restrictive change. This change is consistent with NUREG-1432. M.7 CTS 3.4.2 and 3.4.3 specifies that following the reactor being placed in a hot shutdown condition due to an inoperability, that if the inoperability is not restored "within an additional 48 hours," the reactor shall be placed in a cold shutdown condition within 24 hours. The allowance for the additional 48 hours before a further shutdown is . required is not included in the proposed ITS. In the proposed ITS Condition B specifies the shutdown requirements as discussed in other DOCS and does not include this intermediate completion time to restore an inoperable component. The Completion Time of Required Action B.2 to be in MODE 4 in a total of 30 hours provides a reasonable time to restore an inoperable containment cooling train to OPERABLE status. The deletion of this 48 hour allowance is considered to be a more restrictive change since it decreases the amount of time to restore an inoperable containment cooling train to OPERABLE status. This change is consistent with NUREG-1432 .. Palisades Nuclear Plant Page 4 of 8 05/31/99

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  • M.8 Not used. ---------------

ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS M.9

  • Proposed SR 3.6.6.2 requires the containment cooler fans to be operated for :?: 15 minutes every 31 days. CTS 4.6.5.b only requires that each fan be exercised, with no minimum operating time. The addition of a required duration to this surveillance requirement is a more restrictive change, consistent with NUREG-1432.

LESS RESTRICTIVE CHANGES -REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) LA.1 In the CTS 3.4.la and b, the major components (i.e., Containment Air Coolers and Containment Spray Pumps) of the containment cooling trains are listed with an associated diesel generator and required to be OPERABLE. In CTS 3.4.lc it also requires that "All heat exchangers, valves, piping, and interlocks associated with the above components and required to function during accident conditions are . OPERABLE." In the proposed ITS the LCO will simply read: "Two *containment cooling* trains shall be OPERABLE." In the proposed ITS, the information regarding what will comprise a train of containment cooling is not included in the TS LCO but will be addressed in the Bases. Changes to the Bases are made under.licensee control in accordance with the Bases Change Control Program which is addressed in Chapter 5 of the ITS. This change is consistent with NUREG-1432. LA.2 Not used. LA.3 CTS 4.6.3a for Containment Spray Pumps states in part that "Alternate manual starting between control room console and the local breaker shall be practiCed in the test program." CTS 4.6.3b goes on to require that the pumps " ... operate for at least fifteen minutes. " The level of detail provided by these statements is more appropriate for plant test procedures. The overall is that the pumps be tested to ensure their OPERABILITY. The requirement to alternate manual starting between the control room console and the local breaker and to run the pump for at least fifteen minutes is not included in the proposed ITS and will be addressed by plant test procedures. Changes to plant test procedures are made in accordance with the plant change control process. These changes are consistent with NUREG-1432 .

  • Palisades Nuclear Plant Page 5 of 8 05/31/99 I
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  • LA.4 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS CTS 4.6.4d specifies that ."Each Containment Spray Valve manual control shall be verified to be OPERABLE at least once each refueling by cycling each valve from the control room while observing valve operation at least each 18 months." The requirement to cycle the containment spray valves by manual control are details addressed by the Inservice Testing Program which are not included in the proposed ITS. The Inservice Testing Program is discussed in ITS Section 5.5. The details of the program are located in the plant procedures which implement the program. Therefore, the requirement to manually cycle the containment spray valves will not be included in the proposed ITS but will be located in the plant procedures which implement the Inservice Testing Program. Proposed SR 3.6.6.6 will demonstrate that the valves are OPERABLE and actuate to the correct position on receipt of an actuation signal. This change is consistent with NUREG-1432.

LESS RESTRICTIVE CHANGES (L) L.1 CTS 3.4.2 allows one of the components listed in Specification 3.4.1 to be inoperable for a period of up to seven days. CTS 3.4.3 allows a total of two of the components listed in Specification 3.4.la orb to be inoperable at any one time to be inoperable for up to 24 hours. Specification 3 .4 .1 lists requirements for equipment associated with each diesel generator. For diesel generator 1-2 the equipment specified includes: three Containment Air Coolers (V-lA, V-2A, V-3A); two service water pumps (P-7A, P-7C); one containment spray pump (P-54A); and one component cooling water pump (P-52B). For diesel generator 1-1 the equipment specified includes: one service water pump (P-7B); two containment spray pumps (P-54B, P-54C); two component cooling water pumps (P-52A, P-52C). The requirements for service water pumps and component cooling water pumps are addressed in Technical Specification (TS) Section 3. 7, Plant Systems. The proposed ITS modifies this requirement to allow one or more trains of containment cooling to be inoperable for 72 hours as long as at least 100% of the cooling capability equivalent to a single OPERABLE containment cooling train is available .. One train of containment cooling is defined to be the three safety related containment air coolers and containment spray pump P-54A, or containment spray pumps P-54B and P-54C. Therefore, the proposed ITS would allow four items associated with diesel generator 1-2 to be inoperable for 72 hours, or two items associated with diesel generator 1-1 to be inoperable for a total of 72 hours, or any combination of equipment that leaves 2 spray pumps, or one spray pump with three required containment air coolers OPERABLE. This is considered to be a less restrictive change since a train of containment cooling (see above) is allowed to be inoperable in the proposed ITS'3.6.6 for 72 hours as opposed to the CTS 3.4.2 allowance of one of the components from CTS 3.4.1 for 7 days being inoperable or the CTS 3.4.3 allowance that a total.of two of the components listed in Section 3.4.la or 3.4.lb may be inoperable. (continued) Palisades Nuclear Plant Page 6 of 8 05/31/99

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS LESS RESTRICTIVE CHANGES (L) L.1 (continued)

For CTS 3.4.3 the change to the one or more trains inoperable for 72 hours is also less restrictive since the CTS only allows a total of two of the components from either 3.4.la or 3.4.lb to be inoperable for 24 hours. CTS 3.4.5 also is changed from 24 hours to 72 hours for inoperabilities involving valves, interlocks, or piping associated with the containment cooling system and not covered by CTS 3.4.4. These changes are acceptable since the proposed ITS still requires that the equivalent of one train of containment cooling be OPERABLE and the timeframe for the equipment to.be inoperable equipment is only 72 hours.

  • The support role and requirements of the service water pumps and component cooling water pumps to the containment air coolers and containment spray pumps are addressed . by LCO 3. 0. 6 and the Safety Function Determination Program as well as the definition of OPERABILITY.

This helps to ensure that the containment cooling trains have adequate support equipment to perform their function. This change maintains consistency with the intent of NUREG-1432 . L.2 CTS 3.4.2 and 3.4.3 require that "If the inoperable component (train in ITS) has not been restored to operability within an additional 48 hours the reactor shall be placed in a cold shutdown condition within 24 hours." The proposed ITS requires that for Required Actions and Associated Completion Times not met the plant is eventually required to be placed in MODE 4 within a total of 30 hours. The "additional 48 hours" allowance of the CTS for restoring inoperable components before taking actions is not included.in the proposed ITS. A shutdown to MODE 4 required in the proposed ITS places the plant out of the MODE of Applicability. CTS 3 .4.1 only requires that the. equipment associated with Containment Cooling Systems be OPERABLE prior to making the reactor critical.. As discussed in "more restrictive" change M. l, the proposed ITS will require that the containment cooling equipment be OPERABLE in MODES 1, 2, and 3 which is the MODES for which the equipment is required to function in an accident. The CTS requirement to place the plant in a cold shutdown condition (i.e, < 210 °Fin the CTS) if the required actions and associated completion times are not met is overly restrictive since the equipment is not needed below 300°F. Therefore, only requiring that the plant be placed in MODE 4 is considered to be a less restrictive change. This change maintains consistency with NUREG-1432 . Palisades Nuclear Plant Page 7of8 05/31/99

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  • I -L.3 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.6, CONTAINMENT COOLING SYSTEMS CTS 4.6.2a specifies that for the Containment Spray System test, "The test shall be performed with the isolation valves in the spray supply lines at the containment blocked CTS 4.6.2c specifies that for the Containment Spray System test, "The test will be considered satisfactory if visual observations indicate all components have operated satisfactorily." This level of detail, with respect to how the test is performed and what the testing acceptance criteria are, is more appropriately addressed in plant procedures.

Removal of these details from the Technical Specifications will not affect the requirement to perform the containment spray test, however the methodology and acceptance criteria will not be included in the proposed specifications. This change is consistent with NUREG-1432. L.4 CTS 4.6.3a and 4.6.3b contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. As such, these details are proposed for deletion. CTS 4.6.3a states that "alternate manual starting [of the containment spray pumps] between the control room console and the local breaker shall be practiced in the test program. " The ability to demonstrate the manual starting capability of the pumps from various plant locations is not assumed in the safety analyses and is not relevant to demonstrating that the pumps are capable of meeting their intended safety function. CTS 4.6.3b requires that the containment spray pumps " ... operate for at least fifteen minutes." The ability to demonstrate pump operation for an arbitrary period such as fifteen minutes is not representative of the ability of the pump to operate for an extended period following a postulated accident, as assumed in the safety analyses. Since the above details are not necessary to describe, or are not pertinent to, any actual regulatory requirement, they can be deleted without an impact of public health and safety. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 8 of8 05/31/99

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6. 7, HYDROGEN RECOMBINERS ADMINISTRATIVE CHANGES (A) A.1 All reformatting and renumbering are in accordance with NUREG-1432.

As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is made consistent with NUREG-1432. During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change. A.2 CTS 3.6.4 requires that two hydrogen recombiners be OPERABLE when the plant is in "POWER OPERATION or HOT STANDBY." The proposed ITS has an Applicability of MODES 1 and 2. The CTS reactor operating conditions of POWER OPERATION . and HOT ST AND BY are nominally operationally equivalent as discussed in TS Section 1.1, Definitions. Therefore, this Is considered to be an administrative change to adopt the MODE definitions from NUREG-1432. A.3 CTS Table 4.2.2 Item 11.b. l requires that a channel calibration of all recorilbiner . instrumentation and control circuits be performed at least once per refueling cycle (18 months in ITS). There are no actuation circuits for the hydrogen recombiners as they are manually initiated from a control panel in the auxiliary building. The system is verified to be OPERABLE by a functional test which verifies that the heaters are working properly. Therefore, the CTS wording discussing a "channel calibration of all recombiner instrumentation and controls" is replaced by a system test in proposed ITS SR 3.6.7.1. This is considered to be an administrative change since all of the recombiner functions will be tested to be working properly as part of the system functional test. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 1of4 05/31/99

  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.7, HYDROGEN RECOMBINERS MORE RESTRICTIVE CHANGES (M) M.l CTS 3.6.4 for the hydrogen recombiners requires that the inoperable recombiner be
  • restored to OPERABLE status within 30 days or be in at least "HOT SHUTDOWN within the next 12 hours.,; The proposed ITS requires that if the Required Action and Associated Completion Time not met the plant must be in MODE 3 within 6 hours. The CTS HOT SHUTDOWN is nominally operationally equivalent to the ITS MODE 3 as discussed in TS Section 1.1. However, the proposed ITS requires that MODE 3 be reached in 6 hours as opposed to the 12 hours to reach HOT SHUTDOWN which is allowed by the CTS and therefore this change is considered to be more restrictive.

This 1 change is consistent with NUREG-1437.

  • . M.2 Not used. LESS RESTRICTIVE CHANGES-REMOVAL OF DETAILS TO LICENSEE *CONTROLLED DOCUMENTS (LA) LA. l CTS Table. 4.2-2 Item 11.a specifies the acceptance criteria for the hydrogen recombiner functional test. This acceptance criteria is not included in the proposed ITS SR 3. 6. 7 .1 as it will be contained in the Bases. The proposed ITS SR 3. 6. 7 .1 will require that the functional test be performed but the Bases will contain the more detailed information with respect to performance and acceptance criteria.

Changes to the Bases are made in accordance with the Bases Change Control Program which is specified in TS Chapter 5.0. This change is consistent with NUREG-1432. LA.2 Not used. LA.3 CTS Table 4.2-2 Item b.3 specifies that for the'electrical circuit integrity check "The resistance to ground for any heater element shall be ::: 10,000 ohins." This acceptance criteria is not included in the proposed ITS and will be contained in the Bases.

  • Changes to the Bases are made in accordance with the Bases Change Control Program which is specified in TS Chapter 5.0. This change is consistent with NUREG-1432.

LA.4 Not used. LA.5 CTS 3.6.4 requires two containment hydrogen recombiners to be OPERABLE. Descriptive information about the design of the system is not included in the proposed ITS and is included in the Bases. Changes to the Bases are made in accordance with the Bases Change Control Program which is specified in TS

  • Chapter 5.0. This change is consistent with NUREG-1432.

Palisades Nuclear Plant Page 2 of 4 05/31199

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  • ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.7, HYDROGEN RECOMBINERS LESS RESTRICTIVE CHANGES (L) L.1 In proposed ITS 3. 6. 7 Action A, a note is added which is not contained in the CTS. The note states "LCO 3.0.4 is not applicable." By specifying that LCO 3.0.4 is not applicable, the plant would be allowed to change MODES while relying on an Action statement for an inoperable hydrogen recombiner.

This is acceptable given the long Completion Time of 30 days if one recombiner is inoperable which indicates that this system is not of immediate importance following an accident. Another redundant hydrogen recombiner also is required to be OPERABLE and is sufficient to maintain hydrogen below the flammability limit in the event of a design basis accident. This is considered to be a less restrictive change since in the proposed ITS, the plant would be allowed to change MODES while relying on an Action as opposed to the CTS which would not allow this. This change is consistent with NUREG-1432. L.2 CTS Table 4.2-2 Item 11.a specifies that a hydrogen recombiner unit functional test be performed at least once per 6 months for each unit. The proposed ITS SR 3. 6. 7 .1 specifies that this test be performed every 18 months. The change from 6 months to 18 months is acceptable because the hydrogen recombiners generally pass the surveillance when performed at the 6 month frequency. In*addition, the design of the recombiner is simple in that it is simply relying on heat to recombine hydrogen and oxygen. A 100% capacity fully redundant hydrogen recombiner is also available and maintained OPERABLE. This is considered to be a less restrictive change since the hydrogen recombiner test will be performed every 18 months instead of the current every 6 months. L.3 CTS Table 4.2-2, Footnote *to Item 11.a requires that the minimum recombiner heater sheath temperature increase be measured using installed or portable temperature monitoring instrumentation. This detail is being removed from proposed SR 3. 6. 7 .1 and its associated Bases. This footnote provides no substantive requirement since any measurement of temperature must use either installed or portable temperature monitoring instrumentation. This change will not effect operation of the facility and is consistent with NUREG-1432. L.4 CTS Table 4.2-2, Item 11.b.3 requires the integrity check of the recombiner heater electrical circuits to be performed "immediately following" the system functional test of Item 11.a. This detail is an initial condition for performing the measurements required by proposed SR 3.6.7.3. As such, this detail is more suitable for inclusion in plant procedures and is proposed to be deleted from the specifications. This is consistent

  • with other testing where particular equipment configurations or conditions are required to be met to successfully complete the surveillance testing. This change is consistent with NUREG-1432.

Palisades Nuclear Plant Page 3 of 4 05/31/99

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  • L.5 ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.6.7, HYDROGEN RECOMBINERS CTS Table 4.2-2, Item lLb.2 lists examples of abnormal conditions that are to be sought during the visual examination of the hydrogen recombiners.

These examples are neither exhaustive or explicit. These examples do not provide direction or limitations regarding the conduct of the required examination. As such, these details are more suitable for inclusion in plant procedures and are proposed to be deleted from the specifications. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 4of4 05/31/99

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  • RELOCATED REQUIREMENTS (R) There were no "Relocated" changes m.ade to this section . Palisades Nuclear Plant Page 1of1 ATTAC1ŽENT 3 DISCUSSION OF CHANGES SECTION 3.6, CONTAINMENT 05/31/99
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  • ADMINISTRATIVE CHANGES (A). ATTACHMENT 4 No* SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.6, CONTAINMENT SYSTEMS The Palisades Nuclear Plant is. converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants..

Some of the proposed changes involve reformatting, renumbering, and rewording of Technical Specifications. These changes, since they do not involve technical changes to the Technical Specifications, are administrative. This type of change is connected_ with the movement of requirements within the current requirements, or with the modification of wording which does not affect the technical content of the current Technical Specifications. These changes will also include nontechnical modifications of requirements to conform to the Writer's Guide or provide consistency with the Improved Standard Technical Specifications in NUREG-1432. Administrative changes are not intended to add, delete, or relocate any technical requirements of the current Technical Specifications. In accordance with the criteria set forth in 10 CFR 50.92, Palisades Nuclear Plant staff has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion . 1.. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? The proposed changes involve reformatting, renumbering, and rewording of the existing Techriical Specification. These modifications involve no technical changes to . the existing Technical Specifications. The majority of changes were done in order to be consistent with NUREG-1432. During the*development of NUREG-1432, certain wording preferences or English language conventions were adopted. The changes are administrative in nature and do not impact initiators of analyzed events. They also do not impact the assumed mitigation of accidents or transient events. Therefore, the changes do not*involve a significant increase in. the probability or consequences of an accident previously evaluated.' Palisades Nuclear Plant Page 1of6 05/31/99

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  • 2. 3. ATTAC1ŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.6, CONTAINMENT SYSTEMS Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes involve reformatting, renumbering, and rewording of the existing Technical Specifications. The changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The changes will not impose any new or different requirements or eliminate any existing requirements. Therefore, the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in margin of safety? The proposed changes involve reformatting, renumbering, and rewording of the existing Technical Specifications. The changes are administrative in nature and will not involve any technical changes. The changes will not reduce a margin of safety because they have impact on any safety analysis assumptions. Also, since these changes are administrative in nature, no question of safety is involved. Therefore, the changes do not involve a significant reduction in a margin of safety . MORE RESTRICTIVE CHANGES (M) The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Some of the proposed changes involve adding more restrictive requirements to the existing Technical Specifications by either making current requirements more stringent or by adding new requirements which currently do not exist. These changes may include additional requirements that decrease allowed outage time, increase frequency of surveillance, impose additional surveillance, increase the scope of a specification to include additional plant equipment, increase the applicability of a specification, or provide additional actions. These changes are generally made to conform with the NUREG-1432. In accordance with the criteria set forth in 10 CFR 50.92, Palisades Nuclear Plant staff has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion . Palisades Nuclear Plant Page 2 of 6 05/31/99

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  • 1. 2. 3. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.6, CONTAINMENT SYSTEMS Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
  • The proposed changes provide more stringent requirements than previously existed in the Technical Specifications.

These more stringent requirements do not result in operation that will increase the probability of initiating an analyzed event. If anything, the new requirements may decrease the probability or consequences of an analyzed event by incorporating the more restrictive changes. The changes do not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements conti1.1ue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis. Therefore, the changes do not involve a significant increase in the probability or consequences of an accident previously Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed changes provide more stringent requirements than previously existed in the Technical Specifications. The changes do not alter the plant configuration (no new or different type of equipment will be installed) or make changes in the methods governing normal plant operation. The changes do impose different requirements. However, these changes are consistent with the assumptions in the safety analyses and licensing basis. Therefore, the changes do not create the possibility of a new or different kind of accident froni any accident previously evaluated. Does this change involve a significant reduction in margin of safety? The proposed changes provide more stringent requirements than previously existed in the Technical Specifications. Adding more restrictive requirements either increases or has no impact on the margin of safety. The changes, by definition, provide additional

  • restrictions to enhance plant safety. The changes maintain requirements within the safety analyses and licensing basis. As such, no question of safety is involved.

Therefore, the changes do not involve a significant reduction in a margin of safety . Palisades Nuclear Plant Page 3of6 05/31199

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  • ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.6, CONTAINMENT SYSTEMS LESS RESTRICTIVE CHANGES -REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engmeering Plants." Some of the proposed changes involve moving details (engineering, procedural, etc.) out of the Technical Specifications and into a licensee controlled document.

This information may be moved to the ITS Bases, FSAR, plant procedures or other programs controlled by the licensee. The removal of this information is considered to be less restrictive because it is no longer controlled by the Technical Specification change process. Typically, the information moved is descriptive in nature and its removal conforms with NUREG-1432 for format and content.

  • In accordance with the criteria set forth in 10 CFR 50.92, Palisades Nuclear Plant staff has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration.

The following is provided in support of this conclusion.

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. * * \. The proposed changes move details from the Technical Specifications to a licensee controlled document. The removal of details from the Technical Specifications is not assumed to be an initiator of any analyzed event. The proposed changes do not reduce the functional requirement or alter the intent of any specification. As such, the consequences of an accident remain unchanged. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident

  • previously evaluated . Palisades Nuclear Plant Page 4of6 05/31199
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  • 2. 3. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.6, CONTAINMENT SYSTEMS Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes move detail from the Technical Specifications to a licensee controlled document. The changes will not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. The changes will not impose different requirements, and adequate control of information will be maintained. The changes will not alter assumptions made in the safety analysis and licensmg basis. Therefore, the changes will not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? Margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at wlllch protective or mitigative actions are initiated. There are no.design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. The proposed changes remove details from the Technical Specifications and place them , under licensee control. Removal of these details is acceptable since this information is not directly pertinent to the actual requirement and does not alter the intent of the

  • requirement.

Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to licensee controlled document without a significant impact on safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. RELOCATED CHANGES (R) There were no "Relocated" changes associated with this section Palisades Nuclear Plant Page 5of6 05/31/99

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  • LESS RESTRICTIVE CHANGES (L) ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.6, CONTAINMENT SYSTEMS The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Changes have been proposed which involve making the requirements in the Current Technical Specifications (CTS) less restrictive.

A description of the less restrictive change and corresponding No Significant Hazards Consideration are provided on the following pages for each Specification as applicable . Palisades Nuclear Plant Page 6of6 05/31/99

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  • ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.1, CONTAINMENT LESS RESTRICTIVE CHANGE L.1 The CTS requirements for Type Band C leak rate testing are being revised such that the leakage limit for Type B and C testing is < 0.60 4 only during the first plant startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A. After this, the new limit will now become s 1. 0 L.i. This means that if the testing is performed during a refueling outage, the total Type Band C leakage must be "as-left" at < 0.60 4 prior resuming power operations.

Following this, the leakage limit for the remainder of time until the test is performed again becomes 1.0 L.i for the total containment leakage. Overall containment integrity is maintained because the results of Type B and C testing must be compared against the overall containment leakage limit to ensure that the leakage remains s 1.0 La. CTS 4.5.2.c(l) specifies actions to be taken if .60 L.i is exceeded for Local Leak Detection Tests. The actions m 4.5.2.c(l) to initiate repairs immediately and shut down if the acceptance criteria of 4.5.2.b(l) is not met (ensuring total leakage from all penetrations and isolation valves shall not exceed .60 L.i) will no longer apply. The acceptance will be that the overall containment leakage will riot exceed 1. 0 La. This change is acceptable since the overall containment leakage requirements of ITS 3. 6 .1 remains valid at all times. Any increase in Type B and C leakage will be evaluated against this limit.

  • The change is considered Less Restrictive since the acceptance criteria of . 60 La for Type B and C tests will now become 1.0 La.except during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J. This change is consistent with NUREG-1432.
1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

The proposed change increases the acceptance criteria for Type B and C tests from .60 La to 1.0 La for the total containment leakage. Previously the .60 L.i for Type B and C tests acted as a "trigger point" to ensure actions were taken such _that the overall acceptance criteria of 1.0 La were not violated. In addition the actions to initiate immediate repairs are not required unless total containment leakage exceeds 1. 0 L.i. The 1.0 La limit for total containment leakage remains in the proposed ITS. Any leakage from Type Band C tests which would put total containment leakage over 1.0 La must still be evaluated. Therefore, there is no significant increase in the probability or consequences of an accident previously evaluated . Palisades Nuclear Plant Page 1of3 05/31/99

  • * * , ____ _ 2. ATTAC1ŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.1, CONTAINMENT Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which -govern normal plant operation. The proposed change will continue to ensure total containment leakage is monitored to ensure that it stays within the bounds of the analysis. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety? The proposed change preserves the total containment leakage rate of 1.0 La but does not require actions to be taken once the Type Band C tests exceed .60 Ia as long as the contribution of the Type B and C tests do not make the total containment leakage to exceed 1.0 La*-Therefore, this change does not involve a significant reduction in.a margin of safety. LESS RESTRICTIVE L.2 CTS 3.6.lc requires Containment Integrity be maintained "when positive reactivity changes are made by boron dilution or control rod motion (except for testing one control rod at a time)." In the ITS, containment integrity is ensured by operating the plant within the limits established in LCO 3.6.1, "Containment", LCO 3.6.2, "Containment Air Locks", and LCO 3.6.3, "Containment Isolation Valves." The Applicability for these specifications is MODES 1, 2, 3, and 4. The ITS contains no requirement equivalent to containment integrity in MODE 5, and only a less stringent requirement in MODE 6 referred to as containment closure (LCO 3.9.3, "Containment_Penetrations")

which is only applicable during Core Alteration or movement of irradiated fuel in the containment. As such, the requirement to maintain containment integrity in the ITS is less restrictive than the requirement in the CTS since the ITS would allow positive reactivity changes and control rod motion in MODES 5 or 6. This change is acceptable since in MODES 5 and 6, the probability and consequences of an event which would required containment integrity is reduced due to the pressure and temperature limitations in these MODES. Furthermore, by definition MODE 5 requires the reactor to be shutdown (subcritical) by > 1 3 t:,. k/k. In addition ITS 3 .1.1, "Shutdown Margin (SDM)" requires SDM 23. *In MODE 6 ITS 3.9.1, "Boron Concentration" requires sufficient boron concentration to maintain the reactor subcritical by 5 % Ap with all control rods withdrawn. These requirements ensure that reactivity changes from boron dilutions or control rod motion can be made without approaching a condition (criticality) in which containment integrity would be needed. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 2of3 05/31/99

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  • 1. 2. 3. ATTAC1ŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.1, CONTAINMENT Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change relaxes the plant conditions in which containment integrity must be maintained. This change does not alter any accident precursors or initiators and thereby does not involve a significant increase in the probability of an accident previously evaluated. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. The proposed change does not alter the initial assumptions of any accident analysis, or alter the design assumptions of any system or component relied upon to function in the event of an accident. Therefore, this change does not involve a significant increase in the consequence of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced, and no installed equipment is being operated in a new or different manner. The proposed change only relaxes an administrative requirement associated with containment integrity. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? The margin of safety is determined by the design and qualification of plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. The proposed change eliminates the requirement to maintain containment integrity during reactivity changes from boron dilutions or control rod motion while the plant is in MODES 5 or 6. The purpose of maintaining containment integrity is to prevent the uncontrolled release of radioactive material to the environment in the event of an accident that is capable of generating elevated temperature or pressure changes in the containment atmosphere. Controlled reactivity

  • changes from boron dilutions or control rod motion in MODE 5 or 6 are not events that would directly result in elevated temperatures and pressures in the containment . atmosphere.

As such,' containment integrity is not needed during these evolutions. Therefore, relaxing the requirement to maintain containment integrity in MODE 5 and 6 does not involve a significant reduction in a margin of safety. Palisades Nuclear Plant Page 3of3 05/31199

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  • ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS LESS RESTRICTIVE CHANGE L.1 CTS 4.5.2c(3) requires that repairs be initiated immediately for excessive air lock door seal leakage and repairs be completed within a specified time. CTS 4.5.2c(3) states "If air lock door seal leakage is greater than 0.0234, repairs shall be initiated immediately to restore the door to less than specification 4.5.2.b(2).

In the event repairs cannot be completed within 7 days .... " The proposed ITS Required Actions (R.A.) for Condition A do not require repairs to be initiated immediately since the OPERABLE door can be verified closed in one hour (R.A. A. l) which ensures that the safety function for containment is met. In addition, there is no requirement to complete the repairs in within a* specified time since the safety function of the air lock is met as long as the OPERABLE door is closed. These changes are considered to be less restrictive since the specific repair actions are not specified in the proposed ITS. However, there is no impact on safety since an OPERABLE door is required to be verified closed within one hour (R.A. A. l), IOcked within 24 hours (R.A. A.2) and periodically verified every 31 days to be closed (R.A. A.3) which ensures the containment integrity function is maintained. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures,* systems or components. The proposed change extends the allowed outage time for an inoperable containment air lock door as long as an OPERABLE door is locked. An extension in the allowed outage time for an inoperable component is not assumed to be an initiator of any evaluated accident. Not requiring a inoperable door to be fixed as long as there is an OPERABLE door closed and locked and periodically verified to remain closed and locked ensures that the assumptions with respect to containment integrity are still valid. Therefore, extending the allowed outage time for an inoperable air lock door and not requiring repairs to be initiated immediately does not involve a significant increase in the probability of an accident previously evaluated. The consequences of a previously event are dependent on the initial conditions

  • assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event.

The proposed change does not alter the assumed mitigatory function of the remaining Operable containment air lock *door. Thus, the consequence of an accident occurring during the time that the one containment air lock door is inoperable but the OPERABLE door is locked closed is the same as the consequences for an accident with both air lock doors OPERABLE. Therefore, the proposed change does not favolve a significant increase in the consequences of an accident previously evaluated . Palisades Nuclear Plant Page 1of7 05/31/99

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  • 2. ATTACIŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change still ensures that an OPERABLE containment air lock door is closed. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety? The proposed change increases the allowed outage time for an inoperable containment air lock door and does not require repairs to be initiated immediately as long as an OPERABLE door is closed within 1 hour, locked within 24 hours and periodically verified to be locked closed every 31 days. Therefore, since an OPERABLE door is locked closed and will ensure that containment leakage is within limits and the OPERABLE door is verified to be closed every 31 days, there is no significant reduction in the margin of safety . LESS RESTRICTIVE CHANGE L.2 CTS 4.5.2c(4) specifies requirements for air lock door seal leakage causing the total containment leakage to exceed . 60 La. These requirements are not included in the proposed ITS 3.6.2 relative to exceeding

.60 4, as sufficient actions are provided to ensure containment OPERABILITY is maintained. Proposed ITS 3.6.2 Condition A and C provide ACTIONS for air lock doors being inoperable. In accorda.nce with the acceptance criteria in CTS 4.5.2b(2) which is consistent with proposed SR 3.6.2.1 the acceptance criteria for containment air lock door seal leakage .023 La (See DOC M.3). There is no impact on containment leakage if one air lock door seal leakage is > .60 4 as long as there is a closed OPERABLE door in the air lock. The CTS requirement to test the remaining OPERABLE door, which was required to be locked closed, within 4 hours is also not required to be performed included in the proposed ITS. In accordance .with the usage rules of NUREG-1432, equipment is assumed to be OPERABLE between surveillance periods unless there is some reason to believe otherwise. Therefore, there is no reason to perform a confirmatory test for OPERABILITY if there is no indication that the OPERABLE door seals are degraded . (continued) Palisades Nuclear Plant Page 2of7 05/31/99

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  • ATTAC1ŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS LESS RESTRICTIVE CHANGE L.2 (continued)

The proposed ITS Required Actions and Frequency in ACTION A of verifying the OPERABLE door is closed in the affected air lock within 1 hour AND locking the OPERABLE door closed in the affected air lock within 24 hours AND verifying the OPERABLE door is locked closed in the affected air lock once per 31 days are sufficient to ensure that containment OPERABILITY is maintained. As long as an OPERABLE door exists, the leakage through the other door is irrelevant.

