Text

Proposed Tech Specs,Converting Current TSs to Improved TS Consistent w/NUREG-1433
	ML20078G716
Person / Time
Site:	Hatch
Issue date:	11/01/1994
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20078G706	List: ML20078G705; ML20078G716;
References
	RTR-NUREG-1433 NUDOCS 9411160145
	Download: ML20078G716 (521)
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EDWIN I. IIATCll NUCLEAR PLANT l IMPROVED TECllNICA L SPECtFICATIONS 1 REVISION INSERTION INSTRUCTIONS ) REVISION G I l Page Instruction Application of Selection Criteria Cover Sheet Discard Page 9 of14 (Unit 2) Replace Page 12 of14 (Unit 2) Replace Ul Specifications Cover Sheet Discard 3.1-1 Replace , 3.1-3 Replace i 3.3-57 Replace l 3.3-59 Replace 3.5-7 Replace 3.6-35 Replace , 3.6-37 Replace 3.6-39 through 3.6-45 Discard ) 3.6-39 through 3.6-45 Add 3.8-1 Replace 3.8-19 Replace 3.9-11 Replace 3.9-13 Replace UI Bases Cover Sheet Discard B 3.1-1 Replace B 3.1-3 (backed by B 3.1-3 A) Replace B 3.1-4 Add B 3.1-4A Discard B 3.1-7 Replace , B 3.3-173 Replace B 3.3-175 Replace B 3.3179 Replace B 3.3-181 Replace B 3.5-17 Replace B 3.5-19 Replace 1 B 3.5-20A Replace B 3.5-21 Replace B 3.6-75 through 3.6-93 Discard B 3.6-75 through 3.6-93 Add 9411160145 941103 W fDR ADOCK 05000321 i PDR

Revision G Insertion Instructions (continued) I h Instruction UI Hases (continued) B 3.8-3 Replace B 3.8-42 Replace , B 3.9-21 Replace B 3.9-23 Replace B 3.9-24A Add B 3.9-25 Replace B 3.9-27 Replace B 3.9-28A Replace B 3.9-29 Replace U1 Markup of CTSand DOCN Cover Sheet Discard 2 (DOC ITS: 3.1.1) Replace 1 2A (DOC ITS: 3.1.1) Add 3.7-12 (1 of 3) Replace 3.7-12a (2 of 3) Replace 3.7-13 (3 cf 3) Replace 1 (DOC ITS: 3.6.4.1) Replace 2 (DOCITS: 3.6.4.1) Replace 2A (DOC ITS: 3.6.4.1) Insert 3.7-10b (1 of 3) Replace 3.7-11 (2 of 3) Replace 1 (DOC ITS: 3.6.4.3) Replace 2 (DOC ITS: 3.6.4.3) Replace 2A (DOC ITS: 3.6.4.3) Add 3 (DOC ITS: 3.6.4.3) Replace 3 A (DOC ITS: 3.6.4.3) Add 1 (DOC ITS: 3.8.2) Replace UI NSIID Cover Sheet Discard 4 (DOC ITS: 3.6.4.3) Replace 6 (DOCITS: 3.6.4.3) Add 7 (DOC ITS: 3.6.4.3) Add

a. In replacing each CTS page, reference the upper right corner for the appropriate ITS section.

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l Revision G Insertion Instructions (continued) Pm Instruction Unit 2 Specifications l Cover Sheet Dirrard 1 3.1-1 Replace 3.1 -3 Replace 3.3-59 Replace 3.3-61 Replace 3.3-67 Replace > 3.5-7 Replace ; 3.5-9 Replace 3.6-37 through 3.6-59 Discar3 - 3.6-37 through 3.6-50 Add 3.7-9 Replace 3.7-11 Replace 3.7-13 Replace ; 3.7-15 Replace l 3.8-1 Replace 3.8-19 Replace 3.8-21 Replace l 3.8-27 Replace ) 3.8-33 Replace 3.8-39 Replace -i 3.8-43 Replace 3.9-9 Replace 3.9-11 Replace 3.9-13 Replace 3.10-1 Replace l l U2 Improved Bases Cover Sheet Discard B 3.1-1 Replace l B 3.1-3 Replace ! B 3.1-4 Add B 3.1-4A Discard B 3.1-5 Replace B 3.1-7 Replace B 3.3-173 Replace B 3.3-175 Replace B 3.3-177 Replace B 3.3-179 Replace j B 3.3-181 Replace l B 3.3-183 Replace B 3.5-17 Replace B 3.5-19 Replace (3)

Revision G Insertion Instructions (continued) Pam Instruction U1 Bases (continued) B 3.5-21 Replace B 3.6-81 through 3.6-124 Discard B 3.6-81 through 3.6-105 Add B 3.719 Replace B 3.7-21 Replace l B 3.7-27 Replace l B 3.7-29 Replace : B 3.8-3 Replace l B 3.8-39 Replace B 3.8-43 Replace B 3.8-67 Replace B 3.8-69 Replace B 3.8-89 Replace B 3.8-91 Replace l B 3.9-21 Replace i B 3.9 23 Replace ! B 3.9-24A Add B 3.9-25 Replace B 3.9-27 Replace B 3.9-28A Add B 3.9-29 Replace U2 Markun of CTS and DOCN Cover Sheet Discard I (DOC ITS: 3.1.1) Replace 1 A (DOCITS: 3.1.1) Add l 3/4 3-12 (4 of 9) Replace 1 (DOCITS: 3.3.6.2) Replace 3/4 3-58b (3 of10) Replace 3/4 3-44 (7 of10) Replace 1 (DOC ITS: 3.3.7.1) Replace 3 (DOC ITS: 3.3.7.1) Replace 2 (DOC ITS: 3.5.2) Replace 2A (DOC ITS: 3.5.2) Add 1 3/4 6 36 (1 of 1) thru 1 (ITS DOC: Bases) Discard l 3/4 6-36 (1 of 2) thru 1 (ITS DOC: Bases) Add 1 (ITS DOC: 3.7.5) Replace 3/4 8-9 (1 of1) Replace I

a. In replacing each CTS page, reference the upper right corner for the appropriate ITS l

seCtlOn. (4) 1 I

Revision G Insertion Instructions (continued) Pagg Instruction U2 Markun of CTS and DOC (continuedf 1 (DOC ITS: 3.8.2) Replace 2 (DOC ITS: 3.8.2) Replace 3/4 8-9 (2 of 2) Replace 2 (DOC ITS: 3.8.5) Replace 3/4 8-12 (1 of 2) Replace 2 (DOC ITS: 3.8.8) Replace 2 (DOCITS: 3.9.7) Replace 2 (DOCITS: 3.9.8) Replace 2A (DOC ITS: 3.9.8) Add 1 (DOC CTS '3/4 9.5.2) Replace 1 (DOC CTS 3/4 9.5.3) Replace U2 NSIID Cover Sheet Discard I (NSHD ITS: 3.6.41) thru 1 (NSHD CTS 3/4.6.1.4) up to 3.7 tab Discard , 1 (NSHD ITS: 3.6.4.1) thru 1 (NSIID CTS: 3/4.6.1.4) Add NUREG 1433 Comparison Suces Cover Sheet Discard 1 3.1-1 Replace 3.1-3 Replace 3.3-61 Replace 3.3-65 Replace INSERT A for proposed TS 3.3.7.1 Replace 3.3-73 Replace 3.5-7 Replace INSERT A for proposed TS 3.5.2 Discard 3.6-45 thru 3.6-55 (U2 SGT Sys Refueling) Discard 3.6-45 thru 3.6-55 Add 3.7 9 Replace 3.7-11 Replace 3.7-13 Replace 3.7-15 Replace 3.7-17 Replace

a. In replacing each CTS page, reference the upper right corner for the appropriate ITS section.

(5)

Revision G Insertion Instructions (continued) Page Instruction l NUREG 1433 Comparison Specs (continued) ) INSERT LCO 3.8.1 (U1 Version) Replace l INSERT LCO 3.8.1 (U2 Version) Replace I 3.8-17 Replace I INSERT LCO 3.8.2 (Ul Version) Replace l INSERT LCO 3.8.2 (U2 Version) Replace 3.8-19 D a; lace l INSERT LCO 3.8.4 (U2 Version) Replace i 3.8-27 Replace INSERT LCO 3.8.5 (U2 Version) Replace INSERT LCO 3.8.7 (U2 Version) Replace 3.8-39 Replace INSERT LCO 3.8.8 (U2 Version) Replace l 3.9-11 Replace 3.9-13 Replace 1 3.10-1 Replace l NUREG 1433 Bases Cover Sheet Discard B 3.1-1 Replace B 3.1-3 Replace B 3.1-4a (INSERT B4) Add B 3.1-5 Replace B 3.1-Sa (INSERT BS) Add B 3.1-7 Replace B 3.3-183 Replace Insert A for proposed BASES 3.3.6.2 Replace Insert B2 for proposed BASES B 3.3.6.2 Discard B 3.3-185 Replace B 3.3-187 Replace Insert C for proposed BASES B 3.3.6.2 (Unit 2) Discard Insert D for proposed BASES B 3.3.6.2 (Unit 1) Replace B 3.3-189 Replace B 3.3-191 Replace Insert 11 for proposed B ASES B 3.3.6.2 (Unit 1) Replace ; B 3.3-193 Replace B 3.5-17 Replace B 3.5-19 Replace Insert to B 3.5-19 Replace Insert F to BASES 3.5.2 Replace , B 3.6-97 Unit 1 Version to B 3.7 tab Discard i B 3.6-97 to Insert Ref(Behind B 3.6-13) Add B 3.7-19 Replace l l vo l

l Revision G Insertion Instructions (continued) Page Instruction NUREG 1433 Bases (continued) B 3.7-21 Replace B 3.7-25 Replace B 3.7-27 Replace , B 3.7-29 Replace INSERT LCO 1 UI Version Replace INSERT LCO 2 U2 Version Replace B 3.8-35 Replace INSERT LCO 3.8.2 (Unit 1) Replace B 3.8-39 Replace B 3.8-59 Replace B 3.8-61 Replace B 3.8-87 Replace B 3.8-89 Replace B 3.9-25 Replace B 3.9-27 Replace INSERT B 27 Add INSERT D for proposed B ASES 3.9.7 Replace B 3.9-29 Replace B 3.9-31 Replace INSERT B 3.9-31 Replace INSERT F for proposed B ASES 3.9.8 Replace NUREG 1433 J for Deviation I Cover Sheet Discard 1 (ITS 3.1) Replace l 6 (ITS 3.3) Replace 1 (ITS 3.5) Replace 5 (ITS 3.6) Replace 6 (ITS 3.6) Replace l 9 (ITS 3.6) Replace ; 2 (ITS 3.7) Replace j l (ITS 3.9) Replace i 2 (ITS 3.9) Replace I 1 (ITS 3.10) Replace 1 I i (7) l

O APPLICATIONOF SELECTION CRITERIA P O I l O i

D '

, ~

SttWJY DISPOSITION MATRIX PLANT HATCH UNIT 2 Current Unit 2 New Unit 2 Retained / Criterion TS Number Title TS Number for Inclusion Bases for Inclusion / ExclusionI *MC) None Suppression Pool Spray 3.6.2.4 Yes-3 Suppression pool spray functions to limit the effects of a DBA. 3/4.6.3 Primary Containment Isolation valves 3.6.1.3 Yes-3 Isolation valves function to limit DBA consequences. 3/4.6.4 Vacuum Relief 3/4.6.4.1 Suppression Chamber-Drywell vacuum Breakers 3.6.1.8 Yes-3 Suppression chamber - drywell vacuum breaker operation is assumed in the LOCA analysis to limit drywell pressure thereby ensuring primary contalrunent integrity. 3/4.6.4.2 Reactor Building - Suppression Chamber Vacuum 3.6.1.7 Yes-3 Reactor building - suppression chamber vacuum breaker Break ers uperation is relied on to limit negative pressure differen-tial, secondary to ptimary containment, that could challenge pris.ary containment integrity. 3/4.6.5 Secondary Containment 3/4.6.5.1 Secondary Containment Integrity 3.6.4.1 Yes-3 Secondary contairunent integrity is relied on to limit the offsite dose during an accident by ensuring a release to l containment is delayed and treated prior to release to the environment. 3/4.6.5.2 Secondary Containment Automatic 3.6.4.2 Yes-3 Valve operation within time limits establishes se condary l Isolation Dampers containment and limits offsite dose releases to acceptable l values, p 3/4.6.6 Containment Atmosphere Control 3/4.6.6.1 Standby Gas Treatment System 3.3.6.2 Yes-3 SGT operatir4 following a DBA acts to mitigate the 3.6.4.3 consequ.nres of offaite releases. 3/4.6.6.2 Primary Containment Hydrogen Recombiner Systems 3.6.3.1 Yes-3 Operates, post LOCA, to limit hydrogen and oxygen concen- b trations to below explosive concentrations that might otherwise challenge containment integrity. 3/4.6.6.3 Primary Containment Hydrogen Mixing System 3.6.3.3 Yes-3 Same as above. 3/4.6.6.4 Primary Containment Oxygen Concentration 3.6.3.2 Yes-4 Oxygen concentration is limited such that when ccmbined with hydrogen that is post slated to evolve following a LOCA the total explosive gas concentration remains below explo-sive levels. Therefore, containment integrity is main-tained. 3/4.6.6.5 Primary Containment Purge System 3/4.6.6.5.1 Primary Containment Purge valves 3,6.1.3 Yes-3 Isolation valves function to limit DBA consequences. 3/4.6.6.5.2 Primary Containment Fast Acting Dampers 3.6.1.3 Yes-3 Same as abore. Page 9 of 14

b s s V Sl m ARY DISPOSITION MATRIX FLANT HATCH UNIT 2 Current Unit 2 New Unit 2 Retained / Criterion TS Number Title TS Number for Inclusion Bases for Inclusion / Exclusion (a)(c) 3/4.9.4 Decay Time Relocated No Although this LCO satisfied criterion 2, the activities necessary prier to cocunencing movement of irradiated fuel ensure that there will always be 24 hours of subcriticality before movement of any irradiated fuel. Hence this Speci-fication has been relocated. 3/4.9.5 Secondary Containment 3/4.9.5.1 Refueling Floor 3.6.4.1 Yes-3 Secondary containment integrity is relied on to limit the / offsite dose during a fuel handling accident by ensuring l A) the release to containment is delayed and treated prior to release to the environment. 3/4.9.5.2 Secondary Containment Automatic Isolation Dampers 3.6.4.2 Yes-3 Valve operation within time limits establishes secondary l containment and limits offsite dose releases to acceptable values. 3/4.9.5.3 Standby Gas Treatment System 3.6.a.3 Yes-3 Operation following a fuel handling accident acts to l j'~z mitigate the consequences c~ offsite releases. 3/4.9.6 Comnunications Relocated No See Appendix A. Page 19. 3/4.9.7 Crane and Holst Operability Relocated No See Appendix A Page 20. 3/4.9.8 Crane Travel - Spent Fuel Storage Fool Relocated No See Appendir A. Tage 21. 3/4.9.9 Water Level - Reactor vessel 3.9.6 Yes-2 A minimum amount of water is required to assure adequate scrubbing of fission products following a fuel handling accident. 3/4.9.10 Water Level - Spent Fuel Storage Pool 3.7.8 Yes-2 Same as above. 3/4.9.11 Control Rod Removal 3/4.9.11.1 Single Control Rod Removal 3.10.5 Yes See Note 4. 3/4.9.11.2 Multiple Control Rod Removal 3.10.6 Yes See Note 4. 3/4.9.12 Reactor Coolant Circulation 3.9,7 Yes Does not satisfy the selection criteria, however is being 3.9.8 retained in accordance with the NRC Final Policy Statement on Technical Specification Improvements. 3/4.10 SPECIAL TEST EXCEPTIONS 3.10 3/4.10.1 Primary Containment Integrity Deleted No The latitude of this Special Test Exception is not required at Hatch Unit 2. 3/4.10.2 Rod Worth Minimizer 3.10.7 Yes See Note 4. 3/4.10.3 Shutdown Margin Demonstrations 3.10.8 Yes See Note 4 Page 12 of 14 0

O UNIT 1 IMPROVED TECHNICAL SPECIFICATIONS O l l l l l

I O l l l

l l SDM 3.1.1 ( 3.1 REACTIVITY CONTROL SYSTEMS I 3.1.1 SHUTDOWN MARGIN (SDM) LC0 3.1.1 SDM shall be:

a. 2 0.38% Ak/k, with the highest worth control rod l analytically determined; or

b. 2 0.28% Ak/k, with the highest worth control rod determined by test.

APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limits A.1 Restore SDM to within 6 hours in MODE 1 or 2. limits. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. C. SDM not within limits C.1 Initiate action to Immediately in MODE 3. fully insert all insertable control rods. i D. SDM not within limits D.1 Initiate action to Immediately in MODE 4. fully insert all insertable control rods. AND (continued) , HATCH UNIT 1 3.1-1 REVISION A

SDM 3.1.1 ACTIONS i CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) D.2 Initiate action to 1 hour restore secondary containment to l OPERABLE status. 1 AND 0.3 Initiate action to 1 hour restore required l standby gas treatment (SGT) subsystem (s) to l OPERABLE status. AND D.4 Initiate action to I hour restore isolation capability in each required secondary containment penetration flow path not isolated. E. SDM not within limits E.1 Suspend CORE Immediately in MODE 5. ALTERATIONS except o for control rod insertion and fuel assembly removal. AND E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies. AND (continued) Ol HATCH UNIT 1 3.1-2 REVISIONhG

SDM 3.1.1 t ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. (continued) E.3 Initiate action to I hour restore secondary containment to OPERABLE status. M_D E.4 Initiate action to I hour restore required SGT subsystem (s) to OPERABLE status. AND E.5 Initiate action to I hour restore isolation capability in each required secondary containment penetration flow path not isolated. l O HATCH UNIT 1 3.1-3 REVISION /G

SDM 3.1.1 SURVEILLANCE REQUIREMENTS SVRVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM is: Prior to each in-vessel fuel

a. 2 0.38% M /k with the highest worth movement during control rod analytically determined; fuel loading or sequence

b. 2 0.28% & /k with the highest worth AND control rod determined by test.

Once within 4 hours after criticality following fuel movement within the reactor pressure vessel or control rod repl acement O 1 O HATCH UNIT 1 3.1-4 REVISION A

Secondary Containment Isolation Instrumentation 3.3.6.2 3.3 INSTRUMENTATION 3.3.6.2 Secondary Containment Isolation Instrumentation LC0 3.3.6.2 The secondary containment isolation instrumentation for each Function in Table 3.3.6.2-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.6.2-1. ACTIONS

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

Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in 12 hours for inoperable. trip. Function 2 AND , 24 hours for Functions other than Function 2 B. One or more automatic B.1 Restore isolation 1 hour Functions with capability. isolation capability not maintained. C. Required Action and C.l.1 Isolate the 1 hour associated Completion associated Time of Condition A penetration flow or B not met. path (s). 0_8 (continued) O HATCH UNIT 1 3.3-S7 REVISION \(7

Secondary Containment Isolation Instrumentation 3.3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.I.2 Declare associated I hour secondary containment isolation valves inoperable. AND C.2.1 Place the associated I hour standby gas treatment (SGT) subsystem (s) in operation. 0_8 C.2.2 Declare associated I hour SGT subsystem (s) inoperable. O SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment Isolation Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function A maintains isolation capability. I OD SURVEILLANCE FREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK. 12 hours (continued)

O HATCH UNIT 1 3.3-58 REVISION D

Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.6.2.3 Perform CHANNEL CALIBRATION. 92 days SR 3.3.6.2.4 Perform CHANNEL CALIBRATION. 18 months SR 3.3.6.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months O l l O HATCH UNIT 1 3.3-59 REVISION A

i 1 l Secondary Containment Isolation Instrumentation l 3.3.6.2 i GI t Table 3.3.6.2-1 (page 1 of 1) Secondary Containment Isolation Instrunentation ' l APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE

1. Reactor vessel Water 1,2,3, 2 SR 3.3.6.2.1 2 -47 inches Level -Low Low, Level 2 (a) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.5

2. Drywell Pressure -High 1,2,3 2 SR 3.3.6.2.1 s 1.92 psig SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.5

3. Reactor Building F.xhaust 1,2,3, 2 SR 3.3.6.2.1 5 20 nR/hr Radiation -High (a) SR 3.3.6.2.3 SR 3.3.6.2.5

4. Refueling Floor Exhaust 1,2,3, 2 SR 3.3.6.2.1 5 20 nR/hr R adi at ion -H igh SR 3.3.6.2.3

                                      $(a) (b)                                                             l SR   3.3.6.2.5 (a) During operations with a potential for draining the reactor vessel.

(b) During CORE ALTERATIONS and during movement of irradiated fuel assenblies in secondary containment. e1 l HATCH UNIT 1 3.3-60 REVISION [(f 1 _a

ECCS - Shutdown 3.5.2 O Cj 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION l COOLING (RCIC) SYSTEM l t ! 3.5.2 ECCS - Shutdown LCO 3.5.2 Two low pressure ECCS injection / spray subsystems shall be OPERABLE. i APPLICABILITY: MODE 4, MODE 5, except with the spent fuel storage pool gates removed and water level 2: 22 ft 1/8 inches over the top of the reactor pressure vessel flange. ; ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required ECCS A.1 Restore required ECCS 4 hours injection / spray injection / spray subsystem inoperable, subsystem to OPERABLE

 /~T                                          status.

UI B. Required Action and B.1 Initiate action to immediately associated Completion suspend operations Time of Condition A with a potential for I not met, draining the reactor vessel (0PDRVs). C. Two required ECCS C.1 Initiate action to Immediately injection / spray suspend OPDRVs. subsystems inoperable. AND C.2 Restore one ECCS 4 hours . injection / spray ! subsystem to OPERABLE j status. 1 (continued) HATCH UNIT 1 3.5-7 REVISION A 1

ECCS - Shutdown 3.5.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action C.2 0.1 Initiate action to Immediately and associated restore secondary Completion Time not containment to met. OPERABLE status. AND D.2 Initiate action to Immediately restore required l standby' gas treatment subsystem (s) to l OPERABLE status. AND D.3 Initiate action to Immediately restore isolation capability in each required secondary containment penetration flow path not isolated. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required low pressure 12 hours coolant injection (LPCI) subsystem, the suppression pool water level is 2 146 inches. (continued) l O HATCH UNIT 1 3.5-8 REVISIONhC, l l

Primary Containment Oxygen Concentration 3.6.3.2 l l l 3.6 CONTAINMENT SYSTEMS 3.6.3.2 Primary Containment Oxygen Concentration LCO 3.6.3.2 The primary containment oxygen concentration shall be

                    < 4.0 volume percent.

l 1 APPLICABILITY: MODE 1 during the time period: i

a. From 24 hours after THERMAL POWER is > 15% RTP following startup, to 1

24 hours prior to reducing THERMAL POWER to < 15% RTP b. prior to the next scheduled reactor shutdown. ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME l A. Primary containment A.1 Restore oxygen 24 hours p oxygen concentration not within limit. concentration to within limit. l 1 B. Required Action and B.1 Reduce THERMAL POWER 8 hours i associated Completion to s 15% RTP. Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.2.1 Verify primary containment oxygen 7 days concentration is within limits. O HATCH UNIT 1 3.6-35 REVISION A

Secondary Containment 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment LC0 3.6.4.1 The secondary containment shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs). ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment A.1 Restore secondary 4 hours inoperable in MODE 1, containment to 2, or 3. OPERABLE status. O B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met. B.2 Be in MODE 4. 36 hours C. Secondary containment C.1 --------NOTE--------- inoperable during LC0 3.0.3 is not movement of irradiated applicable. fuel assemblies in the --------------------- l secondary containment, I during CORE Suspend movement of Immediately i ALTERATIONS, or during irradiated fuel OPDRVs. assemblies in the i secondary ' containment, blLD (continued) O HATCH UNIT 1 3.6-36 REVISION A

Secondary Containment l 3.6.4.1 1 i r~ b ACTIONS COMPLETION TIME CONDITION REQUIRED ACTION l i C. (continued) C.2 Suspend CORE Immediately ALTERATIONS. AND C.3 Initiate action to Immediately suspend OPORVs. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify all secondary containment 31 days ' equipment hatches are closed and sealed. N:J SR 3.6.4.1.2 Verify each secondary containment access 31 days door is closed, except when the access opening is being used for entry and exit, then at least one door shall be closed. SR 3.6.4.1.3 ------------------NOTE------------------- The number of standby gas treatment (SGT) subsystem (s) required for this Surveillance is dependent on the secondary containment configuration, and shall be one less than the number required to meet LCO 3.6.4.3, " Standby Gas Treatment (SGT) System," for the given configuration. Verify required SGT subsystem (s) will 18 months on a draw down the secondary containment to STAGGERED TEST 2 0.25 inch of vacuum water gauge in BASIS s 120 seconds. O (continued) HATCH UNIT 1 3.6-37 REVISION q {

Secondary Containment 3.6.4.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.4.1.4 ------------------NOTE------------------- The number of SGT subsystems required for this Surveillance is dependent on the secondary containment configuration, and shall be one less than the number required to meet LC0 3.6.4.3, " Standby Gas Treatment (SGT) System," for the given configuration. Verify required SGT subsystem (s) can 18 months on a l maintain 2 0.25 inch of vacuum water STAGGERED TEST gauge in the secondary containment for BASIS 1 hour at a flow rate s 4000 cfm for each j subsystem. O O HATCH UNIT 1 3.6-38 REVISION K (y

I l SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) LC0 3.6.4.2 Each SCIV shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs). ACTIONS

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

1. 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 systems made inoperable by SCIVs. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected 8 hours penetration flow paths penetration flow path with one SCIV by use of at least inoperable. one closed and de-activated automatic valve, closed manual [ valve, or blind flange. AND (continued) O HATCH UNIT 1 3.6-39 REVISION D

SCIVs 3.6.4.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 --------NOTE--------- Isolation devices in high radiation areas may be verified by use of administrative means. Verify tha affected Once per 31 days penetration flow path is isolated. B. One or more B.1 Isolate the af':cted 4 hours penetration flow paths penetration flow path with two SCIVs by use of at least inoperable. one closed and de-activated automatic _ valve, closed manual valve or blind fl ange. C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND or B not met in MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours (continued) l O HATCH UNIT 1 3.6-40 REVISION D

SCIVs 3.6.4.2 [] ACTIONS (continued) \s CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.I --------NOTE--------- associated Completion LC0 3.0.3 is not Time of Condition A applicable. or B not met during --------------------- movement of irradiated fuel assemblies in the Suspend movement of Immediately secondary containment, irradiated' fuel during CORE assemblies in the ALTERATIONS, or during secondary OPDRVs. containment. A!4D D.2 Suspend CORE Immediately ALTERATIONS. AND D.3 Initiate action to Immediately suspend OPDRVs. O HATCH UNIT I 3.6-41 REVISION A

SCIVs 3.6.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY ) SR 3.6.4.2.1 ------------------NOTES------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for SCIVs that i are open under administrative controls.