  • Therefore, the acceptance criteria of .60 L.i and additional associated actions are no longer required.

This change is considered to be less restrictive since additional actions beyond those . addressed by proposed ACTION A for air lock door seal leakage being > .023 la are not needed to maintain containment OPERABILITY. This change is consistent with NUREG-1432.

1. 2. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

The proposed change deletes the required actions which are to be performed if the containment air lock door leakage would cause total containment leakage to exceed . 60 La. Sufficient actions are already provided if air lock door seal leakage exceeds .023 La which includes ensuring that the OPERABLE door_is closed. By closing the

  • OPERABLE air lock door, the leakage from the other door is irrelevant since containment integrity is ensured by the OPERABLE door which is closed and locked. Therefore, since containment OPERABILITY has not been impacted there is no *significant*

increase in the probability or consequences of !lil accident. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change continues to limit the amount of leakage from containment by ensuring that an OPERABLE door is closed and locked. In addition, overall containment leakage is still limited to 1. 0 L.i and any contribution to this leakage limit from the air lock door seals must still be evaluated. Thus, this change does not create the possibility of a new or differerit kind of accident from any accident

  • previously evaluated . Palisades Nuclear Plant Page 3of7 05/31199
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  • 3. ATTACIWENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.2, CONT AIN1\1ENT AIR LOCKS Does this change involve a significant reduction in a margin of safety? The margin of safety is a function of the overall containment leakage.

The proposed change does) not require that immediate actions be taken to repair the inoperable door or test the OPERABLE door. Failure of one air lock door seal test does not mean the other seal will fail its test. As long as the OPERABLE door is closed and there is every reason to believe the seals in the OPERABLE door are still OPERABLE, then the overall containment leakage should still be within limits. Therefore*, not requiring immediate repairs to the inoperable door or requiring that a test be performed on the OPERABLE door does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE CHANGE L.3 The actions of CTS 4.5.2c have been revised to allow entry and exiting of the containment when one door in both containment air locks are inoperable. Proposed ITS 3.6.2 Condition A, Required Action Note 2 states that "entry and exit is permissible for 7 days under

  • administrative controls if both air locks are inoperable." This purpose of this Note is to allow ingress and egress through an Operable containment air lock door; that is required to be closed and locked to isolated the air lock penetration, to perform technical specification related and non-technical specification related activities on equipment inside containment.

The allowance is permitted for up to 7 days provided appropriate administrative controls are instituted. This allowance is acceptable due to the low probability of an event that could pressurize the containment during the short time that the Operable door is expected to be opened. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed* to be initiated by the failure of plant structures, systems or components. The proposed change does not result in any hardware changes. The containment air locks are not assumed to initiate any analyzed event. The proposed change only allows entry and exiting of containment through the Operable door of a containment air lock when one door in both containment air locks are inoperable. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated . Palisades Nuclear Plant Page 4of7 05/31/99

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  • ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS LESS RESTRICTIVE CHANGE L.3 1. (continued)
2. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The role of the containment air lock is to contain releases from the containment during a design basis accident, and thereby limit the consequences of an accident.

The proposed change does not allow continuous operation such that a single failure would allow a release from containment during a design basis accident. The proposed allowance would limit the time the air lock could be utilized with an inoperable door to only 7 days. During this period, entry and exit are conducted under administrative controls which ensure that, in the event of an accident, the Operable air lock door would be quickly closed and thereby reestablish the containment boundary. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated. , Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change allows the Operable door in a containment air lock with one inoperable door to be utilized for containment ingress and egress for a period of time up to 7 days. The proposed change will continue to ensure the containment boundary is capable of being maintained. Thus, this change does not create the possibility of a new or different kind 1 of accident from any accident previously evaluated . Palisades Nuclear Plant Page S of 7 05/31199

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  • 3. ATTACIŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS Does this change involve .a significant reduction in a margin of safety? The proposed change allows the Operable locked closed air lock door, in an air lock with an inoperable door, to be opened briefly to facilitate entry and exit of containment.

Except for the brief period the Operable door is opened to allow personnel to ingress and egress the air lock, the air lock is capable of performing its intended containment pressure boundary function. Opening the Operable door for a brief period has been found acceptable based on the extreme low probability of an event occurring that would challenge the containment pressure boundary while the door was opened. The proposed change reduces the potential for a plant transient resulting from a unnecessary plant shutdown due to component repairs or to perform technical specification related activities which require entry inside containment. As such, any reduction in a margin of safety during the brief period the Operable air lock door is opened is offset by the reduction in a potential plant transient. Therefore, the proposed change does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE L.4 CTS 4.5.2c(3) provides the corrective actions if the personnel air lock doors exceed their_ leakage limit, or the emergency air lock fails to meet its acceptance criterion for seal contact. In ITS 3.6.2, these same occurrences are addressed by Condition A which states "one or more containment air locks with one containment air lock door inoperable." ITS Condition A represents a relaxation from CTS 4.5.2c(3) since the ITS allows one inoperable door in the personnel air lock to exist concurrently with one inoperable door in the emergency air lock. The CTS does not address multiple air locks with inoperable doors. Thus, if multiple inoperabilities occurred, entry into LCO 3.0.3 would be required. In addition to allowing multiple inoperabilities, the ITS Actions for containment air locks are also modified by a Note which states " Separate Condition entry is allowed for each air lock." The addition of this Note allows separate Completion Time tracking for each Condition starting from the time of discovery of the situation that required entry into the specified Condition. The allowance to have one or more containment air locks with one containment door inoperable, and to permit separate condition entry for each inoperability is acceptable since the Operable door in each air lock ensures that a leak tight containment barrier is established and that the appropriate compensatory measures are taken to ensure the barrier (door) is maintained Operable. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 6of7 05/31/99

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  • 1. 2. 3. ATTAC1ŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS Does tlie change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change allows one personnel air lock door to be inoperable concurrent with one emergency air lock door, and allows separate completion time tracking for each inoperability. Entry into required action in the technical specifications and the method used to track completion times are not assumed to initiate any analyzed events. Therefore, the proposed change does not involve a significant increase in the probability of an accident. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change does not alter the assumed mitigatory function of the remaining Operable containment air lock door. Thus, the consequence of an accident occurring during the time that one containment air lock door is inoperable is the same as the consequences for an accident with multiple air lock doors inoperable. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? I I I I The proposed change does not involve a physical alteration of the plant. No new or I different type of equipment will be installed or changes made to plant parameters which I govern normal plant operation. The proposed change still ensures that an OPERABLE* I containment air lock door is closed. Thus, this change does not create the possibility of . I a new or different kind of accident from any accident previously evaluated. I Does this change involve a significant reduction in a margin of safety? The margin of safety is a function of the overall containment leakage. The proposed change allows one inoperable door in the personnel air lock to exist concurrently with one door in the emergency air lock and allows the completion time for each inoperability to be tracked separately. During the period when multiple inoperabilities exist, the remaining Operable door in each air lock continues to ensure that leakage from the containment atmosphere is within the limits assumed in the safety analysis. As such, there is no increase in the amount of radioactive material released to the environment. Therefore, the proposed change does not involve a significant reduction in a margin of safety . I I I I I I I I I I I I Palisades Nuclear Plant Page 7of7 05/31/99

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  • ATTAC1ŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES LESS RESTRICTIVE CHANGE L.1 CTS Table 4.2.2 item 13.a specifies that the Containment Purge and Ventilation Isolation Valves shall be determined closed "at least once per 24 hours." The proposed ITS SR 3. 6. 3 .1 changes the frequency for verifying the 8 inch containment purge valve and 12 inch air room supply valves are closed from 24 hours to 31 days. This change is reasonable because these valves are not allowed to be in proposed MODES 1-4. These are required to be locked closed, either electrically or by some other means to de-activate the valve. Therefore, the likelihood of the valve being operied is very remote. The 31 day frequency is consistent with the surveillance requirements for other safety related systems which require valve position verification.

This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of * *an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change extends the surveillance frequency for verifying the purge exhaust and air room supply valves from 24 hours to every 31 days. Ari extension in the surveillance frequency is not assumed to be an initiator of any evaluated accident. Therefore, extending the surveillance frequency for verifying that the purge exhaust and air room supply valves are closed does not involve a significant increase in the probability of an accident previously evaluated. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change

  • will extend the surveillance frequency for the purge exhaust and air room supply valves from 24 hours to 31 days. With both valves in a path for purge exhaust or the air room supply open, a direct path outside containment would exist. However, both *valves in each path are required to remain closed during MODES 1,2,3. and 4. The consequence of an accident occurring during the 24 hour surveillance interval presently specified in the technical specifications is the same as the consequences for an accident occurring a 31 day surveillance interval.

Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated . Palisades Nuclear Plant Page 1of10 05/31/99

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  • 2. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Does the change create the possibility of a new or diff kind of accident from any acddent previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue to verify that the purge exhaust and air room supply valves are still closed. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety? . The proposed change increases the surveillance .interval from 24 hours to 31 days. The margin of safety afforded by the purge exhaust and air room supply valves are that they are closed and thereby isolating a direct path from the containment atmosphere.

Since the purge exhaust and air room supply valves are required to be closed in MODES 1, 2, 3 and 4, the likelihood of these valves being open is very small. Therefore, this change does not involve a significant reduction in a margin of safety . LESS RESTRICTIVE CHANGE L.2 CTS 3.6.1 contains the corrective actions for one or more inoperable containment isolation valves. For penetrations associated with a closed piping system and one CIV, CTS 3.6.1 action "c" would require the plant to be placed in at least Hot Shutdown within 6 hours and in Cold Shutdown within the following 30 hours. In ITS 3.6.3, this same inoperability would be addressed by Condition C which would require the affected penetration be isolated in 72 hours, and verified isolated once per 31 days. The Required Actions of the ITS are less restrictive than the CTS since they allow 72 hours to isolate the affected penetration versus a forced shutdown. The 72 hour period provides the necessary time to perform repairs on a failed CIV when relying on an intact closed system. A Completion Time of 72 hours is considered appropriate given the location of certain valves, the reliability of the closed system, and that 72 hours is typically provided for losing one train of redundancy throughout the ITS. If the system and associated CIV were both inoperable (as a containment boundary), the plant would be in LCO 3.0.3 since there is no specific Condition specified. Although ITS Required Action C.2 is an additional restriction on plant operations since it requires a verification that the affected penetration is isolated once per 31 days, the overall change related to the addition of ITS Condition C is characterized as Less Restrictive. Lastly, Required Action C.2 is modified by a Note which states "isolation devices in high radiation areas may be verified by use of administrative means. " This allowance m'inimizes personnel exposure and recognizes that access to high radiation areas is restricted such that the probability of a valve misalignment is small. This change is consistent with NUREG-1432 as modified by TSTF-30. Palisades Nuclear Plant Page 2of10 05/31199

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  • 1. 2. 3. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems

  • or components.

The proposed change extends the allowed outage time for an inoperable containment isolation valve in a penetration that consists of a closed piping system and one containment isolation valve. The proposed change does not involve a change to any accident initiators or precursors. Therefore, the proposed change does not involve a significant increase in the probability of an accident. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change does not alter the assumed mitigatory capabilify of the containment isolation function since the redundant isolation barrier formed by the closed system is required to be Operable. Thus, the consequence of an accident occurring during the time presently allowed in the CTS is the same as the consequences for an accident occurring during the time proposed in the ITS. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated . Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change still ensures that an OPERABLE containment isolation is available whenever the plant is in a condition that requires containment isolation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? The margin of safety is a function of the overall containment leakage. Piping systems that penetrate containment niust be provided with isolation and containment capabilities having redundancy, reliability and performance capabilities which reflect the importance of isolating these penetrations. The proposed change extends the allowed outage time for an inoperable containment isolation valve in a penetration that contains a closed piping system and one containment isolations valve. During this extended period, the only affect on the containment isolation function is a loss of redundancy since the remaining isolation barrier ensures that leakage from the containment atmosphere is within the limits assumed in the safety analysis. As such, there is no increase in the amount of radioactive material released to the environment. Therefore, the proposed change does not involve a significant reduction in a margin of safety. Palisades Nuclear Plant Page 3of10 05/31/99

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  • ATTACIThffiNT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES LESS RESTRICTIVE L.3 CTS 4.5.3d requires that prior to the reactor going critical after a refueling outage, a visual check will be made to confirm that all locked-closed manual containment isolation valves are closed and locked except for valves that are open under administrative controls.

In the ITS, manual valve position verification is required by proposed SR 3.6.3.2 and SR 3.6.3.3. SR 3.6.3.2 and SR 3.6.3.3 are modified by a Note which states "valves and blind flanges in high radiation areas may be verified by use of administrative means." The allowance to verify valve position by use of "administrative means" is a relaxation from the CTS requirement to perform a "visual check." The proposed change allows special provisions for high radiation areas to minimize personnel exposure while still keeping track of containment isolation valve status. This change is considered acceptable since a verification requirement still exists, and because high radiation areas are restricted such that the probability of valve misalignment is small. This change is consistent with NUREG-1432.

1. 2. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure .of plant structures, systems or components. The proposed change allows the periodic verification of manual containment isolation valves in high radiation areas using administrative means. The proposed change does not involve a change to any accident initiators or precursors. Therefore, the proposed change does not involve a significant increase in the probability of an accident. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change does not alter the assumed capability of the containment isolation function since the containment isolation valves in high radiation areas are still assured to be in their correct position. Thus, the consequence of an accident remain unchanged. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be or changes made to plant parameters which govern normal plant operation. The proposed change still ensures that an OPERABLE containment isolation barrier is available whenever the plant is in a condition that *requires containment isolation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Palisades Nuclear Plant Page 4of10 05/31/99

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  • 3. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Does this change, involve a significant reduction in a margin of safety? The margin of safety is a function of the overall containment leakage. The proposed change allows the periodic verification of manual containment isolation valves in high radiation areas using administrative means. This change does not relax the requirement to maintain containment integrity, but recognizes that access to high radiation areas are restricted such that the probability of valve misalignment is small. As such, this change does not result in any activity that would result in an increase in the amount of radioactive material released to the environment.

Therefore, the proposed change does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE CHANGE L.4 CTS 4.5.3d requires that prior to the reactor going critical after a refueling outage, a visual check will be made to confirm that all "locked-closed" manual containment isolation valves are closed and locked except for valves that are open under administrative controls. In the ITS, manual valve position verification is required by proposed SR 3.6.3.2 and SR 3.6.3.3. The requirements of SR 3.6.3.2 and SR 3.6.3.3 are less restrictive than the CTS since they only apply to valves that are "not locked, sealed, or otherwise secured in position." This proposed change is acceptable since these valves are verified closed when they are locked, sealed, or otherwise secured in position. Administrative programs provide the appropriate controls to assure valves that are normally locked, sealed, or otherwise secured in position are in their correct position. This change is consistent with NUREG-1432 as modifie4 by TSTF-45, Rev. 1. 1. Does the change involve a significant increase in the probability or consequen,ce of an accident previously evaluated? Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change limits the periodic verification of manual containment isolation valves to only those vcllves that are not locked, sealed, or otherwise secured in position. The proposed change does not involve a change to any accident initiators or precursors. Therefore, the proposed change does not involve a significant increase in the probability of an accident. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change does not alter the assumed capability of the containment isolation function since the required valves are still assured to be in correct position. Thus, the consequence of an accident remain unchanged. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated . Palisades Nuclear Plant Page 5of10 05/31/99

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  • 2. 3. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change still ensures that an OPERABLE containment isolation barrier is available whenever the plant is in a condition that requires containment isolation. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this involve a significant reduction in a margin of safety? The margin of safety is a function of the overall containment leakage. The proposed change limits the periodic verification of manual containment isoiation valves to orily those valves that are not locked, sealed, or otherwise secured in position. This change does not relax the requirement to maintain containment integrity, but recognizes the misalignment of valves which have been verified closed when they were locked, sealed, or otherwise secured in position is small. As such, this change does not result in any activity that would result in an increase in the amount of radioactive material released to the environment. Therefore, the proposed change does not involve a significant reduction in a margin of safety. * . LESS RESTRICTIVE CHANGE L.5 CTS 4.5.3a, CTS 4.5.3b, and CTS 4.2 Table 4.2.2, Items 13.a and 13.b contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. As such, these details are proposed for deletion. Specifically: CTS 4.5.3a describes the testing necessary for CIVs prior to declaring the valves Operable after maintenance, repairs, or replacement work is performed on the valve or its associated actuator, control, or power circuit. Explicitly stating these tests as they relate to maintenance activities is unnecessary since the technical specifications stipulate the level of performance that must be met for an Operable CIV in the associated Surveillance Requirements. (continued) Palisades Nuclear Plant Page 6of10 . 05/31/99

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  • ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES LESS RESTRICTIVE CHANGE L.5 (continued)

CTS 4.5.3b states that each CIV shall be demonstrated Operable by verifying ... valves actuate to their required position "during Cold Shutdown or at least once per refueling cycle". The phrase "during Cold Shutdown'.' is intended to describe the plant condition which best facilitates testing the applicable (automatic) CIVs. The phrase "at least once per refueling cycle" establishes the frequency for the test. Specifying the plant condition at which CIV testing is perforined (i.e., Cold Shutdown) is a detail which is not pertinent to the actual requirement for testing CIVs. The LCO Applicability for*CIVs stipulates the plant conditions when CIV s are required to be Operable. Testing within the Applicability is governed by valve* Operability, testing in plant conditions outside the Applicability has no impact on safety. CTS 4.2 Table 4.2.2, Item 13a states the containment purge and ventilation isolation valves are determined closed "by checking the valve position indicator in the control room". The intent of verifying valve position is to ensure that the valve is in its correct position. Specifying that valve position be verified "by checking the valve position indicator in the control room" does not constitute a requirement assumed in the safety analyses. Rather, it simply* provides *a method for assuring the valve is in the correct position. Since the valves may be locked closed electrically, mechanically , .or by other physical means, stipulating* "by checking the valve position indicator in the control room" is an inappropriate detail not pertinent to the actual requirement. CTS 4.2 Table 4.2.2, Item 13b states the containment purge and ventilation isolation valves are determined closed by performing a leak.rate test "between the valves." Specifying that a leakage rate test be performed "between the valves" does not constitute a requirement assumed in the safety analyses. Rather, it simply provides a method for conducting the leakage rate test. Since the above details are not necessary to describe, or are not pertinent to, any actual. regulatory requirement, they can be deleted without an impact to public health and safety. These changes are consistent with NUREG-1432 . Palisades Nuclear Plant Page 7of10 05/31/99

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  • 1. 2. 3. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed changes delete details from the Technical Specifications that are not necessary to describe, or are not pertinent to, any actual regulatory requirement. The deletion of details from the Technical Specifications is not assumed to be an initiator of any analyzed event. The proposed changes do not reduce the functional requirement or alter the intent of any specification. As such, the consequences of an accident remain. unchanged. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change deletes detail from the Technical Specifications that are not necess,ary to describe, or are not pertinent to, any actual regulatory requirement. The changes will not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. The changes will not impose different requirements, and adequate control of activities will be maintained. The changes will not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes will not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of Margin of safety is determined by the design and qualification of the plantequipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. The proposed changes delete details from the Technical Specifications. Removal.of these details is acceptable since this information is not directly pertinent to the actual requirement and does not alter the intent of the requirement. Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to licensee controlled document without a significant impact on safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety . Palisades Nuclear Plant Page 8of10 05/31/99

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  • ATTAC1ŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES LESS RESTRICTIVE CHANGE L.6 CTS 4.5.3b requires that each containment isolation valve be demonstrated Operable by verifying it actuates to its required position.

In the ITS, an equivalent test is required by SR 3.6.3.6. However, ITS SR 3.6.3.6 does not require containment isolation valves under administrative controls that are locked, sealed, or otherwise secured in position to be tested. This is because these valves are already in the position necessary to perform the containment isolation function. Thus, there is no need to verify these valve can reposition on an actual or simulated actuation signal. The allowance not to test containment isolation valves that are locked, sealed, or otherwise secured in position is a relaxation from the requirements of the CTS. This change is consistent with NUREG-1432.

1. 2. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change deletes the requirement to perform an actuation test of containment isolation valves that are locked, sealed., or otherwise secured in position since these valves are already in the position necessary to perform their

  • containment isolation function.

The proposed change does not involve a change to any accident initiators or precursors. Therefore, the proposed change does not involve a . significant increase in the probability of an accident. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availabilitY and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change . does not alter the assumed capability of the containment isolation function. Thus, the consequence of an accident remain unchanged. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change still ensures that an OPERABLE containment isolation barrier is available whenever the plant is in a condition that requires containment isolation.

Thus, change does not create the possibility of a new or different kind of accident from any accident previously evaluated . Palisades Nuclear Plant Page 9of10 . 05/31199
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  • 3. ATTAC1ŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Does this change involve .a significant reduction in a margin of safety? The margin of safety is a function of the overall containment leakage. The proposed change deletes the requirement to perform an actuation test of containment isolation valves that are locked, sealed, or otherwise secured in position.

This change does not relax the requirement to maintain containment integrity, but recognizes that valves that are locked, sealed, or otherwise secured in position are already. capable of performing their isolation function. As such, this change does not result in any activity that would result in an increase in the amount of radioactive material released to the environment. Therefore, the proposed change does not involve a significant reduction in a margin of safety . Palisades Nuclear Plant Page 10of10 05/31199

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  • ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.4, CONTAINMENT PRESSURE LESS RESTRICTIVE CHANGES (L) There were no "Less Restrictive" changes associated with this section . Palisades Nuclear Plant Page 1of1 05/31/99 J
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  • ATTACIThlENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION . SPECIFICATION 3.6.5, CONTAINMENT AIR TEMPERATURE LESS RESTRICTIVE CHANGES (L) There were no "Less Restrictive" changes associated with this section . Palisades Nuclear Plant Page 1of1 05/31/99
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  • ATTAC1ŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING LESS RESTRICTIVE CHANGE L.1 CTS 3.4.2 allows one of the components listed in Specification 3.4.l to be inoperable for a period of up to seven days. CTS 3 .4. 3 allows a total of two of the components listed ih Specification 3.4.la orb to be inoperable at any one time to be inoperable for up to 24 hours. Specification 3 .4.1 lists requirements for equipment associated with each diesel generator.

For diesel generator 1-2 the equipment specified includes: three Containment Air Coolers (V-lA, V-2A, V-3A); 2 service water pumps (P-7A, P-7C); one containment spray pump (P-54A); and one component cooling water pump (P-52B). For diesel generator 1-1 the equipment specified includes: one service water pump (P-7B); two containment spray pumps (P-54B, P-54C); two component cooling water pumps (P-52A, P-52C). The requirements for service water pumps and component cooling water pumps are addressed in Technical Specification (TS) Section 3.7, Plant Systems. The proposed ITS modifies this requirement to allow one or more trains of containment cooling to be inoperable for 72 hours as long as at least 100 % of the cooling capability equivalent to a single OPERABLE containment cooling train is available. One train of containment cooling is defined to be the three safety related containment air coolers and containment spray pump P-54A, or containment spray pumps P-54B and P-54C. Therefore, the proposed ITS would allow four items associated with diesel generator 1-2 to be inoperable for 72 hours, or two items assoeiated with diesel generator 1-1 to be inoperable for a total of 72 hours, or any combination of equipment that leaves 2 spray pumps, or one spray pump with three required containment air coolers OPERABLE. This is considered to be a less restrictive change since a train of containment cooling (see above) is allowed to be inoperable in the proposed ITS 3.6.6 for 72 hours as opposed to the CTS 3.4.2 allowance of one of the components from CTS 3.4.1 for 7 days being inoperable or the CTS 3.4.3 allowance that a total of two of the components listed in Section 3.4.la or 3.4.lb may be inoperable. For CTS 3.4.3, the change to the one or more trains inoperable for 72 hours also less restrictive since the CTS only allows a total of two of the components from either 3.4. la or 3.4. lb to be inoperable for 24 hours. CTS 3.4.5 also is changed from 24 hours to 72 hours for inoperabilities involving valves, interlocks, or piping associated with the containment cooling system and not covered by CTS 3.4.4. These changes are acceptable since the proposed ITS still requires that the equivalent of one train of containment cooling be OPERABLE and the timeframe for the equipment to be inoperable equipment is only 72 hours. The support role and requirements of the service water pumps and component cooling water pumps to the containment air coolers and containment spray pumps are addressed by LCO 3.0.6 and the Safety Function Determination Program as well as the definition of OPERABILITY. This helps to ensure that the containment cooling trains have adequate support equipment to perform their function. This change maintains consistency with the intent of NUREG-1432. '* Palisades Nuclear Plant Page 1of8 05/31/99

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  • 1. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change will allow the equipment comprising an entire train of containment cooling (including equipment from separate electrical sources) to be inoperable for up to 72 hours as long as the containment cooling capability equivalent to 100 % of a single containment cooling train is available. The CTS allows a total of 2 components of containment cooling equipment from separate electrical sources to be inoperable for up to 24 hours. An extension in the allowed outage time for an inoperable component is not assumed to be an initiator of any evaluated accident. Therefore, extending the allowed outage time for one train of containment cooling does not involve a significant increase in the probability of an accident previously evaluated.* The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability* and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change will allow one or more trains of containment cooling equipment to be inoperable for 72 hours provided the containment cooling capability equivalent to a single OPERABLE train is available. The CTS would allow one component to be inoperable for 7 days and two components (from either electrical train) to be inoperable for . 24 hours. The only containment cooling train involving more than two components is the train which has three safety related containment air coolers and one containment spray pump. In this situation in the CTS, the plant would be in LCO 3.0.3 which would require that a plant shutdown be initiated within one hour and placed in shutdown in the next 6 hours. However, one train of containment cooling is still provided and is sufficient to maintain containment pressure within the analytical limits in the event of a design basis accident. Thus the consequences of an accident occurring during the allowed outage time presently specified in the technical specifications (LCO 3.0.3) is the same as the consequences for an accident occurring during an allowed outage time of 72 hours. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated . Palisades Nuclear Plant Page 2of8 05/31/99

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  • 2. 3. ATTAC1ŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
  • The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation.

The proposed change will continue to verify that the cooling equivalent to 100% of a single containment cooling train is available. Thus, this change does not create the possibility of a new or different kind of accident from

  • any accident previously evaluated.

Does this change involve a significant reduction in a margin of safety? The proposed change increases the allowed outage time and number of containment cooling components which can be inoperable by allowing one or more containment cooling trains to be inoperable for up to 72 hours provided the containment cooling capability equivalent to 100% of a single containment cooling train is available. The CTS allows a total of two components from separate electrical sources to be inoperable for up to 24 hours. The margin of safety afforded by the containment cooling systems is to ensure that containment pressure and temperature are maintained within the analytical values. By stipulating a 100% containment cooling capability equivalent to a single OPERABLE containment cooling train, the safety function of the containment cooling system is preserved. This is because of the redundancy of the containment cooling trains,. the diversity of the subsystems and recognition of the fact that the inoperability of one component in a train does not necessarily render the containment cooling systems incapable of performing their intended safety function. As such, the margin of safety associated with allowing one or more trains of containment cooling to be inoperable for up to 72 hours provided 100 % of the containment cooling capability equivalent to a single OPERABLE containment cooling train is available is not significantly reduced. *

  • Palisades Nuclear Plant Page 3of8 05/31/99
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  • ATTACIŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING LESS RESTRICTIVE CHANGE L.2 CTS 3.4.2 and 3.4.3 require that "If the inoperable component (train in ITS) has not been restored to operability within an additional 48 hours the reactor shall be placed in a cold shutdown condition within 24 hours. " The proposed ITS requires that for Required Actions and Associated Completion Times not met the plant is eventually required to be placed in MODE 4 a total of 30 hours. The "additional 48 hours" allowance of the CTS for restoring inoperable components before taking actions is not included in the proposed ITS. A shutdown to MODE 4 required in the proposed ITS places the plant out of the MODE of Applicability.

CTS .4.1 only requires that the equipment associated with Containment Cooling Systems be OPERABLE prior to making the reactor critical. As discussed in "more restrictive" change .M.1, the proposed ITS will require that the containment cooling equipment

  • be OPERABLE in MODES 1, 2, and 3 which is the MODES for which the equipment is required to function in an accident.

The CTS requirement to place the plant in a cold shutdown condition (i.e, < 210 °P in the CTS) if the required actions and associated completion times are not met is overly restrictive since the equipment is not needed below 300°P. Therefore, only requiring that the plant be placed_in MODE 4 is considered to be a less restrictive change. This change maintains consistency with NUREG-1432 . 1. Does the change involve a sigmficant increase in the probability or consequence of an accident previously evaluated? Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change only requires that the plant be shutdown to MODE 4 (i.e., < 300°P) as opposed to the CTS cold shutdown condition" (i.e., < 210° P) if the Required Actions.and Associated Completion Time is not met since the equipment is not relied upon in an accident in MODE 4 and below. The CTS shutdown* . requirement is overly restrictive since the equipment is not required to provide protection below 300 °P. Therefore, requiring the plant to be placed in MODE 4, places the unit in a MODE where the equipment is assumed to function.and therefore does not involve a significant increase in the probability or consequence of an accident previously evaluated . Palisades Nuclear Plant Page 4of8 05/31/99

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  • 2. 3. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will continue require that the plant be placed in a MODE which does not require the equipment to be OPERABLE if the,Required Actions and Associated Completion Times are not met. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? The proposed change allows the plant to be placed in MODE 4 ( < 300 °F) as opposed to the CTS "cold shutdown condition" ( <210 °F) is the Required Actions and Associated Completion Times have not been met. The margin of safety afforded by the containment cooling systems is to ensure that containment pressure and temperature are maintained within the analytical values. Since the change still requires that the plant be placed in a MODE where the equipment is not required to operate if the Required Actions and Associated Completion Times have not been met no margin of safety is reduced. Therefore, this change does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE CHANGE L.3 CTS 4.6.2a specifies that for the Containment Spray System test, "The test shall be performed with the isolation valves in the spray supply lines at the containment blocked closed." CTS 4.6.2c specifies that for the Containment Spray System test, "The test will be considered satisfactory if visual observations indicate all components have operated satisfactorily." This level of detail, with respect to how the test is performed and what the testing acceptance criteria are, is more appropriately addressed in plant procedures. Removal of these details from the Technical Specifications will not affect the requirement to perform the containment spray test, however the methodology and acceptance criteria will not be included in the proposed specifications. This change is consistent with NUREG-1432 . Palisades Nuclear Plant Page 5 of 8 05/31/99

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  • 1. 2. 3. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

The proposed change would remove details of how the containment system is aligned and regarding the how test acceptance criteria are to be met .. These directions do not provide direction or limits that must be verified. The details of how the system should be aligned during testing are merely a prerequisite designed to prevent inadvertent actuation of the containment spray system during testing. The acceptance criteria details merely indicate that "visual observation" is an acceptable way to verify system function. Since the details provide no substantive limits, removing this detail from the surveillance testing requirements associated with the containment spray system testing does not effect the testing or operability of the system in any way. Since no effect on the system will occur, the probability of any accident previously evaluated is not effected. The consequences of a previously analyzed event are dependent on the* initial conditions assumed for the analysis, and the availability* and successful functioning of the equipment assumed to operate in response to the analyzed event. The presence or lack of the test details being removed from the specifications does not effect the ability of the containment spray system to perform its safety function. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will merely remove the detail of test alignments and acceptance criteria from the Technical Specifications. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? The proposed change will remove details regarding how the containment spray system is aligned during testing, and a statement that indicates visual observation is an acceptable way of confirming acceptable results. The margin of safety afforded by the containment spray system is to ensure that the containment pressure and temperature are maintained within the analytical values. The removal of this unnecessary detail from the Technical Specifications will not effect the way testing is performed or the availability of the containment spray system. Therefore, this change does not involve a significant reduction in a margin of safety. Palisades Nuclear Plant Page 6of8 05/31/99

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  • ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATiON SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING LESS RESTRICTIVE CHANGE L.4 CTS 4.6.3a and 4.6.3b contain details that are not necessary to describe, or are not pertinent to, any actual regulatory requirement.

As such, these details are proposed for deletion. CTS 4.6.3a states that "alternate manual starting (of the containment spray pumps) between the control room console and the local breaker shall be practiced in the test program." The ability to demonstrate the manual starting capability of the pumps from various plant locations is not assumed in the safety analyses and is not relevant to demonstrating that the pumps are capable of meeting their intended safety function. CTS 4.6.3b requires that the containment spray pumps " ... operate for at least fifteen minutes." The ability to demonstrate pump operation for an arbitrary period such as fifteen minutes is not representative of the ability of the pump to operate for an extended period following a postulated accident, as assumed in the safety analyses. Since the above details are not necessary to describe, or are not pertinent to, any . actual regulatory. requirement, they can be deleted without an impact of public health and safety .. This change is consistent with NUREG-1432.

1. 2. Does the change involve a significant increase in the probability or consequence of an accident .previously evaluated?
  • Analyzed events are assumed to be initiated by the failure of plant structures, systems or components.

Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed* to operate in response to the analyzed event. The proposed change deletes details from the Technical Specifications that are not . necessary to describe, or are not pertinent to, any actual regulatory requirement. The deletion of details from the Technical Specifications is not assumed to be an initiator of any analyzed event. The. proposed changes do not reduce*the functional requirement or

  • alter the intent of any specification.

As such, the consequences of an accident remain unchanged. Therefore, the proposed chl;mges do not involve a significant increase in the probability or consequences of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change deletes detail from the Specifications that are not necessary t6 describe, or are not _pertinent to, any actual regulatory requirement. The changes will not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. The changes will not impose different requirements' and adequate control of information will be maintained. The changes will not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes will not create the possibility of a new or different kind of accident from any accident previously evaluated. Palisades Nuclear Plant Page 7 of 8 05/31/99

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  • ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING 3. Does this change involve.a significant reduction in a margin of safety? I I Margin of safety is determined by the design and qualification of the plant equipment, I the operation of the plant within analyzed limits, and the point at which protective or I mitigative actions are initiated.

There are no design changes or equipment performance I parameter changes associated with this change. No setpoints are affected, and no I change is being proposed in the plant operational limits as a result of this change. The I proposed changes deletes details from the Technical Specifications. Removal of these I. details is acceptable since this information is not directly pertinent to the actual I requirement and does not alter the intent of the requirement. Since these details are not I necessary to adequately describe the actual regulatory requirement, they can be moved I to licensee controlled document without a significant impact on safety. Therefore, the I proposed changes do not involve a significant reduction in a margin of safety. I Palisades Nuclear Plant Page 8of8 05/31/99

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  • ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.7, HYDROGEN RECOMBINERS LESS RESTRICTIVE CHANGE L.1 In proposed ITS 3.6.7 Action A, a note is added which is not contained in the CTS. The note states "LCO 3.0.4 is not applicable." By specifying that LCO 3.0.4 is not applicable, the plant would be allowed to change MODES while relying on an Action statement for an inoperable hydrogen recombiner.

This is acceptable given the long Completion Time of 30 days if one recombiner is inoperable which indicates that this system is not of immediate*importance following an accident. Another redundant hydrogen recombiner also is required to be OPERABLE and is sufficient to maintain hydrogen below the flammability limit in the event of a design basis accident. This is considered to be a less restrictive change since in the proposed ITS, the plant would be allowed to change MODES while relying on an Action as opposed to the CTS which would not allow this. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change would allow the plant to change MODES with an inoperable hydrogen recombiner. The CTS allows one recombiner to be inoperable for 30 days. Therefore, if the plant were operating at 100% power and an inoperable recombiner were found, the plant could still operate for another 30 days before a shutdown would be required. By allowing the plant to change MODES with an inoperable recombiner the probability of having an accident within the 30 days either at power -or starting from a lower MODE without a recombiner and being allowed to increase power are no different. Therefore, extending allowing the plant to change MODES with an inoperable hydrogen recombiner cooling does not involve a significant increase in the probability of an accident previously evaluated. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning. of the equipment assumed to operate in response to the analyzed event. The proposed change will allow the plant to change MODES with an inoperable recombiner. However, a redundant hydrogen recombiner is still OPERABLE and sufficient to minimize the accumulation of hydrogen following a design basis accident., Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated . Palisades Nuclear Plant Page 1of7 05/31199

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  • 2. 3. ATTAC1ŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.7, HYDROGEN RECOMBINERS Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will allow the plant to change MODES with an inoperable hydrogen recombiner and limit the allowed outage time to the same as that for having an inoperable recombiner at power. Thus, this change does not create the possibility of a new or different kind of accident from* any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? The proposed change allows the plant to change MODES with an inoperable hydrogen recombiner. The margin of safety afforded by the hydrogen recombiners are to* ensure that the hydrogen concentration is minimized by burning the hydrogen before it reaches an explosive mixture. An OPERABLE hydrogen recombiner is still required artd is .. sufficient to ensure the hydrogen. concentration does not exceed the flammability limit. In addition, the allowed outage time for the inoperable recombiner is the same 30 days which would be allowed at power. Therefore, this change does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE CHANGE L.2. CTS Table 4.2-2 Item 11.a specifies that a hydrogen recombiner unit functional test be performed at least once per 6 months for each unit The proposed ITS SR 3. 6. 8 .1 that this test be performed every 18 months. The change from 6 months to 18 months is acceptable because the hydrogen recombiners generally pass the surveillance when performed at the 6 month frequency. In addition, the design of the recombiner is simple in that it is simply relying on heat to recombine hydrogen and oxygen. A 100 % capacity fully redundant hydrogen recombiner is also available and maintained OPERABLE. This is considered to be a less restrictive change since the hydrogen recombiner test will be perfornied every 18 months instead of the current every 6 months . Palisades Nuclear Plant Page 2of7 05/31/99

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  • 1. 2 . 3. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.7, HYDROGEN RECOMBINERS Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

Analyzed events are assumed to be initiated by the failure of plant structures, systems or, components. The proposed change extends the surveillance interval for a system functional test on the hydrogen recombiners from 6 months to 18 months. An extension in the surveillance interval is not assumed to be an initiator of any evaluated accident. Therefore, extending the interval for performing the test does not involve a significant increase in the probability of an accident previously evaluated. The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed change ' does not alter the requirement that the system functional test be performed on the hydrogen recombiners it simply increases the interval for performing the required. The consequences for accidents previously evaluated would be the same for either test interval. Therefore, the proposed change does not involve a significant increase fa the consequences of an accident previously evaluated. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will allow the system functional test to be performed every 18 months rather than the existing 6 month frequency. Thus, this change does not create the possibility of a new or different kind _of accident from any accident previously evaluated. Does this change involve a significant reduction in a margin of safety? The proposed change allows the surveillance. interval for performing the system functional test on the hydrogen recombiners to be changed from 6 months to 18 months. The margin of safety afforded by the hydrogen recombiners are to ensure that the hydrogen concentration is minimized by burning the hydrogen before it reaches an explosive concentration. Extending the surveillance interval from 6 months to 18 months is based on the fact that the test is normally successful when performed at a 6 month frequency and is also expected to pass the 18 month frequency based on engineering judgement. If for some reason the hydrogen recombiner was inoperable during the 18 month interval, but the inoperability was not known, an OPERABLE hydrogen recombiner is still available and is sufficient to ensure that the hydrogen concentration does not exceed the flammability limit. Therefore, this change does not involve a significant reduction in a margin of safety. \ Palisades Nuclear Plant Page 3of7 05/31/99

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  • ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.7, HYDROGEN RECOMBINERS LESS RESTRICTIVE CHANGE L.3 CTS Table 4.2-2, Footnote
  • to Item 11.a requires that the minimum recombiner heater sheath temperature increase be measured using installed or portable temperature monitoring instrumentation.

This detail is being removed from proposed SR 3. 6. 7 .1 and its associated Bases. This footnote provides no substantive requirement since any measurement of temperature must use either installed or portable temperature monitoring instrumentation. This change will not effect operation of the facility and is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability*

or consequence of an accident previously evaluated? The removal of this detail does not have an effect on the way the plant is operated. The Hydrogen Recombiners are provided for use following a postulated accident that results in generation of combustible gas in the containment building. The Hydrogen Recombiners are not an* initiator for any previously analyzed accident. Therefore removing this detail from the surveillance testing requirements associated with the Hydrogen Recombiners does not effect the probability of any accident previously evaluated .. The consequences of a previously analyzed-event ate dependent on the initial conditions assumed for the analysis, . and the availability and. successful functioning of the equipment assumed to operate in response to the analyzed event. The method of . . performing the temperature measurement that is part of the surveillance testing for the system does not effect the ability of the Hydrogen Recombiner to perform its safety function. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.

  • Does the change create the possibility of a new or different kind of accident from any accident previously eval_!1ated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will remove the detail provided in footnote

  • that indicates the sheath heater temperature increase is to be measured using installed or portable instrumentation.

Any measurement of temperature requires . . the use of either installed or portable instrumentation, and the proposed detail does not have a substantive effect on the way the required measurements are made. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated . Palisades Nuclear Plant Page 4of7 05/31/99

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  • 3. ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6.7, HYDROGEN RECOMBINERS Does this change involve .a significant reduction in a margin of safety? The proposed change will remove the detail provided in footnote
  • that indicates the sheath heater temperature increase is to be measured using installed or portable instrumentation.

Any measurement of temperature requires the use of either installed or portable instrumentation, and the proposed detail does not have a substantive effect on the way the required measurements are made. The margin of safety afforded by the hydrogen recombiners is to ensure that the hydrogen concentration is minimized by burning the hydrogen before it reaches an explosive mixture. The removal of this unnecessary detail from the Technical Specifications will not effect the way testing is performed or the availability of the required Hydrogen Recombiners. Therefore, this change does not involve a significant reduction in a margin of safety. LESS RESTRICTIVE CHANGE L.4 CTS Table 4.2-2, Item 11.b.3 requires the integrity check of the recombiner heater electrical circuits to be performed "immediately following" the system functional test of Item 11.a. This detail is an initial condition for performing the measurements required by proposed SR 3. 6. 7. 3. As such, this detail is more suitable for inclusion in plant procedures and is proposed to be deleted from the specifications. This is consistent with other testing where particular equipment configurations or conditions are required to be met to successfully complete the surveillance testing. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

The proposed change would remove the detail of an initial condition for performing the integrity check of the recombiner heater electrical circuit. Conditions and special equipment alignments are controlled by plant procedures which describe details of how surveillance testing is performed. The proposed change does not cause a change in the way surveillance testing is performed, it merely removes a level of detail that is unnecessary for inclusion in the specifications. The Hydrogen Recombiners are not an initiator for any previously analyzed accident. Therefore removing this detail from the surveillance testing requirements associated with the Hydrogen Recombiners does not effect the probability of any accident previously evaluated. The consequences of a previously analyzed event are dependent on the initial conditions assiimed for the analysis, and the availability and successful functioning of the

  • equipment assumed to operate in response to the analyzed event. The method of performing the temperature that is part of the surveillance testing for the system does not effect the ability of the Hydrogen Recombiner to perform its safety function.

Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated. Palisades Nuclear Plant Page 5of7 05/31/99

  • 2. ATTAC1ŽENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6. 7, HYDROGEN RECOMBINERS Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation. The proposed change will merely remove the detail of an initial condition to perform a check of the iiltegrity of the electrical heater circuits of the recombiners. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3.* Does this change involve a significant reduction in a margin of safety? The proposed change will remove the detail of an initial condition for performing the integrity check of the electrical heater circuits of the Hydrogen Recombiners. The margin of safety afforded by the hydrogen recombiners is to ensure that the hydrogen concentration is minimized by burning the hydrogen before it reaches an explosive mixture. The removal of this unnecessary detail from the Technical Specifications will not effect the way testing is performed or the availability of the required Hydrogen Recombiners. Therefore, this change does not involve a significant reduction in a * . margin of.safety ..

  • LESS RESTRICTIVE CHANGE L.5 CTS Table 4.2-2, Item 11.b.2 lists examples of abnormal conditions that are to be sought during the visual examination ofthe hydrogen recombiners.

These examples are neither exhaustive or explicit. These examples do not provide direction or limitations regarding the conduct of the required examination. As such, these details are more suitable for inclusion in plant procedures and are proposed to be deleted from the specifications. This change is consistent with NUREG-1432.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

The proposed change would remove examples of conditions that are to be considered during the visual examination of the Hydrogen Recombiners. These examples do not provide direction or limits that must be verified. Neither are they an exhaustive list of conditions that should be considered. Since the examples provide no guidance or limits, removing this detail from the surveillance testing requirements associated with the Hydrogen Recombiners does not effe<i:t the testing or operability of the recombiners in any way. Since no effect on the recombiners will occur, the probability of any accident previously evaluated is not effected. Palisades Nuclear Plant Page 6of7 05/31/99

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  • ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.6. 7, HYDROGEN RECOMBINERS
1. (continued)

The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The presence or lack of example conditions does not effect the ability of the Hydrogen Recombiner to perform its safety function. Therefore, the proposed change does not involve a . significant increase in the consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
3. The proposed change does not involve a physical alteration of the plant. No new or different type of equipment will be installed or changes made to plant parameters which govern normal plant operation.

The proposed change will merely remove the detail of example conditions to be considered during the visual inspection of the recombiners. Thus, this change does not create the possibility q_f a new or different kind of accident from any accident previously evaluated .. Does this change involve a significant reduction in a margin of safety? The proposed change will remove examples of conditions to be considered during the visual examination of the Hydrogen Recombiners. The margin of safety afforded by the hydrogen recombiners is to ensure that the hydrogen concentration is minimized by burning the hydrogen before it reaches an explosive mixture. The removal of this unnecessary detail from the Technical Specifications will not effect the way testing is performed or the availability of the 'required Hydrogen Recombiners. Therefore, this change does not involve a significant reduction in a margin of safety . **Palisades Nuclear Plant Page 7of7 05/31/99

  • ATTACHMENT 5 PALISADES NUCLEAR PLANT
  • SECTION 3.6, CONTAINMENT SYSTEMS MARKUP OF NUREG-1432 TECHNICAL SPECIFICATIONS
  • CTS J'."1.1-*--** I Containment 3.6.l 3.6 CONTAINMENT SYSTEMS 3 l Containment

!(Atmjspheric Dual LJ-(@) LCO 3.6.l Containment shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3; and 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment A-.1 Restore containment 1 hour inoperable. to OPERABLE status. . B. Required Action and* B.l Be in MODE 3. 6 hours associated Completion

  • Time not met. 8HQ B.2 Be in HOOE 5. 36 hours I cf()G sT¥1 3.6-1 J< \'.'"' \;

S Nv--c.\e,,__,. f>\.._,,_+

  • Revised 05/31/99
  • *
  • c;T-S t.t.S:. l c:rS t.t. S-. "2... c::l(l) © c... ,-s '-{ . ), t.+ .y,). S" '-1. ). Ip
  • Containment ai)d Dual
3. 6. l SURVEILLANCE REQUIREMENTS SURVEILLANCE

-:fit A SR 3.6.1.l (' 1 Perform required visual examinations and rate testing feP ee"tai"men! .a 1 r oc e 1 ng, in accordance with SR 3 .6.1. 2 10 CFR 50, ppendix J, as modif "ed by approved e emptions. The lea ge rate acceptance is s 1.0 L. However, during e first unit start following testing erformed in accor ance with 10 CFR SO Appendix J, as modi ied by approved exe ptions, the lea age rate acceptance criteria are < O. L for the Type B and pe C tests, and 0.75 L. for the Ty A test. Verify containment structural in accordance with the Surveillance Program. Struc.t iJ "t n +c. r* ty < lr05c.RT s R / CEOG STS ,, 3.6-2 FREQUENCY_ In accordance with the Containment nTnOOil) 'SUr'Veil 1 ance Program . Rev l, 04/07/95 Revised 05/31/99

  • INSERT SURVIELLANCE SR 3.6.1.3 -----------------------N()TE--------------------------------

Local leak rate tests shall be performed 55 psig. ---------------------------------------------------------------- Perform required Type B and C leakage rate testing, except for containment air lock testing, in accordance with 10 CFR 50, Appendix J, ()ption A, as modified by approved exemptions. The leakage rate acceptance criterion is s 1.0 La. However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, ()ption A, as modified by approved exemptions, the leakage rate acceptance criteria are . < 0.6 La for the Type B and Type C tests . * \\

  • 3.6-2 FREQUENCY

N()TE:-------

SR 3.0.2 is not applicable.

In accordance with 10 CFR 50, Appendix J, ()ption as modified by approved exemptions Revised 05/31/99

  • *
  • c_,-S ?!o.1 3.6 CONTAINMENT SYSTEMS Containment Air Locks j.6.2 3.6.2 Containment Air Locks j(Atmftspheric al)d Dual)j LCO 3.6.2 frwof"containment air shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.* ACTIONS -------------------------------------NOTES------------------------------------

1. Entry and exit is permissible to erform repairs on the affected air lock components.

I 2. Separate Condition entry is allowed for each air lock.

  • 3. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment,*

when leakage results in exceeding the overall containment rate acceptance criteria.

CONDITION A. One or more containment air locks with one containment air lock door inoperable. CEOG STS REQUIRED ACTION ------------NOTES------------

1. Required Actions A.I, A.2, and A.3 are not applicable if both doors in the same air lock are inoperable and Condition C is entered. 2. Entry and exit is permissible for 7 days under adm!9istrative controlsj"'lf locks are

\\ 3.6-3 COMPLETION TIME (j) {continued) Rev 1, 04/07/95 Revised 05/31/99

  • ACTIONS CONDITION A. {continued)
  • B. One or more containment air locks with containment air *lock interlock mechanis*

inoperable. . CEOG STS *

  • Containment Air Locks l(Attlospheri C/and Dua}/) h:1J 3.6.2 REQUIRED ACTION ,. A. l Verify the OPERABLE nu door is closed in the affected air lock. ANO A.2 Lock the OPERABLE (\ .. ,,, door closed in the affected air lock. Mill A.3 ...

NOTE---------(lf..-' Air lock doors in high radiation areas may be verified locked closed by administrative means. ---------------------

Verify the OPERABLE door is locked closed in the affected air lock. ------------NOTES------------

1. Required Actions B.l, tie.; B.2, and 8.3 are not appl1cable if both doors in the same air lock are inoperable and Condition C is entered. 2. Entry and exit of f'..(./ containment is permissible under the control of a dedicated individual.

\ \ 3.6-4 COMPLETION TIME 1 hour 24 hours Once per 31 days {continued) Rev 1, 04/07/95 Revised 05/31/99

  • ACTIONS CONDITION B. (continued) 8.1 AND B.2 AND B.3
  • c. One or more C.1 containment air locks inoperable for reasons other than Condition A or B. Af!D. C.2 Mill C.3 CEOG STS * * , Containment Air Locks QAtTT\({sphe}ic an£oual
3. 6 . REQUIRED ACTION Verify an OPERABLE door is closed in the affected air lock. Lock an OPERABLE door closed in the affected air lock.

Air lock doors in high radiation areas may be verified locked closed by administrative means. --------------------- Verify an OPERABLE door is locked closed in the affected air lock. Initiate action to evaluate overall containment leakage rate per LCO 3.6.l. Verify a door is closed in the affected air lock. Restore lock to OPERABLE status. 3.6-5 COMPLETION TIME 1 hour 24 hours . Once per 31 days Invnediately 1 hour 24 hours (continued) Rev 1, 04/07/95 Revised 05/31/99

  • c..-rs y. s:z... c.{)J l,( ,() c..T.S L\.<;,(.,d.l.k
  • *
  • Containment Air Locks i(AtmoApheric ar)d Dual)f-{j)
3. 6. 2 ACTIONS (continued)

CONDITION REQUIRED ACTION D. Required Action and D. l Be in MODE 3. associated Completion Time not met.

  • ANO D.2 Be in MOOE 5. SURVEILLANCE REQUIREMENTS SR 3.6.2.1 CEOG STS SURVEILLANCE

s -----------------

1. An inoperable air ck door does not invalidate the pr ious successful performance oft overall air lock leakage test. 2. *Results shall e evaluated against acceptance t iteria of 3.6.1.l in accordance ith 10 CFR 50, Appendix J, rnodifi by approved exemptions.

Perfor11 req red air lock leakage rate testing in ccordance with 10 CFR 50, Appendix , as modified by approved exemptio s. The a eptance criteria for air lock test g are: Ovtr*ll air lock leakage rate is (0.05 L.] when tested at P ** For each door, leakage rate is s (.01 L.] when tested at (10.0 psig]. \ \ 3.6-6 COMPLETION TIME 6 hours 36 hours FREQUENCY

NOT ------SR 3.0 is not appli ble In accordance th 10 CFR 50, ppendix J, as modified by approved exemptions (continued)

Rev 1, 04/07 /95 Revised . 05/31/99

  • *
  • INSERT C.TS 4. s. z.d 16 SR 3 .6.2.1 -------------------NOTES-------------------
1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. r-.t.W 2. 3. 4. Results shall be evaluated against acceptance criteria of SR 3.6.1.3 in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions.

A seal contact check shall be perfonned on the emergency escape air lock following each full pressure test. Emergency escape air lock door opening, solely for the purpose of strongback removal and performance of the .seal contact check, does not necessitate additional pressure testing. Local leak rate tests, other than personnel air lock doors, shall be performed 55 psig. Perform required air lock leakage rate testing in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. The acceptance criteria for air lock testing are: a. Overall air lock leakage rate is s 1.0 L. when tested P. and combined with all penetrations and valves subjected to Type Band C tests. However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakage rate

  • acceptance criteria is < 0.6 L, when combined with all penetrations and valves subjected to Type B and C tests. b. For each personnel air lock door, leakage rate is s 0.023 L. when tested 10.0 psig. c. An acceptable emergency escape air lock door seal contact check consists of a verification of continuous contact between the seals and the sealing surfaces.

\ \ 3.6-6 SR 3.0.2 is not applicable In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions Revised 05/31/99

  • * *
  • Containment Air Locks and 3.6.2 SURVEILLANCE RE UIREMENTS continued SR 3.6.2.2 CEOG STS SURVEILLANCE FREQUENCY

NOTE--

Only requ* ed to be perf rmed upon entry or exit thr gh the conta' ment air lock. Verify only one door in the air lock can be 1 'i' opened at a time . \I 3.6-7 Rev 1, 04/07 /95 Revised 05/31/99

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  • Containment Isolation Valves /(Atmo,{Dheric 3.6 CONTAINMENT SYSTEMS 3.6.3 Containment Isolation Valves j(Atmofa'pheric and LCO 3.6.3 Each containment isolation valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS 0.f'J. IL ; ,;..: ,. t'OO""' may be under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.
3. Enter applicable Conditions and Required Actions for system(s) made inoperable by containment valves. 4. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," when leakage results in exceeding the overall containment leakage rate acceptance criteria . A. CONDITION

NOTE---------

Only applicable to . penetration flow paths with two containment isolation valves. One or 110n penetrattoa flow paths with one contain111nt isolation valve xcept for e age o.-a:,.. <"oo.-. U' d1119 s"'-ffl1 ""t-not 1-=><-keJ c..loHcl. CEOG STS

  • A. l \ \ REQUIRED ACTION Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured. 3.6-8 COMPLETION TIME 4 hours (continued)

Rev 1, 04/07 /95 Revised 05/31/99

  • *
  • Containment Isolation Valves I@ . 6. 3 ACTIONS CONDITION A. (continued)

A.2 B. ---------NOTE--------- 8.1 Only applicable to penetration flow paths with two containment isolation valves. ---------------------- CEOG STS

  • REQUIRED ACT ION --------NOTE---------

Isol ation devices in high radiation areas may be verified by use of administrative means. Verify the affected penetration flow path is isolated. COMPLETION TIME Once per 31 days for isolation devices outside containment Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment Isolate the affected 1 hour penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. 3.6-9 (continued) Rev 1, 04/07 /95 Revised 05/31/99

  • *
  • Containment Isolation Valves l(Atijl(}spheric 3.6.3 ACTIONS continued CONDITION C.

Only applicable to penetration flow paths with only one containment isolation valve and a closed system. C.l One or more AND penetration flow paths with one containment C.2 @ isolation valve inoperable. L Seco dary containment byp ss leakage not wi in limit. ,. One or mo penetrat with contat t purge valv .110t within pu valve leakage 1111 s. CEOG STS 0.1 E. l ,, REQUIRED ACTION Isolate the affected penetration flow path by use of at least one closed and de---activated automatic valve, closed manual valve, or blind flange. --------NOTE--------- Isol ation devices in high radiation areas may be verified by use of administrative means. COMPLETION TIME Whours @ I@ Verify the affected Once per 31 days penetration flow path is Is ate the affected p etration flow path use of at least ne [closed and de-activated automatic valve with resilient seals, closed manual valve with resilient seals, or blind flange]. 3.6-10 continued Rev 1, 04/07/95 ----1 r-. "r rnQl'(. + . v. ci<J..,,.,..s '°"" a. ..... \>,\ L bc..1'. c:..\o.sed ... -.ic..\"'-., c.1S' J.b.S

  • ro<>""" c;"'ri1

..... + Revised 05/31/99

  • @ * @ *
  • ACTIONS CONDITION (continued) Requt,.. Actioe and assact.a.l tallpltt1on TiM not Mt. CEOG STS Containment Isolation Valvesl(Atililfsphericjana Dual) t-<I3J 3. 6. 3 ANO REQUIRED ACTION Isolation devices in high radiation areas may be verified by use of administrative means. Verify the affected penetration flow path is isolated.

COMPLETION TIME Once p for is lation devic s outside cont inment P ior to ntering MODE 4 from MOOE 5 if not performed within the previous 92 days for isolation devices inside containment E.3 Perform SR 3.6 .. 6 Once per for the resili nt [ ] days seal purge v ves closed to co ply with Required Ac ion E.l. Be in HOOE 3. Be in MOOE 5. 3.6-11 6 hours 36 hours Rev I, 04/07 /95 Revised 05/31/99 Containment Isolation Valves l(At!i}4spheric/and Dual) f-@ I 3

  • 6
  • 3
  • SURVEILLANCE REQUIREMENTS L 3.6.3.1 SR 3.6.3.CD SR
  • ©
  • a"J fl.o.\-

1 or o-1'-"e .. e.. sec ... ,eJ., CEOG STS

  • SURVEILLANCE FREQUENCY V rify each [42] inch purge valve is 31 ays ealed closed except for o e purge valve in* a penetration flow path ile in Condition E of this LCO. -------------------NOTE--------------------

Valves and blind flanges in high radiation. areas may be verified by use of administrative means. ------------------------------------------- 31 days Verify each containment i so lat ion manua J@ 31 days valve and blind flange that is loca e outside conta1nmentlland is required to be closed during accident conditions 0 is closed, except for containment isolation valves that are*open under administrative controls. ,, 3.6-12 (continued) Rev 04/07/95 Revised 05/31/99 Containment Isolation Valves i{Atmefspheri? and Ou,il )l@ 3.6.3

  • SURVEILLANCE RE UIREMENTS continued L1. s. 3.J SR 3.6.3 0."'ol, ri.dr-\o<...k..