Verify each secondary containment 31 days isolation manual valve and blind flange that is required to be closed during accident conditions is closed. SR 3.6.4.2.2 Verify the isolation time of each power 92 days operated and each automatic SCIV is ' within limits. I O SR 3.6.4.2.3 Verify each automatic SCIV actuates to 18 months the isolation position on an actual or , simulated actuation signal. j l l l 9 HATCH UNIT 1 3.6-42 REVISION A

SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System LC0 3.6.4.3 The Unit 1 and Unit 2 SGT subsystems required to support LC0 3.6.4.1, " Secondary Containment" shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs). ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required Unit 1 A.1 Restore required 30 days from SGT subsystem Unit 1 SGT subsystem discovery of inoperable while: to OPERABLE status, failure to meet O' the LC0

1. Four SGT subsystems required OPERABLE, and

2. Unit I reactor building-to-refuel floor plug not installed.

B. One required Unit 2 B.1 Restore required SGT 7 days SGT subsystem subsystem to OPERABLE i inoper:ble. status. AND l OR 30 days from l discovery of j One required Unit I failure to meet i SGT subsystem the LC0 j inoperable for reasons i other than Condition A. O (continued) i l HATCH UNIT 1 3.6-43 REVISION

SGT System 3.6.4.3 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours l associated Completion Time of Condition A or AND B not met in MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours D. Required Action and ------------NOTE------------- l associated Completion LC0 3.0.3 is not applicable. Time of Condition A or ----------------------------- B not met during movement of irradiated D.1 Place remaining Immediately fuel assemblies in the OPERABLE SGT secondary containment, subsystem (s) in during CORE operation. ALTERATIONS, or during OPDRVs. OR (continued) O O HATCH UNIT 1 3.6-44 REVISION /h

SGT System 3.6.4.3 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) D.2.1 Suspend movement of Immediately l irradiated fuel assemblies in secondary containment. AND D.2.2 Suspend CORE Immediately l ALTERATIONS. AND D.2.3 Initiate action to Immediately l suspend OPDRVs. E. Two or more required E.1 Enter LC0 3.0.3 Immediately l SGT subsystems oper le in MODE I, F. Two or more required F.1 --------NOTE--------- l SGT subsystems LC0 3.0.3 is not inoperable during applicable. movement of irradiated --------------------- fuel assemblies in the secondary containment, Suspend movement of Immediately during CORE irradiated fuel ALTERATIONS, or during assemblies in OPDRVs. secondary containment. AND F.2 Suspend CORE Immediately l ALTERATIONS. AND F.3 Initiate action to Immediately l g suspend OPDRVs. t w HATCH UNIT 1 3.6-45 REVISION \G

SGT System 3.6.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each required SGT subsystem for 31 days 2 10 continuous hours with heaters operating. SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP). SR 3.6.4.3.3 Verify each required SGT subsystem 18 months actuates on an actual or simulated initiation signal. O HATCH UNIT 1 3.6-46 REVISION /Gr

AC Sources - Operating 3.8.1 ['O ) 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating LC0 3.8.1 The following AC electrical power sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the Unit 1 onsite Class lE AC Electrical Power Distribution System;

b. Two Unit I diesel generators (DGs);

c. The swing DG;

d. One Unit 2 DG capable of supplying power to one Unit 2 l Standby Gas Treatment (SGT) subsystem required by LC0 3.6.4.3, "SGT System;" and

e. One qualified circuit between the offsite transmission network and the Unit 2 onsite Class IE AC Electrical Power Distribution subsystem (s) needed to support the Jnit 2 SGT subsystem (s) required by LC0 3.6.4.3. l APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite A.1 Perform SR 3.8.1.1 1 hour circuit inoperable. for OPERABLE required offsite circuits. AND Once per 8 hours thereafter AND (continued) I O l HATCH UNIT 1 3.8-1 REVISION \(, l

AC Sources - Operating 3.8.1 CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 Declare required 24 hours from featurc(s) with no discovery of no offsite power offsite power to available inoperable one 4160 V ESF when the redundant bus concurrent required feature (s) with are inoperable. inoperability of redundant required feature (s) AND A.3 Restore required 72 hours offsite circuit to OPERABLE status. MQ 10 days from discovery of failure to meet LC0 3.8.1.a, b, or c B. One Unit 1 or the B.1 Perform SR 3.8.1.1 1 hour swing DG inoperable. for OPERABLE required offsite circuit (s). AND Once per 8 hours thereafter MQ B.2 Declare required 4 hours from feature (s), supported discovery of by the inoperable DG, Condition B inoperable when the concurrent with redundant required inoperability of feature (s) are redundant inoperable. required feature (s) atLD (continued) HATCH UNIT 1 3.8-2 REVISION A I i l

AC Sources - Operating 3.8.1 O O j O HATCH UNIT 1 3.8-19 REVISION A.'

AC Sources - Shutdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS 3.8.2 AC Sources - Shutdown LC0 3.8.2 The following AC electrical power sources shall be OPERABLE:

a. One qualified circuit between the offsite transmission network and the onsite Unit 1 Class 1E AC electrical power distribution subsystem (s) required by LC0 3.8.8,

                      " Distribution Systems - Shutdown;"

b. One Unit I diesel generator (DG) capable of supplying one subsystem of the onsite Unit 1 Class lE AC electrical power distribution subsystem (s) required by LC0 3.8.8;

c. One qualified circuit between the offsite transmission l network and the onsite Unit 2 Class IE AC electrical power distribution subsystem (s) needed to support the Unit 2 Standby Gas Treatment (SGT) subsystem (s) required l by LC0 3.6.4.3, "SGT System;" and

d. One Unit 2 DG capable of supplying one Unit 2 SGT l subsystem required by LC0 3.6.4.3.

APPLICABILITY: MODES 4 and 5, During movement of irradiated fuel assemblies in the secondary containment. O HATCH UNIT 1 3.8-20 REVISIONKL,

RHR -High Water Level 3.9.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3 Initiate action to Immediately restore required l standby gas treatment subsystem (s) to l OPERABLE status. AND B.4 Initiate action to Immediately restore isolation capability in each required secondary containment penetration flow path not isolated. C. No RHR shutdown C.1 Verify reactor 1 hour from cooling subsystem in coolant circulation discovery of no [ operation, by an alternate reactor coolant method. circulation AND Once per 12 hours thereafter AND C.2 Monitor reactor Once per hour coolant temperature. l l 1 l O HATCH UNIT 1 3.9-11 REVISIONK(.,, '

4 RHR -High Water Level l 3.9.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l SR 3.9.7.1 Verify one RHR shutdown cooling subsystem 12 hours , is operating. l l O HATCH UNIT 1 3.9-12 REVISION A

RHR - Low Water Level 3.9.8 3.9 REFUELING OPERATIONS l 3.9.8 Residual Heat Removal (RHR) - Low Water Level LC0 3.9.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and one RHR shutdown cooling subsystem shall be in operation.

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

The required operating shutdown cooling subsystem may be removed from operation for up to 2 hours per.8 hour period. APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level < 22 ft 1/8 inches above the top of the RPV flange. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME O, A. One or two required A.1 Verify an alternate I hour : RHR shutdown cooling method of decay heat subsystems inoperable, removal is available AND for each inoperable - required RHR shutdown Once per i cooling subsystem. 24 hours thereafter B. Required Action and B.1 Initiate action to Immediately associated Completion restore secondary Time of Condition A containment to , not met. OPERABLE status. ' AND B.2 Initiate action to Immediately restore required l standby gas treatment 4 subsystem (s) to l OPERABLE status. 1 AND . (continued) HATCH UNIT 1 3.9-13 REVISIONKG,

l l RHR - Loe Water Level 3.9.8 i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3 Initiate action to Immediately l restore isolation capability in each jh required secondary containment penetration flow path not isolated. C. No RHR shutdown C.1 Verify reactor 1 hour from cooling subsystem in coolant circulation discovery of no operation. by an alternate reactor coolant method. circulation AND Once per 12 hours thereafter g' AND C.2 Monitor reactor Once per hour I coolant temperature. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 9.8.1 Verify one RHR shutdown cooling subsystem 12 hours is operating. O HATCH UNIT 1 3.9-14 REVISION D 1 I

UNIT 1 IMPROVED BASES O i O

SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM) BASES BACKGROUND SDM requirements are specified to ensure:

a. The reactor can be made subcritical from all operating conditions and transients and Design Basis Events;

b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits; and

c. The reactor will be maintained sufficiently subcritical-to preclude inadvertent criticality in the shutdown condition.

These requirements are satisfied by the control rods, as described in GDC 26 (Ref. 1), which can compensate for the reactivity effects of the fuel and water temperature changes experienced during all operating conditions. APPLICABLE SHUTDOWN MARGIN is an explicit assumption in several of the SAFETY ANALYSES evaluations contained in FSAR Chapter 14. The control rod drop accident (CRDA) analysis (Refs. 2 and 3) assumes the core is subtritical with the highest worth control rod withdrawn. Typically, the first control rod withdrawn has a very high reactivity worth and, should the core be critical during the withdrawal of the first control rod, the consequences of a CRDA could exceed the fuel damage limits for a CRDA (see Bases for LC0 3.1.6, " Rod Pattern Control"). Also, SDM is assumed as an initial condition for the control rod removal error during refueling (Ref. 4) and fuel assembly insertion error during refueling (Ref. 5) accidents. The analysis of these reactivity insertion events assumes the refueling interlocks are OPERABLE when the reactor is in tt.e refueling mode of operation. These interlocks prevent the withdrawal of more than one control rod from the core during refueling. (Special consideration and requirements for multiple control rod withdrawal during refueling are covered in Special Operations LC0 3.10.6,

                    " Multiple Control Rod Withdrawal - Refueling.") The                                          i

) (cortinued) HATCH UNIT 1 B 3.1-1 REVISION A I

1 SDM I B 3.1.1 BASES 1 APPLICABLE analysis assumes this condition is acceptable since the core I SAFETY ANALYSES will be shut down with the highest worth control rod (continued) withdrawn, if adequate SDM has been demonstrated. Prevention or mitigation of reactivity insertion events is necessary to limit energy deposition in the fuel to prevent significant fuel damage, which could result in undue release of radioactivity. Adequate SDM ensures inadvertent criticalities and potential CRDAs involving high worth , control rods (namely the first control rod withdrawn) will ' not cause significant fuel damage. SDM satisfies Criterion 2 of the NRC Policy Statement (Ref. 9). l LCO The specified SDM limit accounts for the uncertainty in the demonstration of SDM by testing. Separate SDM limits are provided for testing where the highest worth control rod is determined analytically or by measurement. This is due to the reduced uncertainty in the SDM test when the highest worth control rod is determined by measurement. When SDM is evaluated by calculations not associated with a test (e.g., to confirm SDM during the fuel loading sequence), additional margin is included to account for uncertainties in the calculation. To ensure adequate SDM during the design process, a design margin is included to account for uncertainties in the design calculations (Ref. 6). APPLICABILITY In MODES 1 and 2, SDM must be provided because subcriticality with the highest worth control rod withdrawn is assumed in the CRDA analysis (Ref. 2). In MODES 3 and 4, SDM is required to ensure the reactor will be held subcritical with margin for a single withdrawn control rod. SDM is required in MODE 5 to prevent an open vessel, inadvertent criticality during the withdrawal of a single control rod from a core cell containing one or more fuel assemblies (Ref.4) or fuel assembly insertion error (Ref. 5). (continued) HATCH UNIT 1 B 3.1-2 REVISION 'A5

SDM l B 3.1.1 j BASES (continued) i ACTIONS M With SDM not within the limits of the LC0 in MODE 1 or 2, i SDM must be restored within 6 hours. Failure to meet the i specified SDM may be caused by a control rod that cannot be inserted. The allowed Completion Time of 6 hours is ; acceptable, considering that the reactor can still be shut l down, assuming no failures of additional control rods to 1 insert, and the low probability of an event occurring during I this interval. l M If tne SDM cannot be restored, the plant must be brought to MODE 3 in 12 hours, to prevent the potential for further ! reductions in available SDM (e.g., additional stuck control rods). The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without I challenging plant systems.

C.1 With SDM not within limits in MODE 3, the operator must ' immediately initiate action to fully insert all insertable control rods. Action must continue until all insertable l control rods are fully inserted. This action results in the least reactive condition for the core. D.l. D.2. D.3. and D.4 With SDM not within limits in MODE 4, the operator must immediately initiate action to fully insert all insertable control rods. Action must continue until all insertable control rods are fully inserted. This action results in the least reactive condition for the core. Action must also be initiated within I hour to provide means for control of potential radioactive releases. This includes ensuring:

1) secondary containment (at least including the Unit I reactor building zone) is OPERABLE; 2) sufficient Standby Gas Treatment (SGT) subsystem (s) are OPERABLE to maintain the secondary containment at a negative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 8; single failure protection is not required while in this ACTION);

and 3) secondary containment isolation capability is (continued) HATCH UNIT 1 B 3.1-3 REVISIONK(; I

SDM B 3.1.1 BASES (continued) available (i.e., at least one secondary containment isolation valve and associated instrumentation 9 (continued) h HATCH UNIT 1 8 3.1-3A REVISIONh6

i SDM B 3.1.1 j l h J BASES ACTIONS D.1. D.2. D.3. and D.4 (continued) : are OPERABLE, or other acceptable administrative controls to assure isolation capability) in each associated secondary containment penetration flow path not isolated that is assumed to be isolated to mitigate radioactive releases. This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE. E.1. E.2. E.3. E.4. and E.5 With SDM not within limits in MODE 5, the operator must immediately suspend CORE ALTERATIONS that could reduce SDM, (e.g., insertion of fuel in the core or the withdrawal of control rods). Suspension of these activities shall not preclude completion of movement of a component to a safe condition. Inserting control rods will reduce the total reactivity and therefore, is excluded from the suspended actions. Removing fuel, while allowable under these Required Actions, should be evaluated for axial reactivity effects before removal. Action must also be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies have been fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and therefore do not have to be inserted. Action must also be initiated within 1 hour to provide means for control of potential radioactive releases. This includes ensuring: 1) secondary containment (at least including the common refueling floor zone) is OPERABLE; (continued) HATCH UNIT 1 B 3.1-4 REVISION 4 (,

SDM B 3.1.1 BASES

2) sufficient SGT subsystem (s) are OPERABLE to maintain the secondary containment at a negative pressure with respect to i the environment (dependent on secondary containment !

configuration, refer to Reference 8; single failure protection is not required while in this ACTION); and 3) secondary containment isolation capability is available (i.e., at least one secondary containment isolation valve and associated instrumentation are OPERABLE, or other acceptable administrative controls to assure isolation capability) in each associated secondary containment penetration flow path not O (continued) HATCH UNIT 1 B 3.1-4A REVISIONd(} j

SDM B 3.1.1 BASES REFERENCES 5. FSAR, Section 14.3.3.4. (continued)

6. FSAR, Section 3.6.5.2.

7. NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel," (revision specified in the COLR).

8. Technical Requirements Manual. l

9. NRC No. 93-102, " Final Policy Statement on Technical l Specification Improvements," July 23, 1993.

O O HATCH UNIT 1 B 3.1-7 REVISION k (,

Reactivity Anomalies l 8 3.1.2 l 1 8 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 Reactivity Anomalies BASES BACKGROUND In accordance with GDC 26, GDC 28, and GDC 29 (Ref. 1), reactivity shall be controllable such that subcriticality is maintained under cold conditions and specified acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Therefore, reactivity anomaly is used as a measure of the predicted versus actual core reactivity during power operation. The continual confirmation of core reactivity is necessary to ensure that the Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity anomaly could be the result of unanticipated changes in fuel reactivity or control rod worth or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus actual core reactivity validates the nuclear methods used in the safety analysis and supports the SDM demonstrations (LC0 3.1.1, " SHUTDOWN MARGIN (SDM)") in . assuring the reactor can be brought safely to cold, subcritical conditions. When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and actual reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable poison, producing zero net reactivity. In order to achieve the required fuel cycle energy output, the uranium enrichment in the new fuel loading and the fuel loaded in the previous cycles provide excess positive reactivity beyond that required to sustain steady state operation at the beginning of cycle (80C). When the reactor is critical at RTP and operating moderator temperature, the excess positive reactivity is compensated by burnable (continued) l!ATCH UNIT 1 B 3.1-8 REVISION A

Primary Containment Isolation Instrumentation I B 3.3.6.1 ( BASES (continued) REFERENCES 1. FSAR, Section 5.2.

2. FSAR, Chapter 14.4.

3. FSAR, Section 3.8.3.

4. NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,"

July 1990.

5. NEDC-30851P-A Supplement 2, " Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.

6. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

(D LJ l HATCH UNIT 1 B 3.3-173 REVISION A

l Secondary Containment Isolation Instrumentation B 3.3.6.2 B 3.3 INSTRUMENTATION B 3.3.6.2 Secondary Containment Isolation Instrumentation BASES BACKGROUND The secondary containment isolation instrumentation automatically initiates closure of appropriate secondary containment isolation valves (SCIVs) and starts the Standby Gas Treatment (SGT) System. The function of these systems, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Refs. I and 2). Secondary containment isolation and establishment of vacuum with the SGT System within the assumed time limits ensures that fission products that leak from primary containment following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be OPERABLE are maintained within applicable limits. The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of secondary containment isolation. Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a secondary containment isolation signal to the isolation logic. Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation logic are (1) reactor vessel water level, (2) drywell pressure, (3) reactor building exhaust high radiation, and (4) refueling floor exhaust high radiation. Redundant sensor input signals from each parameter are provided for initiation of isolation. The outputs of the logic channels in a trip system are arranged into two two-out-of-two trip system logics. Any trip system initiates all SGT subsystems and isolates the automatic isolation valves (dampers) in each secondary containment penetration. Each logic closes at least one of the two valves in each secondary containment penetration and i starts the required SGT subsystems, so that operation of either logic isolates 1 (continued) HATCH UNIT 1 B 3.3-174 REVISION Q l I L

J Secondary Containment Isolation Instrumentation B 3.3.6.2 ( BASES BACKGROUND the secondary containment and provides for the l (continued) necessary filtration of fission products. APPLICABLE The isolation signals generated by the secondary containment SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the LCO, and safety analyses of References 1 and 2 to initiate closure APPLICABILITY of valves and start the SGT System to limit offsite doses. Refer to LC0 3.6.4.2, " Secondary Containment Isolation Valves (SCIVs), and LC0 3.6.4.3, " Standby Gas Treatment (SGT) System," Applicable Safety Analyses Bases for more detail of the safety analyses. The secondary containment isolation instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref. 7). Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the O individual instrumentation channel Functions. Each Function must have the required number of OPERABLE channels with their setpoints set within the specified Allowable Values, as shown in Table 3.3.6.2-1. The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint). A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Each channel must also respond within its assumed response time, where appropriate. Allowable Values are specified for each Function specified in the Table. Nominal trip setpoints are specified in the setpoint' calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated (continued) HATCH UNIT 1 B 3.3-175 REVISION Q

fecondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE device (e.g., trip unit) changes state. The analytic limits SAFETY ANALYSES, are derived from the limiting values of the process LCO, and parameters obtained from the safety analysis. The Allowable APPLICABILITY Values are derived from the analytic limits, corrected for (continued) calibration, process, and some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drif t, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for. In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions when SCIVs and the SGT System are required. The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a function by Function basis.

1. Reactor Vessel Water Level - Low Low. Level 2 Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.

An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The Reactor Vessel Water Level - Low Low, Level 2 Function is one of the Functions assumed to be OPERABLE and capable of providing isolation and initiation signals. The isolation and initiation systems on Reactor Vessel Water Level - Low Low, Level 2 support actions to ensure that any offsite releases are within the limits calculated in the safety analysis (Refs. 3 and 4). Reactor Vessel Water Level - Low Low, level 2 signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel . Four channels of Reactor Vessel Water Level - Low Low, Level 2 Function are available and are required to be OPERABLE to ensure that no (continued) PATCH UNIT 1 8 3.3-176 REVISION A

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 3. 4. Reactor Buildina and Refuelina Floor Exhaust SAFETY ANALYSES, Radiation - Hiah (continued) LCO, and APPLICABILITY The Allowable Values are chosen to ensure radioactive releases do not exceed offsite dose limits. The Reactor Building and Refueling Floor Exhaust Radiation - High Functions are required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, these Functions are not required. The Reactor Building Exhaust Radiation-High Function is also required to be OPERABLE during 0PDRVs (in MODE 4 and MODE 5) because the capability of detecting l radiation releases due to fuel failures (due to fuel uncovery) must be provided to ensure that offsite dose limits are not exceeded. The Refueling Floor Exhaust Radiation-High Function is also required to be OPERABLE during CORE ALTERATIONS, MODE 5 OPDRVs, and movement of l > (N irradiated fuel assemblies in the secondary containment (,) because the capability of detecting radiation releases due to fuel failures (e.g., due to a dropped fuel assembly) must l be provided to ensure that offsite dose limits are not exceeded. ACTIONS A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels. Section 1.3, Completion Times, specifies that once a ; Condition has been entered, subsequent divisions, l subsystems, components, or variables expressed in the ! Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. . Section 1.3 also specifies that Required Actions of the ! Condition continue to apply for each additional failure, I with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel. (continued) HATCH UNIT 1 B 3.3-179 REVISION h l 1

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES l ACTIONS A.1 (continued) Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of '2 hours for l Function 2, and 24 hours for Functions other than Function 2, has been shown to be acceptable (Refs. 5 and 6) to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases). l If the inoperable channel cannot be restored to OPERABLE ' status within the allowable out of service time, the channel i must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where l placing the inoperable channel in trip would result in an undesired isolation), Condition C must be entered and its l Required Actions taken. u O Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped l channels within the same Function result in a complete loss of automatic isolation capability for the associated secondary containment penetration flow path (s) or a complete l loss of automatic initiation capability for the Unit 1 and Unit 2 SGT Systems. A Function is considered to be maintaining secondary containment isolation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate a trip signal from the given Function on a valid signal. This ensures that one of the two SCIVs in each penetration flow path, and the required Unit I and Unit 2 SGT subsystems can be initiated on an isolation signal from the given Function. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. (continued) h HATCH UNIT 1 B 3.3-180 REVISION %

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS C.l.l. C.I.2. C.2.1. and C.2.2 (continued) If any Required Action and associated Completion Time of Condition A or B are not met, the ability to isolate the secondary containment and start the required Unit 1 and Unit 2 SGT Systems cannot be ensured. Therefore, further actions must be performed to ensure the ability to maintain the secondary containment function. Isolating the associated flow paths (closing the ventilation supply and exhaust automatic isolation dampers) and starting the associated SGT subsystem (s) (Required intended function of theActions C.l.1 and C.2.1)lowsperforms instrumentation and al the operation to continue. Alternately, declaring the associated SCIVs or SGT subsystem (s) inoperable (Required Actions C.l.2 and C.2.2) is also acceptable since the Required Actions of the respective LCOs (LC0 3.6.4.2 and LC0 3.6.4.3) provide appropriate actions for the inoperable components. Since each trip system affects multiple SGT subsystems Required Actions C.2.1 and C.2.2 can be performed independently on each SGT subsystem. That is, one SGT subsystem can be started (Required Action C.2.1) while another SGT subsystem l can be declared inoperable (Required Action C.2.2). One hour is sufficient for personnel to establish required plant conditions or to declare the associated components inoperable without unnecessarily challenging plant systems. SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 5 and 6) assumption of the average time required to perform channel surveillance. That analysis demonstrated (continued) HATCH UNIT 1 B 3.3-181 REVISION k (- l l l

I l Secondary Containment Isolation lnstrumentation l B 3.3.6.2 BASES 1 SURVEILLANCE the 6 hour testing allowance does not significantly reduce REQUIREMENTS the probability that the SCIVs will isolate the associated (continued) penetration flow paths and that the SGT System will initiate when necessary. , SR 3.3.6.2.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter : indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or i something even more serious. A CHANNEL CHECK will detect ' gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is ' outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based on operating experience that , demonstrates channel failure is rare. The CHANNEL CHECK l supplements less formal, but more frequent, checks of j channel status during normal operational use of the displays . associated with channels required by the LCO. SR 3.3.6.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required , channel to ensure that the entire channel will perform the ' intended function. Any setpoint adjustment shall be i consistent with the assumptions of the current plant l specific setpoint methodology.

The Frequency of 92 days is based on the reliability analysis of References 5 and 6. l l (continued) g' HATCH UNIT 1 B 3.3-182 REVISION A

ECCS - Shutdown B 3.5.2 O B 3.5 EMERGENCY CORE C0OLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION l COOLING (RCIC) SYSTEM 1 B 3.5.2 ECCS - Shutdown BASES BACKGROUND A description of the Core Spray (CS) System and the low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System is provided in the Bases for LC0 3.5.1, "ECCS - Operating." APPLICABLE The ECCS performance is evaluated for the entire spectrum of SAFETY ANALYSES break sizes for a postulated loss of coolant accident (LOCA). The long term cooling analysis following a design basis LOCA (Ref. 1) demonstrates that only one low pressure ECCS injection / spray subsystem is required, post LOCA, to maintain adequate reactor vessel water level in the event of an inadvertent vessel draindown. It is reasonable to assume, based on engineering judgment, that while in MODES 4 and 5, one low pressure ECCS injection / spray subsystem can maintain adequate reactor vessel water level. To provide redundancy, a minimum of two low pressure ECCS injection /

 \                      spray subsystems are required to be OPERABLE in MODES 4 and 5.