I or .,t\-.e..r \ e.. -'"' S 1 ?_, \ c... T5 't. S'. 5. c.. ** *

  • SR 3.6.3.@ SR CEOG STS SURVEILLANCE Valves and blind flanges in high radiation areas may be verified by use of administrative means. Verify the isolation time of leath po\IJC!r! -ppeJijlted indl each automatic ontainment isolation valve is within limits. . TS Ii=°-4lo, Ii!. I FREQUENCY Prior to entering HOOE 4 from HOOE 5 if not perfonned within the previous 92 days 184 days Verify each automatic containment isolation Jffs}/months valve that is not locked, sealed, or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal. \\ 3.6-13 {continued}

Rev 1, 04/07 /95 Revised 05/31/99

  • * **
  • Containment Isolation Valves\(Atefospheri/?

and L6.3 SURVEILLANCE RE UIREMENTS continued L 3.6.3.8 r 3.6.3.9 CEOG STS SURVEILLANCE Ve ify each ( ] inch c ntainment purge v lve is blocked tor strict the valve pening > (50]%. Verify the combin d leakage rate for al secondary contai ment bypass leakage p ths is s [ L.] whe pressurized [ ig]. 3.6-14 FREQUENCY [!BJ months J @ -----NOTE----- 0-SR 3.0.2 is not applicable In accordance with 10 CFR SO, Appendix J, as modified by approved exemptions Rev 1, 04/07/95 Revised 05/31/99

  • c..15 s (.,, l.. * *
  • Containment Pressure /cAti?ospher}t and DuJl 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure l(AtJli/osphericJnd Oual)t-@ LCO 3.6.4 1.D ps: 5 ; ... MovES S ps: ;,.._ Mo'DE5. APPLICABILITY:

MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION A. Containment pressure A. l Restore containment not within limitt*l{D pressure to within l imi tel. I© B. Required Action and B.l Be in MODE 3. associated Completion ruib Time not met. B.2 Be in MOOE 5. SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.6.4.1 Verify containment pressure is within li1111tf. l@ CEOG STS 3.6-15 COMPLETION TIME 1 hour . 6 hours 36 hours FREQUENCY 12 hours Rev 1, 04/07 /95 Revised 05/31/99

  • *
  • Containment Air Temperature

[{Atr;6spheric an/? Dual) r@ 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature/(At/jl6spheric and/ifual) f-@ LCO 3.6.5 Containment average air temperature shall / (I) (1Ltq} APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS ----------------=r=------------------------ ...... --..... -----====- CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average air temperature not within limit. A. I Restore containment 8 hours average air B. Required Action and B.l associated Completion Time not met. AND B.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE temperature to within limit. Be in MOOE 3. Be in MODE 5. SR 3.i.5.l Verify containment average air temperature 1s within 1 imit. CEOG STS ' ' 3.6-16 6 hours 36 hours FREQUENCY 24 hours Rev 1, 04/07/95 Revised 05/31/99

  • @ * *
  • Containment Sprt?' and Cooling // Systems./(tmospheri c and Dua 1) 3.6.6A / 3.6 CONTAINMENT SYSTEMS I 3.6.6A Containment Cooling Systems (Atiospheric and Dual) / (Credit taken ff( iodine removal by ti 1 containment Spray System);/

I I .* I LCO 3.6.6A Two;containment spray trairs and two containment coo}ing tr 1 ai ns shall be OPERABL/. / ' I I . I I 1 ,. , I APPLICABILITY:/ HODES I, 2, (and] 4]. !' I I , ACTIONS / I / // REQUIRED ACTION // COMPLETION TIME I A. /One containment spray A.1 Restore conttin ent spray train t OPERABLE sta s. I train inoperable/ ./ / I B. and Completion Time/of Condition.A not-'met. I ./ / /' ,e'. One COAtliAMnt / C09l i .. tr&ill / 1noperlb11. CEOG (' I I 8.1 HOOE 3. Mill I B.Jf Be in MODE 5. C.1 Restore conta' ment cooling trai to OPERABLE s tus. 72 hours 10 days fr discovery of failure o meet the LC 84 hours 7 days AHQ 10 days from discovery of failure to meet the LCO (continued) Rev 1, 04/07 /95 Revised 05/31/99

  • *
  • c.1s J.4.1 Canta i nment Cooling Systems J (Atijl6spheri c ajj'd Dua ffi-10 3.6 CONTAINMENT SYSTEMS 3.6.681 LCO 3. 6. 66\ Two containment lsprji trains and/two containffiintl cooling trains shall be OPE BLE. APPLICABILITY:

HODES 1, Z, /and 4f.j ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME c.:\t 3.'\.1. <:.T'S cq.,-,.,.,::o.-e.. l_ ---Onevcontainment A.l Restore containment . I

  • ANQ. A+-\ e<:.st' I oo ").., Qt:
  • C..oJ :..,. J C."-f o. I.: l ,'-/-(

+... .... O\>E 12.AiJ t£ l. o,..,.;-.. : .... _ .. _t c.ocJ:,,, + ... :.. B. One containment cooling train inoperable. C. Two ontainment spray tr ns inoperable. CEOG STS to OPERABLE status. 8.1 I \\ Resto containment cool* g train to OPE BLE status. Restore one containment sp train to OPE status. 3.6-Zl 4 days from (0f{"'* discovery of U:> failure to meet the LCO . 7Z hours (continued} Rev 1, 04/07/95 Revised 05/31/99

  • C.."\:S ]."\.J * *
  • Containment l i ng Sys terns le A tJiiOJilh er i c . .6.6. JD ACTIONS continued CONDITION D. One containment spra train and one containment coolin train I E. Two co ainment cooli trains inop able. f. Any comb* at ion of three o more trains inoper. ble. CEOG STS D. l OR D.2 E.l \\ REQUIRED ACTION Restore con ainment spray tr a i to OPERABLE tatus. Rest e containment ing train to BLE status. Restore one containment cooli train to OPERAB status. Be in MOOE 3. Be in MOOE (l Enter LCO 3.0.3. 3.6-22 COMPLETION TIME I I , 72 hours 6 hours if hours rzD Irrmediately Rev 1, 04/07/95 Revised 05/31/99
  • CTS 11 <...
  • * * :0 Conh i nment Coo 1 i ng Sys terns j (A tm/spheri c /!nd . --3.6.6 ,JD SURVEILLANCE REQUIREMENTS SR 3.6.6'.l SR 3.6.6f.2 SR 3.6 .* &* CEOG STS SURVEILLANCE Verify each containment spray manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position. /-'

Operate each containment t1a;n fan unit for 15 minutes. h<!.J I Verify each automatic containment spray valve 1R the flow path that is not locked, sealed, or otherwise secured in position, actuates to its correct position on an actual or simulated actuation signal. 3.6-23 FREQUENCY 31 days 31 days ( G) 31 days In accordance with the Inservice Testing Program fl Sf months I QJ {continued) Rev 1, 04/07 /95 Revised 05/31/99

  • I '*::-:-I,\
  • * * /Cf) Containment ifn9)Cool ing Systems i(Atf:ll6spheric/and Quall:}-:§ 3 . 6 . 6@:]---{ID SURVEILLANCE RE UIREMENTS continued SURVEILLANCE SR 3.6.si.7 Verify each containment spray pump starts automatically on an actual or simulated actuation signal. SR Verify each containment cooling starts automatically on an actual or simulated actuation signal. SR 3.6.68.9 Verify each spray nozzle is unobstructed.

\\ CEOG STS 3.6-24 FREQUENCY {is( months !18? months 10 years Rev 1, 04/07/95 Revised 05/31/99

  • Spray Additive System (Atmospheric and rfual) 3.6.7 3.6 CONTAINMENT S tive System*(Atmospheric and Dual) The Spray Additive System shall be OP/ / / MODES 1, 2, [and] 3(, and 4]. CONDITION A. Spray Additive System A.I inoperable.

estore Spray Additive System to OPERABLE status. B. Required Action and // B.l associated Completi<>ri Ti me not met. / AND Be in MODE 3. _,/ Be in / B.2 / / / SURVEILLANCE REQUIREMENTS SR 3.6.7.1 SURVEILLANCE Verify each spray additive manual, operated, and automatic valve in path that is not locked, sealed, otherwise secured in position

  • correct position.

*---------

CEOG STS 3.6-25 .. / // power e flow COMPLETION TIME 72 hours 6 hours 84 hours FREQUENCY 31 days (continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • 3.6 CONTAINMENT SYSTEMS 3.6. Hydrogen Recombiners l(Atmo?Pheric and Dual,//( if permanentl/

installed) / 6) LCO 3. 6 .rj -tf wof hydrogen recombi ners sha 11 be OPERABLE. I CD G) & APPLICABILITY: MODES l and 2. ACTIONS CONDITION A. One hydrogen A.l recombiner inoperable.

8. Two hydrogen recombiners inoperable.

8.1 REQUIRED ACTION N).._>) --------NOTE--------- LCO 3.0.4 is not applicable. Restore hydrogen recombiner to OPERABLE status . COMPLETION TIME 30 days Verify by l hour administrativ. means that the hy ogen ANO control fu ction is maintaine . Every 12 hours thereafter 8.2 Rest re one hydrogen 7 days re mbiner to OP. RA8LE status. Required Action and associated Completion Time not met. Be in MOOE 3. 6 hours \ \ CEOG STS

  • 3.6-27 Rev 1, 04/07 /95 Revised 05/31/99
  • . G TS' ""\cJ..le L\.1--L ri-... _.1\.0.. en

'-!.£-1. J:i.. *-11. 6 . l.. C'f'S \el-l" 4.1-1 11.b.3 * *

  • Hydrogen Recombi ners l(Atm.f?spheri c an;d J \ f7'i 3.6. Q) b' SURVEILLANCE REQUIREMENTS SR 3.6.n.l 0J SR SR 3.6t3 CEOG STS SURVEILLANCE Perform a system functional test for each hydrogen recombiner.

Visually examine each hydrogen recombiner enclosure and verify there is no evidence of abnormal conditions. i:..--.A /0 Perform a'4-res1stance to ground test for each heater phase * \ \ 3.6-28 FREQUENCY fl Sf months -l'-'(1ar months (1sr months l CD \© I (J) Rev 1, 04/07 /95 Revised 05/31/99

  • *

*-*---*-*-

/ *' HMs (Atmospheri_c __ // 3 9 / /./' I / / I 3.6 CONTAINMENT -SYSTEM5 3.6.9 Hydrogen Mixing System (Hr%}/ (Atmospheric and Dual) .// 1* 1 LCO 3.6.9 [Two] HMS shall be OPERABLE. // /// and 2. APPLICABILITY: MODES I ACTIONS I COND I_JiION I---,"----+--....,.....___-+---------......., 1 A. One A. l -----./!..NOTE--------- !j inoperable. LCO l.0.4 is not I applicable. ' II ! _/, .. . //' i / .1 Restore HMS train to REQUIRED ACyfoN I COMPLETION TIME / ,.// // OPERABLE status. *' / / Il l

8. Two HMS trains // 8.1 Verify by / 1 hour 1
  • inoperable.

adm1n1strat1ve I , / that the hydrogery' Arfl2 //. control funct1opl is . maintained. / I Arfl2 i I .B.2 Restore one HMS train j; to OP BLE status. ' I I C. "quirM Action and 1 associated Ca11pletion Ti Jiii not met. CEOG STS

  • C.l \ \ 3.6-29 3.fo -30 Every 12 hours thereafter 7 days 6 hours Rev 1, 04/07 /95 Revised 05/31/99
  • * * ... *--------I res
3. s .10 I 3.6 CONTAINMENT SYSTEMS I 3. 6 .10 lad t ne Cleanup Sy/. (!CS) (Atmospheric and Dua 1 j LCO 3.6.10 shall be OPERABi APPLICABILITY:

ODES 1, 2, 3, and 4. / / ACTIONS A. Required Action and associated Completion Ti me not met . I A.I !CS train to 1 OPERABLE status. Be in MODE 3. B.2 Be in MOOE 5. SURVEILLANCE Operate each ICS trai for [;i: 10 continuous hours with heaters o erating or (for systems without he ers) ;i: 15 minutes]. Perfona requ ed ICS filter testing in accordance ith the Ventilation Filter Testing Pr. gram (VFTP). 36 hours (continued) CEOG STS 3.6-31 Rev 1, 04/07/95 I 3, b -3 2.

  • Revised 05/31/99
  • * ** / / ----------*--5h-;-;1d6u
-1ci
  • /.6.11 / // // I 3. 6 CONTAINMENT SYSTEMS // ! 3.6.11 Shield Building (Dualf I LCO 3. 6. II Shi el d .bilridi ng sha 11 be OPERABLE.

/./; ,' / // .. ; APPLICABILITY: MPDES /' 1 , 2 , 3 , and 4 . // / ,,-/' / / ACTIONS I A. $11ield building /inoperable. / 8. Required Action and / associated Completiqn _/ // ACTION /. A.I ,R'estore shield /,,.building to OPERABLE status. * / Be in MODE 3. Time not met. // A.till /' B.2 .* /*' / / SURVEILkANCE REQUIREMENTS , // SURVE I LLANC CEOG STS Verify annulus egative pressure is > [5] inches ater gauge. \ \ 3.6-33 .\-\..'Cl 3.'--:S4 COMPLETION TI 36 hours (continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • Vacuum Relief Valves (Dual) 3.6.12 3.6 CONTAINMENT SYSTEMS 3.6.12 Vacuum LCO 3.6.12 Tj'O"vacuum relief lines shall APPLICAB I, 2, 3, and 4. be OPERABLE.

,/ CONDITION REQUIRED COMPLETION TIME / A. One vacuum relief line A.1 inoperable. relief 72 hours line )'b OPERABLE B. Required Action and as soc i ated Completion Time not met. /-Alli2 ,. .-// 8.2 ' ,,,,,,. .... .' .,/' sta.tUs. . / Be in MODE 3. Be in HOOE 5. SURVEILLANCE SR 3.6.12.1 Verify each vacuum relief li OPERABLE in accordance with the Inserv* e Testing Program. CEOG STS 3.6-35 6 hours FREQUENCY In accordance with the Inservice Testing Program Rev 1, 04/07/95 Revised 05/31/99

  • *
  • 13.6 CONTAINMENT SYSTEMS / ! SBEACS (Dual) 3.6.13 13.6.13 Shield Building Exhaust Air Cl anup System (SBEACS) (Dual) ' I I LCO 3.6.13 Two SBEACS train shall be OPERABLE.

I I APPLICABILITY: I I ACTIONS COMPLETION TIME f REQUIRED ACTION Restore train OPERABLE sta s. A.' A. l B. equired Action and *a.1 Associated Completion I . Time not met. AtiQ 8.2 e in MODE 5. SR 3.6.13.f O erate each SBEACS train for 10 continuous hours with the heaters operating or (for systems without heate s) 15 minutes]. Perform required SBEACS filter testing in accordance with the Ventilat on Filter Testing Program (VFTP). CEOG STS 3.6-36 ""'r o 1.4..1 k ,{,-J 7 7 days 6 hours 36 hours REQUENCY In accordance with the VFTP (continued) Rev 1, 04/07/95 Revised 05/31/99 ( '--* / I

  • *
  • Containment B 3.6.1 B 3.6* CONTAINMENT SYSTEMS B 3.6.1 Containment lcAifuospherlc>

' / ' BASES ( lsm.c.+ur'- ];XJ wi+n '3tcc:.L Pfq:kJ BACKGROUND '. The containment consists of t.fe concrete I reactor buiXd}Ji9l I 8) 100!), 1 ty and the penetrations through this structure. Int c ure is designed to contain radioactive material that may be released from the reactor core following a Design Basis Accident (OBA). Additionally, this structure provides shielding from the fission products that may be present in the containment atmosphere following accident conditions.

  • The containment is a reinforced concrete structure with a ,e:"\:,."'.._J_

vl;+i,.. wall, a flat foundati51n mat, and a dome jl;'.;:'t\* ,__ ___ r_oo,,..,f,.... .... "" .. For co.ritainment§ with ungr,tSuted tendons, ,{;he cyJ inder \..V r>n:\J. --;+eel . .-."\. wall is prestressed with a post tensioning system in the 11-u. : f"'2ss ... vertical and horizontal di dome roof is I@ prestressed utilizing a three way plfst tensioning system. /c:.j_s 0"' 14 sle1.b The inside surface of the containment is lined with a carbon steel liner to ensure *a high degree of leak tightness during b 0-+h -\-1....Q_ operating and accident conditions.

,,: I ('NIJS .,..rt_ . s+.-.........+-e..

Q The is requ1red structural integrity of the o.."'.,l.. J..\....a. -: h "4.W.... ... containment under OBA conditions. The steel liner and its

  • r . h * ;-fe penetrations establish the leakage limiting boundary of the

(.Q"L""' containment. Maintaining the containment OPERABLE limits *,;\ ...... \. the leakage of fission product radioacthity from the containment to the environment. SR 3.6.1.1 leaka e rate Ir-?\ requ1remen s comp y w1 , ppen 1x f. 1 , as -?::J modified by appr9ved exemptions. > ophJ .... c Fe .. T,reA +est,: 1"0 "',.). O fo, T,,. 13....C...+est: The isolation devices for the penetrations 1n t e containment boundary are a part of the containment leak tight barrier. To maintain this leak tight barrier: a. All penetrations required to be closed during accident conditions are_ either: 1. capable of being closed by an OPERABLE automatic containment isolation system, or (continued) B 3.6-1 -"'; . + t-)--.lee:.,. ?I ... -"- o-'1-* Revised 05/31/99

  • * * .------. BASES BACKGROUND (continued)
b. c. Containment ' 8 3 .. 2. closed by manual va_lves, blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in LCO 3.6.3, "Containment Isolation Valves"; e pressurized s ling mechanism as ociated with a exc t as provided in CO 3.6.[ ], *is APPLICABLE The_ safety design basis for the containment is that the SAFETY ANALYSES containment must withstand the pressures and temperatures of' the limiting OBA without exceeding the design leakage rate. f'..,.. _.""-f._\: i: .. J es

,.,....._.__;_..,._.- fee.I::.. c,.,,..+ .. : ... +,...,_ .._ l"\St..i!, .. u.:J,..,.-\-: 1-\o....e'e'i c.cc.;d.,..,_+- "'"' rerSf*t:\-; .M.. IS o.. l..CX:..A, S f>reH .... ; 5 . a..S i' .... Tl-r ......... .. + "is t:"r .. +Wt. . a.,-.. .. .., ... r... ......... .t ."\,. The DBAs that in a release of radioactive material within containment are a loss of coolant accident, a lru\ steam 1 ine break (MSLBho and a control leiijirient as§embly & ejection accident (Ref1YIZ). In the analysis of each of 1(6) these it is assumed.that containment is OPERABLE . such that release of fission products to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate of (O.lO}i of containment air weight per day (Ref. J). This I leakage rate is defined in 10 CFR 50, Appendix J {gef. /11, (1) as L.: the maximum allowable containment leakage rate at .. e+.,,.._....._ I. maximum peak containment (Pa) of which results from the 1 imiting1' , lwlrich !s ldesHjh basu MSLBI I© ldes: 1 ... LCO CEOG STS

  • Satisfactory leakage rate test results are a requirement_

for the establishment of containment OPERABILITY. Containment OPERABILITY is maintained by limiting leakage to s 1.0 L , except prior to the first startup after performing a 10 CFR 50, Appendix J, leakage test. At this " B 3. 6-2 (continued) Rev 1, 04/07/95 Revised 05/31/99 .

  • * *
  • BASES LCO (continued)

APPLICABILITY ACTIONS CEOG STS Containment j (AS'rifosphSfi cf Hol B 3 .6. -Compliance with this LCOJW"ill ensure a containment !Iii configuration, including equipment hatch8}, that I 1_:1.J is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis. ""'"': c..i-. 1--c. Individual leakag;_,,rates specified for t e containment air lock (LCO 3.6.2) 1and purge valves resilient seals j (}) (LCO 3.6.3)t'are not specifically part of the acceptance criteria of 10 CFR 50, Appendix J. Therefore, leakage rates exceeding these individual limits only result in the containment being inoperable when the leaka e results in exceeding criteria of A endi ]I (p '* (.OL<>-In MODES 1, 2, 3, and 4, a OBA could cause a release of radioactive

  • material into containment.

In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature of. these MODES. Therefore, containment is not required to be OPERABLE in MODE 5 to prevent leakage of radioactive material from containment. The.requirements for containment during MODE 6 are addressed in LCO 3.9.3, "Containment Penetrations.n In the event containment is inoperable, containment must be restored to OPERABLE status within I hour. The 1 hour Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining r;\ containmeiil:Vduring MODES 1, 2, 3, and 4. This time period also ensures that the probability of an accident (requiring containment OPERABILITY) periods when \ Cb) containment is inoperable©is minimal. B.l and B.2 If containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a B 3*.6-3 ,, (continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • BASES ACTIONS SURVEILLANCE REQUIREMENTS -tk Lo.-f-.::.:-

.... e ... t-L.e .. ... h +ollow;,.J OIA.+.cy ;ndJeJ 'ire. A +.di ... _...e-+-L"'.,J::.. Tesk.. p,.<YJ'"".--. CEOG STS

  • Containment

!-@ B 3. * . B.1 and B.2 (continued} MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant SR 3. 6 .1.1 SR 3.6.1.2 " B 3.6-4 {continued} Rev 1, 04/07/95 @ Revised / 05/31/99

  • *
  • INSERT SR 3 6 J 3 Maintaining the containment OPERABLE requires compliance with the Type B and C leakage rate test requirements of 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.

Testing is performed at pressures 55 psig. As left leakage prior to the first startup performing a required 10 CFR 50, Appendix J, Option A, leakage test is required to be < 0.6 La for combined Type B and C leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of!> 1.0 La. At !> 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by Appendix J, Option A, as modified by approved exemptions. Thus, SR 3.0.2 (which allows Frequency extensions) does not apply. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis. . SR 3.6.1.3 is modified by a Note which states that local leak tests shall be performed at 55 psig. This value corresponds to the design pressure of the containment and bounds the maximum expected internal pressure resulting.from an MSLB or design basis LOCA . \ \ B 3.6-4 Revised 05/31/99 )

  • * *
  • BASES (continued)

REFERENCES Containment M§osp_h_ij'.i c p--(§) B 3.6.1 I.! 10/CFR 50, APJ{endix J. FSAR, Cho.\lte,r /lf fw 2. FSAR, Section f I ([) 3. FSAR, Section j (i) /4. Re/4ulatory GV'ide 1.35, ;Revision [1]. t- CEOG STS ,, B 3.6-5 Rev 1, 04/07/95 Revised 05/31/99

  • I f-----------------

---*--------....------, Contai ent (Dual) B 3.6.1 . B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Containment (Dual / BASES I I BACKGROUND he containment is a free stan ng steel pressure vessel I surrounded by a reinforced c crete shield building. The ! containment vessel, includi all its penetrations, is a low I leakage steel shell desig d to contain*radioactive ma?lterial 1 that may be rel eased fro the reactor core fo 11 owing a 1 Design Basis Accident BA). Additionally, the containme 1 , and shield building p ovide shielding from the fission products that may b present in the containment atmosp re I following accident conditions. . ' I The containmen vessel is a vertical cylindrical eel 1 pressure vess with hemispherical dome and ell* soidal bottom, com etely enclosed by a reinforced co crete shield building. 4 ft wide annular space exists the wa 11 s an domes of the steel containment l and the. concret shield building to pennit inser.v v)'t'ce inspection and colle ion of containment outleakage. O(ial containments util 'ze an outer concrete building for. shielding and an in r steel containment for leak ti ness. I ontainment piping penetration as emblies provide for the passage of process, service, s ling, and instrumentation I pipelines i.nto the containmen vessel while maintaining containment OPERABILITY. Th shield building provides / biological shielding and a ows controlled release of the ,!' annulus atmosphere under ccident conditions, as well as environmental missile otection for the containment vessel and the Nuclear Steam upply System.

  • l The inner steel co ainment and its penetrations establis the leakage limi ng boundary of the containment.

Maintaining th containment OPERABLE limits the leaka of fission produ radioactivity from the containment t the environment Loss of containment OPERABILITY coul cause ! i

  • site bound ry doses, in the event of a OBA, to e eed values given in he licensing basis. SR 3.6.1.1 leak e rate requir ents comply with 10 CfR 50, Appendix {Ref. 1), as modif' d by approved exemptions. " B 3.6-6

{continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • Containment Air Locks c Dua 1 j / 1) B 3.6.2 . B 3.6 CONTAINMENT SYSTEMS B 3.6.2 Containment Air Locks /(Atm(jSpheric and 9Ual)j BASES BACKGROUND T .....,.,.

,,:.,.. loe..\<.S P"'"'": c..U::L5°>

..,,.\-,,

CEOG STS

  • Containment air locks form part of the containment pressure boundary and provide a means for personnel access during all MODES of operation.

Each air lock is nominally a right circular cylinder, [iQ2fiJ Ii n :a i ametir I with a door at each end. The doors are ln er oc e o preven s1mu aneous opening.

  • During periods when containment is not required to be OPERABLE, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary.

Each air lock door has been designed and tested to certify its ability to withstand a pressure in excess of the maximum expected pressure following a Design Basis Accident (OBA) in containmeot. As such, closure of a single door supports containment OPERABILITY. Each of the doors contains double . II'?\ gasketed seals and local l-e91tll!e nte testing capability to ensure pressure integrity. To effect a leak tight seal, the air lock design uses pressure seated doors (i.e., an increase in containment internal pressure results in increased sealing force on each door). Each pers nel air lock is rovided with lim" switches on both door: that provide c trol room indica on of door fs\ position Additionally, ontrol room indi tion is provided<...;,) to aler the operator w never an air loc door interlock mechan sm is defeated. The containment air locks form part of the containment pressure boundary. As such, air lock integrity and leak tightness is essential for maintaining the containment leakage rate within limit in the event of a OBA. Not *maintaining air lock integrity or leak tightness may in a leakage rate in excess of that assumed in the @Bl!} \3:..; safety analysis. . cf\"e..-'£) B 3.6-11 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • Containment Air Locks /(Atm/Ospheri( and Dut.11-/£ 8 3 .. *
  • BASES (contfn..ied)

APPLICABLE SAFETY ANALYSES § .----------

For a l..tx..Al 02.lcuiJ:.+ul IS s 3 for°"" msl.6, Ca.lculc;.M ffa.x,"'u"'

peo. I( C Cn-k:c, 1 n l'1lclrl' /f t1J4Vrt. IS a.pP, .. (Z 5'-1 /.ld'4Jr.,IJ":" 1 to U'ISVf<. /J (! Hlcir.1'1'1 Q..ri' hou rvJ all D S'Prs 1 TYPt & (Q,ft_ N't1f\\ IS af or a.ic:t#< +nc. f "t SS" fs1d. LCO CEOG STS

  • For dual con ainment, the OBAs that result* in a r lease of radioactive material within c ntainment are a LO A, an MSLB, and a CEA jection accident ef. 2). In the a alysis of each of t se accidents, it is assumed that co ainment is OPERABLE uch that release f fission product to the environ nt is y the rate of con inment leakage The containment was* designed with n allowable leakag rate of [0.50]% f containment air eight per day (Ref. ). This leakage rate is defined in 10 CFR 50, Appe ix J (Ref. 1), a L.: the maximum lowable con inment leakage r e at the calculat ij maximum peak con ainment pressure Pe) of (42.3] psi , which results from th limiting OBA, w ch is a 75% RTP H 8 (Ref. 2). This a owable leakage rte forms the bast for-the acceptance iteria imposed o the SRs associate with the air lock.

.. .. '1 I r;:i Each containment air lock form art of the ontainment l:::!.J pressure boundary. As part of containment, the air lock safety function is related to control of the containment leakage rate resulting from a OBA. Thus, each air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event. '1 B (continued) Rev 1, 04/07/95 Revised 05/31/99

  • * *
  • BASES LCO (continued)

APPLICABILITY ACTIONS Containment Air Locksl(Atli}6spheric Jnd B 3.6.2 Each air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of containment does not exist when containment is required to be OPERABLE. Closure of a in each air lock is sufficient provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is being used for normal entry in.to from . j ([;) containment.

  • . @ In MODES 1, 2, 3, and 4, a OBA could cause a release of radi-0active material to containment.

In MODES 5 and *6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment air locks are not required in MOOE 5 to prevent leakage of radioactive material from containment. The requirements for the containment air locks during MOOE 6 are addressed in LCO 3.9.3, "Containment Penetrations.* three. The. n...+e I The ACTIONS are modified by @J<ot ows entry and \!;.) exit to perform repairs on the affected air lock component. If the outer door is inoperable, then it may be easily accessed for most repairs. It is preferred that the air lock be accessed from inside primary containment by entering through the other OPERABLE air lock. However, if this is not practicable, or if repairs on either door must be performed from the barrel side of the then it is permissible to enter the air lock through the OPERABLE door, (fh.:i) there is a short time during which the . containment boundary is not intact (during access through \.:::J CEOG STS the OPERABLE door). The ability to open the OPERABLE door, even if it means the containment boundary is temporarily not intact, is acceptable because of the low probability of an event that could pressurize the containment during the short time in which the OPERABLE door is expected to be open. After each entry and exit, the OPERABLE door must be eve""'; : s e..<;; 1---_.... -h c..-. \ IU;-h-, AC..Tlc/J S, ,, B 3.6-13 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • BASES ACTIONS (continued)

CEOG STS Containment Air Locks and OualTI-@ B 3.6.2 irrmediately closed. If ALARA conditions permit, entry and exit should be via an OPERABLE air lock. A second Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each air lock. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable air lock. Complying with the Required Actions may allow for continued operation, and a subsequent inoperable air lock is governed by subsequent Condition entry and application of associated Required Actions. A third Note has been included that requires entry into the applicable Conditions and Required Actions of LCO 3.6.1, "Containment,* when leakage results in exceeding the overall containment leakage limit. A.1. A.2. and A.3 With one air lock door inoperable in one or more containment air locks, the OPERABLE door must be verified closed (Required Action A.I) in each affected containment air lock. This ensures that a leak tight containment barrier is maintained by the use of an OPERABLE air lock door. This action must be completed within 1 hour. This specified time period is consistent with the ACTIONS of LCO 3.6.1, which requires containment be restored to OPERABLE status within 1 hour. In addition, the affected air lock penetration must be by locking closed an OPERABLE air lock door within the 24 hour Completion Time. The 24 hour Completion Time is considered reasonable for locking the OPERABLE air lock door, considering the OPERABLE door of the affected air lock is being maintained closed. Required Action A.3 verifies that an air lock with an inoperable door has been isolated by the use of a locked and closed OPERABLE air lock door. This ensures that an acceptable containment leakage boundary is maintained. The Completion Time of once per 31 days is based on engineering judgment and is considered adequate in view of the low likelihood of a locked door being mispositioned and other administrative controls. Required Action A.3 is modified by a Note that applies to air lock doors located in high radiation areas and allows these doors to be verified locked B 3.6-14 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • . BASES ACTIONS
  • CEOG STS *
  • Containment Air Locks!(Atl!Jbspherih and B 3 .. 2 A.I. A.2. and A.3 (continued)

I closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of the door, once it has been verified to be in the proper position, is small. The Required Actions have been modified by two Notes. Note I ensures that only the Required Actions and associated Completion Times of Condition C are required if both doors in the same air lock are inoperable. With both doors in the same air lock inoperable, an OPERABLE door is not ava._!_i.J.Jl to be closed. Actions C.l and C.2 are the appropriate remed ia 1 actions. The ote oes not.affect tracking the Completion Time from the lnitial entry into Condition A; only the requirement to comply with the Required Actions. Note 2 allows use of the air lock for

  • entry and exit for 7 days under administrative controls H both air locks havP. an inoperable door. This 7
  • restriction begins when the second air leek is discovered inoperable.

Containment entry may be required to perform Technical Specifications (TS) Surveillances and Required Actions, as well as other activities on equipment inside containment that are required by TS or activities on equipment that support TS-required equipment. Note is not intended to preclude performing other activities (i.e., non-TS-required activities) if the containment was entered, using the inoperable air lock, to perform an allowed activity listed above. This allowance is acceptable due to* the low probability of an event that pressurize the containment during the short time that the OPERABLE door is expected to be open. B.l. B.2. and 8.3 With an air lock interlock mechanism inoperable in one or more air locks, the Required Actions and associ.ated Completion Times are consistent with those specified in Condition A * . The Required Actions have been modified by two Notes.

  • Note 1 ensures that only the Required Actions and associated Completion Times of Condition C are required if both doors " B 3.6-15 (continued)

Rev 1, 04/07/95 Revised 05/31/99

  • *
  • BASES ACTIONS J:.f-

\ l-o ... + .. lcc.t....;c. v-... h S +k.. I -H--L ,,\: +i.......f Leo ,,...,,,.., +-be. ...,;-tt.._. CEOG STS Containment Air Lock.sj(AtJhospher}c and D{al}-/q' B L6.Z J) B.l. B.2. and B.3 (continued) in the same air lock are inoperable. With both doors in the same air lock inoperable, an OPERABLE door is not available to be closed. Required Actions C.l and C.2 are the appropriate remedial actions. Note 2 allows entry into and exit from containment under the control of a dedicated individual stationed at the air lock to ensure that only one door is opened at a time (i.e., the individual performs the function of the interlock). Required Action B.3 is modified by a Note that applies to air lock doors located in high radiation areas and allows these doors to be verified locked closed by use of administrative means. Allowing verification by means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of the door, once it has been verified to be in the proper position, is small. C. I. C. 2. and C. 3 With one or more air locks inoperable for reasons other than those described in Condition A or B, Required Action C.l requires action to be initiated irrmediately to evaluate previous combined leakage rates using current air lock test results ..... An evaluation is acceptable since it is overly conservative to irrmediately declare the containment inoperable if both doors in an air lock have failed a seal test or if the overall air lock leakage is not within limits. In many instances (e.g., only one seal per door has failed), containment remains OPERABLE, yet only 1 hour (per LCO 3.6.1) would be provided to restore the air lock door to OPERABLE status prior to requiring a plant shutdown. In addition, even with both doors failing the seal test, the overall containment leakage rate can still be within limits. Required Action C.2 requires that one door in the affected containment air lock must be verified.to be closed. This action must be completed within the 1 hour Completion Time. This specified time period is consistent with the ACTIONS of LCO 3.6.1, which requires that containment be restored to OPERABLE status within 1 hour. ,, B 3.6-16 (continued} Rev 1, 04/07/95 Revised 05/31/99

  • *
  • BASES ACTIONS SURVEILLANCE REQUIREMENTS Containment Air Locks /<AtmoWheric aJld B 3. 6. -C'.l. C.2. and C.3 (continued)

Additionally, the affected air lock(s) must be restored to OPERABLE status within the 24 hour Completion Time. The specified time period is considered reasonable for restoring an inoperable air lock to OPERABLE status, assuming that at least one door is maintained closed in each affected air lock. D.l and D.2 If the inoperable containment air lock cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SR 3.6.2.1 Maintaining 1 cks OPERABLE requires compliance with the' le ate test requirements of 10 CFR 50,

  • 1 , as modified by approved exempt1ons.

n1s SR re ec s the leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). The acceptance criteria were established during initial air lock and containment OPERABILITY testing. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall con ainmen e . The Frequency is required by 1<0 .. fu+s, .!?-Hi-c.r fh4.,.. f'rJcf\1u.L. a.irla* o.r< o..t ? SS * (OP+1ou A) Aipendix as modified by approved exemptions. Thus, S 3.0.z hich allows Frequency extensions} does not apply. I© /1Nsi:RT] . fwt The SR has been modified Notes. Note 1 states that : an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a OBA. Note 2 has been added to this SR requiring the results to be evaluated against the acceptance criteria of CEOG STS * (continued) ,, B 3.6-17 Rev 1, 04/07 /95 Q.rntnd1"1e.tits f6 -the. +c.c.""lto. l (e.v1s(cd -+ht d.c.cet+ctnc:.c. Cr1+c.r10. .ror ovue.11* Ty;t 51 c.. J1mit.S a..nd l'ie.w Q.Ctefiancc. @) for fht a.1r Jou< ofooa arid a.1r loc.( . dears {). R *

  • d ev1se 05/31/99
  • *
  • INSERT 1 [From previous paragraph]

... Thus, SR 3.0.2 (which allows Frequency extensions) does not apply. Two exemptions to the requirements of 10 CFR 50, Appendix J have been granted for the containment air locks. The exemption granted by letter dated December 6, 1989 provides partial relief from the requirement of Paragraph III.D.2.(b)(ii) to leak test, at or above the calculated design basis accident peak containment pressure (Pa), containment air locks which were opened during a period when containment integrity was not required. This exemption permits the substitution of a between-the-seal leak test at a reduced pressure, but not less than 10 psig, provided that no maintenance, modification, or other activity has been performed which could affect the sealing capability of the air locks. The exemption granted by letter dated September 30, 1997 applies only to the emergency escape air lock and provides partial relief from the requirement of Paragraph III.D.2.(b)(ii) and Paragraph IIl.D.2.(b)(iii). The requirement of Paragraph 111.D.2.(b)(ii) is discussed above. Paragraph 111.D.2.(b)(iii) requires air locks opened during periods when containment integrity is required to undergo a full air lock pressure test within 3 days after being opened. This exemption permits the performance of a door seal contact verification check in lieu of the final pressure test following the opening of the emergency escape air lock doors for post-test restoration or seal adjustment. This exemption does not affect compliance with .the requirement to perform a full pressure air lock test at 6 month intervals, or the requirement to perform a full pressure air lock test within 72 hours of opening either air lock door during periods when containment integrity is required. \\ B 3.6-17 Revised 05/31/99

  • *
  • BASES SURVEILLANCE REQ IREMENTS Containment Air Locks I (AtJflospheri d and oui!{i) k9J B 3.6.2 *SR 3.6.2.1 (continued) r-----SR 3.6.lat3 This ensures that air lock leakage is properly accounted for in determining the overall containment leakage Il\JSl.I<

T fl;:-----* rate. SR 3.6.

2.2 REFERENCES

Ii. I io CFR sj, Appendif J.1 * & @) FSAR, Section 5f' J:@ I I CD I© e.ve..-"\ 16 ""o ... \"'Ni.. l B fY\<:>"'+h ........... '1 ir o"' +.o ....... ....... e:l\c. ... IA. ..... Je ... c..o ..... k.+ °"ffl1 J ... r'""I f O'-+ ... 1 t;..,,,.,\ \oss c.o,J*....::"' ...... e"'T. Ci'GlZABI UT'# s ...........

\\o.*,,,.

.. 1.14!!""- ..,..&.c:ih, ... cl-po,.je.r, ihc. moo, fh Frr.qVr.n(.y -m.. lt\icJo<.t: .lJJ4f1ficd "' "'-u-(} 'f'tt IC.. CEOG STS \\ B 3.6-18 Rev 1, 04/07/95 Revised 05/31/99 B

  • *
  • INSERT 2 Note 3 clarifies that iterative pressure testing of the emergency escape air lock is not required when the air lock doors are opened solely for the purpose of strongback removal and performance of the seal contact check. Note 4 ensures that air lock testing, other than door seal testing, is performed at a 55 psig consistent with other type Band C tests . B 3.6-18 Revised 05/31/99
  • *
  • Containment Isolation Valves j(Atl1)6spherfc and oAal)
  • B 3.6.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.3 Containment Isolation Valves Dufl) BASES BACKGROUND J<<v."<.iD The containment isolation part of the i s olA.ti* ".) ye.-e+.-c-+& 0 -;:\.ow containment pressure boundary and provide a means pene ra ns no serving c1 en cons uence 1m1 1ng systems to be provided w* h two isola on barrier. that are close on an automatic
  • solation si 1. These isolation devices are elther passive or active (automatic)..

Manual valves, de-activated automatic valves secured 'in their closed position (including check valves with flow through the valve secured), blind flanges, and closed systems are considered passive devices. Check valves, or other automatic valves designed to close without operator action following an accident, are considered active devices. Two barriers in series are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that

_
; I

.,...,,t-,:..l \ ... : .... ,....e"' ' I . *exceeds limits assumed in the safety analysis. One of these barriers may be a closed system. ........ u ... l.A.S \5.,.\ .... s: 0 .. ..Qs. curs upon recei t of a nal or a @ l ,_.,c.-----=--.----. n1pr_.e ... l4ll&'"**e signal. The co ainme 1 o 1 signa"k;D (o..J: ... t-0 .... lcl-4R}f 'C'i'O'Se<<> automatic containment isolation valves in fluid . penetrations not required for operation of Engineered Safety Feature systems in order to revent leakage of radioactive material. U n actuat10 o sa e y 1n c 1on, aut mat1c 1 (2) containment solation v ves also iso te system not re ui"red f containme t or RCS heat emoval. Other @ penetrations are 1solated y the use of valves n the closed position or blind flanges. As a result, the containment isolation valves (and blind flanges) help ensure that the containment atmosphere will be isolated in the event of a r-:::-;sn release of radioactive material to containment atmosphere , c .... 1 .. _i from a Design Basis Accident (OBA). ! t_'t)

  • requirements for containment isolation lf.0 vaives+help ensure that containment is isolated within the time limits assumed in the safety analysis.

Therefore, the CEOG STS OPERABILITY re uirements provide assurance that the containment u ct1 n assumed in the accident analysis will be ne . lea.li::.'.\e I ot c-A '1 B 3.6-19 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • Containment Isolation Valves /(Atm#pheri(/and . 8 3.6 . BASES BACKGROUND (continued) ..... o"°':)k ...j.k_ 1'2. i" 0..*..-\'"00"""'

Si... pp I., vo..\,A.S J1,.....o, f' .._ ca_ e.x: ka.-s +-1.)e..\v-t... """""' 0.. '('ao--. $IA r f ,, \l<-\ 0,:r.e..

  • APPLICABLE The containment isolation valve LCO was derived from the SAFETY ANALYSES assumptions related to minimizing the loss of reactor coolant inventory and establishing the containment boundary during major accidents.

As part of the containment boundary, containment isolation valve OPERABILITY supports leak tightness of the containment. Therefore, the safety analysis of an.y event requiring isolation of containment is applicable to this LCO. CEOG STS The DBAs that res'ult in a release of radioactive material within containment are a loss of coolant accident (LOCA), a " B 3.6-20 (continued) Rev 1, 04/07/95 Revised 05/31/99 16)

  • * * .BASES APPLICABLE SAFETY ANALYSES (continued)

Containment Isolation Valves \(Atm,lispheric B i a lelemjnt assimblYl{3 \flt) eJect1on acc1dent. In the analys1s for each of these *-* accidents, it is assumed that containment isolation valves are either closed or function to close within the required isolation time following event initiation. This ensures that potential paths to the environment through containment isolation valves (including containment purge valve!l__!re minimized. The safety analysis assumes that theCiiiDiii!\ are closed at event initiation. _msJ The OBA_ analysis assumes that, within after i en 1so a 1 n o e on alnm 1s com e e and leaka e terminated except for the design leakage--rate, La. The conta men 1so a 1on a response 1 of 60 seconds""'l includes signal delay, di el generator st tup (for loss of offsite ower), and cont nment isolation alve stroke times. /@ B ... .,.,,.t 11.;,..cJ... \:nes e er sources mo or CEOG STS

  • spring c ose esigne o rec u e co111110n m e f om disablin both valve on a purge line. . rei<k.-st'

... ..,J. 12. ;,....cJ.. o,.;,,.. t'OO"'"' Su.pf'"' I I £1 The purgeivalves may be unable to close in the environment following a LOCA. Therefore, each of the purge valves is I liJ'. (f""c...keaj required to remain isUled\ closed during MODES 1, 2, 3, * & and 4. In this case, the single failure criterion

  • r:--applicable to the containment purge valves due to ailure in (,.'.:!, the control circuit associated with each valve. Again, the purge system valve design precludes a single failure from -compromising the containment boundary as long as the system is operated in accordance with the subject LCO. e minipurge v ves are capa o c osing un r accident (j) c.ond it ions Therefore, t y are a 11 owed be open for > limited iods during p er operation.

B 3.6-21 satisfy Criterion 3.5ti I@ I toe.FR r (continued) Rev 1, 04/07/95 Revised 05/31/99

  • * *
  • Containmei1t Isolation ValvesITA@Osphij'ic BASES (continued)

LCO APPLICABILITY CEOG STS I The normally closed isola va ves are cons1 ered OPERABLE when manual valves are closed, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact. These passive isolation valves or devices are those listed in Reference . . This LCO provides assurance that the containment isolation valves and purge valves will perform their designed safety ,G"i fu11cti.ons to minimize the loss of oce;a(tor) coolant inventory lw and establish the containment oundary during accidents. In MODES l, 2, 3, and 4, a OBA could cause a release of radioactive material to containment. In HODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these HODES. Therefore, the containment isolation valves are not required to be OPERABLE in HOOE 5. The requirements for containment isolation valves during MODE 6 are addressed in LCO 3.9.3, "Containment Penetrations." \ \ B 3.6-22 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • **
  • BASES (continued)

Containment Isolation Valves\(Almosphefic and B .6.3 . o ... ,. Noe;, G) isJd-d The ACTIONS are modified b ene ration flow z_ aths, except for. inch purge valve penetration flow I ry tent 1 y under admi n i strati ve ACTIONS l2;..._c...k controls. These administrative controls consist of C:..r roo"""'-s-....ul-, stationing a dedicated operator at the valve controls, who is in continuous co11111unication with the control room. In CE> fc-'-t 9 p u....l"'-5. o.......,l. this way, the penetration can be rapidly isolated when a need for containment isolation is indicated. ue to the f---7 size o t e-confiliiineiit--*ur-e*1*rne-* ne ra 1on and the fact that those penetrations exhaust directly from the l.:__J --ill'c...k ,,.: . .,.. voo ........ sv..rr'1 'i"'C.'\ l,A-A \ \ * ..j\,..q_ ertv;,...,,,_ ..... e.:i-10 c.. '"' "\ (I.. LoC-1"\ containment atmosphere to the environment, these valves may not be opened under administrative controls. A second Note has been added to provide clarification that, for this LCO, sepa ate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable containment isolation valve. Complying with the Required Actions may allow for continued operation, and subsequent inoperable containment isolation valves are governed by subsequent. Condition entry and application of associated Required Actions. The ACTIONS are further modified by a third Note, which ensures that appropriate remedial actions are taken, if necessary, if the affected systems are rendered inoperable by an inoperable containment isolation valve. A fourth Note has been added that requires entry into the applicable Conditions and Required Actions of LCO 3.6.l when leakage results in exceeding the overall containment leakage limit. A.l and A.2 6 ... t:;:::...._\ 0 In the event one containment isolation valve \.!:) enetration flow paths is inoperable , xcept in one orlmore for a 1 ve or c...*<" .,._o""'t----""711 ea e an le 1 1n ass ea a s ... "'""'\/'C... im't* the affected penetration flow path must be isolated . ...i\.-;c.\....

s e met ad of isolation must. include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Is.olation barriers that meet this criterion are a closed and de-activated automatic containment isolation valve, a closed manual valve, a blind CEOG STS B 3.6-23 * (continued)

Rev 1, 04/07/95 Revised 05/31/99 @)

  • BASES ACTIONS
  • CEOG STS *
  • Containment Isolation Valves /(Atm?spheric

/nd Dual)LlR"' 8 3.6.3 1 U *A.I and A.2 (continued) flange, and a check valve with flow through the valve secured. For penetrations isolated in accordance with Required Action A.I, the device used to isolate the penetration should be the closest available one to containment. Required Action A.I must be completed within the 4 hour Completion Time. The 4 hour Completion Time is reasonable, considering the time required to isolate the penetration and the relative importance of supporting containment OPERABILITY during MODES 1, 2, 3, and 4. For affected penetration flow paths that cannot be restored to OPERABLE status within the 4 hour Completion Time and that have been isolated in accordanca with Required Action A.1, the affected penetration flow paths must be verified to be isolated on a periodic basis. This is necessary to ensure that containment penetrations required to be isolated following an accident and no longer capable of being automatically isolated will be in the isolation position should an event occur. This Required Action does not require any testing or device manipulation.

  • Rather, it involves verification, through a system*walkdown, that those isolation devices outside containment and capable of being mispositioned are in the correct position.

The Completion Time of "once per 31 days for isolation devices outside containment* is appropriate considering the fact that the devices are operated under administrative controls and the probability of their misalignment is low. For the isolation devices inside containment, the time period specified as "prio.r to entering MODE 4 from MODE 5 if not pe:-formed within the previous 92 days* is based on engineering judgment and is considered reasonable in view of the inaccessibility of the isolation devices and other administrative controls that will ensure that isolation device misalignment is an unlikely possibility. Condition A has been modified by a Note that this Condition is only applicable to those penetration flow paths with two containment isolation valves. For penetration flow paths with only one containment isolation valve and a closed system, Condition C provides appropriate actions. Required Action A.2 is modified by a Note that applies to isolation devices located in high radiation areas and allows these devices to be verified closed by use of administrative B 3.6-24 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • * * . BASES ACTIONS Containment Isolation Valves ICAtmd'spheric alii Dual) t@ r B 3.6.3 A.1 and A.2 (continued) means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted.

Therefore, the probability of misalignment of these devices, once they have been verified to be in the proper position, is small. Ll With two containment isolation

  • valves in one or more.) enetration flow paths inoperable xcept for

... .. -....... ----/ ea e u1 ass ea a e no w1 1n rfO\ i "t 1 the af ecte penetrat1on flow path must e isolated 5..v.tr' within 1 hour. The method of isolation must include the use

  • > of at least one isolation barrier that cannot be adversely 1 I affected by a single active failure. Isolation barriers loc."'-f!:*

th.at meet this criterion are a closed and de-activated au.tomatic valve, a closed manual valve, and a b.l ind flange. The 1 Time is consistent with the ACTIONS of LCO the event the affected penetration is isolated in accordance with Required Action B.l, the CEOG STS

  • affected penetration must be verified to be isolated on a periodic basis per Required Action A.2, which remains in effect. This periodic verification is necessary to assure leak tightness of containment and that penetrations
  • requiring following an accident are isolated.

The Completion Time of once per 31 days for verifying each affected penetration flow path is isolated is appropriate considering the fact that the valves are operated ur.der administrative controls and the probability of their misalignment is low. Condition B is. modified by a Note indicating this Condition is only applicable to penetration flow paths with two containment isolation valves. Condition A of this LCO addresses the condition of one containment isolation valve inoperable in this type of penetration flow path. C.l and C.2 With one or more penetration flow paths with one containment isolation valve inoperable, the inoperable valve must be restored to OPERABLE status or the affected penetration fl ow " B 3 .6-25 (continued) Rev 1, 04/07/95 Revised 05/31/99 ,---* * * -BASES ACTIONS Containment Isolation Valves i(Atiij?spheriJi' and Dfal)\-@ B 3.6.3 C.l and C.2 (continued} path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration. Required Action C.1 must be completed @ 7-z_ w1thin the lfAjz-hour Completion Time. The specified time I \5 _ period is reasonable, considering the relative stability of [). \ t I . - C-1"'\cA c ... ,.,,.. I -\,,J*io-.. foo...-.. s .... f'P \ .,_, c.>A-.... t-e.l +o c.\ ... se o.. L-oc..4 .... +.., 6e \...c.bJ c \ox.,\ . L.-\ " "'"-tt<' m(Y('... -H.v.se.. . r" +-\oc..\:..eJ v<--\ ...... 'S c:.l"'tf!J +\....L 1 \:.,,. .Je.\.J< .\., be.. _ I ore""'.l., On<. hwr IS A'bl1"t.J fO I C.IOl<.&l i'f\( 1Al11t.S. 1 7M.. / rnu1 "lime. prr;tdCJ i o. to 7hc.. I ff'o'*""' - 1N1+U. i Df fYla.ittftt..111113 I -fnd' \Alll(S C.\rst.d. I CEOG STS

  • the closed system (hence, reliability}

to act as a penetration isolation boundary and the relative importance of supporting containment OPERABILITY during. MODES 1, 2, 3, and 4. In the event the affected penetration is isolated in accordance with Required Action C.l, the affected penetration flow path must be verified to be isolated on a periodic basis. This is necessary to assure leak tightness of containment and that containment penetrations requiring following an accident are isolated. The Completion Time of once per 31 days for verifying that each affected penetration flow path is isolated is appropriate considering the valves are operated under administrative controls and the probability of their misalignment is low. Condition C is modified by a Note indicating that this Condition is only applicable to those penetration flow paths with only one containment isolation valve and a closed system. This Note is necessary since this Condition is written to specifically address those penetration flow paths in a closed system. Required Action C.2 is modified by a Note that applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the A"i\ probability jflthese once they have\& been verified o be in the proper posit10 , is small. [lhl With the condary containmen within li it, the assumption 'I B 3.6-26 bypass leakage ra of the safety anal (continued} Rev 1, 04/07/95 Revised 05/31/99

  • BASES ACTIONS * @ CEOG STS *
  • Containment Isolation Valves i(At}nosphejfic d3) . 6 .3 met. Therefo , the leakage must limit within hours. Restoratio can be accomplish by isolating t penetration(s) tha caused the limit o be f'i"'.. exceeded b use of one closed ad de-activated aut atic t)
  • valve, cl ed manual valve, or blind flange. Whe a <..:::,.r penetrat*

n is isolated, the eakage rate for th isolated penetra on is assumed to b the actual pathway leakage throug the isolation devi . If two isolatio devices are

  • used isolate the penetr, tion, the leakage ate is assumed to b the lesser actual thway leakage of e two devices. The hour Completion T" e is reasonable c sidering the ti required to resto the leakage by i lating the p etration(s) anc th relative importan of secondary ntainment bypass 1 kage to the overa containment unction. L 1-. E. 2. and E. 3 In the event one or more ontainment purge valves* in one more penetration flow p hs are not within the purge val leakage limits, lve leakage must be restored to within limits, or.the ffected penetration must be iso ted . The method of isolat* n must be by the use of at leas one isolation barrier t t cannot be adversely affected a single active fail e. Isolation barriers that mee this criteri9n are a [ osed and de-activated automatic alve with resilient s ls, a clcsed manual valve with silient seals, 'or a bli flange]. A valve with r ilient seals utilized o satisfy Required E.l m st have been demonstrated meet the leakage requirements SR 3.6.3.6. The specifie Completion .Time is reasonable, onsidering that one co ainment purge valve remains clo ed so that a gross brea of containment does not exist. In accor nee with Required Action t is penetration flow pa must be verified to be isolate on a periodic basis. The periodic verification is n essary to ensure that ontainment penetrations require to be isolated foll wing an accident, which are no. nger capable of being aut matically isolated, will be in e isolation position sh uld an event occur. This Requi d Action does not r quire any testing or valve mani lation. Rather, it nvolves verification, through a ystem walkdown, that those ,, B 3.6-27 (continued)

Rev 1, 04/07/95 Revised 05/31/99

  • * *
  • BASES ACTIONS @ SURVEILLANCE .REQUIREMENTS CEOG STS Containment Isolation Valves . 6 .3 E .1. E. 2. and £. 3 isolation devices outside ontainment capable of being mispositioned are in the orrect position.

For the isolation devices insid containment, the time period specified as nprior to ntering MODE 4 from MODE 5 if perfonned within the p evious 92 days" is based on engineering judgment nd is considered reasonable in the inaccessibility f the isolation devices and ot r administrative cont ls that will ensure that isol device misalignmen is an unlikely possibility. For the containm t purge valve with resilient s isolated in ace dance with Required Action E.l must be perfo d at least once every day . This assures that gradation of the resilient seal is detected and confinns hat the leakage rate of the co ainment purge valve does t increase during the time the enetration is isolated. e nonnal Frequency for SR 3.6 .. 6, 184 days, is based on a NRC initiative, Generic Issue -20 (Ref. Since mor reliance is placed on a single alve while in this Con ition, it is prudent to perfonn he SR more often. Therefo e, a of once per [92) ays was chosen and has be shown to be based operating exper* nee

  • E. E (!11 If the Required Actions and associated Completion Times are not met, the plant must be brought to a MODE which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. Each 42] inch containmen purge valve is re ired to be veri ied sealed closed a 31 day intervals.

This Sur eillance is designe to ensure that a oss breach of co ainment is not cau d by an inadverte spurious r----4----------1---------'--------*---- B 3.6-28 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • BASES SURVEILLANCE REQUIREMENTS Containment Isolation Valves((Atjll'ospheriC/and I B 3.6.3 SR 3.6.3.1 opening of a ontainment purge v ve. Detailed ana ysis of the purge va ves failed to cone sively demonstrat their ability to ose during a LOCA in time to limit o fsite doses. Th refore, these valv s are required to e in the sealed cl sed position durin MODES,l, 2, 3, an 4. A J?l containm t purge valve tha is sealed closed st have motive wer to the valve erator removed. is can be
  • accomp shed by de-energi ng the source of ectric power or by emoving the air s ply to the valve erator. In *this the t "sealed" has no connotation of lea tightness.

The F quency is a resul of an NRC ini iative, Generic I ue 8-24 (Ref. 4), elated to c tainment purge va e use during unit operations. This SR i not required to met while in Con ition E of this LCO. his is reasonable ince the penetrat"on flow path would be isolated. SR 3 .6.3 .iZQ) \ {b) . ) or c.\oS"<lJ b, .. J* net I oc..lc. ei:{ vc..lue.s "'"'"\ lA.>'\. .... µ h c..I !:> se. *,,.... e"";, ...... ,,.,,...e.;i\ .h:.ll-..:j c.. LoC..:\. o+. -tk. vo..\_,e_5 .\--a 1 e""'A:-" MonES 1 1 z..1 i, c..,,..coJ. G."' J. ...,,,,,..\- \oc..\:,.J 1 "1ec. l-4!J, o.- ,,.,: se_ set. ... ,,. CEOG STS

  • lS This SR requires verification that each containment
  • isolation valve and blind flange located outside containmen and required to be closed during accident con itions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the B 3.6-29 (continued)

Rev I, 04/07/95 Revised 05/31/99

  • *
  • BASES SURV[ILLANCE REQUIREMENTS se_

,.,.t °'fPI'\ h vd.,.e..S

  1. .c..+ c...-< l .. sec.le.:l, .,..,.

... "";sli! S" ... c.,_r".l. ""' c..\ose.l pos:4".*.,.,... 1 ... c...e. ._iye.. +,,. l.e.. l!"-c..o,,.,..e.,.+- ros.**ho .......... .. :}, or s.ec.VW":"d* 

"'"Gl ..., ... + 1., ... "-...j 1 sei:..le.l 1 "'"' ot\..e,,.

.... ;re ..... , ... J, CEOG STS

  • Containment Isolation Valves j(Atmoi})heric .SR I@ containment boundary is within design limits. Th i*s SR does not require any testing or valve manipulation.

Rather, it involves verification, through a system walkdown, that those containment isolation valves outside containment and capable of being mispositioned are in the correct position. Since verification of valve position for containment isolation valves outside containment is relatively easy, the 31 day Frequency is based on engineering judgment and was chosen to provide added assurance of *the correct positions. Containment isolation valves that are open under administrative controls are not required to meet the SR during the time the valves are open. The Note applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, for ALARA reasons.1<3d Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, _is small

  • SR 3.6.3.1(!)

This SR requires verification that each containment isolation anu valve and blind flange located inside ccntainme.!!!.tand required _to be closed during accident closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the containment boundary is within design limits. For containment isolation valves inside containment, the

  • Frequency of *prior to entering MOOE 4 from HOOE 5 if not performed within the previous 92 days" is appropriate, since these containment isolation valves are .operated under ' administrative controls and the probability of their misalignment is low. Containment isolation valves that are open under administrative controls are not required to meet the SR during the time that are open. The Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification by ', B 3.6-30 (continued)

Rev 1, 04/07/95 Revised 05/31/99

  • *
  • BASES SURVEILLANCE REQUIREMENTS -fre l.CtiLICS Q.((!

C.IQSJ:.d (SR 3.cl. lk.rl +'l!.S #it. VO.\ V<. 1 S ictJ.uJ e.JoscJJ , Containment Isolation Valves /(Atil¢'spheri(/and DualfH@ B 3.6.3' .SR 3.6.3.continued) \ (D administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified t6 be in their proper position, is small .. SR I_[} Verifying that the isolation time of each (p,owir operfted andj I 0) automatic.,..containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analysis * .sfThe l

  • isolation time and Frequency of this SR are in accordance

([} with the Inservice. Testing Program pr >>2 da;fS\t-' SR 3.6.3.I \"'Z..;,_cJ..,..o..:..-r<>o-.S ... fPl't For containment urge valves with resilient seals, \(0 additional testing the. requirements of 10 *CFR 50, Append1x J (Ref. f1JJJ1s requ1red to ensure \ Operating experience has demonstrated that . li1\ tnis type of seal has the potential to degrade in a shorter '2) time period than do other seal types *. Based on this observation and the importance of maintaining this penetration leak tight (due to the direct path between containment and the environment), a Frequency of 184 days was established as part of the NRC resolution of Generic 1 ___________ "Containment leakage Due to Seal Deterioration" a.s s Pec1Y 1n +he. dJ &hJD.4'1on tor A ...... "..J "'"'" No. 9 D l.:o -tt--(, i+t ,'* . CEOG * *' B 3.6-31 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • * * * .. BASES SURVEILLANCE REQUIREMENTS (continued)

Containment Isolation Valves i(Atmiripheric .6 . .SR 3.6.3.<P© I© Automatic containment isolation valves close on a containment isolation* signal to prevent leakage of radioactive material from containment following a OBA. This SR ensures each automatic containment isolation valve will a)'i a..ctl.L4 or s; M u..\a..-+ et\ actuate to its isolation position on* k(\n --actuation si nal This Surveillance is notrequired for va ves that are locked, sealed, or otherwise secured in the required position under administrative controls. The CEOG STS * '.f18t month Frequency was developed considering it is rudent Im that this SR be performed only during a u age, since isolation of penetrations would eliminate cooling water flow and disrupt normal operation of many critical components. Operating experience has shown that these components usually . pass this SR when performed on the .tiar month Frequency. t GJ Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. SR 3.6.3.8 [Reviewer's Not .. This SR is only re ired for those units with resilien seal purge valves al owed to be open during [HOOE 1, 2, , or 4] and having bl eking devices on the valves that are not permanently

  • stalled. . Verifyin that each [42] inch ontainment purge valve i blocked o restrict opening [50]% is required to nsure 7 ;*. that t e valves can close u er OBA conditions withi the ). time assumed in the analy es of References 2 3 If a LOC occurs, the purge v ves must close to maint n co ainment leakage wit n the valves assumed in he a cident analysis.

At ther times when purge v ves are equired to be capabl of closing (e.g., duri movement of irradiated fuel ass blies), pressurization ncerns are not present, thus the rge valves can be fully open. The [18] month Freque y is appropriate becaus the blocking devices are typi ally removed only durin a refueling outage. sures that the combined eakage rate of all containment bypass le age paths is less than B 3.6-32 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • * * . BASES SURVEILLANCE REQUIREMENTS REFERENCES CEOG STS Containment Isolation Valves i(At:£ospheric/ind B 3.6.3 SR 3.6.3.9 equal to the specif"ed leakage ovides assurance that the assumptions in the saf y analysis are met. The leakag rate of each bypass le age path is assumed to be t maximum pathway leaka e (leakage through the worse of t e two isolation valves) unless the penetration i isolated by use of on closed and de-activate automatic valve, close manual valve, or blind flange. I this case, the leakage rate of the isolated f?:) bypass le age path is assumed t be the actual pathway leakage rough the isolation d ice. If both isolatio valves n the penetration are osed, the actual leak e rate
  • the lesser leakage r e of the two valves. is meth of quantifying maxim pathway leakage is on to be use for this SR (i.e., Ap ndix J maximum pathwaY. leakage li its are to be quantifi in accordance with A endix J). e Frequency is requir by 10 CFR 50, Appendi J, as dified by approved e mptions (and therefore the Frequency extensions SR 3.0.2 may not be plied), since the testing is an AP. endix J, Type C test. his SR simply imposes additional cceptance criteria.