The low pressure ECCS subsystems satisfy Criterion 3 of the NRC Policy Statement (Ref. 3). l LCO Two low pressure ECCS injection / spray subsystems are required to be OPERABLE. The low pressure ECCS injection / spray subsystems consist of two CS subsystems and two LPCI subsystems. Each CS subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or condensate storage tank (CST) to the reactor pressure vessel (RPV). Each LPCI subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool to the RPV. Only a single LPCI pump is required per subsystem because of the larger injection capacity in relation to a CS subsystem. In MODES 4 and 5, the RHR System cross tie valve is not required to be closed. The necessary portions of the Plant (continued) HATCH UNIT 1 B 3.5-17 REVISION A (. l r l l

1 ECCS - Shutdown l B 3.5.2 l BASES h LC0 Service Water System are also required to provide (continued) appropriate cooling to each required ECCS subsystem. One LPCI subsystem may be aligned for decay heat removal and ; considered OPERABLE for the ECCS function, if it can be manually realigned (remote or local) to the LPCI mode and is not otherwise inoperable. Because of low pressure and low temperature conditions in MODES 4 and 5, sufficient time will be available to manually align and initiate LPCI subsystem operation to provide core cooling prior to postulated fuel uncovery. APPLICABILITY OPEPsABILITY of the low pressure ECCS injection / spray subsystems is required in MODES 4 and 5 to ensure adequate coolant inventory and sufficient heat removal capability for the irradiated fuel in the core in case of an inadvertent draindown of the vessel. Requirements for ECCS OPERABILITY during MODES 1, 2, and 3 are discussed in the Applicability section of the Bases for LC0 3.5.1. ECCS subsystems are not required to be OPERABLE during MODE 5 with the spent fuel storage pool gates removed and the water level maintained at 2 22 ft 1/8 inches above the RPV flange (equivalent to 21 ft of water above the top of irradiated fuel assemblies seated in the spent fuel storage pool racks). This provides sufficient coolant inventory to allow operator action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown. The Automatic Depressurization System is not required to be OPERABLE during MODES 4 and 5 because the RPV pressure is s 150 psig, and the CS System and the LPCI subsystems can provide core cooling without any depressurization of the primary system. The High Pressure Coolant Injection System is not required to be OPERABLE during MODES 4 and 5 since the low pressure ECCS injection / spray subsystems can provide sufficient flow to the vessel. I l (continued) l HATCH UNIT 1 B 3.5-18 REVISION A

l ECCS - Shutdown B 3.5.2 BASES (continued) ACTIONS A.1 and B.1 If any one required low pressure ECCS injection / spray subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status in 4 hours. In this condition, the remaining OPERABLE subsystem can provide sufficient vessel flooding capability to recover from an inadvertent vessel draindown. However, overall system reliability is reduced because a single failure in the remaining OPERABLE subsystem concurrent with a vessel draindown could result in the ECCS not being able to perform its intended function. The 4 hour Completion Time for restoring the required low pressure ECCS injection / spray subsystem to OPERABLE status is based on engineering judgment that considered the remaining available subsystem and the low probability of a vessel draindown event. With the inoperable subsystem not restored to OPERABLE status in the required Completion Time, action must be immediately initiated to suspend operations with a potential for draining the reactor vessel (0PDRVs) to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must Ox continue until OPDRVs are suspended. C.l. C.2. D.I. D.2. and 0.3 With both of the required ECCS injection / spray subsystems inoperable, all coolant inventory makeup capability may be unavailable. Therefore, actions must immediately be initiated to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. One ECCS injection / spray subsystem must also be restored to OPERABLE status within 4 hours. The 4 hour Completion Time to restore at least one low pressure ECCS injection / spray subsystem to OPERABLE status ensures that prompt action will be taken to provide the required cooling capacity or to initiate actions to place the plant in a condition that minimizes any potential fission product release to the environment.

(continued) l HATCH UNIT I B 3.5-19 REVISION A l

l

ECCS - Shutdown B 3.5.2 BASES h ACTIONS C.l. C.2. D.l. D.2. and D.3 (continued) If at least one low pressure ECCS injection / spray subsystem is not restored to OPERABLE status within the 4 hour Completion Time, additional actions are required to minimize any potential fission product release to the environment. This includes ensuring: 1) secondary containment (at least including: the Unit I reactor building zone if in MODE 4; or the common refueling floor zone if in MODE 5) is OPERABLE; 2) sufficient standby gas treatment subsystem (s) are OPERABLE to maintain the secondary containment at a negative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 2; single failure protection is not required while in this ACTION); and 3) secondary containment isolation capability is available (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability) in each associated secondary containment penetration flow path not isolated that is assumed to be isolated to mitigate radioactivity releases. OPERABILITY may be verified by an administrative check, or by examining logs or other information, to determine whether the components are out of service for maintenance or other T reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, the Surveillance may need to be performed to restore the component to OPER'.BLE status. Actions must continue until all required components are OPERABLE. SURVEILLANCE SR 3.5.2.1 and SR 3.5.2.2 REQUIREMENTS The minimum water level of 146 inches required for the suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the CS System and LPCI subsystem pumps, recirculation volume, and vortex prevention. With the suppression pool water level less than the required limit, all ECCS injection / spray subsystems are inoperable unless they are aligned to an OPERABLE CST. l l (continued) ] l HATCH UNIT 1 B 3.5-20 REVISIONfQ ) l l

ECCS - Shutdown B 3.5.2 lv j BASES When suppression pool level is < 146 inches, the CS System is considered OPERABLE only if it can take suction from the CST, and the CST water level is sufficient to provide the required NPSH for the CS pump. Therefore, a verification that either the suppression pool water level is 2146 inches or that CS is aligned to take suction from the CST and the CST contains 2 150,000 gallons of water, equivalent to O O

(continued)

HATCH UNIT 1 B 3.5-/f J(JA REvlSiON.4(7 l l i

ECCS - Shutdown B 3.5.2 O BASES 'd SURVEILLANCE SR 3.5.2.1 and SR 3.5.2.2 (continued) REQUIREMENTS 12 ft, ensures that the CS System can supply at least 50,000 gallons of makeup water to the RPV. The CS suction is uncovered at the 100,000 gallon level. However, as noted, only one requirad CS subsystem may take credit for the CST option during OPDRVs. During OPDRVs, the volume in the CST may not provide adequate makeup if the RPV were completely drained. Therefore, only one CS subsystem is allowed to use the CST. This ensures the other required ECCS subsystem has adequate makeup volume. The 12 hour Frequency of these SRs was developed considering operating experience related to suppression pool water level and CST water level variations and instrument drift 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 suppression pool or CST water level condition. q SR 3.5.2.3. SR 3.5.2.5 and SR 3.5.2.6 b The Bases provided for SR 3.5.1.1, SR 3.5.1.7, and SR 3.5.1.10 are applicable to SR 3.5.2.3, SR 3.5.2.5, and SR 3.5.2.6, respectively. However, the LPCI flow rate requirement for SR 3.5.2.5 is based on a single pump, not the two pump flow rate requirement of SR 3.5.1.7. SR 3.5.2.4 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be (continued) HATCH UNIT 1 8 3.5-21 REVISION A i

l ECCS - Shutdown l B 3.5.2 l l BASES I SURVEILLANCE SR 3.5.2.4 (continued) l REQUIREMENTS inadvertently misaligned, such as check valves. The 31 day Frequency is appropriate because the valves are operated i under procedural control and the probability of their being I mispositioned during this time period is low. In MODES 4 and 5, the RHR System may operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, RHR valves that are required for LPCI subsystem operation may be aligned for decay heat removal. Therefore, this SR is modified by a Note that allows one LPCI subsystem of the RHR System to be considered OPERABLE for the ECCS function if all the required valves in the LPCI flow path can be manually realigned (remote or local) to allow injection into the RPV, and the system is not otherwise inoperable. This will ensure adequate core cooling if an inadvertent RPV draindown should occur. REFERENCES 1. NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 i SAFER /GESTR-LOCA Loss-of-Coolant Analysis," December 1986.

2. Technical Requirements Manual. l

3. NRC No. 93-102, " Final Policy Statement on Technical l Specification Improvements," July 23, 1993. i 1

i I O HATCH UNIT 1 B 3.5-22 REVISION

Secondary Containment B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment BASES BACKGROUND The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA). In conjunction with operation of the Standby Gas Treatment (SGT) System and closure of certain valves whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the ' fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment. The secondary containment is a structure that completely : encloses the primary containment and those components that may be postulated to contain primary system fluid. This structure forms a control volume that serves to hold up and dilute the fission products. O It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g., due to pump and motor heat load additions). The secondary containment encompasses three separate zones: the Unit I reactor building (Zone I), the Unit 2 reactor building (Zone II), and the common refueling floor (Zone III). The secondary containment can be modified to exclude the Unit 2 reactor building (Zone II) provided the following requirements are met:

a. Unit 2 Technical Specifications do not require OPERABILITY of Zone II;

b. All hatches separating Zone III from Zone II are closed and sealed; and

c. At least one door in each access path separating Zone III from Zone II is closed.

4 (continued) l HATCH UNIT 1 B 3.6-75 REVISIONkl-i i 1

l l l Secondary Containment B 3.6.4.1 BASES BACKGROUND Similarly, other zones can be excluded from the secondary (continued) containment OPERABILITY requirement during various plant operating conditions with the appropriate controls. For example, during Unit I shutdown operations, the secondary containment can be modified to exclude the Unit I reactor building (Zone I) (either alcne or in combination with excluding Zone II as described ai>6ve) provided the following requirements are met:

a. Unit 1 is not conducting operations with a potential for draining the reactor vessel (0PDRV);

b. All hatches separating Zone III from Zone I are closed l and sealed; and

c. At least one door in each access path separating Zone III from Zone I is closed.

O l l l (continued) HATCH UNIT 1 B 3.6-7(7 5 A REVISIONh(y

Secondary Containment B 3.6.4.1 l BASES BACKGROUND To prevent ground' level exfiltration while allowing the

         -(continued)    secondary containment to be designed as a conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LC0 3.6.4.2, " Secondary Containment Isolation Valves (SCIVs)," and LC0 3.6.4.3, " Standby Gas Treatment-(SGT) System." When one or more zones are excluded from the secondary containment, the_ specific requirements for the support systems will also change (e.g.,

securing particular SGT or drain isolation valves). APPLICABLE There are two principal accidents for which credit is taken SAFETY ANALYSES for secondary containment OPERABILITY. These are a loss of coolant accident (LOCA) (Ref.1) and a fuel handling l accident inside secondary containment (Ref. 2). The secondary containment performs no active function in response to each of these limiting events; however, its leak ' tightness is required to ensure that the release of ' radioactive materials from the primary containment is. restricted to those leakage paths and associated leakage O rates assumed in the accident analysis and that fission products entrapped within the secondary containment structure will be treated by the Unit 1 and Unit 2 SGT Systems prior to discharge to the environment. Postulated LOCA leakage paths from the primary containment into secondary containment include those into both the reactor building and refueling floor areas (e.g., drywell head leakage). Secondary containment satisfies Criterion 3 of the NRC Policy Statement (Ref. 4). LC0 An OPERABLE secondary containment provides a control volume , into which fission products that bypass or leak from primary ' containment, or are released from the reactor coolant ' pressure boundary components located in secondary containment, can be diluted and processed prior to release to the environment. For the secondary containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and (continued) HATCH UNIT 1 B3.6-2776 REVISIONf(h

i Secondary Containment B 3.6.4.1 l BASES maintained. The secondary containment boundary required to be OPERABLE is dependent on the operating status of both units, as well as the configuration of doors, hatches, refueling floor plugs, SCIVs, and available flow paths to SGT Systems. The required boundary encompasses the zones which can be postulated to contain fission products from accidents required to be considered for the condition of i each unit, and furthermore, must include zones not isolated from the I i l l O (Continued) HATCH UNIT 1 B3.6-K7(p REVISION A

Secondary Containment B 3.6.4.1 BASES [v1 LC0 SGT subsystems being credited for meeting LC0 3.6.4.3. (continued) Allowed configurations, associated SGT subsystem requirements, and associated SCIV requirements are detailed in the Technical Requiremercs Manual (Ref. 3). APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment (the reactor building zone and potentially the refueling floor zone). Therefore, secondary containment l OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY. In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODEA. Therefore, maintaining secondary containment OPEMBLE is not required in MODE 4 or 5 to ensure a control folume, except for other situations for which significant releases of radioactive material can be postulated, such as during OPDRVs, during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment, fNote, moving ( (q) irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3.) Since CORE ALTERATIONS l and movement of irradiated fuel assemblies are only postulated to release radioactive material to the refueling floor zone, the secondary containment configuration may consist of only Zone III during these conditions. I Similarly, during OPDRVs while in MODE 4 (vessel head ' bolted) the release of radioactive materials is only postulated to the associated reactor building, the secondary containment configuration may consist of only Zone I. ACTIONS A_d l If secondary containment is inoperable, it must be restored ] to OPERABLE status within 4 hours. The 4 hour Completion Time providas a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal. (continued) HATCH UNIT 1 B 3. 6-78' -~ 7 REVISIONf(h

Secondary Containment B 3.6.4.1 i BASES h I secondary containment OPERABILITY) occurring during periods where secondary containment is inoperable is minimal. O (continued) < HATCH UNIT 1 B3.6-Xth/;A REVISION'4C 1 ! l

 .- .                         ~                           .                       .

Secondary Containment B 3.6.4.1 BASES ACTIONS B.1 and B.2 (continued) If secondary containment cannot be restored to OPERABLE l

status within the required Completion Time, the plant must be brought to a MODE-in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 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. C.1. C.2. and C.3

                                                                                          'l Movement of irradiated fuel assemblies in the secondary containment, CORE ALTERATIONS, and OPDRVs can be postulated to cause fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the environment. CORE ALTERATIONS and movement of irradiated fuel assemblies must be immediately suspended if the secondary containment is inoperable.

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until 0PDRVs are suspended. Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel ' assemblies while in MODE 4 or 5, LC0 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor ! operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown, i (continued) HATCH UNIT 1 B 3.6 'BQ / g REVISIONK/._

Secondary Containment B 3.6.4.1 BASES (continued) SURVEILLANCE SR 3.6.4.1.1 and SR 3.6.4.1.2 REQUIREMENTS Verifying that secondary containment equipment hatches and access doors are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary containment will not occur. SR 3.6.4.1.1 also requires equipment hatches to be sealed. In this application, the term

                 " sealed" has no connotation of leak tightness. Maintaining secondary containment OPERABILITY requires verifying each door in the access opening is closed, except when the access opening is being used for normal transient entry and exit (then at least one door must remain closed). When the secondary containment configuration excludes Zone I and/or Zone II, these SRs also include verifying the hatches and doors separating the common refueling floor zone from the reactor building (s). The 31 day Frequency for these SRs has been shown to be adequate, based on operating experience, and is considered adequate in view of the other indications of door and hatch status that are available to the operator.

SR 3.6.4.1.3 and SR 3.6.4.1.4 O The Unit 1 and Unit 2 SGT Systems exhausts the secondary l containment atmosphere to the environment through l appropriate treatment equi ent. To ensure that all fission products are treated, SR ' ,.4.1.3 verifies that the appropriate SGT System (s) will rapidly establish and l maintain a pressure in the secondary containment that is , less than the lowest postulated pressure external to the i secondary containment boundary. This is confirmed by demonstrating that the required SGT subsystem (s), will draw l down the secondary containment to 2 0.25 inch of vacuum 1 water gauge in s 120 seconds. This cannot be accomplished l l if the secondary containment boundary is not intact. SR 3.6.4.1.4 demonstrates that the required subsystem (s) can l maintain 2 0.25 inch of vacuum water gauge for 1 hour at a flow rate s 4000 cfm for each SGT subsystem. The 1 hour l test period allows secondary containment to be in thermal equilibrium at steady state (continued) HATCH UNIT 1 B 3.641 q REVISIONh(_

l l Secondary Containment B 3.6.4.1 BASES SURVEILLANCE. SR 3.6.4.1.3 and SR 3.6.4.1.4 (continued) REQUIREMENTS conditions. Therefore, these two tests are used to ensure secondary containment boundary integrity. Since these SRs are secondary containment tests, they need not be performed with each SGT subsystem. The SGT subsystems are tested on a-STAGGERED TEST BASIS, however, to ensure that in addition to the requirements of LC0 3.6.4.3, each SGT subsystem or combination of subsystems will perform this test. The number of SGT subsystems and the required combinations are I dependent on the configuration of the secondary containment ! and are detailed in the Technical Requirements Manual (Ref. 3). 'The Note to SR 3.6.4.1.3 and SR 3.6.4.1.4 specifies that the number of required SGT subsystems be one less than the number required to meet LC0 3.6.4.3, " Standby Gas Treatment System," for the given configuration. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. FSAR, Section 14.4.3.

2. FSAR, Section 14.4.4.

3. Technical Requirements Manual. l

4. NRC No. 93-102, " Final Policy Statement on Technical l Specification Improvements," July 23, 1993.

O HATCH UNIT 1 B 3.6-$ $ o REVISION h 6 i 1 I

SCIVs B 3.6.4.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) g BASES BACKGROUND The function of the SCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs). Secondary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that fission products that leak from primary containment following a DBA, or that are released during certain operations when primary containment is not required to be OPERABLE or take place outside primary containment, are maintained within the secondary containment boundary. The OPERABILITY requirements for SCIVs help ensure that an 4 adequate secondary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. These isolation devices consist of either

'~

passive devices or active (automatic) devices. Manual valves, de-activated automatic valves secured in their closed position, check valves with flow through the valve secured, and blind flanges are considered passive devices. Automatic SCIVs close on a secondary containment isolation signal to establish a boundary for untreated radioactive material within secondary containment following a DBA or other accidents. Other penetrations are isolated by the use of valves in the closed position or blind flanges. APPLICABLE The SCIVs must be OPERABLE to ensure the secondary SAFETY ANALYSES containment barrier to fission product releases is established. The principal accidents for which the secondary containment boundary is required are a loss of coolant accident (Ref.1) and a fuel handling accident inside secondary containment (Ref. 2). The secondary , containment performs no active function in response to l either of these limiting events, but the boundary l (Continued) q HATCH UNIT 1 B 3.6-81 REVISION A l

                   - - . - _ _ _ - _ . - . _                                 _                 ._    _   __.__m._.__m_--_-._-__-----___mw

SCIVs B 3.6.4.2 BASES APPLICABLE established by SCIVs is required to ensure that leakage from SAFETY ANALYSES the primary containment is processed by the Standby Gas (continued) Treatment (SGT) System before being released to the environment. Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped inside secondary containment so that they can be treatei by the SGT System prior to discharge to the environment. SCIVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 4). l LC0 SCIVs form a part of the secondary containment boundary. l The SCIV safety function is related to control of offsite radiation releases resulting from DBAs. The power operated isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves covered by this LCO, along with their associated stroke O times, are listed in Reference 3. The normally closed isolation valves or blind flanges are considered OPERABLE when manual valves are closed, or open in accordance with appropriate administrative controls, automatic SCIVs are de-activated and secured in their closed position, and blind flanges are in place. These passive isolation valves or devices are listed in Reference 3. The SCIVs required to be OPERABLE are dependent on the configuration of the secondary containment (which is dependent on the operating status of both units, as well as the configuration of doors, hatches, refueling floor plugs, and available flow paths to SGT Systems). The required boundary encompasses the zones which can be postulated to contain fission products from accidents required to be considered for the condition of each unit, and furthermore, must include zones not isolated from the SGT subsystems being credited for meeting LCO 3.6.4.3, " Standby Gas Treatment (SGT) System." The required SCIVs are those in penetrations connunicating with the zones required for secondary containment OPERABILITY and are detailed in Reference 3. (continued) HATCH UNIT 1 B 3.6-82 REVISION ( (.

i l SCIVs l B 3.6.4.2 BASES APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore, the OPERABILITY of SCIVs is required. I Ol I (continued) HATCH UNIT 1 B 3.6-BQ 5- A REVISIONh(,.

SCIVs B 3.6.4.2 BASES APPLICABILITY In MODES 4 and 5, the probability and consequences of a LOCA (continued) are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant radioactive releases can be postulated, such as during operations with a potential for draining the reactor vessel (0PDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment. (Note, moving irradiated fucl assemblies in the secondary containment may also occur in MODES 1, 2, and 3). ACTIONS The ACTIONS are modified by three Notes. The first Note allows penetration flow paths to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator, who is in continuous communication with the control room, at the controls of the isolation device. In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated. The second Note provides clarification that for the purpose of this LC0 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 SCIV. Complying with the Required Actions may allow for continued operation, and subsequent inoperable SCIVs are governed by subsequent Condition entry and application of associated Required Actions. The third Note ensures appropriate remedial actions are taken, if necessary, if the affected system (s) are rendered inoperable by an inoperable SCIV. A.1 and A.2 In the event that there are one or more penetration flow paths with one SCIV inoperable, 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 (continued) HATCH UNIT 1 B 3.6-b e 3 REVISION h (,

SCIVs , B 3.6.4.2 BASES 1 ACTIONS A.I and A.2 (continued) , l failure. Isolation barriers that meet this criterion are a closed and de-activated automatic SCIV, a closed manual i valve, and a blind flange. For penetrations isolated in I accordance with Required Action A.1, the device used to i isolate the penetration should be the closest available device to secondary containment. The Required Action must be completed within the 8 hour Completion Time. The specified time period is reasonable considering the time required to isolate the penetration, and the probability of a DBA, which requires the SCIVs tc close, occurring during this short time is very low. For affected penetrations that have been isolated in accordance with Required Action A.1, the affected penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that secondary : containment penetrations required to be isolated following l an accident, but no longer capable of being automatically ' isolated, will be in the isolation position should an event occur. The Completion Time of once per 31 days is appropriate because the isolation devices are operated under administrative controls and the probability of their misalignment is low. This Required Action does not require any testing or device manipulation. Rather, it involves verification that the affected penetration remains isolated. Required Action A.2 is modified by a Note that applies to devices located in high radiation areas and allows them to be verified closed by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment once they have been verified to be in the proper position, is low. B.I With two SCIVs in one or more penetration flow paths l inoperable, the affected penetration flow path must be isolated within 4 hours. The method of isolation must (continued) HATCH UNIT 1 B 3. 6-84 el REVISIONK(,

SCIVs l B 3.6.4.2 I O BASES -.U ACTIONS B.1 (continued) 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 4 hour Completion Time is reasonable considering the time required to isolate the penetration and the probability of a DBA, which requires the SCIVs to close, occurring during this short time, is very low. C.1 and C.2 If any Required Action and associated Completion Time of Condition A or B cannot be met, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 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. D.I. D.2. and D.3 If any Required Action and asso:iated Complet'.on Time of Condition A or 8 are not met, the plant must be placed in a condition in which the LC0 does not apply. I? applicable, CORE ALTERATIONS and the movement of irradiated fuel assemblies in the secondary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until 0PDRVs are suspended. Required Action D.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LC0 3.0.3 would not specify any action. If moving fuel while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. (continued) HATCH UNIT 1 B 3. 6 '85 5L REVISION A

SCIVs B 3.6.4.2 BASES hj 1 ACTIONS D.l. D.2 and D.3 (continued) Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.6.4.2.1 REQUIREMENTS This SR verifies that each secondary containment manual l isolation valve and blind flange that is required to be i closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the secondary containment boundary is within design limits. This SR does not require l any testing or valve manipulation. Rather, it involves i verification that those isolation devices in secondary containment that are capable of being mispositioned are in the correct position. Since these isolation devices are readily accessible to personnel during normal operation and verification of their position is relatively easy, the 31 day Frequency was chosen

                                                                                    &l W  '

to provide added assurance that the isolation devices are in the correct positions. Two Notes have been added to this SR. The first Note applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative controls. Allowing verification by administrative controls 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 isolation devices, once they have been verified to be in the propar position, is low. A second Note has been included to clarify that SCIVs that are open under administrative controls are not required to meet the SR during the time the SCIVs are open. (continued) HATCH UNIT 1 B 3.6- % & REVISION \(_

l SCIVs B 3.6.4.2 ( BASES SURVEILLANCE SR 3.6.4.2.2 REQUIREMENTS (continued) Verifying that the isolation time of each power operated and each automatic SCIV is within limits is required to l demonstrate OPERABILITY. The isolation time test ensures that the SCIV will isolate in a time period less than or equal to that assumed in the safety analyses. The Frequency of this SR was developed based upon engineering judgment and the similarity to PCIVs. SR 3.6.4.2.3 Verifying that each automatic SCIV closes on a secondary l containment isolation signal is required to prevent leakage of radioactive material from secondary containment following a DBA or other accidents. This SR ensures that each automatic SCIV will actuate to the isolation position on a secondary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this SR to provide complete testing of the safety function. The 18 month Frequency is based on the need to perform this Surveillance ,/ under the conditions that apply during a plant outage and ( the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. FSAR, Section 14.3.3.

2. FSAR, Section 14.3.4.

3. Technical Requirements Manual.

4. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

O HATCH UNIT 1 B 3.6- $ ) REVISIONk(, 1 l

SGT System B 3.6.4.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.3 Standby Gas Treatment (SGT) Systrm BASES BACKGROUND The SGT System is required by 10 CFR 50, Appendix A, GDC 41,

                    " Containment Atmosphere Cleanup" (Ref. 1). The function of the SGT System is to ensure that radioactive materials that leak from the primary containment into the secondary containment following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment.

The Unit 1 and Unit 2 SGT Systems each consists of two fully redundant subsystems, each with its own set of dampers, charcoal filter train, and controls. The Unit 1 SGT subsystems' ductwork is separate from the inlet to tin filter train to the discharge of the fan. The rest of the ductwork is common. The Unit 2 SGT subsystems' ductwork is separate except for the suction from the drywell and torus, which is common (however, this suction path is not required l for subsystem OPERABILITY). Each charcoal filter train consists of (components listed in order of the direction of the air flow):

a. A demister or moisture separator;

b. An electric heater;

c. A prefilter;

d. A high efficiency particulate air (HEPA) filter;

e. Two charcoal adsorbers for Unit I subsystems and one charcoal adsorber for Unit 2 subsystems;

f. A second HEPA filter; and

g. An axial vane fan for Unit I subsystems and a centrifugal fan for Unit 2 subsystems.

The sizing of the SGT Systems equipment and components is based on the results of an infiltration analysis, as well as an exfiltration analysis of the secondary containment. The internal pressure of the SGT Systems boundary region is (continued) HATCH UNIT 1 B 3.6-8 REVISIONf(f

SGT System B 3.6.4.3 BASES BACKGROUND maintained at a negative pressure of 0.25 inches water gauge (continued) when the system is in operation, which represents the , internal pressure required to ensure zero exfiltration of air from the building when exposed to a 10 mph wind. The demister is pr vided to remove entrained water in the air, while the electric heater reduces the relative humidity of the airstream to < 70% (Refs. 2 and 3). The prefilter removes large particulate matter, while the HEPA filter removes fine particulate matter and protects the charcoal from fouling. The charecal adsorbers remove gaseous elemental iodine and organic iodides, and the final HEPA filter collects any carbon fines exhausted from the charcoal adsorber. The Unit I and Unit 2 SGT Systems automatically start and operate in response to actuation signals indicative of conditions or an accident that could require operation of the system. Following initiation, all required charcoal filter train fans start. Upon verification that the - required subsystems are operating, the redundant required subsystem is normally shut down. O APPLICABLE The design basis for the Unit 1 and Unit 2 SGT Systems is to SAFETY ANALYSES mitigate the consequences of a loss of coolant accident and fuel handling accidents (Refs. 2 and 3). For all events analyzed, the SGT Systems are shown to be automatically initiated to reduce, via filtration and adsorption, the radioactive material released to the environment. ' The SGT System satisfies Criterion 3 of the NRC Policy Statev at (Ref. 5). l LC0 Following a DBA, a minimum number of SGT subsystems are . required to maintain the secondary containment at a negative pressure with respect to the environment and to process gaseous releases. Meeting the LCO requirements for OPERABLE subsystems ensures operation of a minimum number of SGT ' subsystems in the event of a single active ftilure. The . required number of SGT subsystems is dependent on the configuration required to meet LCO 3.6.4.1, " Secondary Containment." For secondary containment OPERABILITY r (continued) HATCH UNIT 1 B 3.64Q 3 g REVISION A (- l l l

SGT System B 3.6.4.3 BASES consisting of all three zones, the required number of SGT subsystems is four. With secondary containment OPERABILITY consisting of one reactor building and the common refueling floor zones, the required number of SGT subsystem is three. Allowed configurations and associated SGT subsystem requirements are detailed in the Technical Requirements Manual (Ref. 4). 9 (continued) HATCH UNIT 1 B 3.6-91Yi /\ REVISIONk(,

SGT System. B 3.6.4.3 BASES LC0 In addition, with secondary containment in modified (continued) configurations, the SGT System valves to excluded zone (s) are not included as part of SGT System OPERABILITY (i.e., the valves may be secured closed and are not required to open on an actuation signal). APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, Unit 1 and Unit 2 SGT Systems OPERABILITY are required during these MODES. In MODES 4 and 5, the probability and consequences of a LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT Systems in OPERABLE status is not required in MODE 4 or 5, except for-other situations under which significant releases of-radioactive material can be postulated, such as during i operations with a potential for draining the reactor vessel (OPDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment. O ACTIONS A.1 and B.1 With one required Unit 1 or Unit 2 SGT subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status. In this condition, the remaining required OPERABLE l SGT subsystems are adequate to perform the required radioactivity release control function. However, the overall system reliability is reduced because a single failure in one of the remaining required OPERABLE subsystems could result in the radioactivity release control function , not being adequately performed. The 7 and 30 day Completion ! Times are based on consideration of such factors as the availability of the OPERABLE redundant SGT subsystems and the low probability of a DBA occurring during this period. Additionally, the 30 day Completion Time of Required Action A.1 is based on three remaining 0PERABLE SGT subsystems, of which two are Unit 2 subsystems, and the secondary-containment volume in the Unit I reactor building being open . to the common refueling floor where the two Unit 2 SGT ' subsystems can readily provide rapid drawdown of vacuum. Testing and analysis has shown that in this configuration, (continued) HATCH UNIT 1 B3.6-[90 REVISION [bj

SGT System B 3.6.4.3 BASES h even with an additional single failure (which is not necessary to assume while in ACTIONS) the secondary containment volume may be drawn to a vacuum in the time required to support assumptions of analyses. C.1 and C.2 l If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, the 9 (continued) TCH UNIT 1 B 3.6,93'ciC A REVISION [6J

SGT System B 3.6.4.3 BASES v ACTIONS C.1 and C.2 (continued) l plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 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. In the event that a Unit 2 SGT subsystem is the one not restored to OPERABLE status as required by Required Action A.1 or B.1, and:

1. All three zones are required for secondary containment OPERABILITY; arid

2. Unit 2 is shutdown with its Technical Specifications not requiring secondary containment OPERABILITY (i.e.,

not handling irradiated fuel, performing CORE ALTERATIONS, or conducting OPDRV), operation of Unit I can continue provided that the Unit 2 % reactor building zone is isolated from the remainder of secondary containment and the SGT System. In this modified secono'ary containment configuration, only three SGT subsystems are required to be OPERABLE to meet LC0 3.6.4.3, and no limitation is applied to the inoperable Unit 2 SGT subsystem. This in effect is an alternative to restoring the inoperable Unit 2 SGT subsystem, i.e., shut down Unit 2 and isolate its reactor building zone from secondary containment and SGT System. D.l. D.2.1. D.2.2. and D.2.3 During movement of irradiated fuel assemblies in the ; secondary containment, during CORE ALTERATIONS, or d2 ring OPDRVs, when Required Action A.1 or B.1 cannot be completed within the required Completion Time, the remaining required OPERABLE SGT subsystems should immediately be placed in ! operation. This action ensures that the remaining ' subsystems are OPERABLE, that no failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected. (continued) HATCH UNIT 1 B 3.6-M Fl{ REVISIONKL

l SGT System l B 3.6.4.3 l BASES 1 An alternative to Required Action 0.1 is to immediately l l i suspend activities that represent a potential for releasing I radioactive material to the secondary containment, thus placing the plant in a condition that minimizes risk. If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies must immediately be suspended. Suspension of these activities must not preclude completion of movement of a component to a safe position. Also, if applicable, 1 actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions l must continue until OPDRVs are suspended. l l The Required Actions of Condition D have been modified by a l l Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LC0 3.0.3 i would not specify any action. If moving irradiated fuel l assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. O < l l l (continued) HATCH UNIT 1 B 3.6-R'r i A REVISION % l

1 i SGT System B 3.6.4.3 O V BASES ACTIONS L1 l (continued) If two or more required SGT subsystems are inoperable in l MODE 1, 2 or 3, the Unit 1 and Unit 2 SGT Systems may not be capable of supporting the required radioactivity release control function. Therefore, LC0 3.0.3 must be entered immediately. F.1. F.2. and F.3 l When two or more required SGT subsystems are inoperable, if l applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in secondary containment must immediately be suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. Q V Required Action F.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel l assemblies while in MODE 4 or 5, LC0 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiatcd fuel assemblies would not be a sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating each required Unit I and Unit 2 SGT subsystem for 1 10 continuous hours ensures that they 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. Operation with the heaters on for > 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system. (continued) HATCH UNIT 1 B3.6-wiz REVISION \(- .

SGT System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.2 REQUIREMENTS (continued) This SR verifies that the required Unit 1 and Unit 2 SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP. SR 3.6.4.3.3 This SR verifies that each required Unit I and Unit 2 SGT subsystem starts on receipt of an actual or simulated initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this SR to provide complete testing of the safety function. While this Surveillance can be performed with the reactor at power, operatina experience has shown that these components usually pass tne Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was found to be acceptable from a reliability standpoint. REFERENCES 1. 10 CFR 50, Appendix A, GDC 41.

2. FSAR, Section 5.3.

3. Unit 2 FSAR, Section 6.2.3.

4. Technical Requirements Manual l

5. NRC No. 93-102, " Final Policy Statement on Technical l Specification Improvements," July 23, 1993.

HATCH UNIT 1 B 3.6-9(q3 REVISION 'A (_ 1 i

AC Sources - Operating B 3.8.1 p BASES BACKGROUND A description of the Unit 2 onsite power sources is provided (continued) in the Bases for Unit 2 LC0 3.8.1. APPLICABLE The initial conditions of DBA and transient analyses in the SAFETY ANALYSES FSAR, Chapters 5 and 6 (Refs. 3 and 4, respectively) and Chapter 14 (Ref. 5), assume ESF systems are OPERABLE. The AC electrical power sources are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolart System (RCS), and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for Section 3.2. Power Distribution Limits; Section 3.5, Emergency Core Cooling System (ECCS) and Reactor Core Isolation Cooling (RCIC) System; and Section 3.6, Containment Systems. The OPERABILITY of the AC electrical power sources is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the unit. This includes maintaining the onsite or offsite AC sources OPERABLE during accident conditions in the event of: ID (j a. An assumed loss of all offsite power sources or all onsite AC power sources; and

b. A postulated worst case single failure.

AC sources satisfy Criterion 3 of the NRC Policy Statement (Ref. 14). LC0 Two qualified circuits between the offsite transmission network and the onsite Unit 1 Class lE Distribution System and three separate and independent DGs (IA, IB, and IC) ensure availability of the required power to shut down the reactor and maintain it in a safe shutdown condition after an anticipated operational occurrence (A00) or a postulated DBA. In addition, since some components required by Unit 1 are powered from Unit 2 sources (i.e., Standby Gas Treatment (SGT) System). one qualified circuit between the offsite transmission aetwork and the onsite Unit 2 Class 1E Distribution System, and one Unit 2 DG (2A or 2C), capable of supplying power to one required Unit 2 SGT subsystem, l must also be OPERABLE. (continued) HATCH UNIT 1 B 3.8-3 REVISIONkL

AC Sources - Operating B 3.8.1 BASES O LC0 Qualified offsite circuits are those that are described in (continued) the FSAR, and are part of the licensing basis for the unit. Each offsite circuit must be capable of maintaining rated frequency and voltage, and accepting required loads during an accident, while connected to the ESF buses. For the purpose of this LCO, each Unit 1 offsite circuit consists of incoming breaker and disconnect to the respective 1C and ID SATs, the IC and ID transformers, and the respective circuit path including feeder breakers to 4.16 kV ESF buses. (However, for design purposes, the offsite circuit excludes the feeder breakers to each 4.16 kV ESF bus). Feeder breakers from each circuit to the IF ESF bus are required to be OPERABLE; however, only one feeder breaker per bus to the IE and 1G ESF buses is required to be OPERABLE, but they must be from different SATs (e.g.,1E feeder breaker from the IC SAT and the IG feeder breaker from the ID SAT). With IE and 1G ESF buses both fed from one SAT (normal line up is both buses fed from 10 SAT), both feeder breakers to each of these ESF buses are required to be OPERABLE. The Unit 2 offsite circuit also consists of the incoming breaker and disconnect to the 4.16 kV ESF buses required to be OPERABLE to provide power to the Unit 2 equipment required by LC0 3.6.4.3. Each DG must be capable of starting, accelerating to rated frequency and voltage, and connecting to its respective ESF bus on detection of bus undervoltage. This sequence must be accomplished within 12 seconds. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and must continue to operate until offsite power can be restored to the ESF buses. These capabilities are required to be met from a variety of initial conditions, such as DG in standby with the engine hot and DG in standby with the engine at ambient condition. Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY. The AC sources must be separate and independent (to the extent possible) (Ref.1) of other AC sources. For the DGs, the separation and independence are complete. For the offsite AC sources, the separation and independence are to the extent practical. A circuit may be connected to more than one ESF bus, with automatic transfer capability to the (continued) HATCH UNIT 1 B 3.8-4 REVISION A

AC Sources - Shutdown B 3.8.2 l BASES-l LC0 associated with a Distribution System Engineered Safety l (continued) Feature (ESF) bus required OPERABLE by LC0 3.8.8, ensures that a diverse power source is available for providing electrical power support assuming a loss of the offsite circuit. In addition, some components that may be required by Unit 1 are powered from Unit 2 sources (i.e., Standby Gas 1 Treatment (SGT) System). Therefore, one qualified circuit between the offsite transmission network and the onsite Unit 2 Class 1E Distribution System, and one Unit 2 DG capable of supplying power to one required Unit 2 SGT subsystem, must l also be OPERABLE. Together, OPERABILITY of the required 1 offsite circuits and DGs ensures the availability of , sufficient AC sources to operate the plant in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents and reactor vessel draindown). The qualified offsite circuits must be capable of maintaining rated frequency and voltage while connected to their respective ESF buses, and of accepting required loads during an accident. Qualified offsite_ circuits are' those that are described in the FSAR and are part of the licensing basis for the unit. The Unit I and Unit 2 offsite circuits O consist of incoming breaker and disconnect to the IC or ID and the 2C or 2D startup auxiliary transformers (SATs), associated 1C or ID and 2C or 2D SATs, and_the respective circuit path including feeder breakers to all 4.16 kV ESF - buses required by LC0 3.8.8. (However, for design purposes, the offsite circuit excludes the feeder breakers to each 4.16 kV ESF bus). The required DGs must be capable of starting, accelerating to rated frequency and voltage, connecting to their respective ESF bus on detection of bus undervoltage, and accepting required loads. This sequence must be accomplished within 12 seconds. Each DG must also be capable of accepting required loads within the assumed ; loading sequence intervals, and must continue to operate ! until offsite power can be restored to the ESF buses. These j capabilities are required to be met from a variety of initial conditions such as DG in standby with engine hot and DG in standby with engine at ambient conditions. Additional DG capabilities must be demonstrated to meet required Surveillances, e.g., capability of the DG to revert to standby status on an ECCS signal while operating in parallel test mode. (continued) l l HATCH UNIT 1 B 3.8-42 REVISION 1f., I l l

AC Sources - Shutdown 3 3.8.2 BASES h Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY. O l (continued) HATCH UNIT 1 B 3.8-42A REVISION D

RPV Water Level B 3.9.6 BASES SURVEILLANCE SR 3.9.6.1 (continued) REQUIREMENTS The Frequency of 24 hours is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls on valve positions, which make significant unplanned level changes unlikely. REFERENCES 1. Regulatory Guide 1.25, March 23, 1972.

2. FSAR, Section 14.4.4.

3. NUREG-0800, Section 15.7.4.

4. 10 CFR 100.11.

5. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

O lO l HATCH UNIT 1 B 3.9-21 REVISION A

l l RHR -High Water Level B 3.9.7 i B 3.9 REFUELING OPERATIONS h B 3.9.7 Residual Heat Removal (RHR) -High Water Level BASES BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34 (Ref. 1). Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled. In addition to the RHR subsystems, the volume of water above the reactor pressure vessel (RPV) flange provides a heat sink for decay heat removal. O APPLICABLE With the unit in "0DE 5, the RHR System is not required to SAFETY ANALYSES mitigate any events or accidents evaluated in the safety analyses. The RHR System is required for removing decay heat to maintain the temperature of the reactor coolant. The RHR System satisfies Criterion 4 of the NRC Policy Statement (Ref. 3). l LC0 Only one RHR shutdown cooling subsystem ic required to be OPERABLE and in operation in MODE 5 with irradiated fuel in the RPV and the water level 2 22 ft 1/8 inches above the RPV flange (equivalent to 21 ft of water above the top of irradiated fuel assemblies in the spent fuel storage pool racks). Only one subsystem is required because the volume of water above the RPV flange provides backup decay heat removal capability. (continued) HATCH UNIT 1 8 3.9-22 REVISIONK/-.

l l RHR - High Water Level B 3.9.7 BASES LC0 An OPERABLE RHR shutdown cooling subsystem consists of an (continued) RHR pump, a heat exchanger, an RHRSW pump providing cooling to the heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path. In MODE 5, the RHR cross tie valve is not required to be closed; thus, the valve may be opened to allow RHR pumps in one loop to discharge through the opposite recirculation loop to make a complete subsystem. In addition, the RHRSW cross tie valves may be open to allow RHRSW pumps in one loop to provide cooling to a heat exchanger in the opposite loop to make a complete subsystem. Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. A Note is provided to allow a 2 hour exception to shut down the operating subsystem every 8 hours. O APPLICABILITY One RHR shutdown cooling subsystem must be OPERABLE and in operation in MODE 5, with irradiated fuel in the RPV and the water level 2: 22 ft 1/8 inches above the top of the RPV flange, to provide decay heat removal. RHR shutdown cooling subsystem requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS). RHR Shutdown Cooling subsystem requirements in MODE 5 with irradiated fuel in the RPV and the water level < 22 ft 1/8 inches above the RPV flange are given in LC0 3.9.8, " Residual Heat Removal (RHR) - Low Water Level." ACTIONS Ad With no RHR shutdown cooling subsystem OPERABLE, an alternate method of decay heat removal must be established within 1 hour. In this condition, the volume of water above the RPV flange provides adequate capability to remove decay heat from the reactor core. However, the overall reliability is reduced because loss of water level could (continued) HATCH UNIT 1 8 3.9-23 REVISION A

RHR - High Water Level B 3.9.7 BASES ACTIONS A.1 (continued) result in reduced decay heat removal capability. The 1 hour Completion Time is based on decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of these alternate method (s) must be reconfirmed every 24 hours thereafter. This will ensure continued heat removal capability. Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures. For example, this may include the use of the Fuel Pool Cooling System, the Reactor Water Cleanup System, operating with the regenerative heat exchanger bypassed, or any other subsystem that can remove heat from the coolant. The method used to remove the decay heat should be the most prudent choice based on unit conditions. B.l. B.2. B.3. and 8.4 If no RHR shutdown cooling subsystem is OPERABLE and an alternate method of decay heat removal is not available in accordance with Required Action A.1, actions shall be taken W immediately to suspend operations involving an increase in reactor decay heat load by suspending loading of irradiated i fuel assemblies into the RPV. Additional actions are required to minimize any potential fission product release to the environment. This includes ensuring: 1) secondary containment (at least including the i common refueling floor zone) is OPERABLE; 2) sufficient standby gas treatment subsystem (s) are OPERABLE to maintain i the secondary containment at a negative pressure with ' respect to the environment (dependent on secondary i containment configuration, refer to Reference 2; single failure protection is not required while in this ACTION); ; and 3) secondary containment isolation capability is available (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability) in each secondary containment penetration flow l path not isolated that is assumed to be isolated to mitigate radioactive releases. This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of service for maintenance or other reasons. It is not necessary to (continued) HATCH UNIT 1 B 3.9-24 REVISIONf((

I RHR - Hign Water Level B 3.9.7 BASES perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored O l (continued) HATCH UNIT 1 B 3.9-76 6 4 A Revision'as l l

                                                                                            .wam   -4 a *---

RHR - High Water Level B 3.9.7 BASES ACTIONS B.1. B.2. B.3. and B.4 (continued) to OPERABLE status. In this case, a Surveillance may need 4 to be performed to restore the component to OPERABLE status. Actions must continue until all required components are ; OPERABLE. C.1 and C.2 If no RHR shutdcan cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour. The Completion Time is modified , such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation. During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem), the reactor 3 coolant temperature must be periodically monitored to ensure

<                       proper functioning of the alternate method. The or e per hour Completion Time is deemed appropriate.

l SURVEILLANCE SR 3.9.7.1 REQUIREMENTS This Surveillance demonstrates that the required RHR shutdown cooling subsystem is in operation and circulating j reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours is sufficient 3 in view of other visual and audible indications available to ;

the operator for monitoring the RHR subsystem in the control i room.

i REFERENCES 1. 10 CFR 50, Appendix A, GDC 34.

2. Technical Requirements Manual. l f

3. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993. l O

HATCH UNIT 1 B 3.9-2)ic REVISION'Af l

l Q i RHR - Low Water Level l B 3.9.8 ) l B 3.9 REFUELING OPERATIONS f B 3.9.8 Residual Heat Removal (RHR) - Low Water Level I BASES 1 BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay j heat and sensible heat from the reactor coolant, as required by GDC 34 (Ref. 1). Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation l loop. Each pump discharges the reactor coolant, after it l has been cooled by circulation through the respective heat i exchangers, to the reactor via the associated recirculation l loop. The RHR heat exchangers transfer heat to the RHR j Service Water System. The RHR shutdown cooling mode is ' manually controlled.

1 APPLICABLE With the unit in MODE 5, the RHR System is not required to SAFETY ANALYSES mitigate any events or accidents evaluated in the safety l analyses. The RHR System is required for removing decay I heat to maintain the temperature of the reactor coolant. The RHR System satisfies Criterion 4 of the NRC Policy Statement (Ref. 3). l l LC0 In MODE 5 with irradiated fuel in the reactor pressure ) vessel (RPV) and the water level < 22 ft 1/8 inches above the RPV flange, two RHR shutdown cooling subsystems must be j OPERABLE. l An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, an RHRSW pump providing cooling to the heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path. The two subsystems have a common suction source and are allowed to have a common heat exchanger and common discharge piping. (continued) HATCH UNIT I B3.9-27 REVISION 'A /-

1 I RHR - Low Water Level B 3.9.8 O BASES LC0 Since the piping and heat exchangers are passive components (continued) that are assumed not to fail, they are allowed to be common to both subsystems. Thus, to meet the LCO, both pumps in one loop or one pump in each of the two loops must be OPERABLE. In MODE 5, the RHR cross tie valve is not required to be closed; thus, the valve may be opened to allow pumps in one loop to discharge through the opposite recirculation loop to make a complete subsystem. In addition, the RHRSW cross tie valves may be open to allow RHRSW pumps in one loop to provide cooling to a heat exchanger in the opposite loop to make a complete subsystem. Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. A Note is provided to allow a 2 hour exception to shut down the operating subsystem every 8 hours. APPLICABILITY Two RHR shutdown cooling subsystems are required to be OPERABLE, and one must be in operation in MODE 5, with irradiated fuel in the RPV and the water level < 22 ft 1/8 inches above the top of the RPV flange, to provide decay heat removal. RHR shutdown cooling subsystem requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS). RHR shutdown cooling subsystem requirements in MODE 5 with irradiated fuel in the RPV and the water level 2: 22 ft 1/8 inches above the RPV flange are given in LCO 3.9.7, " Residual Heat Removal (RHR) - High Water Le;el." ACTIONS A.1 With one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem is capable of providing the required decay heat removal. However, the overall reliability is reduced. Therefore an alternate method of decay heat removal must be provided. With both required RHR (continued) HATCH UNIT 1 B 3.9-27 REVISION A

RHR - Low Water Level B 3.9.8 BASES ACTIONS Ad (continued) shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO. The 1 hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of this alternate method (s) must be reconfirmed every 24 hours thereafter. This will ensure continued heat removal capability. Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures. For example, this may include the use of the Reactor Water Cleanup System, operating with the regenerative heat exchanger bypassed. The method used to remove decay heat should be the most prudent choice based on unit conditions. B.l. B.2 and B.3 O With the required RHR shutdown cooling subsystem (s) inoperable and the required alternate method (s) of decay heat removal not available in accordance with Required i Action A.1, additional actions are required to minimize any potential fission product release to the environment. This includes ensuring: 1) secondary containment (at least i including the common refueling floor zone) is OPERABLE; ;

2) sufficient standby gas treatment subsystem (s) are I OPERABLE to maintain the secondary containment at a ncgative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 2; j single failure protection is not required while in this ACTION); and 3) secondary containment isolation capability is available (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability) in each associated secondary containment penetration flow path not isolated that is assumed to be isolated to mitigate radioactive releases. This may be performed as an administrative check, by examining logs or (continued)

HATCH UNIT 1 R 3.9-28 REVISIONh(,

1 l l l RHR -- Low Water Level B 3.9.8 I) V BASES other information to determine whether the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, the Surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE. 1 l l O l 1 1

I (Continued) HATCH UNIT 1 REVISIONh,(_' B 3.9-EFAI. , A Zt ,

RHR - Low Water Level B 3.9.8 O v BASES ACTIONS C.1 and C.2 (continued) If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour. The Completion Time is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation. During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate. SURVEILLANCE SR 3.9.8.1 REQUIREMENTS This Surveillance demonstrates that one required RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by ( the flow rate necessary to provide sufficient decay heat V] removal capability. The frequency of 12 hours is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystems in the control room. REFERENCES 1. 10 CFR 50, . dix A, GDC 34.

2. Technical Requirements Manual. l

3. NRC No. 93-102, " Final Policy Statement on Technical i Specification Improvements," July 23, 1993.

O HATCH UNIT 1 B 3.9-25 REVISION R (., l l i

 .$  A4 m mama m. ed a ia.1,,+ #SAaf --.Are. ~4a --*-- mA e 44.n. --,ma-2 m- -- .m.-r --   *A a .%,. w..