[Bypass leakage *s considered part Unless specifi lly exempted].]

1. FSAR, Section {<ff. l CJ I 2. / FSAR, seCftion [ ] . I (f) @ @. Generic Issue B-20. /4. /Genericftssue s-i4.I OJ 10 CFR 50, Appendix J. " B 3 .6-33 [Reviewer's Note: Rev 1, 04/07/95 Revised 05/31/99
  • Containment .Pressure

/(At)i{ospheri<ij B 3.6 . B 3.6 CONTAINMENT SYSTEMS B Containment BASES BACKGROUND The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident LOCA or main steam line break MSLB . These imi a so prevent the containment p essure from exc ding the contain nt design J;"\ negative pre sure differenti with respect to he outside \J.J atmosphere n the event of *nadvertent actuat'on of the Containme S ra S stem. Containment pressure is a process variable that is monitored and controlled. The containment pressure limits are derived from the input conditions used in the containment functional anal ses an econ a1n n s rue ure x erna r sur an s1s. S ou operation occur ou s1 e ese im1 s co1nc1 ent with a Design Basis Accident (OBA), post accident containment pressures could exceed calculated values. ,..f S '-I l'r*i +k ..; ... l.... " ("(.5;.U'.t!.-"U... I 1-----. h .fl-.. .....il...Je. .,._ _

'f'.......,f.5 5;,,c.e..

().. c..:.,,,;"""_"' ""..J, ... +.:.:1 .........

! w. o.f +iv-r-"'fs)

I --Wk,\¢ thL ""'<>-'!-; .... .. :"' ... e-* i,,T-er"....Q_ f'>'"E:.SS1,.o...r:.a.. ..-es ... \h h-o.-(continued) B 3.6-34 Rev 1, 04/07/95 ... .r* f.-_;+ ;t all .. weJ Moe>ES L.f ...._ ______

  1. -.q_
{ ,.._+ '-,.,*+.<.e.Q.

'-"-cl * .... l-k"j ...... h c.""'"_.i-e.:"' i" c. ;s Revised 05/31/99

  • *
  • BASES APPLICABLE SAFETY ANALYSES (continued)

Containment Pressure B occurred hile the contai ent internalpr.es .. re was at the LCO val e of [1.5] psig, total pressure o [57.3] psig would esult. This val e is still below e desi n value o 60 si . Th containment was a so desi ne for an/ 1nterna pres ure equal to [5.0 psig below externa,Y pressure in rder to withstand he resultant presscire drop from an ac dental actuation the Containment r::ri System. e LCO limit of [-.3] psig ensures t.Kat operation l.!..J within e design limit of -0.5] psig is main.fained. The maximu calculated extern pressure that wo d occur as a resul of an inadvertent;actuation of the C ntainment Spray Sys m is [2.8] psig. Containment pressure satisfies Criterion 2 of the NRC lsWteme:f(tJ. LCO Maintaining containment pressure less than or equal to the LCO-upper pressure limit that, in the event of a

  • l ...... 0 L-:+<. -h.,,..

OBA, the resultant peak containment accident pressure wi 11 o...-e.. praJ:JeJ remain below the containment desi n ressure. Maintaining Re.. a._ ..... ...ih:J... con al nm pressure grea er an or equa o the LCO 1 ower pressur limit ensures at containmen will not exceed cd\o.... -h..-... 'r-:j\.J..... the de ign negative pr sure differential following the 1.-h-e"'" -tk inadv rtent actuation f the Containmen Spray System. ; .s. ""'* . . .L _ -+.,. c_ ..... °f" .... "'"""**"*11-4 . ..+ .._ 't>SA. .. '"("'-.. ! _____________ __.. APPLICABILITY CEOG STS

  • In MODES 1, 2, 3, and 4, a OBA could cause a release of radioactive material to containment.

Since maintaining containment pressure within limits is essential to ensure initial conditions assumed in the accident analysis are maintained, the LCO is applicable in MODES 1, 2, 3, and 4. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining containment pressure within the limits of the LCO is not required in HOOE 5 or 6. '* B 3.6:...35 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • SECTION 3.6 INSERT (B 3.6.4 Applicable Safety Analysis)

The external design pressure of the containment shell is 3 psig. This value is approximately 0.5 psig greater than the maximum external pressure that could be developed if the containment were sealed during a period of low barometric pressure and high temperature and, subsequently, *the containment atmosphere were cooled with a concurrent major rise in barometric pressure. Vacuum breakers are, therefore, not provided and no minimum containment pressure specification is required (Ref. 2) . B 3.6-35 Revised 05/31/99

  • * *
  • Containment Pressure 1@ B 3.6.4 I *BASES (continued)

ACTIONS SURVEILLANCE REQUIREMENTS REFERENCES CEOG STS When containment pressure is not within the limits of the LCO, containment pressure must be restored to within these limits within 1 hour. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment,* which requires that containment be restored to OPERABLE status within 1 hour. B.1 and B.2 If containment pressure cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems . SR 3.6.4P1.l Verifying that containment pressure is within limits ensures that operation remains within the limits assumed in the accident The 12 hour Frequency of this SR was IE!) developed after taking into consideration operating experience related to trending of containment pressure variations during the applicable HODES. Furthermore, the 12 hour Frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal containment pressure condition. " B 3. 6-36 Rev I, 04/07/95 Revised 05/31/99

  • * * \ \ \ \. \ I \ \ B 3.6 CONTAINMENT SYSTEMS B 3.6.4B Containment Pressure (Dual) Containment Pressure (Dual) B 3.6.48 The containment pressure is limited, during normal operation, to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident (LOCA) or main steam line break (MSLB). These.limits also prevent the containment pressure from exceeding the containment design negative pressure differential, with respect to the outside atmosphere, in the event of inadvertent actuation of the Containment Spray System. ntainment pressure is a process variable that is monitored an The containment pressure limits are derived from the input conditions used in the containment functional analy s and the containment structure external pressure analysi * *should operatfon occur outside these limits
  • coincide with a Design Basis Accident (OBA), post accident containmen exceed calculated values . / j 1 I I *I l i I
  • APPLICABLE SAFETY ANALYSES i I I / / Containment inter al pressure is an initial condition used in the OBA analyse to establish the maximum peak containment internal ressure. The limiting OBAs considered for determining tha ma imum containment internal pressure (Pa) are the LOCA and M 8. An HSLB at 75% RTP results in the highest calculated in ernal containment pressure of 42.3 psig, which is below e internal design pressure of 44 psig. The postulated OBA are analyzed assuming degraded containment Engineered Safety eature (ESF} systems (i.e., assuming the loss of one ESF bu which is the worst case single active failure, resulting
  • n one train of the Containment Spray System and. one t ain of the Containment Cooling System being rendered inope ble). It is this maximum containment pressure, that is sed :to ensure that the licensing basis dose limitations are m
  • The initial pressure condition used in th containment analysis was [14.7] psig. The maximum cont inment pressure resulting from the limiting OBA, [42.3] psig, does not exceed the containment design pressure, [44] p ig. The containment was also designed for an internal p ssure equal to [0.65] psid below external pressure to withsta d the B 3.6-37 / i .. C EOG. .. SIS __ -***----__ I ! ' --------------*---**-.

Revised 05/31/99 ! I I I

  • *
  • Containment Air Temperature

/(Atli}4spheric 3 .6 . . B 3.6 CONTAINMENT SYSTEMS B 3.6.5 Containment Air Temperaturel(Atll}6spheric ¥\d BASES BACKGROUND The containment structure serves to contain radioactive material that may be released from the reactor core following a Design Basis Accident (OBA). The containment average air temperature is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident {LOCA) or main steam 1 i ne break {MSLB). , G""e.c.ic:A e7+) c..:,,-c..

S .1 I I c_ ......

f"\1lll'\. .. '-""-""""'-e;A.. The containment average air temperature limit is derived from the in ut conditions used in the containment

  • @ anal;ses c nta1 ment s ru ure externa ressure \\...\ lanil stsL Thh LCO ensures that initial conditions assumed in the analysis of containment response to a OBA are not v1olated dur1ng The total amount of energy* _ to removed fro_m cont a i. nment by the Conta_i nment Spray and Cooling systems during post accident conditions is dependent on the energy released to the containment due to*the event, as well as the initial containment temperature and pressure.

The higher the initial temperature, the more energy that must be removed, resulting in*a higher peak pressure and temperature. Exceeding containment design pressure may result in leakage greater than that assumed in the accident analysis (Ref. 1). Operation with containment in excess of the LCO limit violates an initial condition assumed i.n the accident analysis.

  • APPLICABLE SAFETY ANALYSES CEOG STS Containment average air temperature is*an initial condition used in the OBA analyses that establishes the containment environmental qualification operating envelope for both pressure and temperature.

The limit for containment average air temperature ensures that operation is maintained within the assumptions used in the OBA analysis for containment. The accident analyses and evaluations considered both LOCAs and MSLBs for determining the .maximum peak containment pressures and temperatures. The worst case MSLB generates larger mass and energy releases than the worst case LOCA. Thus, the HSLB event bounds the LOCA event from the containment peak pressure and temperature standpoint. The '1 B 3.6-40 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • * * . BASES APPLICABLE SAFETY ANALYSES (continued) b ... : \ .L: "") cl es:"\....

...\--e""" of "l.8 0 F. jl...a. e.J+c.;t--#-..- l.s "'esLa-d)\.Q..

44. S.ho..-+-,..J -h*-. .fl-.._ + ""'t""" ...._+ __._

e.J . -tk.'. . LCO APPLICABILITY CEOG STS

  • For dual cont nment, the initial (120]°F resu ed in a.maximum vap temperature in containment f [413.5]°F.

The t perature of the containmen steel pressure vess also reaches appro mately (413.5]°F The containment. av rage temperature of [120]°F sures that, in the vent of an accident, the maximum temperature fr containment of (26 .3]°F during OCA conditions and 413.5]°F during MSL conditions is no exceeded. The con equences of exceedin this design temp rature may be the p tential for degradat*on of the co ainment structure u der accident loads. During a OBA, with an initial containment average air temperature less than or equal to the LCO temperature limit, the resultant peak accident e erat re is maintained below f/'""i the containment design e era e. As a result, the abi 1 i ty of containment to* erform 1 ts function is ensured. t"ess: .... ,.e. In MODES 1, 2, 3, and 4, a OBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and* consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining containment average air temperature within the limit is not required in MODE 5 or 6. B 3.6.:.41 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • Containment Air Temperature i(Atn}Ospheric an{ Dual) H"§'l 3.6.5 -*BASES (contin.ued)

ACTIONS SURVEILLANCE REQUIREMENTS -n.a.. l'iO 0 f" \; ..... j a..c...t ... I: ........ c..S> ..... ....-<!J f,,,,... 4 t-.cc..:Je ..... <t a"'...l, clae.S .,_.,+ "'-CC-0........,... t i>"s+-r.,.. ....... e ...... ; ,,._e-c.c... ... ,'t CEOG STS

  • When containment average air temperature is not within the limit of the LCO, it must be restored to within limit within 8 hours. This Required Action is necessary to return operation to within the bounds of the containment analysis.

The 8 hour Completion Time is acceptable considering the sensitivity of the analysis to variations

  • in this parameter and provides sufficient time to correct minor problems.

B.1 and B.2 If the containment average air temperature cannot be restored to within its limit within the required Completion Time, the plant must be brought to a MODE in the LCO does not apply. To achieve this status, the plant must be brought to at least.MODE 3 within'6 hours and to MOOE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems . SR 3.6.5.1 Verifying that containment average air temperature is within the LCO limit ensures that containment o eration remains _w1t 1n the limit assumed for the containment analyses. In order to determine the containment average air temperature, O::C an arithmetic average is calculated using measurements taken at locations-within the containment selected to provide a representative sample of the overall containment atmosphere. The 24 hour Frequency of this SR is considered acceptable based on the observed slow rates of temperature increase within containment as a result of environmental heat sources (due to the lar e volume of containment

  • Furt ermore, e our r, quency 1s cons1 re a equa e 1n vi of other indicati s available int e control room,. in uding alarms, (j) to aler the operator to n abnormal contain nt temperature condit*on.

,, B 3. 6-42 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • BASES (continued)

REFERENCES

  • CEOG STS *
  • 2 . Containment Air Temperature J<Atmispheric

/nd Dual) t@ B 3.6.5 FSAR, Section "f l (j) FSAR, Section f' \ I B 3 .6-43 Rev 1, 04/07/95 Revised 05/31/99

  • *
  • Cont a i ng Sy;tems' s/her; 3 B 3.6 CONTAINMENT SYSTEMS I B 3.6.6A Containment Spray/and Cooling Systems and Dual) (Credit taken fpr iodine removal by the Con inment Spray System) BASES !' / BACKGROUND rhe Containment Spray and Co ainment Cooling systems containment atmosph e to post 1 accident pressure and temp ature in containment to less // than the design values. eduction of containment pressur and the iodine removal pability of the spray reduce th 1 release of fission pro ct radioactivity from containm t to 1 the environment, in t event of a Design Basis Accid t I (OBA), to within lim* s. The Containment Spray and / Containment Cooling systems are designed to the re irements , of 10 CFR 50, Appe dix A, GDC 38, "Containment He // Removal," GDC 39 "Inspection of Containment Hea Removal Systems," GOC 4 , "Test iilg of Containment Heat emova l "Inspection o Containment Atmosphere Cleanu Systems9" and I Systems," GOC , "Containment Atmosphere Clea up," GDC 42, I , GDC 43, "Tes ing of Containment Atmosphere eanup Systems" ; 1 1 (Ref. 1), other documents that were app opriate at the time of l
  • ensing (identified on a unit s ecific basis). l CEOG STS The Con ainment Cooling System*and Con inment Spray System are En ineered Safety Feature (ESF) s stems. They are desi ed to ensure that the heat rem val capability required dur* g the post accident period ca be attained.

The Co ainment Spray System and the ntainment Cooling System p ovide redundant methods to lim" and maintain post ccident conditions to less th the containment design values. Containment Spray System The Containment Spray S tern consists of two separate trains of equal capacity, eac capable of meeting the design bases. Each train includes containment spray pump, spray headers, nozzles, valves, an piping. Each train is powered from a separate ESF bus. he refueling water tank (RWT) supplies borated water to e containment spray during the injection phase of operati n. In the recirculation mode of operation, containment spr, y pump suction is transferred from the RWT to the contai ent sump{s). (continued) " B 3.6-44 Rev 1, 04/07/95 --------;-;-' ___ B

  • Revised 05/31/99
  • *
  • ing Systems/

ouauJ-@ B 3.6.@ B 3.6 CONTAINMENT SYSTEMS B 3.6.61 BASES BACKGROUND i\: ... a..-J R-.l; y.Jes f'k:..,,,.f- ..f.wo I li QI)'"' 'jA-.. ..... p s"'d-;o.,, l: r S":t:°et-1 "I.,....jet:i-: o,,... tt,...J. g_e.f. .. (:. .* + ....... l o.. .... J H**.i c.<>"'..\-4!:.:"" ....... e..rr ,. .* ... \. ..... ,. ...... ( .... ?: { vr.1..rt .:;o ... n.a *

  • 1 ,\, _ 1 c.o ..... -i-r..-\S

\ AJ \r.., . c-..-....,eA, CEOG STS * '1 B 3.6-55 ual Spray 16) (coiltin.ued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • SECTION 3.6 INSERT (B 3.6.6 Background)

The systems are arranged with two spray pumps and one air cooler fan powered from one diesel generator, and with one spray pump and three air cooler fans powered from the other diesel generator. The Containment Spray System was originally designed to be redundant to the Containment Air Coolers (CACs) and fans. These systems were originally designed such that either two containment spray pumps or thfee CACs could limit containment pressure to less than design. However, the current safety analyses take credit for one containment spray pump when evaluating cases with three CACs, and for one air cooler fan in cases with two spray pumps. To address this dependency between the Containment Spray System and the Containment Air Recirculation and Cooling System the title of this Specification is "Containment Cooling Systems, " and includes both systems. The LCO is written in terms of trains of containment cooling. One train of containment cooling is associated with Diesel Generator 1-1 and includes Containment Spray Pumps p.:.54s and P-54C, and air cooler fan V-4A. The other train of containment cooling is associated with Diesel Generator 1-2 and includes Containment Spray Pump P-54A along with CACs VHX-1, VHX-2, and VHX-3 and their associated safety related fans, V-lA, V-2A, and V-3A .. Additional details of the required equipment and its operation is discussed with the containment cooling system with which it is associated . B 3.6-55 Revised . 05/31/99

  • *
  • BASES BACKGROUND 1:-

.-D <"" J+*' *, ...... C.<>A;--c:..-h'G- .. , "'\ ... ..,:.\-l.---f +,..:s .. p\.-ospk.:..+-< (U.o ? . !:, Pl.osf\..,.,+t12. ), se ... ,e +o .,. e..-0.1 e.. .. ...,I..:""" .... \ie v-e\ ..... sed f .. \\o .... : A) ... li.Jo ... '""e""t Sp"' ...,: l\

,,, +k

.,f e-

  • . /2 Containment

-f-2(-r-y-ifl--....::d) Cooling Systems I (Atm(i4pheri (/and DupH-fo) . 8 Containment Sorav System (continued). The Containment Spray System provides a spray of cold borated water into the upper regions of containment to reduce the containment pressure and temperature during a r,;--, OBA. solution temperature is an important factor in J (:!.,,* determining the heat removal capability of the Containment Spray System during the injection phase. In the recirculation mode of operation, heat is removed from the containment sum water b the shutdown cooling heat exchan ers. ac trun o on -S"ystem _____ r:-"o provi es a quate spray cove age to meet 50% o the system f Ides i n re i rements for con a i nment heat The Containment System is actuated either automaticall b a ,ibntainment Hi h /rq'l. coincide wit a sa e in ec ion ua ion s na \....!/ or manua y. n auto atic actuation opens the containment lsok-h-;;Jspray"'"lrum¥dischij'gi)valves, starts the ontainment .lf'Lt1* . spray umps, and begins the injection phase. e \..J:.) con ain n spray e er iso a va ves en on contai ent s ra a uation si 1. A manual actuation of the Containment Spray System is available on the main control board to beg.in the same sequence. The injection

6)

Th.... ce.c..:.-c. ... \,-t-.. .l 5,,...-.p "'"" c, is b"'\ tk .5\.. ..... h<,,.+ e.cc..he."'j"rS. phase continues until an,1RWT J\vel Low signal is received. The Low :l'flivel signal for generates aA"ecirculation i'ctuat ion ,i1i na .that aligns valves from the. containment . spray pump suction.to the containment sump. The Containment Spray System in recirculation mode maintains an equilibrium tem erature between the containment atmosphere and the recircu ated sump water. Operation of the Containment Spray System in the recirculation mode is controlled by the operator in accordance with the emergency operating procedures. V ... .. i r'C) Cont inment C li <)_tJS£RT J..7--?'rwo trai s of containment cooling, each capacit to supply 50% o the design co ing requirem ts, CEOG STS

  • pr vided. Two trai s with two fa units each ar suppl"ed with cooling ater from as arate train o servic wate . All four fan are required o furnish the esign coo ing capacity.

r is drawn i o the the f and discharged o the steam nerator compar. ments and essuri zer comp tments. \ \ B 3.6-56 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • SECTION 3.6 INSERT 1 (B 3.6.6 Background)

The containment spray pumps also provide a required support function for the High Pressure Safety Injection pumps as described in the Bases for specification 3.5.2. The High Pressure Safety Injection pumps alone may not have adequate NPSH after a postulated accident and the realignment of their suctions from the SIRWT to the containment sump. Provision is made to manually provide flow from the discharge of the containment spray pumps to the suction of the High Pressure Safety Injection pumps after the change to recirculation mode has occurred. The additional suction pressure ensures that adequate NPSH is available for the High Pressure Safety Injection pumps. INSERT 2 The Containment Afr Recirculation and Cooling System includes four air handling and cooling units, referred to as the Containment Air Coolers (CACs), which are located entirely within the containment building. Three of the CACs (VHX-1, VHX-2, and VHX-3) are safety related coolers and are cooled by the critical service water. The fourth CAC (VHX-4) is not taken credit for in an DBA for maintaining containment temperature limit, but is used during normal operation along with the three CACs to maintain containment temperature below design

  • limits. The fan associated with VHX-4, V-4A, is assumed in the safety analysis as assisting in the containment atmosphere mixing function.

Each CAC has two vaJieaxial fans with direct connected motors which draw air through the cooling coils. Both of these fans are normally in operation, but only one fan and motor for each CAC is rated for post DBA operation. The post-DBA rated "safety related" fan units, V-lA, V-2A, V-3A, and V-4A, serve not only to provide forced flow for the ass9ciated cooler, but also provide mixing of the containment atmosphere. A single operating safety related fan unit will provide enough air flow to assure that there is adequate mixing of unsprayed containment areas to assure the assumed iodine removal by the containment spray. The fan units also support the functioning of the hydrogen recombiners, as discussed in the Bases for LCO 3.6.7, "Hydrogen Recombiners." In post accident operation following a SIS, all four Containment air coolers are designed to change automatically to the emergency mode. The CACs are automatically changed to the emergency mode by a Safety Injection Signal (SIS). This signal will trip the normal rated fan motor in each unit, open the high-capacity service water inlet valve to VHX-1, VHX-2, and VHX-3, and close the high-capacity service water discharge valve from VHX-4. The test to verify the service water valves actuate to their correct position upon receipt of an SIS signal is included in the surveillance test performed as

  • part of Specification 3.7.8, "Service Water System." The safety related fans are normally in operation and only receive an actuation signal through the DBA sequencers following an SIS in conjunction with a loss of off site power. This actuation is tested by the surveillance which verifies the energizing of loads from the DBA sequencers in Specification 3.8.l, "AC Sources-Operating. " B 3.6-56 Revised 05/31/99
  • * ** . BASES BACKGROUND APPLICABLE SAFETY ANALYSES Tk fo.st-...1

.. h ! lli?A i S <>."di / ; "' re.J c.rJ +o c.o,_-h::.;,_..,.. ... + E Ci F' s1 .,..sc: .... -: ... j \05s .. .. * (o"hi: ......... i ... t Sf"t:.\ l.J\...;c.."'-...;o,,.s+ .'.>;"\let. ... t;v-t. ...... Lr<.&1) (!NstJ?.i

  • 1) 1:\,,k lit. lg.\-5 6,<XA) j,J.\e' 1y.1'i.l-I (t..sUJ) CEOG STS
  • 1:) Containment Systems [AtmPiP"heric . B 3.6.6f,_.1I) .Containment Cooling System (continued)

'1 B 3.6-57 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • SECTION 3.6 INSERT 1 The analyses and evaluations considered a range of power levels and equipment configurations as described in Reference
2. The peak containment pressure case is the 0 % power MSLB with initial (pre-accident) co_nditions of 140°F and 16.2 psia. The peak temperature case is the 102% power MSLB with initial (pre-accident) conditions of 140°F and 15.7 psia. INSERT 2 (B 3.6.6 Applicable Safety Analyses)

The external design pressure of the containment shell is 3 psig. This value is approximately 0.5 psig greater than the maximum external pressure that could be developed if the containment were sealed during a period of low barometric and high temperature and, subsequently, *the containment atmosphere was cooled with a concurrent major rise in _ barometric pressure. I\ B 3.6-57 Revised 05/31/99

  • *
  • i::-* * .;i.Z.. C.. L \ BASES APPLICABLE SAFETY ANALYSES (continued) 5 fr""' *1

... :..).!...:""' ... }; II }o e I *J&-.J, 'o"' Tl f+ J... h.r r&.f,J-s-pr"'l

.....

I LCO CEOG STS

  • Containment

(:_'?> . Cooling Systems 1° B 3.6.61D---QD

  • cooling feet in the interio of the air tight c ntainment.

Additfo al di scussic*n is pro ded in the Bases *r * /v Specif, cations 3.6.4A, 3.6. , and 3.6.12, "Va uum Relief (_:: Valv ." The model Containment Cool *ng System actuation from the 1 contain t analysis is bas on the unit speci ic response rfl\ time as ciated with excee ing the CCAS. to ach eve full Contai ent Cooling Syste air and safety gra e cooling water low. The Containment Spray System and the Containment Cooling System satisfy Criterion 3 of &he/NRC Policy/StatemenU. 10 ci::z. ?(p [c,,)fJ..) c§g.) During a OBA, a minimum containment cool in trairvJ wo c n ainmen s ra rains, r one o ac , require maintain the pressure an te perature below the design limits To ensure that these of-. \G;) requfrements are met, two a:onWnment waY> trains \tu\ containment cooling be OPERABLE. Therefore, in 'the event of an accident, the minimum requirements are met, assumjng the worst case single active failure occ Spray System ic or .. l8:'J pump, spray headers, nozzles, valves, piping, instruments, and controls to ensure an OPERABLE flow path capable of \ \ B 3.6-58 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • SECTION 3.6 INSERT (B 3.6.6 LCO) One train of containment cooling is associated with Diesel Generator 1-1 and includes Containment Spray Pumps P-54B and P-54C, and air cooler fan V-4A. The other train of containment cooling is associated with Diesel Generator 1-2 and includes Containment Spray Pump P-54A along with CACs VHX-1, VHX-2, and VHX-3 and their associated safety related fans, V-lA, V-2A, and V-3A . *,I B 3.6-58 Revised 05/31/99
  • *
  • BASES LCO (continued)

APPLICABILITY ACTIONS or'\e. ar .,,;ore.. a.:,..5 c_ .. ,,...+.-: c "'"' \' ir ' I 00 lJo of -}k c,o.J-C.:,. ..... ;..') . l..J c.-c;* .... 5\<... c-..-h:_:..,.,...tJ c.ool:

s c. J..
\ -!.' ... } CEOG STS * (,) ;:--......--'"i --Containment Sprl rut\ Cooling Systems J(Atm,?spherii' and D\l"aU/-f)

B }!' taking suction from upon an ESF actuation signal and automatically transferring suction to the containment sump. *Each ontainmen!F.ITf_ .. _ oling typicall)i includes } JU) dem* ters, cooli coils, da!JI _rs,_ fans L0 co trols to en re an OPERA flow pa ./ * --JB In MODES l_, 2.ff a OBA could cause a release of I {f) radioactive material to containment and an increase in containment pressure and temperature requiring the operation of the containment spray trains and containment cooling trains. In and 6,. the probability and consequences of these ! events are reduced due to the pressure and temperature limitations of these MODES. Thus, the Spray and Containment Cooling systems are not required to be OPERABLE JO') in.Moo6Js and 6.. . Iv he 14 day po ion of the Completio Time for Required Action A. l i based upon engineeri judgment. It takes into accoun the low probability coincident entry in o two Condit ons in this Specifica ion coupled with the ow probabil i y of an accident ace ring during this tim

  • Refer t Section 1.3, "Comple on Times," for a mor !detail discussion of the p pose of the ",from di covery of :failu e to meet the LCO" po ion of the Completi Time. '1 B 3.6-59 ) (continued)

Rev 1, 04/07/95 Revised 05/31/99 '((;' ' (--

  • *
  • SECTION 3.6 INSERT (B 3.6.6 LCO) The Containment Air Recirculation and Cooling System portion of the containment cooling train which must be OPERABLE includes the three safety related air coolers which each consist of four cooling coil banks, the safety related fan which must be in operation to be OPERABLE, gravity-operated fan discharge dampers, instruments, and controls to ensure an OPERABLE flow path. CAC fans V-lA, V-2A, V-3A, and V-4A must be in operation to be considered OPERABLE.