O UNIT 1 MARKUP OF CURRENT TECHNICAL SPECIFICATIONS AND DISCUSSION OF CHANGES O 1 O

l DISCUSSION OF CHANGES ITS: SECTION 3.1.1 - SHUTDOWN MARGIN TECHNICAL CHANGE - MORE RESTRICTIVE M.1 Currently, if SDM is not met, Specification 3.3.H requires that the unit ) be placed in Cold Shutdown (MODE 4). Proposed Action B required the unit ' to be placed in MODE 3 if SLM is not met. While this change could be considered operationally less restrictive, it is considered to be more conservative with respect to providing proper reactivity controls. In addition, no actions are provided if SDM is not met while in MODES 4 and

5. Therefore, the following changes have been made:

If SDM is not met while the unit is in MODE 1 or 2, the proposed ACTIONS (ACTIONS A and B) require SDM to be restored within 6 hours or to be in MODE 3 within the following 12 hours. This places the unit in a shutdown condition and does not require a cooldown to MODE 4, which would add positive reactivity during a time when SDM is not met. (A cooldown could result in re-criticality.) In addition, once in MODE 3, proposed ACTION C requires action to be initiated to fully insert all insertable control rods. This action further reduces core reactivity. If SDM is not met in MODE 4 or 5, new ACTIONS (ACTIONS D and E) are p provided to initiate action to: 1) insert all insertable control rods (in the core cells containing fuel for MODE 5); 2) suspend CORE ALTERATIONS, except for control rod insertion and fuel assembly removal (if in MODE 5), and 3) initiate actions within I hour to restore to OPERABLE status secondary containment, necesary SGT l subsystems and one SCIV and associated instrumentation in each required secondary containment penetration flow path not isolated. The first two actions ensure SDM is not further reduced, while the last three actions provide some protection from radioactive release if an inadvertent criticality is experienced. In addition, the Bases include a discussion acknowledging the possibility of various configurations for the secondary containment boundary. Given the specific configuration, either one, two, or three SGT subsystems may be required to assure the necessary negative pressure when required. As discussed in the Bases, these details are provided in the Technical Requirements Manual (TRM). M.2 An additional Surveillance Frequency for SDM verification has been added to cl ari fy the requirements necessary for assuring SDM during the refueling process. Because SDM is assumed in several refueling mode analyses in the FSAR, assurance that intermediate fuel loading patterns during refueling have adequate SDM is necessary. This change imposes a requirement where none is explicitly provided in the existing Technical Specifications. However, this new requirement does not require introducing tests or modes of operation of a new or different nature. As discussed in the Bases corresponding to this Surveillance Frequency, the SDM verification in these situations is best accomplished by analysis HATCH UNIT 1 2 REVISION / b

i l l DISCUSSION OF CHANGES / V) ITS: SECTION 3.1.1 - SHUTDOWN MARGIN l l TECHNICAL CHANGE - MORE RESTRICTIVE : l (continued) M.1. 1 (continued) (rather than in-sequence criticals) because of the many changes in core i loading during a typical refueling. Bounding analyses may be used to demonstrate adequate SDM for the most reactive configurations during refueling, thereby showing acceptability of the entire fuel movement sequence. O V O HATCH UNIT 1 J'2.A REVISION [(,--

1 f 9 Me. Ice 3-(,L1.) LIMillNG CONDITIONS FOR OPERATION SURVEILtANCE REQUIREMENTS C. Secondary Containment s pip ( C. Secondary Containment

1. Normal Unit *1Intearity Secondary t@' E 1. Surveillance While Inteority Containment ~~ Maintained g

a. Otonnal UnitJ secondary con- (@rnal Unit Csecondary containment titTimeit9*~~

[(O D 44,\ malatMned 110P@1+=s shall be ggt surveUlanWshall be perfonned as indicated below: 4 of Unit 1 plant operation except when all of the , following conditions are met -- L /) . Z' (.5'/ I 1 I

a. AErmeL_Qn_113. secondary contain- 1 (1) The reactor is subcritical and7 ment capability test shall be Specification 3.3.A. is set. conducted Ti isolating th ,

it I s ary contai t j (2) The reactor water tenperature cing the st gas treah Appl ud,k is below 212*F and the reactor 3 (and ment s em filter tr in opera \_ LA.] h-i coolant systen is vented. ltinn. Such tests shall demonstrate i the capability to maintainft wini- ~ ( y ,n. Q l M- (3) No activity is being performed mum 1/4 inch of water vacune under eich can reduce the shutdown gg3.q,q.),3 marpn belas that stated in calm wind (< 5 w h) conditions with Specification 3.3.A. Teacn filter In in rlow rate no g, more than 4000 cf=- . (4) The fuel cask or hvadiated fuel b. realUnit_.f nment is not being nmed in the re- ] capability to maintain a minimme f(ok tpML actor tuildire. 1/4 inch of water vacuum under calm SR RRty w.Mb --* wind (< 5 mph) conditions wit)1iiich 5

          #               1
                            ,()__ All hatches betsmen the nonstl 3 filter train griow r w not no j

thit 1

mntaituent (than 4000 cfm shall bo demonstrated

( a losed sea . l [6) At least one door in I ' path betneen t nanaal thit secondary cantai Ik and th 2 seccrdary irnent f ts el q (7) Inservice c or g y 49 i n progniss. Y'/5 'I 5 As 3.vi 1. I conta b N 16 1.t. L sha intained cining alkpodes of iki4 2 plant operation}4xcgt 7 Oper mal Conditim 4 aWined in the it 2 Technical A . 3/ Specific t, ions. w-I*hormal Unit I secondary containment includes the Unit I reactor building area below the refueling floor and the connon Unit 1 and Unit I area above the A. (refueling specificationfloor. 3.7.C.2.For modified Unit I secondary containment conditions see / HATCH - UNIT 1 3.7-12 heendment No. E2, 40, 56, M , M B, 160 hp3

hfCri$ctt% u % % p) i V LIMITING CONDITIONS FOR OPERAT105 SURVEILLANCE RE0VIREMENTS

2. Modified Unit 1 Secondary Containment * \ 2. Surveillance While Inteority Maintained Intearity g,@ l h

a. Operation with(aIodified Unitl OG3 TUO, sect de/ containment secondary containment ktegrMy (PebesuTy surveillance snall be performed

( 4 0 % 4'g is permissible provided all of the following coditions are met: as indicated below: N

a. A(modified Unit 1l secondary contain-hl) The reactor is subcritical and 3 ment capability test shall be per-Specification 3.3.A. is met. fo " iter iwii ' y tne modified Qibbk (2) The reactor water temperature is pl 1s ondary con inment and th standby ga reatment Q,3 g

below 212*F and the reactor ; syste filte trains in pration. coolant system is ventet - ' ucn tests shall demons mAia no g, capability to maintain inimum N3) All hatches between he modified 1/4 inch of water vacuum under WA'YI 3 Unit I secondary con inment calm wind (< 5 mph) conditions and Unit 2 secondary c ntainment lWith eacn riiter train flow rate m. re closed and sealed. lnot more than 4000 efm (4) At least one door in each ccess . If normal Unit I se ary contain 3 pat between the modified U t1 .nt integrity should equired as seco ary containment and un 2 st ed in Specification .C.I., secon y containment is close perf' surveillance as sta d in Spe fication 4.7.C.I.a. (5) All hatch separating the mod- , If modif Unit I secondary to ified Unit secondary contain- tainment is ubsequently require ment from th Unit I reactor as stated in ecification 3.7.C.2. A .)} building area low the refueling performsurveilknceasstatedin floor are close and sealed. < Specification 4 R 2 = v (6 At least one door n each access ath separating th modified it I secondary to ainment f m the Unit I react r building are below the refuel g floor is c osed. (7) The SG valves to the Un I reactor uilding area below the refu ing floor, to the Unit I dr ell, and to the Unit I tor are secured closed.

               % Integrity of the y dified Unit ] l N econdary containme          shall be m ' tained during all odes of Unit plant operations xcept Operat        al Condition 4 a defined gh in the Un 2 Technical                              f

( Specificatio ft" Refueling rations may cqnunue in the modifi Unit I secondary LA.) ntainmentpro'dedallcondkions/ ecification M .C.2.a. are %et t in Mo[MiedUnitIsecondary ntainment includes thehommon Unit I and Unit 2 area ah ve the refueling fl For normal Unit I s'eqndary containment L A

1

( conditionb see Specification .l. N HATCH - UNIT 1 3.7-12a Amendment No. 94, 158 243

Spec,A.+ f A 4\ LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS O g 3.7.C 3. Violation of Secondary a.~., /d 4.7.C.3. Surveillancebfter Intearity Violated Containment Intearity . S )

a. Without(tiatesunit [. l Aft a seconda containment v' la-seconda ntainment ope.4 . tion 4 determin the standby ga bCTldd b i store

{TaFc-UnitJsecondary g { treatment < system wi i be operated inmediatel after th affected zones containment integrity y re isolat from the inder of within 4 hours, or the seconda ontainmen . The' perforn the following (as abi ity to mai in the inder 9 applicable): of t seconda ntainee at L4 2, 1/4 in of water uum pr ute (1) Suspend irradiated fuel under ca A (< 5 mph and/or fuel cask. handling' indcono(tions W shall be cohf~irned. ftLTW' in theVfalch Unft L- 3

 ,3                                secondary containment.      p/o('6 3
   'profcw A 2*tWi            (2) Be in at least Hot Shutdown         y       C 1-          b ws    c                        within the next 12 hours and p ph                     meet the Conditions of 3.7.C.I.a. within the next 24 hours.
                         . Without Hatch-Unit         econTar P ontainment, refer to he follow-     '

I ihg Hatch-Unit 2 Techni i Spec 4(ication, for LCOs t be w follow for Hatch-Unit 2: (1) Sectioll .6.5.1. ( (2) Section 3.911. fD. Primary containment Isolation Valves D primary Containment isolation Valv

1. Valves Recutred to be Operable 1. Surveillance of Operable Valves During reactor power operation, Surveillance of the primary con-all primary containment isolation tainment isolation valves shall be valves and all reactor coolant performed as follows:

system instrument line excess flow check valves shall be operable except a. At least once per operating as stated in Specification 3.7.D.2. cycle the operable isolation valves that are power operated and automatically initiated shall be tested for simulated automatic initiation and the closure times. k Disc +.,.o a q TT$ 4 1 1 pe,g

                                                         / /N
                                   \                N no.

HATCH - 1 3.7-13 Amendment No. 40, 56, H, MO, H9,15E 3cd

DISCUSSION OF CHANGES 1 ITS: SECTION 3.6.4.1 - SECONDARY CONTAINMENT ADMINISTRATIVE A.1 The definition of SECONDARY CONTAINMENT INTEGRITY has been deleted from ' the proposed Technical Specifications. It is replaced with the requirement for secondary containment to be OPERABLE. This was done i because of the confusion associated with these definitions compared to its ' use in the respective LCO. The change is editorial in that all the requirements are specifically addressed in the proposed LC0 for the ; secondary containment and in the Secondary Containment Isolation Valves ; and Standby Gas Treatment System Specifications. The Applicability has been reworded to be consistent with the new definitions of MODES and to : have a positive statement as to when it is applicable, not when it is not J applicable. Parts 1 and 2 form the MODES 1, 2 and 3 requirements, Part 3 forms the CORE ALTERATIONS requirement, and Part 4 forms the movement of irradiated fuel assemblies in the secondary containment requirement. In addition, a Required Action has been added to suspend CORE ALTERATIONS (Required Action C.2). Therefore the change is a presentation preference adopted by the BWR Standard Technical Specifications, NUREG 1433. A.2 The technical content of this Applicability is being moved to Section 3.10 of the proposed Technical Specifications. Any technical changes to this requirement are addressed in the Discussion of Changes associated with proposed Specification 3.10.1. < A.3 This requirement has been deleted since it is applicable to the operation of Unit 2, and this Technical Specification applies only to Unit 1. Thus, any needed requirements for Unit 2 are located in the Unit 2 Technical Specifications. TECHNICAL CHANGE - MORE RESTRICTIVE

M.1 This Surveillance has been broken into two separate Surveillances, SR 3.6.4.1.3 and SR 3.6.4.1.4. The tests will ensure the ability of the :

secondary containment to maintain 1/4 inch vacuum, and SR 3.6.4.1.3 will ensure the vacuum is attained in 120 seconds. SR 3.6.4.1.4 will ensure l the SGT system maintains the vacuum for 1 hour. These new requirements ' are additional restrictions on plant operation. O HATCH UNIT 1 1 REVISIONh(.,

r i DISCUSSION OF CHANGES , ITS: SECTION 3.6.4.1 - SECONDARY CONTAINMENT TECHNICAL CHANGE - MORE RESTRICTIVE (continued) 1 M.2 The analysis for secondary containment drawdown assumes two SGT subsystems are required for two secondary containment zones (i.e., one reactor ; building and common refuel floor), and three SGT subsystems for all three zones (see also M.5)). Thus, the test specifies the " required" number of operating SGT subsystems and the flow rate for each subsystem of s 4000 , cfm. " Required" number of SGT subsystems is dependent on the existing secondary containment configuration, with the explicit details provided in ' Note 2 to SR 3.6.4.1.3 and the Note to SR 3.6.4.1.4. To ensure all SGT subsystems are tested (since the test does not specify that all SGT ' subsystems must be tested) the Frequency is on a STAGGERED TEST BASIS, which will ensure all appropriate combinations of SGT subsystems are tested. These are additional restrictions of plant operation. M.3 An Applicability has been added. Secondary Containment is now required to be OPERABLE during operations with a potential for draining the reactor vessel to provide mitigation if an inadvertent vessel draindown event occurs. An appropriate Required Action has also been added . (Required ' Action C.3). This is an additional restriction on plant operation. M.4 Two new Surveillance Requirements have been added. SR 3.6.4.1.1 will verify that all secondary containment hatches are closed every 31 days. O SR 3.6.4.1.2 will verify that each access door is closed, except when used for opening, and then one door is closed, every 31 days. These are additional restrictions on plant operation. M.5 Based on the design of Unit 2 SGT System, any reliance on Unit 2 SGT for Unit I secondary containment requirements necessarily requires the Unit 2 reactor. building boundary be included as part of the secondary containment. (This is from a specific single failure which results in Unit 2 SGT suction always from both the Unit 2 reactor building and the > common refueling floor; an exception can be made if Unit 2 is shut down and its reactor building, including SGT suction, is isolated from the I refueling floor zone.) As such, in order to assure the secondary , containment is drawn down to a negative pressure in the required time ' assuming a worst case single failure, all four SGT subsystems are required to be OPERABLE during Unit 1 MODES 1, 2, and 3, handling irradiated fuel, l and CORE ALTERATIONS (unless in " modified" secondary containment). Refer i to discussion LA.2 for the details of presentation of this added l restriction. l TECHNICAL CHANGE - LESS RESTRICTIVE

    " Generic" LA.1 The details comprising operability of the secondary containment are proposed to be located in the Surveillances, the Technical Requirements
  ,         Manual (TRM), and the Bases to this Specification. Secondary Containment operability requirements are explicitly required in SR 3.6.4.1.1 and SR 1

HATCH UNIT 1 2 REVISIONy

~ DISCUSSION OF CHANGES ! ITS: SECTION 3.6.4.1 - SECONDARY CONTAINMENT TECHNICAL CHANGE - LESS RESTRICTIVE

      " Generic" LA.1 (continued)                                                               ,

3.6.4.1.2, and the details regarding allowable configurations, including associated SGT and SCIV requirements, are in the TRM. Therefore, this information does not need to be repeated in the LCO. Any change to the information located in the Bases will be controlled by the provision of the Bases Control Program described in Chapter 5 of the Technical Specifications. Changes to procedures and the TRM will be controlled by the provisions of 10 CFR 50.59. LA.2 The general description of the modified aad normal secondary containment has been relocated to the Bases to this Specification, with explicit ! details located in the Technical Requirements Manual (TRM). This information describes what Secondary Containment comprises, thus, is more appropriately located in the Bases and TRM. Any change to the Bases will l be controlled by the provisions of the Bases Control Program described in Chapter 5 of the Technical Specifications. Changes to the TRM will be controlled by the provisions of 10 CFR 50.59. , The secondary containment boundary will typically consist of more than just the current Hatch-Unit 1 Secondary Containment (refer to discussion . M.5 for this change). Given the number of variations of secondary containment boundaries, dependent on variations in the operational status of both units, the resultant complexity of providing the details of all . options within the ITS is deemed detrimental to operator usefulness. The ; simplified presentation provides adequate requirements to assure proper implementation without unwarranted complexity. ' LA.3 The details for the methods of performing this Surveillance are relocated to procedures. Changes to the procedures will be controlled by the provisions of 10 CFR 50.59. 4 4 i O HATCH UNIT 1 \cl A REVISION %

. = - . . . . - .-. 4i6.Sco 3 .( . 4 3 LIMITING CONDITIONS FOR OPEPATION O SURVE1LLANCE RE0VIREMENTS B. Standby Gas Treatment System B. Standby Gas Treatment System fwd 00* i. Doerability Reauirements 1. Surveillance When System

Doerablg 3 &51.a.() a5d R i1-oTT~inreei]2ofrinunik tinit 2)f of the fou L,5 At least once per operating cycle, not ((0 independent standby gas treatment to exceed 18 months, the following system trains shall be operable at 3 f,/g))(alltimeswhenUnitIsecondary g conditions shall be demonstrated: PI+ fem kontainment integrity is required. a. Pressure drop across the combined cPW L'3 HEPA filters and charcoal absorber Applid'(b With one of the Unit standby gas bank is less than 6 inches of treatment syste rable, for water at the system design flow any reason, Unit I reactor rate (+10% -05). . operation and fuel handling and/or k f M3,3 $ handling of casks in the vicinity b. Operability of inlet heater at f of the spent fuel pools is rated power when tested in , pemissible for a period of seven accordance with ANSI N510-1975. (7) days; pro cea in s sii a ive - omponents in e remaining c. Air distribution is uniform within o rable standby as treatment 20% across the filter train when sys s in each unt (minimum of \ tested in accordance with in un 1 and I in U t 2) shall b demonstr ed to be ope ble within (N510-1975. 4 hours, a daily there ter, l _ l j gpyg g3.s.4.D-

tg 6 4 % % % 4 3's 7 O l l l h l l O eA1Ee - e~I1 1 2.1-lee Am.nem.n, o. lie w l l l

l i I FecL M 3(.43

    %            LIMITING CONDITIONS FOR OPERATION                                           SURVEILLANCE RE0VIREMENTS B. Standby Gas Treatment System                                  B. Staniv Cas Treatnent System

[. .$

1. Operability Recutrements (Cont'd) 1. Surveillance When Systen Goer D If the Inoperable Unit I standby gas F "

l,bff D.* p 'd M , treatment system is not made fully d. Autanatic initiationef each train of. $(3.g.,y,3 O.1 operable within the seven (7) day j,i 4d the Unit I and Unit 2 stanty gas ! gg %d# tmatment systens. h shutdan period, the Unit the I reactor shall be 1 g and placedrin cold shutdow) d (, condition within the next 36 hours and Q' mYjn ual oper ty of the ass ; M unit 1 or Unit 2 fuel handling

       /                     operations shall be teminated wtthe-4 ve for fil       .coolino.                      -

l

    .\/{ gQ                  howes.                                           p rope ,3 u p ,4 W soz.s 3

Unit I reactor operation and thit 1 or M o . L. 3 8 j j%4-M j Unit 2 fuel handlingJhall not be

                                                                                                                      ~~[~

p A allmed if both of the thit I standby N b gas tr3tatment systems am inoperable or p reIo, 4 ' bC jog'3 hty (T60th of the thit 23tandby am. g, ! g, j g9.ju p,[y p-) yl treatment systens are inoperablefExcept } as allowed Vy 4.s.o.a.o.

                                                                                                                         ,,,      7_     ,

1. With both Unit 2 SGTS inoperattle for ,

surveillance of the Unit 2 pri ' tairrent excess flw isolation dangers, Unit I reactor operation i ; per%ssible for a period of 12 hours if thh allwing conditions are net: (1) Maintain in Unit I least 1/4" H,0 vacutsn ry containnent by AM O using nomal SGTS as necess, tilation and Unit 1 (2) Assure operability f both Unit 1 SGTS filter trains Assure Unit 2 SGTS val s to l fueling floor cannot opened i (4) Al aw no fuel novement in un is 1 or ; (5) Unit 2 s ondary contairrent ! integrity intact except for Unit 1 ; 2 SGTS oper lity requirstents. g : i

1. If the requirtr'oqts of 3.7.B.I.b. I cannot be met, anNrderly shutdan ,

all be initiated W the reactor N I shh be brought to HA Shutdown withi 2 hours and shah be in Cold .t . Shutdan ' thin the foll ! 24 hours. ] & s w E Q 1 HATCH - UNIT 1 3.7-11 Amendent No. M, 40, M, %, MO, 118 M6 t

DISCUSSION OF CHANGES [ ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM ADMINISTRATIVE A.1 The technical content of this requirement is being moved to Section 5 of the proposed Technical Specifications in accordance with the format of the BWR Standard Technical Specifications, NUREG 1433. Any technical changes to this requirement are addressed in the Discussion of Changes associated with proposed Specification 5.5.7. A surveillance requirement (proposed SR 3.6.4.3.2) is added to clarify that the tests of the Ventilation Filter Testing Program must also be completed and passed for determining OPERABILITY of the SGT System. Since this is a presentation preference that maintains current requirements, this change is considered administrative. A.2 The description of the signal used to automatically initiate the SGT System " actual or simulated initiation signal" has been added for clarity. This is consistent with the BWR Standard Technical Specifications, NUREG 1433, and no change is intended. This Surveillance has been deleted since there is no bypass valve in the A.3 system. The system has internal orifices for filter cooling. A.4 A new ACTION is proposed (ACTION E) which directs entry into LC0 3.0.3 if l two or more required standby gas treatment subsystems are inoperable in O MODES 1, 2, or 3. This avoids confusion as to the proper action if in MODES I, 2, or 3 and simultaneously handling irradiated fuel or conducting operations with a potential for draining the vessel. Since this proposed ACTION effectively results in the same action as the current specification, this change is considered administrative. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 An Applicability has been added. The SGT System is now required to be OPERABLE during operations with a potential for draining the reactor vessel to provide mitigation if an inadvertent vessel draindown event occurs. Appropriate Required Actions have also been added (Required Actions C.2.3 and E.3. In addition, Required Actions have been added (proposed Required Actions D.2.2 and F.2) to suspend CORE ALTERATIONS, l consistent with the Applicability of Secondary Containment (and SGT System). These are additional restrictions on plant operation. M.2 An additional shutdown action has been added (Required Action C.1) to not l only be in Cold Shutdown (MODE 4) within 36 hours, but to also be in Hot i Shutdown (MODE 3) within 12 hours. This is an additional restriction on I plant operation. O HATCH UNIT 1 1 REVISIONfh

I l DISCUSSION OF CHANGES ( ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM TECHNICAL CHANGE - MORE RESTRICTIVf (continued) M.3 The time to suspend fuel handling has been changed from 4 hours to immediately. This is an additional restriction on plant operation. M.4 This allowance has been deleted since it is not needed in the proposed Specifications. The new Specifications will allow Unit I reactor operations for 7 days with both Unit 2 SGT subsystems inoperable (see comment L.3) based on only three SGT subsystems being required OPERABLE (which is the case when the. Unit 2 secondary coretainment configuration is as required in the CTS). Thus, an additional 12 hours, as provided by this allowance, is not needed. The deletion of this allowance is more restrictive on plant operation. M.5 SR 3.6.4.3.1 requires the SGT System to be run 10 continuous hours each 31 days, while the CTS state a total of 10 hours. This is an additional restriction on plant operations. M.6 ITS 3.6.4.1, discussion M.5, details added requirements for the Unit 2 reactor building to be included in the Unit 1 secondary containment

-         requirements. In support of this addition, a fourth SGT subsystem is also required to be OPERABLE to assure that, including a single failure, the SGT System can drawdown and maintain a vacuum on this combined secondary containment. With the addition of the fourth SGT subsystem, a 30 day Completion Time is also proposed (ACTION A) for one inoperable Unit 1 SGT subsystem. (Note this ACTION is proposed to be consistent for both units; refer to Unit 2 ITS 3.6.4.3, discussion L.7.)

TECHNICAL CHANGE - LESS RESTRICTIVE

  " Specific" L.1    The proposed change will delete the requirement to test the other SGT subsystems when one subsystem is inoperable.             The requirement for demonstrating operability of the redundant subsystems was originally prescribed because there was a lack of plant operating history and a lack of sufficient equipment failure data. Since that time, plant operating experience has demonstrated that testing of the redundant subsystems when one subsystem is inoperable is not necessary to provide adequate assurance of system operability.

This change will allow credit to be taken for normal periodic Surveillances as a demonstration of operability and availability of the remaining components. The periodic frequencies specified to demonstrate operability of the remaining components have been shown to be adequate to ensure equipment operability. As stated in NRC Generic Letter 87-09, "It is overly conservative to assume that systems or components are inoperable when a surveillance requirement has not been performed. The opposite is HATCH UNIT 1 2 REVISIONk6

DISCUSSION OF CHANGES ( ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM TECHNICAL CHANGE - LESS RESTRICTIVE

   " Specific" L.1 (continued) in fact the case; the vast majority of surveillances demonstrate the systems or components in fact are operable." Therefore, reliance on the specified surveillance intervals does not result in a reduced level of confidence concerning the equipment availability. Also, the original General Electric Standard Technical Specifications (STS), NUREG 123, and, more specifically, all the Technical Specifications approved for recently licensed BWR's accept the philosophy of system operability based on satisfactory performance of monthly, quarterly, refueling interval, post maintenance or other specified performance tests without requiring additional testing when another system is inoperable (except for diesel generator testing, which is not being changed).

/N U o V HATCH UNIT 1 MM REVISIONh(,

l DISCUSSION OF CHANGES l O d ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM i i l TECHNICAL CHANGE - LESS RESTRICTIVE l (continued) L.2 An alternative is proposed to suspending operations if a SGT subsystem cannot be returned to OPERABLE status within the time allowed, and l movement of irradiated fuel assemblies, CORE ALTERATIONS, or operations with the potential for draining the reactor vessel are being conducted. The alternative is to initiate the remaining OPERABLE subsystems of SGT and continue to conduct the operations. Since the remaining subsystems are sufficient for any accident, the risk of failure of the subsystems to perform their intended function is significantly reduced if they are running. L.3 The proposed ACTIONS will allow the one required Unit 2 SGT subsystem, when three subsystems are initially required, to be inoperable (thus both Unit 2 SGT subsystems are inoperable) for up to 7 days without requiring a Unit 1 shutdown or suspension of operations. This is consistent with the 7 days allowed for an inoperable Unit 1 SGT subsystem. With two OPERABLE subsystems (two Unit 1, or one Unit 1 and one Unit 2), the safety analysis assumptions are met, provided no single active failure occurs. Thus, since 7 days has been found to be acceptable for one of the two cases (one Unit 1 SGT subsystem inoperable), it is considered justifiable , for the other case (one required Unit 2 SGT subsystem inoperable). Refer ( to M.6 for changes involving the requirement for four SGT subsystems to be initially OPERABLE. L.4 Comment number not used. L.5 (Refer also to ITS 3.6.4.1 discussions M.5 and LA.2.) The required number of OPERABLE SGT subsystems is proposed to be dependent on the configuration of the secondary containment. The current requirement for three OPERABLE SGT subsystems is based on only two zones of secondary containment being required for secondary containment OPERABILITY (Zone I - Unit I reactor building and Zone III - common refueling floor (Zone II - Unit 2 reactor building]). M.5 discusses the addition of the third zone and fourth SGT subsystem, effective when Zone II is part of secondary containment. However, based on the specific operational status of each of the two units, certain zones may be isolated from the secondary containment boundary; thereby reducing the volume of the secondary containment. With a reduced secondary containment volume, a reduced number of SGT subsystems would be required to drawdown to and maintain that volume at 0.25 inches vacuum in the required time. There are two specific combinations of secondary containment and minimum SGT requirements that have been evaluated for additional flexibility. (Note two other new and unique configurations, i.e., Zone I only and Zone II only, for MODE 4 OPDRV, are addressed as a more restrictive change; see O HATCH UNIT 1 y REVISION A'Gy-

DISCUSSION OF CHANGES O V ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM TECHNICAL CHANGE - LESS RESTRICTid (continued) L.5 (continued) M.1 for this Specification.) These configurations each can be adequately supported, given the single failure of one subsystem, by less than the currently required SGT subsystems. A brief discussion of each follows: A. With Unit 1 secondary containment required, and Unit 2 shutdown with its reactor building (Zone II) isolated from Zone III such that these volumes do not communicate (i.e. consistent with the existing CTS

            " normal" Hatch-Unit I secondary containment requirement), Zones I and III are the secondary containment. In this configuration only three SGT subsystem are required, however CTS only allows one to be a Unit 2 subsystem.       Analysis has shown that two Unit 2 subsystems can adequately perform the required secondary containment drawdown function. Based on this, and to accommodate a single failure, any three subsystems (including two Unit 2 and one Unit 1) are adequate to support secondary containment OPERABILITY.

/ B. With both Units shutdown, with refueling floor activities requiring \ secondary containment OPERABILITY (CORE ALTERATIONS or handling irradiated fuel), Zone I and II can be isolated from Zone III. In this configuration, only Zone III need be drawndown and maintained at a negative pressure. Testing and analysis has shown that any single SGT subsystem can perform the necessary function. The analyses performed in support of the above configurations and minimum SGT subsystems also confirmed no significant impact on probabilities or consequences of an accident, and no reduction in any margin of safety. Additionally, the proposed presentation of SGT requirements in ITS does not explicitly detail the minimum number of OPERABLE SGT subsystems. Rather it states, "The Unit I and Unit 2 SGT subsystems required to support LC0 3.6.4.1, " Secondary Containment," shall be OPERABLE." The Bases provides an outline of the above discussed configurations and number of required SGT subsystems, with reference to the Technical Requirements Manual for complete details. Given the number of variations of secondary containment boundaries, dependent on variations in the operational status of both units, the resultant complexity of providing the details of all options within the ITS is deemed detrimental to operator usefulness. The simplified presentation provides adequate requirements to assure proper implementation without unwarranted complexity. HATCH UNIT 1 y3[Y REVISIONgf

DISCUSSION OF CHANGES ITS: SECTION 3.8.2 - AC SOURCES - SHUTDOWN O d ADMINISTRATIVE A.1 This requirement has been deleted since the DGs do not provide power to the fuel pool cooling pumps. These pumps receive power from the non-safety related buses. As such, this change is considered administrative. A.2 The Applicability has been rewritten to be MODES 4 and 5, and during ' movement of irradiated fuel assemblies in the secondary containment. This effectively encompasses all the current requirements, and based on the current secondary containment Applicability, is a little more restrictive. However, requirementsbased areupon the proposed Secondary Containment Applicability, the the same. Therefore, this change is considered administrative and is made to provide clarity. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 ACTIONS have been added to provide proper Required Actions to take when a required AC source is inoperable. The new ACTIONS (ACTIONS A and B) will eitherCurrently, no actions are provided.

1) require declaring the affected components inoperable and taking the ACTIONS of the applicable system LC0 (Required Action A.1) or 2) will require suspending CORE ALTERATIONS, OPDRVs, and irradiated fuel movement in the secondary containment, and initiating action to restore the inoperable source O (Required Actions A.2.1, A.2.2, A.2.3, A.2.4, B.1, B.2, 8.3, and B.4). In addition,AC required Surveillances sources. have been added to ensure the OPERABILITY of the restrictions on plant operation.These new ACTIONS and Surveillances are additiona TECHNICAL CHANGE - LESS RESTRICTIVE L.1 The AC Sources Specification, while in Shutdown or Refuel, has been modified to only require one Unit 1 DG to be OPERABLE. However, a Unit 1 offsite circuit requirement has been added to replace the DG. This qualified circuit must be OPERABLE between the offsite transmission network subsystem and(s)the onsiteby required Unit LC01 3.8.8.

Class 1E AC electrical power distribution This means that all Unit 1 buses ' needingsource. offsite AC power must be capable of being powered from a qualified Unit 1 In addition, since certain Unit 2 equipment may be needed to meet Unit 1 accident analysis, this LC0 also requires an OPERABLE Unit 2 DG (capable of powering required equipment), and a Unit 2 offsite l source (which is available to power to the required Unit 2 equipment). These additional requirements adequately compensate for the reduction in the number of required DGs (from two to one). i f) % l

HATCH UNIT 1 1 REVISIONAsQ

O UNIT 1 NO SIGNIFICANT IIAZARDS CONSIDERATION O 3 I O 1

f N0 SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM L.3 CHANGE ' In accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company , has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following: -

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

This change would allow an additional 7 days to restore one Unit 2 SGT subsystem, when three subsystems are initially required, when it is found l inoperable. The SGT System is not considered an initiator of any previously analyzed accident. Therefore, this change does not significantly increase the probability of such accidents. The proposed change would allow additional limited operation with less than the required SGT subsystems OPERABLE. However, since the only change is in the allowed outage time, the consequences of an event that may occur during the outage time would not be any different than during the currently allowed outage time for a Unit 1 SGT subsystem. Therefore, this change does not significantly increase the consequences of any previously analyzed accident.

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

The proposed change does not involve any design changes, plant ' modifications, or changes in plant operation. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The change increases the allowed outage time. The margin of safety considered in determining the allowed outage time is based on engineering judgement and probability of occurrence of an event requiring the unavailable capabilities. The proposed extension is based on similar allowed outage times for the Unit 1 SGT System and the redundancy of the Unit 1 SGT subsystems which accomplish similar functions in the mitigation of the event. Therefore, the change does not involve a significant reduction in the margin of safety. O HATCH UNIT 1 4 REVISIONkG

i NO SIGNIFICANT HAZARDS DETERMINATI0h ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT 5YSTEM L.5 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following: j l

1. Does the change involve a significant increase in the probability or l consequences of an accident previously evaluated? l The required number of OPERABLE SGT subsystems is proposed to be dependent on the configuration of the secondary containment. Based on the specific operational status of each of the two units, certain secondary containment zones may be isolated from the secondary containment boundary; thereby ,

reducing the volume of the secondary containment. With a reduced secondary containment volume, a reduced number of SGT subsystems would be required to drawdown to and maintain that volume at 0.25 inches vacuum in the required time. Since the SGT System is not assumed to be an initiator of any previously analyzed accident, the change does not significantly increase the probability of such accidents. The change will not increase the consequences of an accident previously analyzed since sufficient. SGT subsystems remain OPERABLE to mitigate the previously evaluated accidents accounting for a single active failure. Additionally, the proposed change relocates the details of required number of SGT subsystems for various secondary containment configuration from the Technical Specifications to the Bases and Technical Requirements Manual (TRM). The Bases, and TRM containing the relocated information are subject to the change control provisions in the Administrative Controls section of Technical Specifications, and will be maintained in accordance with 10 CFR 50.59. Since any changes to the Bases or TRM will be evaluated per the requirements of 10 CFR 50.59, no increase (significant or insignificant) in the probability or consequences of an accident previously evaluated will be allowed. Therefore, this change does not involve a significant increase in the probability or 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 involve new configurations of available SGT - subsystems on various secondary containment boundaries; however, the secondary containment function to contain fission products released during accidents, and the necessary support provided by the SGT System to assure fission products released to the secondary containment are filtered prior to release, remains unaffected. The system will continue to function in the same way as before the change. Therefore, the proposed changes do not , create the possibility of a new or different kind of accident from any previously evaluated. HATCH UNIT 1 6 REVISION M

N0 SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM L.5 CHANGE (continued)

3. Does this change involve a significant reduction in a margin of safety?

With the minimum SGT subsystems OPERABLE, sufficient SGT subsystems are available to mitigate the consequences of an accident and account for a single active failure of one of the subsystems. Therefore, since sufficient SGT subsystems are still required to be OPERABLE to meet the analysis assumptions, the change will not result in a significant reduction in a margin of safety. Since any future changes to these requirements in the Bases or TRM will be evaluated per the requirements of 10 CFR 50.59, no reduction (significant or insignificant) in a margin of safety will be allowed. (o ./ HATCH UNIT 1 7 REVISION M g

1 O UNIT 2 IMPROVED TECHNICAL SPECIFICATIONS O O i

SDM l 3.1.1 1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) LCO 3.1.1 SDM shall be:

a. 2 0.38% Ak/k, with the highest worth control rod analytically determined; or

b. 2 0.28% Ak/k, with the highest worth control rod determined by test.

APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limits A.1 Restore SDM to within 6 hours in MODE 1 or 2. limits. C\ U B. Required Action and B.1 Be in MODE 2. 12 hours associated Completion Time of Condition A not met. C. SDM not within limits C.1 Initiate action to Imraediately in MODE 3. fully insert all insertable control rods. D. SOM not within limits D.1 Initiate action to Immediately in MODE 4. fully insert all insertable control rods. AND (continued) < O ; HATCH UNIT 2 3.1-1 REVISION A I

SDM 3.1.1 l ACTIONS I CONDITION REQUIRED ACTION COMPLETION TIME

l D. (continued) 0.2 Initiate action to 1 hour l restore secondary l containment to OPERABLE status. . l AND l D.3 Initiate action to 1 hour restore required l i standby gas treatment ! (SGT) subsystem (s) to l OPERABLE status. AND l D.4 Initiate action to I hour l restore isolation ! capability in each 1 required secondary l containment penetration flow path not isolated. E. SDM not within limits E.1 Suspend CORE Immediately in MODE 5. ALTERATIONS except for control rod insertion and fuel assembly removal. AND E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies. AND (continued) O HATCH UNIT 2 3.1-2 REVISION \(,

SDM 3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. (continued) E.3 Initiate action to I hour restore secondary l containment to OPERABLE status. AND E.4 Initiate action to I hour restore required SGT subsystem (s) to OPERABLE status. AND E.5 Initiate action to I hour restore isolation capability in each required secondary containment penetration flow path not isolated. m b HATCH UNIT 2 3.1-3 REVISION %

SDM l 3.1.1 I l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM is: Prior to each in-vessel fuel

a. 2 0.38% Ak/k with the highest worth movement during control rod analytically determined; fuel loading or sequence

b. 2 0.28% Ak/k with the highest worth AND control rod determined by test.

Once within 4 hours after criticality following fuel movement within the reactor pressure vessel or control rod replacement O O HATCH UNIT 2 3.1-4 REVISION A l

~ Secondary Containment Isolation Instrumentation l 3.3.6.2 ; 3.3 INSTRUMENTATION 3.3.6.2 Secondary Containment Isolation Instrumentation l LC0 3.3.6.2 The secondary containment isolation instrumentation for each ; Function in Table 3.3.6.2-1 shall be OPERABLE. ! APPLICABILITY: According to Table 3.3.6.2-1. ACTIONS

     --------__-_-_-----------------------NOTE-------------------------------------                                                              ,

Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in 12 hours for inoperable. trip. Function 2 AND 24 hours for Functions other than Function 2 B. One or more automatic B.1 Restore isolation 1 hour Functions with capability. isolation capability not maintained. C. Required Action and C.1.1 Isolate the 1 hour associated Completion associated Time of Condition A penetration flow or B not met. path (s). 08 (continued) O HATCH UNIT 2 3.3-59 REVISIONh(,

Secondary Containment Isolation Instrumentation 3.3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.I.2 Declare associated I hour secondary containment isolation valves inoperable. AND C.2.1 Place the associated I hour standby gas treatment (SGT) subsystem (s) in operation. OB C.2.2 Declare associated I hour SGT subsystem (s) inoperable. O SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment Isolation Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains isolation capability. I D SURVEILLANCE FREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK. 12 hours (continued)

O HATCH UNIT 2 3.3-60 REVISION D

Secondary Containment Isolation Instrumentation 3.3.6.2 [d

SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.6.2.3 Perform CHANNEL CALIBRATION. 92 days SR 3.3.6.2.4 Perform CHANNEL CALIBRATION. 18 months SR 3.3.6.2.5 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months O

i l 1 O HATCH UNIT 2 3.3-61 REVISION A

i l l Secondary Containment Isolation Instrumentation l 3.3.6.2 l l l Table 3.3.6.2-1 (page 1 of 1) secondary Containment Isolation Instrumentation 9Il l APPLICABLE i MODES OR REQUIRED i OTHER CHANNELS l SPECIFIED PER SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE i

1. Reactor Vessel Water 1,2,3, 2 SR 3.3.6.2.I e -47 inches Level -Low Low, Level 2 (a) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.5

2. Drywell Pressure -High 1,2,3 2 SR 3.3.6.2.1 5 1.92 psig I SR 3.3.6.2.2 I SR 3.3.6.2.4 1 SR 3.3.6.2.5 l 1

3. Reactor Building Exhaust 1,2,3, 2 SR 3.3.6.2.1 5 60 mR/hr j Radiat ion -H i gh (a) SR 3.3.6.2.3 i SR 3.3.6.2.5 ;

4. Refueling Floor Exhaust 1,2,3, 2 SR 3.3.6.2.1 5 20 mR/hr i R adi at ion -H i gh

                                    $(a)'(b)                       SR    3.3.6.2.3                        l SR   3.3.6.2.5 (a) During operations with a potential for draining the reactor vessel.

(b) During CORE ALTERAfl0NS and during movement of irradiated fuel assemblies in secondary containment. l l 1 1 l l 1 l l y l l HATCH UNIT 2 3.3-62 REVISIONff(7

MCREC System Instrumentation 3.3.7.1 r"'si i 3.3 INSTRUMENTATION %) 3.3.7.1 Main Control Room Environmental Control (MCREC) System Instrumentation LC0 3.3.7.1 Two channels of the Control Room Air Inlet Radiation-High Function shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the l secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs). ACTIONS

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

Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or both channels A.1 Declare associated I hour from inoperable. MCREC subsystem (s) discovery of inoperable. loss of MCREC initiation capability in both trip systems AND A.2 Place channel in 6 hours trip. (continued) O HATCH UNIT 2 3.3-67 REVISION /4 I

MCREC System Instrumentation 3.3.7.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME j B. Required Action and B.1 Place the associated I hour associated Completion MCREC subsystem (s) in Time not met. the pressurization mode of operation. OE B.2 Declare associated I hour MCREC subsystem (s) inoperable. SURVEILLANCE REQUIREMENTS

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

When a Control Room Air Inlet Radiation-High channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the other channel is OPERABLE. SURVEILLANCE FREQUENCY SR 3.3.7.1.1 Perform CHANNEL CHECK. 24 hours SR 3.3.7.1.2 Perform CHANNEL FUNCTIONAL TEST. 31 days SR 3.3.7.1.3 Perform CH%r ti. CALIBRATION. The 92 days Allowable Value shall be s 1 mr/ hour. SR 3.3.7.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months e HATCH UNIT 2 3.3-68 REVISION A

1 l l ECCS - Shutdown ) 3.5.2 (- 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION I COOLING (RCIC) SYSTEM 3.5.2 ECCS - Shutdown LC0 3.5.2 Two low pressure ECCS injection / spray subsystems shall be OPERABLE. l APPLICABILITY: MODE 4, MODE 5, except with the spent fuel storage pool gates removed and water level 2 22 ft 1/8 inches over the top of the reactor pressure vessel flange. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required ECCS A.1 Restore required ECCS 4 hours injection / spray injection / spray subsystem inoperable. subsystem to OPERABLE status. B. Required Action and B.1 Initiate action to Immediately associated Completion suspend operations Time of Condition A with a potential for not met, draining the reactor vessel (0PDRVs). C. Two required ECCS C.1 Initiate action to Immediately injection / spray suspend OPORVs. subsystems inoperable. AND C.2 Restore one ECCS 4 hours injection / spray subsystem to OPERABLE status. 1 (continued) O l HATCH UNIT 2 3.5-7 REVISION A

ECCS - Shutdown 3.5.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME i D. Required Ar. tion C.2 D.1 Initiate action to Immediately and i.ssaciated restore secondary l l Comtl atian Time not containment to 1 met. OPERABLE status. AND D.2 Initiate action to Immediately I restore required l ; standby gas trectment l subsystem (s) to l l OPERABLE status. AND D.3 Initiate action to Immediately restore isolation j capability in each required secondary l containment penetration flow path & not isolated. W 0 HATCH UNIT 2 3.5-8 REVISION

ECCS - Shutdown 3.5.2 s i (U I 1 SURVEILLANCE REQUIREMENTS I SURVEILLANCE FREQUENCY j SR 3.5.2.1 Verify, for each required low pressure 12 hours coolant injection (LPCI) subsystem, the suppression pool water level is 2 146 inches. SR 3.5.2.2 Verify, for each required core spray (CS) 12 hours subsystem, the:

a. Suppression pool water level is 2 146 inches; or

b. -----------------NOTE-----------------

Only one required CS subsystem may O' take credit for this option during OPDRVs. Condensate storage tank water level is 2 12 ft. (continued) O HATCH UNIT 2 3.5-9 REVIS10N/(6

ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.2.3 Verify, for each required ECCS injection / 31 days spray subsystem, the piping is filled with water from the pump discharge valve to the injection valve. SR 3.5.2.4 -------------------NOTE------------------ One LPCI subsystem may be considersi OPERABLE during alignment and opera' ion for decay heat removal if capable of being manually realigned and not otherwise inoperable. I Verify each required ECCS injection / spray 31 days subsystem 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. O, SR 3.5.2.5 Verify each required ECCS pump develops the In accordance specified flow rate against a system head with the corresponding to the specified reactor Inservice pressure. Testing Program SYSTEM HEAD N0. CORRESPONDING I 0F TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF CS 2 4250 gpm 1 2 113 psig LPCI 2 7700 gpm 1 2 20 psig i SR 3.5.2.6 -------------------NOTE-------------------- Vessel injection / spray may be excluded. Verify each required ECCS injection / spray 18 months subsystem actuates on an actual or simulated automatic initiation signal. 1 1 HATCH UNIT 2 3.5-10 REVISION A l

Drywell Cooling System Fans 3.6.3.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i SR 3.6.3.3.1 Operate each required drywell cooling 92 days l system fan for 2: 15 minutes. l l I i O O HATCH UNIT 2 3.6-37 REVISION A

Secondary Containment l 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment l LC0 3.6.4.1 The secondary containment shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment, During CGRE ALTERATIONS, During operations with a potential for draining the reactor vessel (OPDRV). ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment A.1 Restore secondary 4 hours inoperable in MODE 1, containment to 2, or 3. OPERABLE status. O B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND l not met. B.2 Be in MODE 4. 36 hours C. Secondary containment C.1 --------NOTE-------- inoperable during LC0 3.0.3 is not movement of irradiated applicable. l fuel assemblies in the -------------------- secondary containment, during CORE Suspend movement of Immediately ALTERATIONS, or during irradiated fuel OPDRVs. assemblies in the secondary containment.

                                   @_D                                    (continued)

O HATCH UNIT 2 3.6-38 REVISION [/ l

Drywell Cooling System Fans 3.6.3.3 ( SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.3.1 Operate each required drywell cooling 92 days system fan for 215 minutes. O l O l HATCH UNIT 2 3.6-37 REVISION A

1 l Secondary Containment l 3.6.4.1 3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment l LC0 3.6.4.1 The secondary containment shall be OPERABLE. l l l APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the l secondary containment, i During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRV). ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment A.1 Restore secondary 4 hours , inoperable in MODE 1, containment to l 2, or 3. OPERABLE status. 1 B. Required Action and B.1 Be in MODE 3. 12 hours 0 ! associated Completion Time of Condition A AND l not met. B.2 Be in MODE 4. 36 hours C. Secondary containment C.1 --------NOTE-------- inoperable during LC0 3.0.3 is not movement of irradiated applicable. fuel assemblies in the -------------------- secondary containment, l during CORE Suspend movement of Immediately ALTERATIONS, or during irradiated fuel OPDRVs. assemblies in the secondary containment. AND (continued) O HATCH UNIT 2 3.6-38 REVISION [/

Secondary Containment l 3.6.4.1 /~% Q ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 Suspend CORE Immediately ALTERATIONS. AND C.3 Initiate action to Immediately suspend OPDRVs. O HATCH UNIT 2 3.6-39 REVISION D

Secondary Containment l 3.6.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify all secondary containment 31 days l equipment hatches are closed and sealed. 1 SR 3.6.4.1.2 Verify each secondary containment access 31 days l door is closed, except when the access opening is being used for entry and exit, then at least one door shall be closed. SR 3.6.4.1.3 -------------------NOTE------------------ The number of standby gas treatment (SGT) subsystem (s) required for this Surveillance is dependent on the secondary containment configuration, and shall be one less than the number required to meet LC0 3.6.4.3, " Standby Gas Treatment (SGT) System," for the given configuration. T Verify required SGT subsystem (s) will 18 months on a draw down the secondary containment to STAGGERED TEST 2 0.25 inch of vacuum water gauge in BASIS s 120 seconds. (continued) O HATCH UNIT 2 3.6-40 REVISION)T'

Secondary Containment l 3.6.4.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.4.1.4 ------------------NOTE------------------- The number of SGT subsystems required for this Surveillance is dependent on the secondary containment configuration, and shall be one less than the number required to meet LC0 3.6.4.3, " Standby Gas Treatment (SGT) System," for the given configuration. Verify required SGT subsystem (s) can 18 months on a l maintain 2 0.25 inch of vacuum water STAGGERED TEST gauge in the secondary containment for BASIS 1 hour at a flow rate :s 4000 cfm for each subsystem. v I i HATCH UNIT 2 # 3.6-41 REVISION D

                                                                                /        1 1

I l

.1

SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 1 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) l l LCO 3.6.4.2 Each SCIV shall be OPERABLE. l l l l APPLICABILITY: MODES I, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment, ; During CORE ALTERATIONS, ' During operatios with a potential for draining the reactor vessel (0PDRV). ACTIONS

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

1. Penetration flow paths may be unisolated intermittently under l administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made

____ _!$!__ _$_ ___ I__________________. ___________________________________ . 1 l CONDITION REyVIRED ACTION COMPLETION TIME i I A. One or more A.I Isolate the affected 8 hours penetration flow paths penetration flow path with one SCIV by use of at least inoperable, I one closed and de- ! activated automatic l valve, closed manual I valve, or blind i flange. i l AND (continued) O HATCH UNIT 2 3.6-42 REVISION

SCIVs 3.6.4.2 f i) g ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 --------NOTE--------- Isolation devices in high radiation areas ' may be verified by use of administrative means. Verify the affected Once per penetration flow path 31 days is isolated. B. One or more B.1 Isolate the affected 4 hours penetration flow paths penetration flow path with two SCIVs by use of at least inoperable, one closed and de-activated automatic valve, closed manual (. valve, or blind flange. C. Required Action and C.1 Be in MOD 3. 12 hours associated Completion Time of Condition A or _AtLQ B not met in MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours (continued) O HATCH UNIT 2 3.6-43 REVISIONfq i

SCIVs , 3.6.4.2 l l ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 --------NOTE----- associated Completion LCO 3.0.3 is not Time of Condition A applicable. I or B not met during ----------------- movement of irradiated fuel assemblies in the Suspend movement of Immediately secondary containment, irradiated fuel during CORE assemblies in the i ALTERATIONS, or during secondary ; OPDRVs. containment. I AND D.2 Suspend CORE Immediately ALTERATIONS. ' AND D.3 Initiate action to Immediately suspend OPDRVs. l O HATCH UNIT 2 3.6-44 REVISION ((j

1 l l SCIVs l 3.6.4.2 l r 'g ),. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.2.1 ------------------NOTES------------------ l l

1. Valves and blind flanges in high l radiation areas may be verified by !

use of administrative means. '

2. Not required to be met for SCIVs that !

are open under administrative i controls. l Verify each secondary containment 31 days isolation manual valve and blind flange that is required to be closed during accident conditions is closed. SR 3.6.4.2.2 Verify the isolation time of each power 92 days operated and each automatic SCIV is within limits.