These fans only receive a start signal from the DBA sequencer; they are assumed to be in operation, and are not started by either a CHP or an SIS signal. (B 3 .6.6 ACTIONS) This condition allows for the loss of any two containment spray pumps, or the three required CACs and one containment spray pump, even if they are supplied from different electrical trains of power as long as two containment spray pumps and one cooling fan, or the three required CACs and one containment spray pump are available. ,, B 3.6-59 Revised 05/31/99

  • * * * ---:-' --* -=. -*' i Containment

{Ipr!y fan(j Cooling Systems jCAtii)6spheri/ and or.1 D-;@ . B 3 . . 66't:--(fD BASES ACTIONS B.1 (continued) CEOG STS With one required containme cooling train inoperable, the inoperable containment coo ng train must be restored to OPERABLE status within 7 ays. The components in this degraded condition are pable of providing greater than 100% of the heat remov needs (for the condition of one containment cooling t ain inoperable) after an accident. The 7 day Completion ime was developed based on the same reasons as those f Re*quired Action A.I. The 14 day porti of the Completion Time for Required Action B.1 is b sed upon engineering judgment. It ta es into account e low probability of coincident entr into two Conditio in .his Specification coupled with e low probability of an accident occurring during this ime. Refer to S ction 1.3 for a more detailed discus ion of the purpose the "from discovery of failure to et the LCO" portion f the Completion Time. two required containment spray ains inoperable, one I f the required containment spray t ins must be restored to OPERABLE status within 72 hours. e components in this degraded condition are capable o providing greater than 100% of the heat removal needs fter an accident. The 72 hour Completion Time was d eloped taking into account the redundant heat removal pabilities afforded by combinations of the Contai ent Spray System and Containmen Cooling System and the 1 probability of a OBA occurring during this period. D.1 and D.2 With one requir containment spray train inoper le and one of the require containment cooling trains ino rable, the inoperable c tainment spray train or the in erable containment cooling train must be restored OPERABLE status wi in 72 hours. The components i this degraded conditio are capable of providing at le st 100% of the heat removal needs after an accident. The 7. hour Completion Time s developed based on the same easons as those for Req *red Action C.l. \ ' B 3.6-60 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • * *
  • BASES ACTIONS (continued)

SURVEILLANCE REQUIREMENTS CEOG STS 1J Cooling Systems l<At!ii6spheric . 3.6.6 ----------.--------.E.1 With two contain ent* cooling trains required contai nt cooling *trains ust be restored t OPERABLE statu within 72 hours. T e components in t s degraded condi ion are capable of roviding greater an 100% of the h at removal needs af er an accident. e 72 hour Comp etion Time was deve oped based on the reasons as hose for Required A tion C.1. If Required and associated Completion /.h"'l TimelJof this LCO are not met, the piant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be to at least MODE 3 within !O\ 6 hours and to within The a 11 owed Completion Times ar'e'reasonable, based on operating eoxperience, to.reach the required plant conditions. from full power conditions in an orderly manner and without challenging plant systems . With a combination of th ee or more Containme t Spray Syste and Containment Co ing System trains i operable, unit s in a condition o tside the accident alysis. The fore, LCO 3.0.3 mu t be entered i11111edi ely. SR 3.6.6j.l (@ the {f) Verifying the correct alignment for manual, power operated, and automatic valves, excluding check valves, in the Containment Spray System provides assurance that the proper flow path exists for Containment Spray System operation. This SR also does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct positions prior to being secured. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown, that " B 3.6-61 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • BASES SURVEILLANCE REQUIREMENTS

/0 Cooling SR 3.6.sj.I (continued) l those valves outside containment and capable of potentially being mispositioned, are in the correct position. SR 3. 6. 6i). 2 I. \1) I 'A'a.t'f rc.U:a.t!d e,.,,..+ ... =--e-+ ft.',. c.,Je..- Operat1ng unit for 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. The 31 day Frequency was developed considering the known reliability of the fan units and controls, the two train redundancy available, and the low probability of a significant degradation of the containment cooling train occurring between surveillances. '1-( ,.Al 4 SR 3.6.0l!d° (D water flow rate to@ I fG'\ 1./1\-1<-I, assurance the design rate assumed CJ in the safety analyses will ba achieved Also CEOG STS

  • considered in selectin this Frequency were the known rel iabi 1 ity of the

)'Ystem, the ! fj} redundancy, and low probability of a significant degradation of flow occurring between surveillances. SR 3.6.6IK3 Ci) --, > ,.L.14-\}Fl-,.....Ll"I the containment spray header is full of water to the mln1m1zes the time required to fill the header. This ensures that spray flow will be admitted to the containment atmosphere within the time frame assumed in the cont a i nrnent ana 1 ys is. , The 31 day Frequency is based on the static nature of the fill header and the low probability of a significant degradation of the water level in the piping occurring between surveillances. \I B 3 .6-62 (c;ontinued) 1, 04/07/95 Revised 05/31/99

  • **
  • BASES SURVEILLANCE REQUIREMENTS (continued)

SR 1...1..lo e.i:.ch ,,;.__t.-... -1.* ... \l ..:\""L. I +<::> *, C.o rr fo ,._ '-'-fb" .. 1:.- 0.,. ......... 1 .. ci.cf114+:e-.A of'.. c.o ... .+c.:.""""e ... +-.srn., o.. .... \-oJ: C..&..\\ \ 0" a....... cu...-t ..... ...t_ o.r !..: o.ct.1A-A-+:C"\...- i.e.> C...H P. Contain110nt Systems i<AtmC)&pheric /rld B 3.6. @ he.:d ika_

, 4*ec.
h.r

..... o;o re :reJ de.xl! .. feJ. SR 3.6.§1.5 . I Veri f in that each i __ lo @ 250 sid diffi ential res re on reci ulation ensures t at spray pump per ormance as not egra e ur1ng the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by Section XI of the ASHE Code (Ref.). Since the containment spray pumps f (1) cannot be tested with flow through the spray headers, they are tested on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this.SR is in accordance with the Inservice Testing Program. SR 3.6.61.6 and SR* 3.6.61.7 These SRs ve fy each automati containment spra valve actuates to its correct posi on and that* each ontainment spray pump starts u on rece* t of an actual or simulated actuatio si nal. This Surveillance is not required for valves that are ocked, or otherwise secured in the re uired osition under administrative controls. The 8 mont requency 1 s ase on t e nee .to perform these Surveillances under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillances were performed with the reactor at power. experience has shown that these components usually pass the Survei 11 ances when performed at the .fl8i' month Frequency. Therefore, the Frequency was concluded to be *acceptable from a reliability standpoint. surveillance of containment sump valves is also required*b1* SR 3.5.2.5, _.A" single surveillance may be .used to* satisfy both requirements. (j) CEOG STS

  • SR 3.6.tf.s . This SR verifies each containment cooling r actuates upon receipt of an actual or simulated actuation signal.

Frequency is based on engineering judgment and has been shown to be acceptable through operating \I B 3.6-63 (continued) Rev 1, 04/07/95 Revised 05/31/99

  • * *
  • BASES SURVEILLANCE REQUIREMENTS REFERENCES CEOG STS ,---.,

Containment Cooling Systems/(Atl1)6'spheric

3. 6. -:-\\') SR (continued) experience.

See SR and SR above, for further discussion of the basis for the i18¥month Frequency. fi fl' SR 3.6.§1.9 .. __, With the. containment spray inlet valves closed and the spray header drafoed of any solution, low pressure air or smoke can be blown through test connections. Performance of this SR demonstrates that each spray nozzle is unobstructed and provides assurance that spray coverage of the containment during an accident is not degraded. Due to the assive design of the nozzle, a test at t e 1rst r fue in d at Ff\ 10 year intervals is considered a equa e o e ec obstruction of the spray nozzles. 10 CFR 0, Appendix A, GD 38, GDC 39, GDC and GOC 43. 2. FSAR,

3. FSAR, Sectiorj ql..1y. 14. I FSAR, Secfion r1]. (g}@. FSAR, Section fb.1}2;-@ ASHE, Boiler and Pressure Vessel Code, Section XI. {g, 1, FSA?, l,..\..\c..

!If.Ii,\-3 f= 1 lc.\ak 14. If. 2..-I F'$Alt, q_\ ,, B 3 .6-64 Rev 1, 04/07/95 Revised 05/31/99

  • * ** -----------*

Spray Additive System (Atmospheric and Dual) B 3.6.7 -----------* -* -*--------.. -----. --*-.-**----. . . ... B 3.6 CONTAINMENT SYSTEMS *s 3.6.7 Spray Additive System {Atmospheric and Dual) / The Spray Additive System is a subsystem of Containment Spray System that assists in reducing the todine fission product inventory in the containment in the event of an accident such as a loss of coolany'accident {LOCA). qe addition of a spray additive to tie boric acid spray increases the pH of the sppay solution and . maintains the containment sump pH,above 8.0 during the phase of an An elevated pH is .desired \ince it enhances the todine removal capacity of the sprays an aids in the retenti'On of iodine in the water in the contain ent sump. / // The Spray Addi ]Ve Systeoi consists of a single spray* additive tank an two r,edundant 100% capacity trains. Each train contain*s a em}t:al addition pump, an injection valve, isolation valves, a ow meter, and a flow controller. Upon receipt of a contai,ri nt spray _actuation signal (CSAS), the chemical additif.on pump start and the injection valves open in each redunda train. The spray additive is then injected into e Contain nt Spray System at the suction of the spray pumps at metered amounts corresponding to the containmen spray pump discharge flow . rate. Th rate at which the s ray additive is added is en a recirculation ac ation signal is generated and thiy'Containment Spray System ters the mode ot" operation. The pH of the ntainment spray solution is ma:1ntained between 9.0 and 10.0 ring the injection mode and/between 8.0 and 9.0 in the recirc ation mode. Upon r,aching a Low-Low level in the spray emical addition /ank, the spray chemical addition pumps top and the injection and isolation. valves close {Ref. 1). -The Spray Additive System aids in reducing t e iodine . fission production inventory in the containme t atmosphere. I i I ' SAF/ANALYSES / ( The Spray Additive System is essential to the effe tive removal of airborne iodine within containment follo ing a Design Basis Accident (OBA). // (conti I . I // CEOG STS --------.C--' --* --* -** ,. ________ ----Q, 3 . lP ... 1-0 Revised 05/31/99 ! i I !

  • *-** Hydrogen Recombi ners j(Atn\?spheri c B 3.6 CONTAINMENT SYSTEMS B Hydrogen Recombiners

/(Atmoslefheric and Di!l)(If permarli'ntly installed)h'_B-) BASES BACKGROUND

  • vJhtA a.

i S rL:.t.eJ ; .... orL ... ,* .. I ""'LS J,/'t:.,...J""' b\ "-"'-.L . ., J .... re.. ** s ..... f'* .. .... <ec.o""" o"' of D)C."\ '!"°' ..\-o oC.C.1..v-I

  • APPLICABLE SAFETY ANALYSES CEOG STS The hydrogen recombiners provide for controlling the bulk hydrogen concentration in containment to less than the lower fla11111able concentration of 4.1 v/o following a OBA. This control would prevent a containmentwide hydrogen burn, thus ensuring the pressure and temperature assumed in the analysis are not exceeded and minimizing damage to safety , , B 3 .6-71 (continued)

Rev 1, 04/07/95 Revised 05/31/99

  • *
  • SECTION 3.6 INSERT (B 3.6.7 Background)*

In order for a Hydrogen Recombiner to be considered OPERABLE, at least one containment cooling safety related fan powered from the same electrical train must be in operation or available for operation. Fan operation is necessary to ensure that the post-accident containment atmosphere is adequately mixed preventing local hydrogen buildups in excess of the flammability liniit. The supporting fan must be powered from the same electrical train as the recombiner, to assure that it would be available in the event of an accident combined with a Loss of Offsite Power and failure of the opposite Diesel Generator (DG). The fan must be . started or verified to be in operation when the recombiners are placed in operation because these fans do not receive an SIS start signal (although they do receive a DBA sequencer start signal). If there is no qualified fan available, the associated recombiner would not have all of its required support equipment and would have to be declared inoperable. Only fans V-lA, V-2A, V-3A, and V-4A (associated with VHX-1, VHX-2, VHX-3, and VHX-4 respectively) are qualified for the post-accident containment environment . Recombiner M-69a, powered from the Left Train (DG 1-1), is supported by* V-4A, which is the only qualified fan powered from the Left Train. If V-4A was unavailable, Hydrogen Recombiner M-69B would have to be declared inoperable. Reconibiner M-69A, powered from the Right Train (DG 1-2), can be supported by V-lA, V-2A, or V-3A, all of which are from the Right Train. If V-lA, V-2A, and V-3A were all unavailable, Hydrogen Recombiner M-69A would have to be declared inoperable. LCO 3. 6. 6, "Containment Cooling Systems," also contains requirements for containment cooling fan OPERABILITY. The restoration time specified for Containment Air Coolers in LCO 3.6.6 is more restrictive than that specified for Hydrogen Rec.ombiners in LCO 3.6.7.

  • B 3.6-71 Revised 05/31/99
  • * * * . BASES APPLICABLE SAFETY ANALYSES (continued)

APPLICABILITY CEOG STS .related equipment located in containment. The limiting OBA relative to hydrogen. generation is a LOCA. Hydrogen may accumulate within containment following a LOCA as a result of: a. b. c. d. A metal steam reaction between the zirconium fuel rod -cladding and the (rj(ct.efjccoolant;(fr,,....,.. .... I Radiolytic of water in the \R@'(c£8e{f..-*""'"""J) {jJ Coolant System (fjJ:S) and the containment sump;

  • Hydrogen in the(BC'S at the time of the LOCA (i.e., hydrogen dissolved in the r c coolant and hydrogen l gas in the pr..:Ssurizer vapor space ; or r ... : ......... , Corrosion of metals exposed to Containment Spray System and Emergency Core Cooling Systems solutions.

to evaluate the potential for hydrogen accumulation in containment following a LOCA, the hydrogen generation as a function of time following the initiation of the .... *--"':"I calculated. Conservative assumptions(fij¢011111etldedl

  • Reference 3 are used to maximize the amount of hy rogen calculated.
  • satisfy Criterion 3 of lti?e NRCI 1oc..FR so.1 -q In HODES 1 and 2, two hydrogen recombfoers are required to control the post LOCA hydrogen concentration within containment below its fla11111ability limit of 4.1 v/o, assuming a worst case single failure. B 3.6-72 (continued)

Rev 1, 04/07/95 Revised 05/31/99

  • * * \ SECTION 3.6 INSERT (B 3.6.7 LCO) In addition, one safety related containment cooling fan associated with each train must be in operation.

These requirements ensure operability of at least one hydrogen recombiner and adequate mixing of the containment atmosphere in the event of a worst case single active failure . ,, B 3.6-72 Revised 05/31/99

  • * *
  • BASES APPL I CAB IL ITV (continued)

ACTIONS CEOG STS I Hydrogen Recombi ners /(Atp!ospherfc and D\ia 116 B 3.6. 1 0 _In MODES 3 and 4, both the hydrogen production rate and the total hydrogen produced after a LOCA would be less than that calculated for the OBA LOCA. Also, because of the limited time in these MODES, the probability of an accident requiring the hydrogen recombiners is low. Therefore, the hydrogen recombiners are not required in MODE 3 or 4. In MODES 5 and 6, the probability and consequences of a LOCA -are low, due to the pressure and temperature limitations. Therefore, hydrogen recombiners are not required in these MODES. A...! With one containment hydrogen recombiner inoperable, the inoperable recombiner must be restored to OPERABLE status within 30 days. In this condition, the remaining OPERABLE hydrogen recombiner is adequate to perform the hydrogen control function. The 30 day Completion Time is based on the availability of the other hydrogen recombineri the small probability of a LOCA or MSLB occurring (that would generate an amount of hydrogen that the flan111ability limit), and the amount of time available after a LOCA or MSLB (should one occur) for operator action to prevent hydrogen accumulation from exceeding the flilllll1ability limit. Required Action A.I has been modified by a Note stating that the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when one recombiner is inoperable. This allowance is based on the availability of the other hydrogen recombiner, the small probability of a LOCA or MSLB occurring (that would generate an amount of hydrogen that exceeds the flan111ability limit), and the amount of time available after a LOCA or MSLB (should one occur) for operator action to prevent hydrogen accumulation from exceeding the flan111ability limit. [Reviewer's ote: This Condi ion is only allow d for units J 6) with an a ernate hydrogen ontrol system ace ptable to the technica staff. I 1 B 3 .6-73 (continued) Rev 1, 04/07/95 -Revised 05/31/99

  • BASES ACTIONS
  • CEOG STS *
  • Hydrogen Recombi ners /(AtmoJpheri c and /6ua 1) I 8 3.6.8 B'.l and B.2 (continu ) With two hydrogen r ombiners inoperable, the ab' ity to perform the hydrog control function via alter ate capabilities must e verified by administrativ means within 1 hour. The alt rnate hydrogen control capabjlities are provided by [th containment Hydrogen Purge recombiner/Hyd ogen Igniter System/Hydroge Mixing System/Contai ment Air Dilution System/Co ainment Inerting System]. T 1 hour Completion Time all s a reasonable

/?l period of me to verify that a loss of ydrogen control function es not exist. [Reviewer's ate: The following is to be sed if a non-Technical Spec fication alternate hydroge control function is used to justify this Condition: In add" ion, the alternate hydroge control system capab" ity must be verified every 2 hours thereafter to ensu its continued availabilit .] [Both] the [initial] ver* ication [and all subsequen verifications] may be pe formed as an administrative check, by examining logs o o er information to determin the availability of the ternate hydrogen control s stem. It does not mean t perform the Surveillances n eded to demonstrate OPERA LITY of the alternate hydrogen antral system. If the ab' ity to perform the hydrogen con ol function is maintained continued operation is mitted with two hydrogen recombiners inoperable or up to 7 days. Seven d ys is a reasonable time to al w two hydrogen recombiner to be inoperable because t e hydrogen contror functi is maintained and beca se of the low probability f the occurrence of a LO A that would generate hyd gen in amounts capable of exceed ng the fla11111ability limit ©cru\@ If the inoperable hydrogen be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power condit'ions in an orderly manner and without challenging plant systems. " B 3.6-74 . (continued) Rev 1, 04/07/95 Revised 05/31/99

  • *
  • Hydrogen Recombi ners UAtmoiP!\eri c and{ua ir,:f 8 .6. ! :..,""\ /.'.!\ I( L.J V,/ I*-. . * *BASES (continued)

SURVEILLANCE REQUIREMENTS l"h:s SR a.lso ..\".-..Q.

  • 0.C o._

+.:s-"<T-.\-o. ; s "'"' c .... f\...*tC. -\:..._ ... 1-1'-or "-"'\ .S ' o'° 11.. s: b,.,_ \c., ""f1-.c. ... :'t"l +es+ .... e ... i .... .cS 1-k re 'i +.c nC.C. 0 + ca. cA..-. i._ .. h. .... c ... REFERENCES CEOG STS * *

  • Irv . SR

\:::.; Performance of a system functional test for each hydrogen recombiner ensures that the recombiners are operational and can attain and sustain the temperature necessary for hydrogen recombination. In particular, this SR requires verification that the minimum heater sheath temperature increases to 700°F in s 90 minutes. After reaching 700°F, the power is increased to maximum for a roximatel 2 minutes and verified to be 60 kW. Operating ex:ze i ence has s own a ese componen s u a y pass the Surveillance hen performed at t e fl8] month Freq ncy. m Therefore, e Frequency was c eluded to be acce able from I

  • a reliabi ty standpoint.

This SR ensures that there are no physical problems that could affect recombiner operation. Since the recombiners are mechanically passive, they are not subject to mechanical failure. The only credible failures involve loss of power, blockage of the internal flow path, missile. impact, etc. A* visual inspection is sufficient to determine abnormal conditions that could cause such failures. The tIBf month I (D Frequency for this SR was developed considering that the incidence of hydrogen recombiners failing the SR in the past is low. This SR requires performance of a resistance to ground test for each heater phase to ensure that there are no detectable grounds in any heater phase. This is accomplished by . verifying that the resistance to ground for any heater phase .r.t 10 000 ohms. The fl8tmonth Frequency for this SR was I tJJ deve ope cons1 ering*that the incidence of hydrogen recombiners failing the SR in the past is low. 1. 10 CFR 50.44. 2. l 10 cFiso, AppEtfidix A. GDC/41. I I

5. \ 1-G 3. Regulatory Guide 1.7, Revision flt:' \ {S) I' B 3.6-75 Rev 1, 04/07/95 Revised 05/31/99
  • *
  • B 3.6 CONTAINMENT SYSTEMS HHS B 3.5.9 Hydrogen Mixing System (HMS} (Atmospheric and Dual) BASES BACKGROUND The HMS reduces the potential for breac of containment due to a hydrogen oxygen reaction by pro'vi ing a uniformly mixed post accident containment atmosphere, thereby minimizing the potential for local hydrogen burns e to a local pocket of hydrogen above the flanmable conce ration and giving the operator the capability of preve ing the occurrence of a bulk hydrogen burn inside conta* ment per 10 CFR 50.44, "Standards for Combustible Gas ontrol Systems in Water-Coo.led Reactors" (Ref. ), and 10 CFR 50, GDC "Containment Atmosrhcre Cle upn (Ref. 2). \ The post acctdent HHS is Engineered Safe'ty Feature and is designed to withstand a oss of coolant accident {LOCA) without loss of functi
  • The system has two independent trains, each of wh..ich onsists of two dome air circulation fans, motors, and co rols. Each train is sized for [37,000] cfm. The wo trains are initiated automatically on a containment coo ng*actuation.

signal (CCAS) or can be manually started from control room. Each train is powered *from a eparate emergency power supply. Since each train can de 100% of the mixing requirements, the system will ovide*its destgn function with a limiting single acti e failure. \"

  • The HMS celerates the air mixfng process between the uppe*r dome sp ce of the containment atm(>sphere during LOCA opera ons. It also prevents any h'Qt spot air pockets duri the containment cooling mode 'a{ld avoids any hydrogen con ntration in pocket areas. \ \ \. drogen mixing within the containment by he Containment Spray System, the emergency fan coolers, and the containment internal design, which permits convective miXing and The HMS, operating in conjunction with the Spray System and the emergency fan coolers, prevents'\ocalized accumulations of hydrogen from exceeding the fla ability limit of 4.1 volume percent {v/o). ,, B 3.6-76 (continued)

Rev 1, 04/07/95 Revised 05/31/99

  • *
  • ICS (Atmospheric and Dual) B 3.6.10 ----*----.-....

... CONTAINMENT SYSTEMS ---------------Iodine Cleanup System (ICS) (Atmospheric and Dual) :\ !//! / I BACKGROUND i l \ \ (f) j CEOG ST The ICS is provided per GDC 41, "Containment Atmospher,I Cleanup," GDC 42, "Inspection of Containment Atmosphere leanup Systems," and GDC 43, "Testing of ContainmeJJi mosphere Cleanup Systems" (Ref. 1), to reduce th}!' co entration of fission products released to cont inment atmosphere following a postulated a<;Cident. The ICS w ld function together with the Containmept Spray and Cooling ystems following a Design Basis Accident (OBA) to reduce t potential release of principall iodine, from the containment/t'o the environment. The ICS consi ts of two 100% capacity independent, and redundant ains. Each train a heater, [cooling coils,] a prefilter, a moi$ture separator, a high efficiency partic ate air (HEPA) filter, an activated charcoal adsorber s ction for removal of radioiodines, and a fan. Ductwork, valv and/or dCllfipers, and instrumentation also form part of the stem. /The moisture separators function to reduce the ist11re content of the airstream. A second bank of HEPA filt the adsorber section to collect carbon fines and R ovide backup in case of failure of the main HEPA filter Only the upstream HEPA filter and the charcoal adsorber sec ion are credited in the analysis. The system jnitiate filtered recirculation of the containment foll ing receipt of a containment actuations* nal. The sy::.tem design is described in

2. The primary purpose of the heaters is to ensure that the relative humid,lty of the airstream ent ring the charcoal adsorbers is maintained below 70%, whic is consistent with the assigneq/iodine and iodide removal e ficiencies as per Regulatory/Guide 1.52 (Ref. 3). _ The separator is included for moist re (free water) from the gas stream. Heaters are use to heat the gas 1tream, which lowers the relative humidity.

Continuous ope,ration of each train for at least 10 hours p month with heaters on reduces moisture buildup on the H A filters adsorbers. Both the moisture separator and he ter are portant to the effectiveness of the charcoal adso ers. \ \ B 3.6-81 Rev Revised 05/31/99 I

  • *
  • Shield Building (Dual) B 3.6.11 -r---a 3.6 CONTAINMENT SYSTEMS B '3_.6.11 Shield Bu-ilding (Dual) i I BASES *, ) i '-I I The shield building is a concrete structure that surl?u'nds steel containment vessel. Between the contain)l(ent

'-Yessel and the shield building inner wall is an ,rinular s ace that collects any containment leakage may occur fo lowing a loss of coolant accident (LOCA). /This space also allows for periodic inspection of the 90ter surface of the s el containment vessel. . / LOCA, the Shield Building Air Cleanup System CSBEACS) establishes a negativ,...pressure in the annulus betWeen the shield building and the steel containment v'e.ssel. Filters in th system then .control the release of radf&active contamina s to the environment. A description of the SBEACS is pr ided in the Bases for Specification 3.6.'1_3, "Shield uilding Exhaust Air Cleanup System (SBEACS)." Shield bu' ding OPERABILITY is.required to ensure retention o pri ry containment leakage and proper operation of th S CS. \ APPLICABLE The design basis f shield ilding OPERABILITY is a /I '\ SAFETY ANALYSES large break LOCA. Maintaining hield building OPERABILITY ensures that th release of rad active material from the \ primary contai ment atmosphere is estricted to those leakage.path and associated leaka rates assumed in the accident a ysis. * * * / . I . d building satisfies Criterion of the NRC Policy t. I ield building OPERABILITY must be maintaine to ensure l proper operation of the SBEACS and to limit ra *oactive 1 1 eakage from the containment to those paths and 'eakage I rates assumed in the accident analysis. \ j l l ' / ( CEOG STS I I B 3.6-86 & :?..(.-<<gC, {continu d) I Rev 1, 04/07/95 Revised 05/31/99

  • *
  • Vacuum Relief Valves (Dual) B 3.6.12 3.6 CONTAINMENT SYSTEMS 6.12 Vacuum Relief Valves (Dual} BASES BACKGROUND"\

The vacuum relief valves protect the containmen vessel 1 "\ against negative pressure (i.e., a lower press e inside \ than outside}. Excessive negative pressure

  • side APPLICABLE SAFETY ANALYSES \. EOG STS \\containment can occur if there is an inadve ent actuation

'of the Containment Cooling System or the ntainment Spray Multiple equipment failures or uman errors are to have inadvertent actuatio . The pressure vessel con ins two 100% vacuum relief )Jnes installed in paralle that protect the containmeQt from excessive exter l loading. The vacuum relief lin'e.$ are 24 inch penetr. tions that connect the shield building annulus to th containment. Each vacuum relief line iS\isolated by pneumatically operated butterfly series th a check valve located on the containment side the P. netration. i Each butterfly valve

  • actuated by a separate pressure 1* controller that sens he differential pressure between the'* containment and the nnu s. Each butterfly valve is I provided with an r accum lator that allows the valve to
  • open following a oss of air. The combined essure drop at). ted flow through either vacuum relie line will not exce d the containment pressure vessel desi n external pressure d ferential of [0.65] psig with any evailing atmospheric pre sure. Desi n of the vacuum relief lines involve calculating the ef ct of an inadvertent containment spray tuation that c n reduce the atmospheric temperature

{and h ce pressure) nside containment {Ref. 1). Conservative ass ptions are used for all the pertinent parameters in the cal lation. ; For example, the minimum spray water temperature 1 assumed, I as well as maximum initial containment ! spray flow, all trains of spray operating, etc. The ! resulting containment pressure versus time is calcula d, ! including the effect of the vacuum relief valves openin { . I a a.& ge f I Rev I, 04/01 /,9s Revised 05/31/99

  • * *** SBEACS (Dual} B 3.6.13 B 3.6 SYSTEMS B;'-3..6.13 Shi'eld Building Exhaust Air Cleanup System (SBEACS) (Dual) .""'-., ..... _ BASES ."-. I ACKGROUND I I \. l \ . @I ' ' i
  • I I '*"-,. The SBEACS is required by 10 CFR 50, Append i /A, GDC 41, \ *-."Containment Atmosphere Cleanup" (Ref. 1),):0 ensure that L radioactive material leaking from the pri ry containment O!f a containment into the shield build"ng (secondary i contalnment) following a Design Basis cident are and prior to exhausting to t environment.

I The containment has a secondary co ainment, the 'shield 1* building, whi'c.h is _a concrete str ture that surrounds the steel primary Clintainment vessel Between the containment vessel and the sh{eld building nner wall is an annular space that collects,_ any conta* ment leakage that may occur following a loss of'<\oolant ccident (LOCA). This space also allows for perio(i,'!c i pection of the outer surface o the steel containment 'tfe..s 1. Following a LOCA, the EA establishes a negative pressur in the annulus betwee the s ield building and the steel containment vessel. ilters the system then control the release of radioac ve contamin ts to the environment

  • Shield building O RABILITY is re ired to ensure retention of primary conta"nment leakage and roper operation of the SBEACS. The SBEACS c sists of two separate and edundant trains. Each train
  • eludes a heater, cooling coi , a ?refilter, moisture s parator*, a high efficiency part ul ate afr (HEP ) filter, a activated charcoal adsorber secti for removal of.radio odines, and a fan. Ductwork, valves nd/or damper , and instrumentation also form part of he system. The sture separators function to *reduce the m sture cont t of the airstream.

A second bank of HEPA "lters fol ws the adsorber section to collect carbon fine and pr ide backup in case of failure of the main .HEPA fi er b nk. Only the upstream HEPA filter and the charcoal section are credited in the analysis. The syste initiates and maintains a negative air pressure in the shield building by means of filtered exhaust ventilation the shield building following receipt of a safety injecti actuation*signal (SIAS). The system is described in Reference

2. (continu d) *Revised 05/31/99
  • *
  • ATTACH1\1ENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.1, CONTAINMENT Change* Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has . been provided.
2. . Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of intent. 3. The requirement/statement has been deleted since it is not applicable to this facility.

4.
  • The following requirements have been renumbered, where applicable, to reflect this deletion . Chang_es have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
5. This change reflects the current licensing basis/technical specification.
6. This change reflects the implementation of 10 CFR 50 Appendix J Option B for Type A containment leakage tests and the incorporation of the Containment Leak Rate Testing Program which is found in TS Chapter 5.0, Administrative Controls.

This change is consistent with TSTF-52 for the portions of containment testing which implement

7. 10 CFR 50 Appendix J Option B. Since the Type B and C testing is done in accordance with 10 CFR 50 Appendix J Option A, the appropriate wording is changed where appropriate.

In addition, some editorial changes were also made as part of TSTF-52 and reflected in the proposed Bases. The proposed ITS includes a discussion of the Main Steam Line Bre.ak (MSLB) accident to clarify the basis of the term "Pa" since the MSLB accident results in the maximum calculated containment pressure for the design basis accidents. However, as discussed in the Bases, the limiting accident from an offsite dose perspective (i.e., affected by containment leakage) is the Loss of Coolant Accident (LOCA). Therefore, the maximum calculated LOCA pressure will be the Pa pressure. The Pa provided in the Bases represents the analytical limit provided in the FSAR (currently

52. 64 psig) rounded up to 53 psig. Palisades Nuclear Plant Page 1of2 05/31/99
  • *
  • Change Discussion ATTAC1ŽENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.1, CONTAINMENT
8. The Palisades Nuclear Plant used Regulatory Guide 1.35 as a reference for the Containment Structural Integrity Surveillance Program but does not fully commit to Regulatory Guide 1.35. Therefore, this reference is not included in the proposed ITS. 9. Not used. 10. The Palisades Nuclear Plant is considered to be an "Atmospheric" containment.

The heading titles referring to either "Atmospheric" or "Atmospheric and Dual" are deleted since they add no value to the usage of the plant specific Palisades Nuclear Plant Technical Specifications. In addition, the portions of NUREG-1432 which are provided for "dual" containments are not applicable to the Palisades Nuclear Plant. 11. A new SR (SR 3.6.1.3) has been added to address equipment and penetrations subject to Type B and C leakage rate tests in accordance with 10 CFR 50, Appendix J, Option A. The creation of an additional SR is necessary since the Type A testing specified in SR 3.6.1.1 is performed in accordance with 10 CFR 50, Appendix J, _ Option B. The requirement and Frequency of ITS SR 3. 6 .1. 3 is similar to ISTS SR 3.6.1.1 with the exception that it contains an explicit requirement to perform leak rate tests 55 psig consistent with the CTS. Conforming changes have been made to the Bases. ISTS SR 3.6.1.1 was also modified by deleting the sentence which reads, "Failure to meet air lock and purge valve with resilient seal leakage limits .... does_ not invalidate the acceptability of these overall leakage determination unless their contribution to the overall Type A, B, and C leakage causes that to exceed limits." The basis for this deletion is that the limit for air-lock leakage is the same limit as the limit for combine Type B and C leakage and because there is no unique leakage limit specified for purge valves with resilient seals. ' . Palisades Nuclear Plant Page 2of2 05/31/99 . I

  • *
  • ATTACH1\1ENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS Change Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has been provided.
  • 2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of intent. 3. The requirement/statement has been deleted since it is not applicable to this facility.

4. The following requirements have been renumbered, where applicable, to reflect this deletion.

Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.