%d SR  3.6.4.2.3     Verify each automatic SCIV actuates to           18 months the isolation position on an actual or                               i simulated actuation signal.                                          l 1

l

I HATCH UNIT 2 3.6-45 REVISION /6 l 1 l

SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System l LC0 3.6.4.3 The Unit I and Unit 2 SGT subsystems required to support , LC0 3.6.4.1, " Secondary Containment," shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in secondary containment, ; During CORE ALTERATIONS, ' During operatios with a potential for draining the reactor vessel (0PDRV). ACTIONS

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

When two Unit 1 SGT subsystems are placed in an inoperable status solely for inspection of the Unit I hardened vent rupture disk, entry into associated Conditions and Required Actions may be delayed for up to 24 hours, provided both Unit 2 SGT subsystems are OPERABLE. CONDITION REQUIRED ACTION COMPLETION TIME A. One required Unit 1 A.1 Restore required 30 days from SGT subsystem Unit 1 SGT subsystem discovery of inoperable while: to OPERABLE status. failure to meet the LCO

1. Four SGT subsystems required OPERABLE, and

2. Unit I reactor building-to-refuel floor plug not installed. ;

1 (continued) G' HATCH UN!T 2 3.6-46 REVISION /G

SGT System 3.6.4.3 i ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. One required Unit 2 B.1 Restore required SGT 7 days SGT subsystem subsystem to OPERABLE inoperable. status. AND E 30 days from discovery of One required Unit 1 failure to meet SGT subsystem the LC0 inoperable for reasons other than Condition A. C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A or AND 8 not met in MODE 1, 2, or 3. C.2 Be in MODE 4. 36 hours (~ ( D. Required Action and --------------NOTE----------- associated Completion LC0 3.0.3 is not applicable. Time of Condition A ----------------------------- or B not met during movement of irradiated D.1 Place remaining Immediately fuel assemblies in the OPERABLE SGT secondary containment, subsystem (s) in during CORE operation. ALTERATIONS, or during OPDRVs. @ (continued) O HATCH UNIT 2 3.6-47 REVISION A -

'Y

SGT System 3.6.4.3 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) D.2.1 Suspend movement of Immediately irradiated fuel assemblies in secondary containment. i AND Immediately D.2.2 Suspend CORE ALTERATIONS. bl!D Immediately D.2.3 Initiate action to suspend OPDRVs. I E. Two or more required E.1 Enter LCO 3.0.3. Immediately SGT subsystems inoperable in MODE 1, ) 2, or 3. l I l

i

l 1 i e HATCH UNIT 2 3.6-48 REVISION [h

SGT System 3.6.4.3 D Q ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME F. Two or more required F.1 --------NOTE----- SGT subsystems LC0 3.0.3 is not inoperable during applicable. movement of irradiated ----------------- fuel assemblies in the secondary containment, Suspend movement of Immediately during CORE irradiated fuel ALTERATIONS, or during assemblics in OPDRVs. secondary containment. AND F.2 Suspend CORE ALTERATIONS. AND F.3 Initiate action to suspend OPDRVs. 9 a l i O HATCH UNIT 2 3.6-49 REVISION I I

SGT System 3.6.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.3.1 Operate each required SGT subsystem for 31 days l 2 10 continuous hours with heaters operating. SR 3.6.4.3.2 Perform required SGT filter testing in In accordance l accordance with the Ventilation Filter with the VFTP Testing Program (VFTP). SR 3.6.4.3.3 Verify each required SGT subsystem 18 months l actuates on an actual or simulated l I initiation signal. O l i l l O\ HATCH UNIT 2 3.6-50 j REVISION [(7'

MCREC System 3.7.4 i [' 3.7 PLANT SYSTEMS ; 3.7.4 Main Control Room Environmental Control (MCREC) System l LC0 3.7.4 Two MCREC subsystems shall be OPERABLE. j APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the l secondary containment, During CORE ALTERATIONS, ! During operations with a potential for draining the reactor I vessel (0PDRVs). ! ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One MCREC subsystem A.1 Restore MCREC 7 days inoperable. subsystem to OPERABLE status. 13 1 V B. Required Action and 8.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours (continued) f% U HATCH UNIT 2 3.7-9 REVISION \(-

MCREC System 3.7.4 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and ------------NOTE------------- associated Completion LC0 3.0.3 is not applicable. Time of Condition A ----------------------------- not met during movement of irradiated C.1 Place OPERABLE MCREC Immediately fuel assemblies in the subsystem in secondary containment, pressurization mode. l during CORE ALTERATIONS, or during OR OPDRVs. C.2.1 Suspend movement of Immediately irradiated fuel assemblies in the secondary l containment. AND C.2.2 Suspend CORE Immediately ALTERATIONS. AND C.2.3 Initiate action to Immediately suspend OPDRVs. D. Two MCREC subsystems D.1 Enter LC0 3.0.3. Immediately inoperable in MODE 1, 2, or 3. l (continued) 1 l 9 HATCH UNIT 2 3.7-10 REVISION \(.,

MCREC System ! 3.7.4 '

) ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

E. Two MCREC subsystems ------------NOTE------------- inoperable during LC0 3.0.3 is not applicable, movement of irradiated ----------------------------- fuel assemblies in the secondary containment, E.1 Suspend movement of Immediately l during CORE irradiated fuel ALTERATIONS, or during assemblies in the OPDRVs. secondary l containment. AND E.2 Suspend CORE Immediately ALTERATIONS. AND E.3 Initiate action to Immeciately suspend OPDRVs. O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate each MCREC subsystem 2: 15 minutes. 31 days SR 3.7.4.2 Perform required MCREC filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP). SR 3.7.4.3 Verify each MCREC subsystem actuates on an 18 months actual or simulated initiation signal. (continued) l HATCH UNIT 2 3.7-11 REVISIONh(f 1

MCREC System 3.7.4 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.7.4.4 Verify each MCREC subsystem can maintain 18 months on a a positive pressure of 2 0.1 inches STAGGERED water gauge relative to the turbine TEST BASIS building during the pressurization mode of operation at a subsystem flow rate of s 2750 cfm and an outside air flow rate s 400 cfm. l Oll l I O HATCH UNIT 2 3.7-12 REVISION A

Control Room AC System 3.7.5 ( ) 3.7 PLANT SYSTEMS

%)

3.7.5 Control Room Air Conditioning (AC) System l LC0 3.7.5 Three control room AC subsystems shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, I During movement of irradiated fuel assemblies in the l secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs). ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME j l A. One control room AC A.1 Verify outside air 1 hour subsystem inoperable. temperature s 65 F. AND Once per 12 hours thereafter . AND A.2 Verify maximum I hour outside air ! temperature in the I previous 24 hours s 65 F. B. Required Action and B.1 Restore control room 30 days associated Completion AC subsystem to Time of Condition A OPERABLE status. not met. (continued) O 1 HATCH UNIT 2 3.7-13 REVISION &(,

Control Room AC System 3.7.5 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Two control room AC C.1 Verify outside air 1 hour subsystems inoperable. temperature is s 65 F. AND Once per 12 hours thereafter AND l C.2 Verify maximum I hour outside air temperature in the previous 24 hours s 65 F. AND C.3 Restore one control 30 days room AC subsystem to OPERABLE status. D. Required Action and 0.1 Be in MODE 3. 12 hours associated Completion Time of Condition 8 or AND C not met in MODE 1, l 2, or 3. D.2 Be in MODE 4. 36 hours (continued) O HATCH UNIT 2 3.7-14 REVISION A

Control Room AC System 3.7.5 ( ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and ------------NOTE------------- associated Completion LC0 3.0.3 is not applicable. Time of Condition B or ----------------------------- C not met during movement of irradiated E.1 Place necessary Immediately fuel assemblies in the OPERABLE control room secondary containment, AC subsystems in l during CORE operation. ALTERATIONS, or during OPDRVs. 0_R E.2.1 Suspend movement of Immediately irradiated fuel assemblies in the secondary l containment. AND E.2.2 Suspend CORE Immediately ALTERATIONS. AND E.2.3 Initiate action to Immediately suspend OPDRVs. F. Three control room AC F.1 Enter LC0 3.0.3. Immediately subsystems inoperable in MODE 1, 2, or 3. (continued) l I l l HATCH UNIT 2 3.7-15 REVISIONA6,

4 Control Room AC System 3.7.5 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME G. Three control room AC ------------NOTE------------- subsystems inoperable LC0 3.0.3 is not applicable. during movement of ----------------------------- irradiated fuel assemblies in the G.1 Suspend movement of Immediately l secondary containment, irradiated fuel during CORE assemblies in the ALTERATIONS, or during secondary l OPDRVs. containment. AND G.2 Suspend CORE Immediately ALTERATIONS. b!LQ G.3 Initiate actions to Immediately suspend OPDRVs. O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each control room AC subsystem has 18 mo- tu; the capability to remove the assumed heat load. O HATCH UNIT 2 3.7-16 REVISION %(., 1

AC Sources - Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating i LC0 3.8.1 The following AC electrical power sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the Unit 2 onsite Class 1E AC Electrical '

Power Distribution System;

b. Two Unit 2 diesel generators (DGs);

c. The swing DG;

d. One Unit 1 DG; and

e. One qualified circuit between the offsite transmission network and the Unit 1 onsite Class IE AC Electrical Power Distribution subsystem (s) needed to support the ,

Unit 1 equipment required to be OPERABLE by LCO 3.6.4.3,

                          " Standby Gas Treatment (SGT) System," LC0 3.7.4, " Main         ;

Control Room Environmental Control (MCREC) System," and.

LC0 3.7.5,-" Control Room Air Conditioning (AC) System." ' O APPLICABILITY: MODES 1, 2, and 3. l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite A.1 Perform SR 3.8.1.1 1 hour circuit inoperable. for OPERABLE required offsite circuits. AND Once per 8 hours , thereafter AND (continued) O ! HATCH UNIT 2 3.8-1 REVISION Q j

_ - _ . ~~ _ ______________U

AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i A. (continued) A.2 Declare required 24 hours from feature (s) with no discovery of no l offsite power offsite power to 1 available inoperable one 4160 V ESF when the redundant bus concurrent required feature (s) with are inoperable. inoperability of , redundant I required , feature (s) l AND A.3 Restore required 72 hours l offsite circuit to OPERABLE status. l AND 10 days from discovery of failure to meet LC0 3.8.1.a, b, or c B. One Unit 2 or the B.1 Perform SR 3.8.1.1 1 hour swing DG inoperable. for OPERABLE required offsite circuit (s). AND Once per 8 hours thereafter AND B.2 Declare required 4 hours from feature (s), supported discovery of by the inoperable DG, Condition B inoperable when the concurrent with redundant required inoperability of feature (s) are redundant inoperable, required feature (s) MQ (continued) O HATCH 'JNIT 2 3.8-2 REVISION A

AC Sourt.es - Operating 3.8.1 0 r O h O C. HATCH UNIT 2 3.8-39 REVISION l

AC Sources - St.utdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS 3.8.2 AC Sources - Shutdown h LCO 3.8.2 The following AC electrical power sources shall be OPERABLE:

a. One qualified circuit between the offsite transmission network and the onsite Unit 2 Class 1E AC electrical power distribution subsystem (s) required by LC0 3.8.8,

                              " Distribution Systems - Shutdown;"

b. One Unit 2 diesel generator (DG) capable of supplying one subsystem of the onsite 0,..t 2 Class IE AC electrical power distribution subsystem (s) required by LC0 3.8.8;

c. One qualified circuit between the offsite transmission l network and the onsite Unit 1 Class 1E AC electrical l power distribution subsystem (s) needed to support the Unit 1 equipment required to be OPERABLE by LCO 3.6.4.3,

                              " Standby Gas Treatment (SGT) System," LC0 3.7.4, " Main Control Room Environmental Control (MCREC) System," and LC0 3.7.5, " Control Room Air Conditioning (AC) System;"

and

d. One Unit 1 DG capable of supplying one subsystem of each of the Unit I equipment required to be OPERABLE by LC0 3.6.4.3, LC0 3.7.4, and LC0 3.7.5. l l APPLICABILITY: MODES 4 and 5, During movement of irradiated fuel assemblies in the l secondary containment.

1 9 4 HATCH UNIT 2 3.8-RAO REVISIONq(

 . _ _       _ _ _ _ _ _                                                                                                      \

i l AC Sources - Shutdown 3.8.2 l O ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME l A. One or more required ------------NOTE------------- offsite circuit (s) Enter applicable Condition l inoperable. and Required Actions of LC0 3.8.8, with one required 4160 V ESF bus de-energized as a result of Condition A. A.1 Declare affected Immediately required feature (s), with no offsite power available, inoperable. 08 A.2.1 Suspend CORE Immediately ALTERATIONS. AND ()) A.2.2 Suspend movement of Immediately irradiated fuel assemblies in the secondary l containment. AND A.2.3 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel (0PDRVs). AND A.2.4 Initiate action to Immediately restore required offsite power circuit (s) to OPERABLE status. (continued) O HATCH UNIT 2 3.8- M Al REVISION ((_.

l AC Sources - Shutdown 3.8.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. One or more required B.1 Sc<oer.d CORE Immediately DG(s) inoperable. A!TERATIONS. AND B.2 Suspend movement of Immediately irradiated fuel assemblies in l secondary containment. AND B.3 Initiate action to Immediately suspend OPDRVs. AND B.4 Initiate action to Immediately restore required DG(s) to OPERABLE status. I l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.2.1 -------------------NOTE-------------------- The following SRs are not required to be performed: SR 3.8.1.2.b, SR 3.8.1.7 through SR 3.8.1.9, SR 3.8.1.11 through SR 3.8.1.14, SR 3.8.1.16, and SR 3.8.1.17. For required Unit 2 AC sources, the SRs of In accordance LC0 3.8.1, except SR 3.8.1.6, SR 3.8.1.15, with applicable and SR 3.8.1.18, are applicable. SRs (continued) HATCH UNIT 2 3.8- 4 1 1 REVISIONk(- m

Diesel Fuel Oil and Transfer, Lube Oil, and Starting Air 3.8.3 l ] SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify each Unit 2 and swing DG fuel oil 31 days storage tank contains 2 33,000 gallons of fuel. SR 3.8.3.2 Verify each required DG lube oil inventory 31 days is a 400 gallons. SR 3.8.3.3 Verify fuel oil total particulate In accorance concentration of Unit 2 and swing DG stored with the Diesel fuel oil are tested in accordance with, and Fuel Oil maintained within the limits of, the Diesel Testing Program Fuel Oil Testing Program. 3.8.3.4 Verify each required DG air start receiver 31 days O SR pressure is 2 225 psig. SR 3.8.3.5 Verify each Unit 2 and swing DG fuel oil 31 days transfer subsystem operates to automatically transfer fuel oil from the storage tank to the day tank. SR 3.8.3.6 Check for and remove accumulated water from 184 days each Unit 2 and swing DG fuel oil storage tank. SR 3.8.3.7 Verify each Unit 2 and swing DG fuel oil 18 months transfer subsystem operates to manually transfer fuel from the associated fuel oil storage tank to the day tank of each required DG. O HATCH UNIT 2 3.8-27 REVISION A i

1 l DC Sources - Operating i 3.8.4 ! l 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources - Operating l l LC0 3.8.4 The following DC electrical power subsystems shall be ; OPERABLE: i l l ! a. The Unit 2 Division 1 and Division 2 station service DC I l electrical power subsystems;

b. The Unit 2 and swing DGs DC electrical power subsystems; l c. The Unit 1 DG DC electrical power subsystems needed to l support the Unit 1 equipment required to be OPERABLE by l

LC0 3.6.4.3, " Standby Gas Treatment (SGT) System," l LC0 3.7.4, " Main Control Room Environmental Control (MCREC) System," LCO 3.7.5, " Control Room Air Conditioning (AC) System," and LC0 3.8.1, "AC Sources-0perating." i APPLICABILITY: MODES I, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Swing DG DC electrical A.1 Restore DG DC 7 days power subsystem electrical power inoperable due to subsystem to OPERABLE i performance of status. SR 3.8.4.7 or SR 3.8.4.8. 0.R One or more required Unit 1 DG DC electrical power - subsystems inoperable. 1 1 (continued) O HATCH UNIT 2

3. 8-2(c REVISION,A'(;

DC Sources - Shutdown l' 3.8.5 3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources - Shutdown LCO 3.8.5 The following DC electrical power subsystems shall be ' OPERABLE:

a. The Unit 2 DC electrical power subsystems needed to support the DC electrical power distribution subsystem (s) required by LC0 3.8.8, " Distribution Systems - Shutdown"; and

b. The Unit 1 DG DC electrical power subsystems needed to support the equipment required to be OPERABLE by LC0 ,

3.6.4.3, " Standby Gas Treatment (SGT) System," and LC0 l 3.7.4, " Main Control Room Environmental Control- (MCREC) ' System, "LCO 3.7.5, " Control Room Air Conditioning (AC) : System," and 3.8.2, "AC Sources ~- Shutdown." l t APPLICABILITY: MODES 4 and 5, ; During movement of irradiated fuel assemblies in the l secondary containment. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i A. One or more required A.1 Declare affected Immediately ' DC electrical power required feature (s) subsystems inoperable. inoperable. ' OB i A.2.1 Suspend CORE Immediately ALTERATIONS. AND A.2.2 Suspend movement of Immediately irradiated fuel i assemblies in the , secondary l containment. AND (continued) O ' HATCH UNIT 2 3.8-343 REVISION M l

l l DC Sources - Shutdown 3.8.5 ACTIONS i CONDITION REQUIRED ACTION COMPLETION TIME I A. (continued) A.2.3 Initiate action to Immediately i suspend operations with a potential for l draining the reactor i vessel. MQ l l A.2.4 Initiate action to Immediately i restore required DC ' electrical power I subsystems to ) OPERABLE status. I O i l l l 9 HATCH UNIT 2 3.8-34 REVISION A

Battery Cell Parameters 3.8.6 Ov Table 3.8.6-1 (page 1 of 2) Battery Cell Parameter Requirements CATEGORY A: CATEGORY B: CATEGORY C: LIMITS FOR EACH LIMITS FOR EACH LIMITS DESIGNATED PIL0T CONNECTED CELL FOR EACH PARAMETER CELL CONNECTED CELL Electrolyte > Minimum level > Minimum level Above top of Level indication mark, indication mark, plates, and not and s % inch above and s % inch above overflowing maximum level maximum level indication mark (a) indication mark (a) Float Voltage 2 2.13 V 2 2.13 V > 2.07 V Specific 2 1.200 2 1.195 Not more than Gravity (b)(c) 0.020 below . /^ ' MLD average of all ! connected cells Average of all connected cells AND

                                               > 1.205 Average of all connected cells 2 1.195 (a)   It is acceptable for the electrolyte level to temporarily increase above the specified maximum level during equalizing charges provided it is not overflowing.

(b) Corrected for electrolyte temperature and level. Level correction is n'ot required, however, when on float charge battery charging is < 1 amp for station service batteries and < 0.5 amp for DG batteries. (c) A battery charging current of < 1 amp for station service batteries and

        < 0.5 amp for DG batteries when on float charge is acceptable for meeting specific gravity limits following a battery recharge, for a maximum of 7 days. When charging current is used to satisfy specific gravity requirements, specific gravity of each connected cell shall be O        measured prior to expiration of the 7 day allowance.

HATCH UNIT 2 3.8-39 REVISION D

l 1 Distribution Systems - Operating 3.8.7 3.8 ELECTRICAL POWER SYSTEMS h I 3.8.7 Distribution Systems - Operating LC0 3.8.7 The following AC and DC electrical power distribution subsystems shall be OPERABLE:

a. Unit 2 AC and DC electrical power uistribution subsystems comprised of:

1. 4160 V essential buses 2E, 2F, and 2G;

2. 600 V essential buses 2C and 2D;

3. 120/208 V essential cabinets 2A and 2B;

4. 120/208 V instrument buses 2A and 28;

5. 125/250 V DC station service buses 2A and 2B;

6. DG DC electrical power distribution subsystems; and

b. Unit 1 AC and DC electrical power distribution subsystems needed to support equipment required to be OPERABLE by LCO 3.6.4.3, " Standby Gas Treatment (SGT)

System," LC0 3.7.4, " Main Control Room Environmental Control (MCREC) System," LC0 3.7.5, " Control Room Air Conditioning (AC) System," and LC0 3.8.1, "AC Sources-0perating." APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required Unit 7 days Unit 1 AC or DC 1 AC and DC electrical power subsystem (s) to distribution OPERABLE status. subsystems inoperable. , 1 (continued) O HATCH UNIT 2 3.8-410 REVISION %r- , l

1 Distribution Systems - Shutdown 3.8.8 3 3.8 ELECTRICAL POWER SYSTEMS 3.8.8 Distribution Systems - Shutdown *

LC0 3.8.8 The necessary portions of the following AC and DC electrical power distribution subsystems shall be OPERABLE:

a. The Unit 2 AC and DC electrical power distribution subsystems needed to support equipment required to be OPERABLE; and

b. The Unit 1 AC and DC electrical power distribution subsystems needed to support equipment required to be OPERABLE by LCO 3.6.4.3, " Standby Gas Treatment (SGT)

System," LC0 3.7.4, " Main Control Room Environmental Control (MCREC) System," LC0 3.7.5, " Control Room Air Conditioning (AC) System," and LC0 3.8.2, "AC Sources-Shutdown. " APPLICABILITY: MODES 4 and 5, During movement of irradiated fuel assemblies in the j : secondary containment. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare associated Immediately AC or DC electrical supported required power distribution feature (s) subsystems inoperable. inoperable. 0E A.2.1 Suspend CORE Immediately ALTERATIONS. AND (continued) O HATCH UNIT 2 3.8-4% 3 REVISION h(

1 Distribution Systems - Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2 Suspend handling of Immediately irradiated fuel assemblies in the secondary l containment. AllD A.2.3 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel. AND A.2.4 Initiate actions to Immediately restore required AC and DC electrical power distribution subsystem (s) to OPERABLE status. AND I i A.2.5 Declare associated Immediately l required shutdown ! cooling subsystem (s) inoperable and not in operation. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.I Verify correct breaker alignments and 7 days voltage to required AC and DC electrical power distribution subsystems. O HATCH UNIT 2 3.8-4% REVISIONK(_.

RPV Water level 3.9.6 l 3.9 REFUELING OPERATIONS 'V 3.9.6 Reactor Pressure Vessel (RPV) Water Level LC0 3.9.6 RPV water level shall be 2 23 ft above the top of the irradiated fuel assemblies seated within the RPV. APPLICABILITY: During movement of irradiated fuel assemblies within the RPV, During movement of new fuel assemblies or handling of control rods within the RPV, when irradiated fuel assemblies are seated within the RPV. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RPV water level not A.1 Suspend movement of Immediately within limit. fuel assemblies and handling of control rods within the RPV. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify RPV water level is 2 23 ft above the 24 hours top of the irradiated fuel assemblies seated within the RPV. O HATCH UNIT 2 3.9-9 REVISION A

RHR -High Water Level 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Residual Heat Removal (RHR) - High Water Level LC0 3.9.7 One RHR shutdown cooling subsystem shall be OPERABLE and in operation.

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

The required RHR shutdown cooling subsystem may be removed from operation for up to 2 hours per 8 hour period. APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level 2: 22 ft 1/8 inches above the top of the RPV flange. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required RHR shutdown A.1 Verify an alternate 1 hour cooling subsystem inoperable. method of decay heat removal is available. AND W ; Once per 24 hours thereafter l B. Required Action and B.1 Suspend loading Immediately associated Completion irradiated fuel Time of Condition A assemblies into the , not met. RPV. l MLD l B.2 Initiate action to Immediately l restore secondary l containment to OPERABLE status. AND l (continued) 9 HATCH UNIT 2 3.9-10 REVISION h(,

RHR -High Water Level 3.9.7 f3 g ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3 Initiate action to Immediately restore required l standby gas treatment subsystem (s) to l OPERABLE status. h!LD B.4 Initiate action to Immediately restore isolation capability in each required secondary l containment penetration flow path not isolated. C. No RHR shutdown C.1 Verify reactor 1 hour from cooling subsystem in coolant circulation discovery of no operation. by an alternate reactor coolant method. circulation AND One per 12 hours thereafter AND C.2 Monitor reactor Once per hour coolant temperature. O HATCH UNIT 2 3.9-11 REVISION %G

RHR -High Water Level 3.9.7 SURVEILLANCE REQUIREMEt65 SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify one RHR shutdown cooling subsystem 12 hours is operating. i O l I i HATCH UNIT 2 3.9-12 REVISION A m

RHR - Low Water Level 3.9.8 A Q 3.9 REFUELING OPERATIONS I 3.9.8 Residual Heat Removal (RHR) - Low Water Level LC0 3.9.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and ; one RHR shutdown cooling subsystem shall be in operation.

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

The required operating shutdown cooling subsystem may be removed from operation for up to 2 hours per 8 hour period. APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level < 22 ft 1/8 inches above the top of the RPV flange. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ( ,N A. One or two required A.1 Verify an alternate 1 hour RHR shutdown cooling method of decay heat subsystems inoperable. removal is available AND for each inoperable required RHR shutdown Once per cooling subsystem. 24 hours thereafter B. Required Action and B.1 Initiate action to Immediately associated Completion restore secondary l Time of Condition A containment to not met. OPERABLE status. AND B.2 Initiate action to Immediately restore required l standby gas treatment subsystem (s) to l OPERABLE status. AND (continued) HATCH UNIT 2 3.9-13 REVISIONh(_,

RHR - Low Water Level 3.9.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3 li.itiate action to Immediately restere isolation capabil ty in each required secondary l containment penetration flow path not isolated. C. No RHR shutdown C.I Verify reactor I hour from cooling subsystem in coolant circulation discovery of no operation. by an alternate reactor coolant method. circulation AND Once per 12 hours thereafter AND C.2 Monitor reactor Once per hour coolant temperature. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l l SR 3.9.8.1 Verify one RHR shutdown cooling subsystem 12 hours 1 is operating. ! l 9 HATCH UNIT 2 3.9-14 REVISION %4

Inservice Leak and Hydrostatic Testing Operation 3.10.1 / 3.10 SPECIAL OPERATIONS 3.10.1 Inservice Leak and Hydrostatic Testing Operation LC0 3.10.1 The average reactor coolant temperature specified in Table 1.1-1 for MODE 4 may be changed to "NA," and operation considered not to be in MODE 3; and the requirements of LC0 3.4.8, " Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown," may be suspended, to allow performance of an inservice leak or hydrostatic test provided the follewing MODE 3 LCOs are met:

a. LC0 3.3.6.2, "Secona.:y Containment Isolation Instrumentation," Functiens 1, 3, and 4 of Table 3.3.6.2-1;

b. LC0 3.6.4.1, " Secondary Containment;" l

c. LC0 3.6.4.2, " Secondary Containment Isolation Valves (SCIVs);" and l

d. LC0 3.6.4.3, " Standby Gas Treatment (SGT) System." l APPLICABILITY: MODE 4 with average reactor coolant temperature > 212 F.