5. This change reflects the current licensing basis/technical specification.

The Palisades CTS note is .more explicit to clarify that a door which has been "locked" in order to* comply with ACTIONS can be used to perform repairs on the affected components. This is consistent with the application of the note and removes any confusion when applying this note to a door which has locked. 6. ISTS SR 3.6.2.1 (ITS SR 3.6.2.1) has been revised to reflect certain aspects of air lock testing presently contained in the CTS. This includes;

1) the addition of Note 3 which allows the emergency escape doors to be opened for strongback removal and performance of a seal contact check following testing, 2) explicit acceptance criteria for the emergency escape air lock doors, and 3) reference to Option A of 10 CFR 50, Appendix J. The Bases for ISTS SR 3.6.2.1 (ITS SR 3.6.2.1) were revised to provide conforming changes to the actual SR, and to include a discussion on the NRC staff approved exemptions to 10 CFR 50, Appendix J as they relate to the containment air locks . Palisades Nuclear Plant Page 1of2 05/31/99
  • *
  • Change 7. Not used. Discussion ATTACIThfENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.2, CONTAINMENT AIR LOCKS 8. The Palisades CTS does not contain a surveillance to test the interlock mechanism of the airlock. SR 3.6.2.2 is added in the proposed ITS. TSTF-17, Rev. 1, revises the Frequency of performing the surveillance from 184 days to 24 months to reflect that the surveillance is performed in the conditions that apply in a plant outage where the airlock is primarily used. For Palisades the Frequency is changed to 18 months to correspond with the current cycle lengths. The proposed changes to NUREG-1432 reflect the wording to perform the surveillance at the 18 month Frequency.

This change is consistent with TSTF-17, Rev. 1. 9. The Palisades Nuclear Plant is considered to be an "Atmospheric" containment. The heading titles referring to either "Atmospheric" or "Atmospheric and Dual" are deleted since they add no value to the usage of the plant specific Palisades Nuclear Plant Technical Specifications. In addition, the portions of NUREG-1432 which are provided for "dual" containments are not applicable to the Palisades Nuclear Plant . 10 . The NUREG-1432 Bases for the Applicable Safety Analysis is modified to reflect the appropriate information for the Palisades Nuclear Plant. The air lock testing (Type ,B testing) is performed to 10 CFR 50 Appendix J Option A. The wording for the Option A Pa only refers to the peak pressure from a design basis accident. The Bases have been revised to include the maximum calculate peak pressures for both the LOCA and MSLB and to clarify tqat Type "B" testing is performed at containment pressure 55 psig . Palisades Nuclear Plant Page 2of2 05/31/99

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  • Change ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARK.UPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has been provided.
2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of intent. 3. The requirement/statement has been deleted since it is not applicable to this facility.

4. The following requirements have been renumbered, where applicable, to reflect*this . deletion . Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
5. This change reflects the current .licensing basis/technical specification.
  • 6. The Palisades purge and exhaust system consists of a supply through a 12 inch air room supply path and two 8 inch exhaust paths. Previously the Palisades Nuclear Plant design contained "large" purge valves but they have been removed from the plant design. The NUREG-1432 portions dealing with the [42] inch purge valves is not included in the proposed ITS. Each path of the existing purge exhaust and air room supply contains two isolation valves with resilient seals. The primary difference between the use of the purge and exhaust path at Palisades and NUREG-1432 portions dealing with the "mini-purge" valves is that at Palisades the purge exhaust valves and air room supply valves are locked closed in MODES 1-4 and are not allowed to be opened. Therefore, for Palisades the valves are inoperable if the valves are open, or closed but not locked, since they have not been qualified to be able to close in the event of a Design Basis Accident.

Verification that the containment purge exhaust and air room supply valves are closed is accomplished by performing a leakage rate test every 6 months. NUREG-1432 also that a leakage rate test be performed within 92 days after opening the valves. That requirement has not been adopted in the ITS since the containment purge exhaust and air room supply valves are not allowed to be opened in MODES 1 -4. (continued) Palisades Nuclear Plant Page 1of5 05/31/99

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  • ATTAC1ŽENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Change Discussion
6. (continued)
7. 8. ISTS SR 3. 6. 3 .1 requires a verification that each [ 42] inch purge valve is sealed closed I
  • except for one purge valve in a penetration flow path while in Condition E of the LCO. I That SR was not adopted in the ITS since the Palisades plant design does not include I 42" valves. Rather, ISTS SR 3.6.3.2 was revised to agree with LCO 3.6.3 (CTS 3.q.5) I which requires the purge exhaust valves and air room supply valves to be closed in I Modes 1, 2, 3, and 4. If one of these valves were open, the valve would be declared
  • I inoperable and the Required Actions of Condition D taken (See JFD 12). Proposed ITS I Condition D, unlike ISTS Condition E, does not allow unrestricted operation in I Modes 1, 2, 3, or 4. Thus, the exception contained in ISTS SR 3.6.3.l related to I Condition E does,not apply. Since the purge valves are not used in MODES 1, 2, 3, I or 4 for pressure control, ALARA, surveillance testing, or air quality considerations, I that aspect of ISTS SR 3.6.3.2 has been deleted from the SR. I The purge exhaust and air room supply valves at Palisades close on a high pressure or a high radiation signal. Proposed ITS SR 3.6.3.2 and SR 3.6.3.3 (NUREG-1432 SR 3.6.3.3 and SR 3.6.3.4) incorporate the following statement with respect to verification of manual containment isolation valve and blind flange position: " ... and not locked, sealed, or otherwise secured." In addition, the Bases for proposed SR 3.6.3.2 and 3.6.3.3 has the following statement added: "This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing." This allowance is made since the purpose of the. surveillance is to verify valves which might become mispositioned.
  • If the manual containment isolation valve or blind flange is locked, sealed, or otherwise secured, this helps to ensure the valve or blind flange is not inadvertently mispositioned.

This generic change to NUREG-1432 was proposed by the industry owner's groups in TSTF-45, R.1 and subsequently approved by the NRC on 12/19/96 . Palisades Nuclear Plant Page 2 of 5 05/31/99

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  • ATTAC1ŽENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Change Discussion
9. Proposed ITS SR 3.6.3.4 (NUREG-1432 SR 3.6.3.5) revises the SR from "Verify the isolation time of each power operated and each automatic containment isolation valve is within limits" to "Verify the isolation time of each automatic power operated containment isolation valve is within limits." The Bases for the SR are also revised. This change is made since there may be valves credited as containment isolation valves which are power operated (i.e., can be remotely operated) that do not receive a containment isolation signal. These power operated valves do not have an isolation time as assumed in the accident analyses since they require operator action. Therefore, deleting reference to power operated isolation valve time testing reduces the potential for misinterpreting the requirements of thjs SR while maintaining the assumptions of the accident analysis.

This change is consistent with industry owner's group generic change TSTF-46, R. l. 10. CTS 3.6.5 requires "The containment purge exhaust and air room supply isolation valves shall be locked closed whenever the plant is above COLD SHUTDOWN." This portion of the CTS is.addressed by the proposed ITS LCO 3.6.3 for the containment isolation valves being OPERABLE and the ITS MODES 1-4 are substituted for the "above COLD SHUTDOWN." If one of these valves were not locked shut the CTS requires "With one containment purge exhaust or air room supply isolation valve not locked closed, lock the valve closed within 1 hour." This becomes ACTION Din the proposed ITS 3.6.3. This requirement is included in the proposed ITS because the purge exhaust and air room supply isolation valves are have not been qualified to close following a LOC{\. The proposed Conditions A and B have been modified to reflect this change with wording which states "(except for purge exhaust valve or air room supply valve not locked closed.)" NUREG-1432 includes an exclusion for the purge valve leakage limits to instruct the user that another condition should be entered and so

  • similar wording is added to address the case where a purge exhaust or air room supply valve is not locked closed. These changes represent a plant specific change to reflect the Palisades Nuclear Plant CTS and analysis.

Since purge valve leakage limits are not applicable to Palisades, reference to leakage limits in Conditions A and B have been deleted. 11. The Palisades Nuclear Plant design includes actuating the containment isolation valves by a Containment High Pressure (CHP) signal, or a Containment High Radiation (CHR) signal, or both. Most containment isolation valves receive both signals, however there are some valves which only receive one signal. For example, the component cooling water is only closed op a CHP signal. Therefore, in the proposed ITS, the Bases are revised to repeat the actual SR wording and then identify the applicable signals for clarity. This is a plant specific change to reflect the Palisades Nuclear Plant design. Palisades Nuclear Plant Page 3 of 5 05/31/99

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  • ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Change Discussion
12. CTS 3.6.5 requires that the containment purge exhaust and air room supply isolation valves shall be locked closed whenever the plant is above COLD SHUTDOWN.

This requirement is an implicit requirement in the proposed ITS 3.6.3 since it states "Each containment isolation valve shall be OPERABLE" and the proposed Bases state that the purge exhaust and air room supply valves must be locked closed. Therefore, an additional Action must also be provided in the proposed ITS if this requirement is not met. Proposed ITS Action D addresses the situation where "One purge exhaust or air room supply valve not locked closed" and requires that within 1 hour, the affected valve must be locked closed. This Action is added to the proposed ITS to reflect a plant specific change based on the Palisades Nuclear Plant CTS. Proposed Condition D replaced ISTS 3.6.3 Condition E since failure of the leakage rate test for the containment purge exhaust and air room supply valves indicates the affected valve may 13. not be fully closed.

  • NUREG-1432 Condition Chas a "bracketed" 4 hours as a Completion Time for an inoperable containment isolation valve in a penetration flow path with only one containment isolat.ion valve and a closed system.

Rev. 2, revises this 4 hours to 72 hours. A Completion Time of 72 hours is appropriate since it involves a closed system which minimizes the potential of a leakage pathway. Since the Palisades Nuclear Plant CTS did not provide an explicit Completion Time for this condition, the 72 hours is included in the proposed ITS. TSTF-30, Rev. 2, also included a change to the Bases for Required Action C.l and C.2 which stated: "The closed system must meet the requirements of Reference 3 ("Reference 3" is SRP 6.2.4). Such a Bases statement is not consistent with the current licensing basis and was not incorporated. SRP 6.2.4 does not provide "requirements," but rather provides review guidance for acceptability of the design for containment isolation capability. The containment penetrations which utilize closed systems have previously been determineq to be acceptable. The acceptability may have been based on compliance with SRP 6.2.4 or may have been justified on some other basis. Whatever the basis, the system design has been determined acceptable for use as a containment isolation feature, and it is inappropriate to now prevent such use unless it

  • complies with SRP 6.2.4. Therefore, this change is consistent with TSTF-30, Rev. 2, except where the generic change is inconsistent with the current licensing basis . Palisades Nuclear Plant Page 4 of 5 05/31/99
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  • Change ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.3, CONTAINMENT ISOLATION VALVES Discussion
14. The Palisades Nuclear Plant is considered to be an "Atmospheric" containment.

The heading titles referring tO' either "Atmospheric" or "Atmospheric and Dual" are deleted since they add no value to the usage of the plant specific Palisades Nuclear Plant Technical Specifications. In addition, the portions of NUREG-1432 which are provided for "dual" containments are not applicable to the Palisades Nuclear Plant. 15. The purge exhaust and air room supply isolation valves are all outside containment. Therefore, the Bases is reworded to reflect this where appropriate.

  • Palisades Nuclear Plant Page 5 of 5 05/31199
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  • Change Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.4, CONTAINMENT PRESSURE Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS. " The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has been provided.
2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of intent. 3. The requirement/statement has been deleted since it is not applicable to this facility.

4. The following requirements have been renumbered, where applicable; to reflect this deletion . Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
5. This reflects the current licensing basis/technical specification.
6. The Palisades analysis has two pressure limits depending on whether or not the reactor critical.

The allowed containment pressure is primarily dependent upon the heat addition in the event of a Design Basis Accident (DBA). If the reactor is not critical, then the resulting heat addition to containment as a result of a DBA is lower which allows for a higher initial containment pressure and results in greater operational flexibility. Conversely, once the reactor is critical, the heat addition to containment as , a result of a DBA will be higher and results in the need to maintain a lower initial containment pressure in order to ensure that the peak containment pressure remains below the design value. This analysis is reflected in the TS as a limit of 1.0 psig for containment pressure in MODES 1 and 2, and 1.5 psig in MODES 3 and 4. This analysis was performed to provide a higher margin in MODES 3 and 4 which would provide greater operational flexibility . Palisades Nuclear Plant Page 1of2 05/31/99

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  • Change Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.4, CONTAINMENT PRESSURE 7. The Palisades Nuclear Plant analysis does not address an inadvertent containment spray. The external design pressure of the containment shell is 3 psig. This value is approximately 0.5 psig beyond the maximum external pressure that could be developed if the containment were sealed during a period of low barometric pressure and high temperature and, subsequently, the containment atmosphere were cooled with a concurrent rise in barometric pressure (FSAR 5.8.1). For these reasons, no minimum pressure limit is provided in the CTS and is not proposed in the ITS. This discussion is included in the Bases. This is a plant specific change to reflect the Palisades Nuclear Plant CTS and analysis.
8. The Palisades Nuclear Plant is considered to be an "Atmospheric" containment.

The . heading titles referring to either "Atmospheric" or "Atmospheric and Dual" are deleted since they add no value to the usage of the plant specific Palisades Nuclear Plant Technical Specifications. In addition, the portions of NUREG-1432 which are provided for "dual" containments are not applicable to the Palisades Nuclear Plant . Palisades Nuclear Plant Page 2 of 2 05/31/99

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  • ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.5, CONTAINMENT AIR TEMPERATURE Change Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has been provided.
2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of intent. 3. The requirement/ statement has been deleted since it is not applicable to this facility.

4. The following requirements have been renumbered, where to reflect this deletion . Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
5. This change reflects the current licensing basis/technical specification.
6. The NUREG-1432 ITS LCO Bases discusses that with an_ initial containment average air temperature less than or equal to the LCO temperature limit, "the resultant peak accident temperature is maintained below the containment design temperature." In the proposed ITS, this _is reworded to state "the resultant peak accident pressure is maintained below the containment design pressure." As discussed in the proposed ITS Applicable Safety Analysis Section, containment air temperature does exceed the containment design temperature for a limited period of time. The primary reason for limiting containment air temperature is to ensure that the containment pressure will remain within limits. Therefore, the NUREG-1432 is revised to change temperature to pressure in this Bases discussion.
7. The Palisades Nuclear Plant design does not include alarms to alert the operator to an abnormal containment temperature condition.

Therefore, this statement in the Bases is deleted since it is does not apply . Palisades Nuclear Plant Page 1of2 05/31/99

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  • ATTAC1ŽENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.5, CONTAINMENT AIR TEMPERATURE Change Discussion
8. The Palisades Nuclear Plant is considered to be an "Atmospheric" containment.

The heading titles referring to either "Atmospheric" or "Atmospheric and Dual" are deleted since they add no value to the usage of the plant specific Palisades Nuclear Plant Technical Specifications. In addition, the portions of NUREG-1432 which are provided for "dual" containments are not applicable to the Palisades Nuclear Plant . Palisades Nuclear Plant Page 2 of 2 05/31/99

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  • ATTAC1ŽENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING SYSTEMS Change Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has been provided.
2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of intent. 3. The requirement/statement has been deleted since it is not applicable to this facility. 4 . 5. 6. The following requirements have been renumbered, where applicable, to reflect this deletion. Changes have been made (additions, deletions, and/or changes to. the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description. This change reflects the current licensing basis/technical specification. The Palisades Nuclear Plant design for the Containment Spray System utilizes three Containment Spray Pumps. The Containment Air Recirculation and Cooling System utilizes four Containment Air Coolers but only three of these coolers are safety related. The original design of the system considered the containment spray pumps and the containment air.coolers to be redundant systems. Subsequent analysis required that one containment spray pump be used when the three containment air coolers were being relied upon for heat removal. Therefore, the title of the LCO in the proposed ITS is "Containment Cooling Systems," and the LCO itself refers to a train of containment cooling. A train of containment cooling refers to either the three containment air coolers and containment spray pump 54A which are electrically aligned, or the other two containment spray pumps and air cooler fan V-4A which are electrically aligned. The proposed ITS Condition A is written to reflect that one or more trains of containment cooling can be inoperable as long as the containment cooling capability equivalent to one train exists. This was written to reflect the Palisades Nuclear Plant design that any two containment spray pumps can provide sufficient cooling capability even though the two pumps may be aligned on different "electrical" trains. This is consistent with the application of. the current Palisades Technical Specifications. Palisades Nuclear Plant Page 1of5 05/31/99

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  • Change ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING SYSTEMS Discussion
6. (continued)
7. 8. In addition, this condition is similar to that in NUREG-1432 Section 3.5.2 which would allow one or more trains to be inoperable as long as at least 100 % of the ECCS flow equivalent to a single OPERABLE ECCS train is available.

Specification 3.6.6 requires two trains of containment cooling to be OPERABLE with the components for each train discussed above. The trains alignments are dictated by the associated diesel generator which would provide electrical power in the event of a loss of offsite power. However, from a heat removal standpoint, any' containment spray pump can provide sufficient cooling in conjunction with the containment air coolers to satisfy the analytical requirements. Likewise, any two containment spray pumps can provide sufficient cooling without reliance on the containment air coolers. (Although at least one air cooler fan is required for mixing of the containment atmosphere.) The Completion Time of 72 hours is reduced from the NUREG-1432 allowance of 7 days since the original design diversity does not exist. In addition, the NUREG-1432 Conditions B, C, D, E, and Gare not applicable to the Palisades design since complete redundancy between the containment spray system and the containment air coolers at the system level does not exist as discussed above. The appropriate changes are made to the TS and Bases to reflect these changes. This change is a plant specific change to reflect the Palisades Nuclear Plant design and analysis. Not all containment sump isolation valves are addressed by NUREG-1432 SR 3.6.6.6 as is implied in the Bases. For the Palisades design, only some of the containment sump valves are actuated for the spray pump, and some are actuated for the HPSI which is addressed by proposed SR 3.5.2.5. Where these are the same valves, the Bases statement is appropriate. For Palisades the Applicability of the Containment Spray and Cool,ing Systems is MODES 1, 2, and 3. This is due to the fact that the Shutdown Cooling (SDC) system must be placed service in MODE 4 which requires that the containment spray pump flow path be closed to the SDC heat exchanger and the SDC path be aligned to the SDC heat exchanger .. The TS and Bases are changed to reflect this Applicability and proposed ITS Action Bis revised to require MODE 4 to be reached in 30 hours which places the plant out of the MODE of Applicability. This change is made to reflect the

  • Palisades Nuclear Plant CTS and design . Palisades Nuclear Plant Page 2 of 5 05/31/99
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  • ATTAC1ŽENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING SYSTEMS Change Discussion
9. NUREG-1432 SR 3.6.6B.5 specifies the containment spray pump differential pressure requirements on recirculation flow. The industry owners group's sponsored a generic change TSTF-78 to modify the wording in the brackets contained in SR 3.6.6B.5 to state "developed head at the flow test point is greater than or equal to the required developed head." Therefore, the specific value of the required developed head will not be in the technical specifications but will be contained in the Inservice Testing Program. The Palisades proposed ITS has been revised to incorporate this change. 10. , The Palisades Nuclear Plant is considered to be an "Atmospheric" containment.

The heading titles referring to either "Atmospheric" or "Atniospheric and Dual" are deleted since they add no value to the usage of the plant specific Palisades Nuclear Plant Technical Specifications. lL NUREG-1432 includes LCO 3.6.6A for Containment Spray Systems which take credit for iodine removal and LCO 3.6.6B for Containment Spray Systems which do not take credit for iodine removal. LCO 3.6.6A, which takes credit for iodine removal, is based on the use of sodium hydroxide additjon to the containment spray water as a means of iodine removal. The Palisades Nuclear Plant no longer utilizes the addition of sodium hydroxide to containment spray, but rather uses Trisodium Phosphate (TSP) in containment which is dissolved when fluid is released into containment and then recirculated by the Containment Spray System. Therefore, the proposed Palisades ITS is based on NUREG-1432 3.6.6.B(Credit not taken for iodine removal by the Containment Spray System) because the iodine removal function is facilitated by TSP as opposed to sodium hydroxide which is the model for NUREG-1432 3.6.6.A. The "B" designation is removed from the titles since it serves no purpose in the proposed ITS. The proposed Bases includes wording to reflect the iodine removal function provided by the Containment Spray System in conjunction with the TSP. 12.

  • The Palisades safety analyses identify two limiting events for containment peak pressure and peak temperature respectively.

This information is incorporated into an insert to 13. B 3.6.6, Applicable Safety Analyses, that replaces the STS sentence deleted. The inserted information reflects the intent of the STS bases in that the initial conditions assumed in the limiting events are described. The Palisades Nuclear Plant analysis does not include an explicit evaluation of an inadvertent containment spray event. There is an analysis regarding the external design pressure of containment which is in the proposed Bases for ITS specification 3.6.4. Therefore, the NUREG-1432 discussion of an inadvertent containment spray actuation is not included in the proposed ITS. Palisades Nuclear Plant Page 3 of 5 05/31/99

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  • ATTAC1ŽENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING SYSTEMS Change *Discussion
14. NUREG-1432 requires a verification of the cooling water flow rate to the containment air coolers every 31 days. For the Palisades Nuclear Plant design, this flow rate can only be verified during a shutdown condition due to the valve alignments involved.

Therefore, the NUREG-1432 31 day Frequency is replaced with an 18 month Frequency in the proposed ITS to reflect the test being performed during shutdown conditions. This change is a plant specific change to reflect the Palisades Nuclear Plant design. 15. The Palisades plant was designed prior to issuance of the General Design Criteria (GDC) in 10 CFR 50. Therefore, reference to the GDCs is omitted and appropriately replaced by reference to "Palisades Nuclear Plant design criteria . " The Palisades Nuclear Plant design was compared to the GDCs as they appeared in 10 CPR 50 Appendix A on July 7, 1971. It was this updated discussion, including the identified exemptions, which formed the original plant Licensing Basis for future compliance with 16. 17. the GDCs. , ISTS SR 3.6;6.3 requires the verification that the required flow rate through each containment cooling train is greater than the specified limit. The proposed Palisades ITS modifies this requirement to verify that the total flow rate through the containment air coolers is greater than the specified limit through the,specifled coolers when the system is aligned for accident conditions. This change is required because the Palisades design does not include flow instrumentation on each cooler individually and significant changes to the flow distribution occur as the result of automatic realignment of service water system valves. The flow distribution through the containment air coolers when the accident alignment is governed by the piping configuration of the system, i.e. valves are placed in their wide-open state, flow is isolated to one heat exchanger, and safety related portions of the .system are automatically realigned to support the system safety CTS does not contain a surveillance requirement that corresponds to ISTS SR 3.6.6.3. ISTS 3.6.6.3 has been modified into ITS 3.6.6.4 that requires flow testing of the containment air coolers in a manner that is consistent with the Palisades design, installed instrumentation, and operating practice. The Containment Spray pumps at Palisades provide a required support function for the High Pressure Safety Injection (HPSI) pumps. A new insert to the Background of 3.6.6 is provided to describe the HPSI pump support function associated with the Containment Spray pumps . Palisades Nuclear Plant Page 4 of 5 05/31/99

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  • Change ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.6, CONTAINMENT SPRAY AND COOLING SYSTEMS Discussion
18. The sixth paragraph in the Applicable Safety Analyses discussion has been deleted since this information has been adequately addressed in the fourth paragraph of the Applicable Safety Analyses discussion.

As stated in the Background discussion, the "Containment Air Recirculation and Cooling System" and the "Containment Spray System" are generally addressed as the "Containment Cooling Systems." Since the assumed response times in the containment analysis are the same for the containment air coolers as they are for the containment spray pumps (following a loss of offsite power), the discussion in the Applicable Safety Analyses has been revised to address the generic term "Containment Cooling Systems." For event without a loss of offsite power, no delay time is assumed for the containment air coolers since they are already in operation . Palisades Nuclear Plant Page 5 of 5 05/31/99

  • * ** Change Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.7, SPRAY ADDITIVE SYSTEM Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has been provided.
2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of intent. 3. The requirement/statement has been deleted since it is not applicable to this facility.

4. The following requirements have been renumbered, where applicable, to reflect this deletion . Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
5. This change reflects the current licensing basis/technical specification.
6. ISTS LCO 3.6.7, "Spray Additive System," is not used. Palisades design does not include a Spray Additive System. Therefore, this specification is not applicable to the Palisades Nuclear Plant. Palisades Nuclear Plant Page 1of1 05/31/99
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  • ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.8, HYDROGEN RECOMBINERS Change Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup ofthe NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has been provided.
2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are. editorial in nature and do not involve technical changes or changes of intent. 3. The requirement/statement has been deleted since it is not applicable to this facility.

4. The following requirements have been renumbered, where applicable; to reflect this deletion . Changes have been made (additions, deletions, *and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
5. This change reflects the current licensing basis/technical specification.
6. The Palisades plant was designed prior to issuance of the General Design Criteria (GDC) in 10 CFR 50. Therefore, reference to the GDCs is omitted and appropriately replaced by reference to "Palisades Nuclear Plant design criteria." The Palisades Nuclear Plant design was compared to the GDCs as they appeared in 10 CFR 50 Appendix A on July 7, 1971. *It was this updated discussion, including the identified exemptions, which formed the original plant Licensing Basis for future compliance with the GDCs. 7. The Palisades Nuclear Plant is considered to be an "Atmospheric" containment.

The heading titles referring to either "Atmospheric" or "Atmospheric and Dual" are deleted since they add no value to* the usage of the plant specific Palisades Nuclear Plant Technical Specifications . Palisades Nuclear Plant Page 1of2 05/31/99

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  • ATTAC1ŽENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.8, HYDROGEN RECOMBINERS Change Discussion
8. The Hydrogen Recombiners at the Palisades Nuclear Plant are permanently installed and therefore NUREG-1432 Specification 3.6.8 is used as a model. The parenthetical information in the title regarding "if permanently installed" is also removed since it adds no value to the usage of the plant specific Palisades Nuclear Plant Technical Specifications.
9. The Palisades Nuclear Plant analysis takes credit for the Containment Air Cooler (CAC) post-DBA fans as providing a mixing function of the post-accident containment atmosphere to ensure that local accumulations of hydrogen do not exceed the flammability limit. Therefore, requirements will be discussed in the Bases for the LCO as being required support equipment in order for the Hydrogen Recombiner to be considered OPERABLE.

This is a plant specific change to reflect Palisades Nuclear Plant design and analysis.

10. The proposed ITS adds a "continuity" test as part of proposed ITS SR 3. 6. 7. 3 which is not included in NUREG-1432.

The continuity test measures the resistance of each phase to neutral, permitting a determination of a single phase fault or the opening of a single heater bank. This requirement is adqed to reflect the Palisades Nuclear Plant CTS . Palisades Nuclear Plant Page 2of2 05/31/99

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  • ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS
  • SPECIFICATION 3.6.9, HYDROGEN MIXING SYSTEM Change Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has . been provided.
2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of intent. 3. The requirement/statement has been deleted since it is not applicable to this facility.

4. The following requirements have been renumbered, where applicable, to reflectthis deletion . .Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
5. This change reflects the current licensing basis/technlcal specification.
6. ISTS LCO 3.6.9, "Hydrogen Mixing System (HMS)," is not used. Palisades design does not include a Hydrogen Mixing System. Therefore, this specification is not applicable to the Palisades Nuclear Plant. Palisades Nuclear Plant Page 1of1 05/31/99
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  • L ATTACHMENT 6 JUSTIFICATION

]fOR DEVIATIONS SPECIFICATION 3.6.10, IODINE CLEANUP SYSTEM Change Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS: The Change Numbers correspond to the respective deviation shown on the "NUREG MARK.UPS." . The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.

1. The brackets have been removed and the proper plant specific information or value has been provided.
2. Deviatioris have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of intent. 3. The requirement/statement has been deleted since it is not applicable to this facility.

4. The following requirements have been renumbered, where applicable, to reflect this deletion . Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
5. This 'change reflects the current licensing basis/technical specification.
6. ISTS LCO 3.6.10, "Iodine Cleanup System," is not used. Palisades design does not include an Iodine Cleanup System. Therefore, this specification is not applicable to the Palisades Nuclear Plant. Palisades Nuclear Plant Page 1of1 05/31/99 ' . '
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  • ATTAC1ŽENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.11, SHIELD BUILDING (DUAL) Change Discussion Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific information or value has been provided.
2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of intent.

  • 3. The requirement/statement has been deleted since it is not applicable to this facility.
4. T.he following requirements have been renumbered, where applicable, to reflect this deletion . Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or . analysis description.
5. This change reflects the current licensing basis/technical specification.
6. ISTS LCO 3.6.11, "Shield Building," is not used. Palisades design does not include a Shield Building.

Therefore, this specification is not applicable to the Palisades Nuclear. Plant. Palisades Nuclear Plant Page 1of1 05/31/99

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  • l ATTACIWENT 6 JUSTIFICATlON FOR DEVIATIONS SPECIFICATION 3.6.12, VACUUM RELIEF VALVES (DUAL) Change Discussion Note: This attachment provides a brief discussion of the deviatiorui from NUREG:...1432 that were made to support the development of the Palisades Nuclear Plant ITS. the Change _Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
1. The brackets have been removed and the proper plant specific inforimitiori or value has been provided.
2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of intent. 3. The requirement/statement has been deleted since it is not applicable to this facility.

4. The following requirements have been renumbered, where applicable,*

to reflect this deletion . Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.

5. This change reflects the current. licensing basis/technical specification.
6. ISTS LCO 3.6.12, "Vacuum Relief Valves (Dual)," is not used. Palisades design does not include Vacuum Relief Valves. Therefore, this specification is not applicable to the Palisades Nuclear Plant. Palisades Nuclear Plant Page 1of1 05/31/99
  • * * .. . .. . . ATTAC1ŽENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.6.13, SHIELD BUILDING SYS'.tEM OltTAL) Change Discussion Note: This attachment provides a brief discussion of the deviations froi,ii N-1$.Ed-1432 t,hat were made to support the development of the Palisades Nuclear tlia,nge Numbers correspond to the respective deviation shown on the_"NDREG MARKUPS." The first five justifications were used generically throughout lnarkllp .6f the NUREG. Not all generic justifications are used in each spedtication.
1. The brackets have been removed and the proper plant specific:

inforirtatioh bf vah1e has been provided.

2. Deviations have been made for clarity, grammatical prefererice, or to establish consistency within the Improved Technical Specifications.

These deviations are editorial in nature and do not involve technical changes or changes of iriteiit 3. The requirement/statement has been deleted since it is not to this fadlitf. 4. The following requirements have been renumbered, whete applicable, to refied this deletion . '* ., . Changes have been made (additions, deletions, and/or changes* to the NUREG) to reflect the facility specific nomenclature, number, reference; system description, or analysis description.

  • 5. This change reflects the current licensing basis/technical spedfication*. . 6. ISTS LCO 3.6.13, "Shield Building Exhaust Cleanup System (Dual)," is not used. Palisades design does not include Shield Building EXhaust Cleanup System. Therefore, this specification is not applicable to the Palisades Nuclear Plant. Palisades Nuclear Plant Page 1of1 OS/31V99}}
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