O HATCH UNIT 2 3.10-1 REVISIONk(_

Insertfice Leak and Hydrostatic Testing Operation ; 3.10.1 l l ACTIONS

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

Separate Condition entry is allowed for each requirement of the LCO. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more of the A.1 --------NOTE--------- ! above requirements not Required Actions to j met. be in MODE 4 include I reducing average reactor coolant l temperature to s 212 F. Enter the applicable Immediately Condition of the affected LCO. _0_E A.2.1 Suspend activities immediately that could increase the average reactor coolant temperature or pressure. AND A.2.2 Reduce average 24 hours reactor coolant temperature to s 212 F. O HATCH UNIT 2 3.10-2 REVISION A

a O unir 21=enovto 8^sts e* O i l l l O I

SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM) BASES BACKGROUND SDM requirements are specified to ensure:

a. The reactor can be made subcritical from all operating conditions and transients and Design Basis Events;

b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits; and

c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

These requirements are satisfied by the control rods, as described in GDC 26 (Ref. 1), which can compensate for the reactivity effects of the fuel and water temperature changes experienced during all operating conditions. APPLICABLE SHUTDOWN MARGIN is an explicit assumption in several of the SAFETY ANALYSES evaluations contained in FSAR Chapter 15. The control rod drop accident (CRDA) analysis (Refs. 2 and 3) assumes the core is subcritical with the highest worth control rod withdrawn. Typically, the first control rod withdrawn has a very high reactivity worth and, should the core be critical during the withdrawal of the first control rod, the consequences of a CRDA could exceed the fuel damage limits for a CRDA (see Bases for LC0 3.1.6, " Rod Pattern Control"). Also, SDM is assumed as an initial condition for the control ! rod removal error during refueling (Ref. 4) and fuel ; assembly insertion error during refueling (Ref. 5) I accidents. The analysis of these reactivity insertion events assumes the refueling interlocks are OPERABLE when i the reactor is in the refueling mode of operation. These interlocks prevent the withdrawal of more than one control rod from the core during refueling. (Special consideration and requirements for multiple control rod withdrawal during refueling are covered in Special Operations LC0 3.10.6,

                   " Multiple Control Rod Withdrawal - Refueling.") The 1

(continued) l HATCH UNIT 2 B 3.1-1 REVISION A

                                                         . .                        ._A

SDM B 3.1.1 BASES APPLICABLE analysis assumes this condition is acceptable since the core SAFETY ANALYSES will be shut down with the highest worth control rod (continued) withdrawn, if adequate SDM has been demonstrated. Prevention or mitigation of reactivity insertion events is necessary to limit energy deposition in the fuel to prevent significant fuel damage, which could result in undue release of radioactivity. Adequate SDM ensures inadvertent criticalities and potential CRDAs involving high worth control rods (namely the first control rod withdrawn) will not cause significant fuel damage. SDM satisfies Criterion 2 of the NRC Policy Statement (Ref. 9). l LC0 The specified SDM limit accounts for the uncertainty in the demonstration of SDM by testing. Separate SDM limits are provided for testing where the highest worth control rod is determined analytically or by measurement. This is due to the reduced uncertainty in the SDM test when the highest worth control rod is determined by measurement. When SDM is evaluated by calculations not associated with a test (e.g., to confirm SDM during the fuel loading sequence), additional margin is included to account for uncertainties in the calculation. To ensure adequate SDM during the design process, a design margin is included to account for uncertainties in the design calculations (Ref. 6). APPLICABILITY In MODES 1 and 2, SDM must be provided because subcriticality with the highest worth control rod withdrawn is assumed in the CRDA analysis (Ref. 2). In MODES 3 and 4, SDM is required to ensure the reactor will be held subcritical with margin for a single withdrawn control rod. SDM is required in MODE 5 to prevent an open vessel, , inadvertent criticality during the withdrawal of a single l control rod from a core cell containing one or more fuel assemblies (Ref.4) or fuel assembly insertion error (Ref. 5). l i l (continued) O HATCH UNIT 2 B 3.1-2 REVISIONf6

M-SDM B 3.1.1 BASES (continued) ACTIONS A.1 With SDM not within the limits of the LCO in MODE 1 or 2, SDM must be restored within 6 hours. Failure to meet the specified SDM may be caused by a control rod that cannot be inserted. The allowed Completion Time of 6 hours is acceptable, considering that the reactor en still be shut down, assuming no failures of additional control rods to : insert, and the low probability of an event occurring during this interval. fld If the SDM cannot be restored, the plant must be brought to MODE 3 in 12 hours, to prevent the potential for further reductions in available SDM (e.g., additional stuck control rods). The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. O u , With SDM not within limits in MODE 3, the operator must immediately initiate action to fully insert all insertable control rods. Action must continue until all insertable control rods are fully inserted. This action results in the least reactive condition for the core. D.l. D.2. D.3. and D.4 With SDM not within limits in MODE 4, the operator must immediately initiate action to fully insert all insertable control rods. Action must continue until all insertable control rods are fully inserted. This action results in the least reactive condition for the core. Action must also be initiated within I hour to provide means for control of potential radioactive releases. This includes ensuring:

1) secondary containment (at least including the Unit 2 reactor building zone) is OPERABLE; 2) sufficient Standby Ga.s Treatment (SGT) subsystem (s) are OPERABLE to maintain the secondary containment at a negative pressure with respect to the environment (depe,ident on secondary l

(continued) HATCH UNIT 2 8 3.1-3 REVISION M

l l SDM B 3.1.1 BASES (continued) containment configuration, refer to Reference 8; single j failure protection is not required while in this ACTION); and 3) secondary containment isolation capability is available (i.e., at least one secondary containment isolation valve and associated instrumentation 1 l l l l l l O (continued) HATCH UNIT 2 B 3.1-3A REVISIONk(,

1 l l SDM l 8 3.1.1 1 BASES ACTIONS D.l. D.2. D.3. and D.4 (continued) are OPERABLE, or other acceptable administrative controls to assure isolation capability) in each associated secondary l containment penetration flow path not isolated that is assumed to be isolated to mitigate radioactivity releases. This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE. l E.1. E.2. E.3. E.4. and E.5 { With SDM not within limits in MODE 5, the operator must immediately suspend CORE ALTERATIONS that could reduce SDM, ) p s (e.g., insertion of fuel in the core or the withdrawal of j control rods). Suspension of these activities shall not 1 preclude completion of movement of a component to a safe i condition. Inserting control rods will reduce the total j reactivity and therefore, is excluded from the suspended I actions. Removing fuel, while allowable under these I Required Actions, should be evaluated for axial reactivity I effects before removal. Action must also be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies have been fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and therefore do not have to be inserted. Action must also be initiated within 1 hour to provide means for control of potential radioactive releases. This includes ensuring: 1) secondary containment (at least in.cluding the common refueling floor zone) is (continued) HATCH UNIT 2 B 3.1-4 REVISIONh(- 1

SDM B 3.1.1 BASES OPERABLE; 2) sufficient SGT subsystem (s) are OPERABLE to maintain the secondary containment at a negative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 8; single failure protection is not required while in this ACTION); and 3) seccndary containment isolation capability is available (i.e., at least one secondary containment isolation valve and associated instrumentation are OPERABLE, or other acceptable administrative controls to assure isolation capability) in each associated 9 (continued) h HATCH UNIT 2 B 3.1-4A REVIS10N'&/_

SDM B 3.I.I I o BASES l l ACTIONS E.1. E.2. E.3. E.4. and E.5 (continued) secondary containment penetration flow path not isolated l that is assumed to be isolated to mitigate radioactivity releases. This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status. Action must continue until all required components are OPERABLE. SURVEILLANCE SR 3.1.1.1 REQUIREMENTS Adequate SDM must be verified to ensure that the reactor can be made subcritical from any initial operating condition. This can be accomplished via a test, an evaluation, or a combination of the two. Adequate SDM is demonstrated by testing before or during the first startup after fuel movement r. shuffling within the reactor pressure vessel, or control rod replacement. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a coK al rod from another core location. Since core reactivity wili vhg during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle. Therefore, ta obtain the SDM, the initial value must be changed by the value, "R", which is the difference between the calculated value of minimum SDM during the operating cycle and the calculated B0C SDM. If the value of R is positive (that is, B0C is the point in the cycle with the minimum SDM), no correction to the BOC measured value is required (Ref. 7). For the SDM demonstrations where the highest worth rod is determined solely on calculation, additional margin (0.10% Ak/k) must be added to the SDM limit of 0.28% Ak/k to account for uncertainties in the calculation of the highest worth control rod. (continued) HATCH UNIT 2 B 3.1-5 REVISION Q

SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued) REQUIREMENTS The SDM may be demonstrated during an in-sequence control rod withdrawal, in which the highest worth control rod is analytically determined, or during local criticals, where the highest worth control rod is determined by testing. Local critical tests require the withdrawal of out of sequence control rods. This testing would therefore require bypassing of the Rod Worth Minimizer to allow the out of sequence withdrawal, and therefore additional requirements must be met (see LC0 3.10.7, " Control Rod Testing - Operating"). The Frequency of 4 hours after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification. During MODE 5, adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals. An evaluation of each in-vessel fuel movement during fuel loading (including shuffling fuel within t' a core) is required to ensure adequate SDM is maintained during refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern. For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence. These bounding analyses include additional margins to the SDM limit to account for the associated uncertainties. Spiral offload / reload sequences inherently satisfy the SR, provided the fuel assemblies are-reloaded in the same configuration analyzed for the new cycle. Removing fuel from the core will always result in an increase in SDM. REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Section 15.1.38.

3. NEDE-240ll-P-A-US, " General Electric Standard Application for Reactor Fuel," Supplement for United States, (revision specified in the COLR).

(continued) HATCH UNIT 2 B 3.1-6 REVISION A l

1 SDH B 3.1.1 th ( BASES l REFERENCES 4. FSAR, Section 15.1.13. l (continued) 1 FSAR, Section 15.1.14. 5.

6. FSAR, Section 4.3.2.4.1.

7. NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel," (revision specified in the COLR).

8. Technical Requirements Manual. l

9. NRC No. 93-102, " Final Policy Statement on Technical i Specification Improvements," July 23, 1993.

O i O i HATCH UNIT 2 B 3.1-7 REVISION k(. l 1

r l Reactivity Anomalies B 3.1.2 l 1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 Reactivity Anomalies BASES BACKGROUND In accordance with GDC 26, GDC 28, and GDC 29 (Ref. 1), reactivity shall be controllable such that subtriticality is maintained under cold conditions and specified acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Therefore, reactivity anomaly is used as a measure of the predicted versus actual core reactivity during power operation. The continual confirmation of core reactivity is necessary to ensure that the Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity anomaly could be the result of unanticipated changes in fuel reactivity or control rod worth or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus actual core reactivity validates the . nuclear methods used in the safety analysis and supports the l SDM demonstrations (LC0 3.1.1, " SHUTDOWN MARGIN (SDM)") in assuring the reactor can be brought safely to cold, subcritical conditions. When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and actual reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable poison, producing zero net reactivity. In order to achieve the required fuel cycle energy output, the uranium enrichment in the new fuel loading and the fuel i loaded in the previous cycles provide excess positive i reactivity beyond that required to sustain steady state operation at the beginning of cycle (B0C). When the reactor is critical at RTP and operating moderator temperature, the ex' cess positive reactivity is compensated by burnable (continued) HATCH UNIT 2 B 3.1-8 REVISION A

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.7 (continued) REQUIREMENTS measurement, or in overlapping segments, with verification that all components are tested. A Note to the Surveillance states that the radiation detectors may be excluded from ISOLATION SYSTEM RESPONSE TIME testing. This Note is necessary because of the difficulty of generating an appropriate detector input signal and because the principles of detector operation virtually ensure an instantaneous response time. Response times for radiation detector channels shall be measured from detector output or the input of the first electronic component in the channel. ISOLATION SYSTEM RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. This Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience that shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. O REFERENCES 1. FSAR, Section 6.3.

2. FSAR, Chapter 15.

3. FSAR, Section 4.2.3.4.2. ,

4. NEDC-31677P-A, " Technical Specification Improvement I Analysis for BWR Isolation Actuation Instrumentation," i July 1990.

5. NEDC-30851P-A Supplement 2, " Technical Specifications !

Improvement Analysis for BWR Isolation Instrumentation : Common to RPS and ECCS Instrumentation," March 1989. l l

6. Technical Requirements Manual.

7. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

HATCH UNIT 2 B 3.3-173 REVISION A '

e Secondary Containment Isolation Instrumentation B 3.3.6.2 1 8 3.3 INSTRUMENTATION B 3.3.6.2 Secondary Containment Isolation Instrumentation BASES BACKGROUND The secondary containment isolation instrumentation automatically initiates closure of appropriate secondary containment isolation valves (SCIVs) and starts the Standby Gas Treatment (SGT) System. The function of these systems, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Refs. I and 2). Secondary containment isolation and establishment of vacuum with the SGT System within the assumed time limits ensures that fission products that leak from primary containment ' following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be OPERABLE are maintained within applicable limits. The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of secondary containment isolation. Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, j which then outputs a secondary containment isolation signal i to the isolation logic. Functional diversity is provided by l monitoring a wide range of independent parameters. The i input parameters to the isolation logic are (1) reactor I vessel water level, (2) drywell pressure, (3) reactor ! building exhaust high radiation, and (4) refueling floor exhaust high radiation. Redundant sensor input signals from each parameter are provided for initiation of isolation. The outputs of the logic channels in a trip system are arranged into two two-out-of-two trip system logics. Any trip system initiates all SGT subsystems and isolates the automatic isolation valves (dampers) in each secondary containment penetration. Each logic closes at least one of the two valves in each secondary containment penetration and starts the required SGT subsystems, so that operation of either logic (continued) HATCH UNIT 2 8 3.3-174 REVISION A

Secondary Containment Isolation Instrumentation B 3.3.6.2 (m () BASES BACKGROUND isolates the secondary containment and provides l (continued) for the necessary filtration of fission products. APPLICABLE The isolation signals generated by the secondary containment SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the LCO, and safety analyses of References 1 and 2 to initiate closure APPLICABILITY of valves and start the SGT System to limit offsite doses. Refer to LCO 3.6.4.2, " Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3, " Standby Gas Treatment (SGT) System," Applicable Safety Analyses Bases for more detail of the safety analyses. t The secondary containment isolation instrumentation l satisfies Criterion 3 of the NRC Policy Statement (Ref. 7). Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the O individual instrumentation channel Functions. Each Function must have the required number of OPERABLE channels with their setpoints set within the specified Allowable Values, as shown in Table 3.3.6.2-1. The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint). A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Each channel must also respond within its assumed response time, where appropriate. Allowable Values are specified for each Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor (continued) HATCH VNIT 2 B 3.3-175 REVISIONKG

1 l Secondary Containment Isolation Instrumentation l B 3.3.6.2 BASES l vessel water level), and when the measured output value of APPLICABLE SAFETY ANALYSES, the process parameter exceeds the setpoint, the associated LCO, and device (e.g., trip unit) changes state. The analytic limits APPLICABILITY are derived from the limiting values of the process (continued) parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for. In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions when SCIVs and the SGT System are required. The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below cn a Function by Function basis.

1. Reactor Vessel Water Level-Low low. Level 2 Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.

An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The Reactor Vessel Water Level-Low Low, level 2 Function is one of the Functions assumed to be OPERABLE and capable of providing isolation and initiation signals. The isolation and initiation systems on Reactor Vessel Water Level-Low Low, Level 2 support actions to ensure that any offsite releases are within the limits calculated in the safety analysis (Refs. 3 and 4). Reactor Vessel Water Level-Low Low, Level 2 signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of (continued) HATCH UNIT 2 B 3.3-176 REVISION A

1 l l Secondary Containment Isolation Instrumentation l B 3.3.6.2 im (v) BASES APPLICABLE 1. Reactor Vessel Water Level-Low Low. Level 2 SAFETY ANALYSES, (continued) LCO, and APPLICABILITY Reactor Vessel Water Level-Low Low, level 2 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The React Vessel Water Level-Low Low, Level 2 Allowable Value was chosen to be the same as the High Pressure Coolant Injection / Reactor Core Isolation Cooling (HPCI/RCIC) Reactor Vessel Water Level-Low Low, Level 2 Allowable Value (LC0 3.3.5.1 and LC0 3.3.5.2), since this could indicate that the capability to cool the fuel is being threatened. The Reactor Vessel Water Level-Low Low, Level 2 Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus,.there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature p d limitations of these MODES; thus, this Function is not required. In addition, the Function is also required to be OPERABLE during operations with a potential for draining the reactor vessel (0PDRVs) because the capability of isolating potential sources of leakage must be provided to ensure that offsite dose limits are not exceeded if core damage occurs. l

2. Drywell Pressure-Hiah High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The isolation on high drywell pressure supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis.

However, the Drywell Pressure-High Function associated with isolation is not assumed in any FSAR accident or transient analyses. It is retained for the overall redundancy and diversity of the secondary containment isolation (continued) l l HATCH UNIT 2 B 3.3-177 REVISION N (a

Secondary Containment Isolation Instrumentation i B 3.3.6.2 BASES APPLIC?,BLE 2. Drywell Pressure-Hiah (continued) SAFETY ANALYSES, LCO, and instrumentation as required by the NRC approved APPLICABILITY licensing basis. High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High Functions are available l and are required to be OPERABLE to ensure that no single l instrument failure can preclude performance of the isolation function. l The Allowable Value was chosen to be the same as the ECCS l Drywell Pressure-High Function Allowable Value j (LC0 3.3.5.1) since this is indicative of a loss of coolant I accident (LOCA). The Drywell Pressure-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy I exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES. l

3. 4. Reactor Buildina and Refuelina Floor Exhaust Radiation-Hiqh High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding.

The release may have originated from the primary containment l due to a break in the RCPB or the refueling floor due to a ' fuel handling accident. When Exhaust Radiation-High is detected, secondary containment isolation and actuation of the SGT System are initiated to limit the release of fission products as assumed in the FSAR safety analyses (Ref. 4). The Exhaust Radiation-High signals are initiated from radiation detectors that are located near the ventilation exhaust ductwork coming from the reactor building and the refueling floor zones, respectively. The signal from each detector is input to an individual monitor whose trip (continued) HATCH UNIT 2 B 3.3-178 REVISIONkC, L _

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 3. 4. Reactor Buildina and Refuelino Floor Exhaust SAFETY ANALYSES, Radiation-Hiah (continued) LCO, and APPLICABILITY outputs are assigned to an isolation channel. Four channels of Reactor Building Exhaust Radiation-High Function and four channels of Refueling Floor Exhaust Radiation-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Values are chosen to ensure radioactive releases do not exceed offsite dose limits. The Reactor Building and Refueling Floor Exhaust Radiation-High Functions are required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, these Functions are not required. The Reactor Building Exhaust Radiation-High O Function is also required to be OPERABLE during OPDRVs (in MODE 4 and MODE 5) because the capability of detecting l radiation releases due to fuel failures (due to fuel uncovery) must be provided to ensure that offsite dose limits are not exceeded. The Refueling Floor Exhaust Radiation-High Function is also required to be OPERABLE during CORE ALTERATIONS, MODE 5 OPDRVs, and movement of-irradiated fuel assemblies in the secondary containment because the capability detecting radiation releases due to fuel failures (e.g., due to a dropped fuel assembly) must be l provided to ensure that offsite dose limits are not exceeded. l ACTIONS A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, (continued) HATCH UNIT 2 B 3.3-179 REVISIONh,(.,

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS will not result in separate entry into the Condition. (continued) Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel. A.1 Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Function 2, and 24 hours for Functions other than Function 2, has been shown to be acceptable (Refs. 5 and 6) to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required ; Action A.1. Placing the inoperable channel in trip would i conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where ) placing the inoper'ble channel in trip would result in an ' undesired isolation), Condition C must be entered and its l , Required Actions taken. j l B.1 Required Action B.1 is intended to ensure that ap)ropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic isolation capability for the associated secondary containment penetration flow path (s) or e complete 1.oss of automatic initiation capability for the Unit I and Unit 2 l (continued) h HATCH UNIT 2 B 3.3-180 REVISION M

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS _ Dd (continued) SGT Systems. A Function is considered to be maintaining l secondary containment isolation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate a trip signal from the given Function on a valid signal. This ensures that one of the two SCIVs in each penetration flow path, and the required Unit'I and Unit 2 SGT subsystems can be initiated on an isolation - signal from the given Function. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. C.I.1. C.I.2. C.2.1. and C.2.2 If any Required Action and associated Completion Time of Condition A or 8 are not met, the ability to isolate the secondary containment and start the required Unit 1 and O- Unit 2 SGT Systems cannot be ensured. Therefore, further l actions must be performed to ensure the ability to maintain the secondary containment function. Isolating the associated flow paths (closing the ventilation supply and l exhaust automatic isolation dampers) and starting the associated SGT subsystem (s) (Required Actions C.1.1 and C.2.1) performs the intended function of the instrumentation and allows operation to continue. Alternately, declaring the associated SCIVs or SGT subsystem (s) inoperable (Required Actions C.1.2 and C.2.2) is also acceptable since the Required Actions of the respective LCOs (LC0 3.6.4.2 and LCO 3.6.4.3) provide l appropriate actions for the inoperable components. Since-each trip system affects multiple SGT subsystems, Required l Actions C.2.1 and C.2.2 can be performed independently on each SGT subsystem. That is, one SGT subsystem can be started (Required Action C.2.1) while another SGT subsystem l can be declared inoperable ~(Required Action C.2.2). (continued) HATCH UNIT 2 B 3.3-181 REVISION h C

Secondary Containment isolation Instrumentation B 3.3.6.2 BASES ACTIONS C.1.1. C.l.2. C.2.1. and C.2.2 (continued) One hour is sufficient for personnel to establish required plant conditions or to declare the associated components inoperable without unnecessarily challenging plant systems. SURVEILLANCt' As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required 0.ctions taken. This Note is based on the reliability analycis (Refs. 5 and 6) assumption of the l average time required to perform channel surveillance. That I analysis demonstrated the 6 hour testing allowance does not ! significantly reduce the probability that the SCIVs will isolate the associated penetration flow paths and that the i SGT System will initiate when necessary. l SR 3.3.6.2.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. (continued) HATCH UNIT 2 B 3.3-182 REVISION 'IQ

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR '3.3.6.2.1 (continued) REQUIREMENTS Agreement criteria are determined by the plant staff based , on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the ; instrument has drifted outside its limit. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCO. SR 3.3.6.2.2 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability analysis of References 5 and 6. SR 3.3.6.2.3 and SR 3.3.6.2.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology. The Frequencies of SR 3.3.6.2.3 and SR 3.3.6.2.4 are based on the assumption of the magnitude of equipment drift in the setpoint analysis. (continued) HATCH UNIT 2 8 3.3-183 REVISION A

Secondary Containment Isolation Instrumentation I B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2.5 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel. The system functional testing performed on SCIVs and the SGT System in LC0 3.6.4.2 and LC0 3.6.4.3, l respectively, overlaps this Surveillance to provide complete testing of the assumed safety function. While this Surveillance can be performed with the reactor at power for some of the Functions, operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was found to be acceptable from a reliability standpoint. REFERENCES 1. FSAR, Section 6.3.

2. FSAR, Chapter 15.

3. FSAR, Section 15.1.40.

4. FSAR, Sections 15.1.39 and 15.1.41.

5. NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR It$lation Actuation Instrumentation,"

July 1990.

6. NEDC-30851P-A Supplement 2, " Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.

7. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

I HATCH UNIT 2 B 3.3-184 REVTSION/6

ECCS - Shutdown i B 3.5.2 h L' B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.2 ECCS - Shutdown BASES BACKGROUND A description of the Core Spray (CS) System and the low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System is provided in the Bases for LC0 3.5.1, "ECCS - Operating." APPLICABLE The ECCS performance is evaluated for the entire spectrum of SAFETY ANALYSES break sizes for a postulated loss of coolant accident (LOCA). The long term cooling analysis following a design basis LOCA (Ref. 1) demonstrates that only one low pressure ECCS injection / spray subsystem is required, post LOCA, to maintain adequate reactor vessel water level in the event of an inadvertent vessel draindown. It is reasonable to assume, based on engineering judgment, that while in MODES 4 and 5, one low pressure ECCS injection / spray subsystem can maintain adequate reactor vessel water level. To provide

 \                     redundancy, a minimum of two low pressure ECCS injection /

spray subsystems are required to be OPERABLE in MODES 4 and 5. The low pressure ECCS subsystems satisfy Criterion 3 of the NRC Policy Statement (Ref. 3). l LCO Two low pressure ECCS injection / spray subsystems are required to be OPERABLE. The low pressure ECCS injection / spray subsystems consist of two CS subsystems and two LPCI subsystems. Each CS subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool or condensate storage tank (CST) to the reactor pressure vessel (RPV). Each LPCI subsystem consists of one motor driven pump, piping, and valves to transfer water from the suppression pool to the RPV. Only a single LPCI pump is required per subsystem because of the larger injection capacity in relation to a CS subsystem. In M0. DES 4 and 5, the RHR System cross tie valve is not required to be closed. The necessary portions of the Plant (continued) l HATCH UNIT 2 B 3.5-17 REVISIONk(_,

ECCS - Shutdown B 3.5.2 BASES LC0 Service Water System are also required to provide (continued) appropriate cooling to each required ECCS subsystem. One LPCI subsystem may be aligned for decay heat removal and l considered OPERABLE for the ECCS function, if it can be manually realigned (remote or local) to the LPCI mode and is not otherwise inoperable. Because of low pressure and low ' temperature conditions in MODES 4 and 5, sufficient time will be available to manually al!gn and initiate LPCI ) subsystem operation to provide core cooling prior to i postulated fuel uncovery. j APPLICABILITY OPERABILITY of the low pressure ECCS injection / spray subsystems is required in MODES 4 and 5 to ensure adequate coolant inventory and sufficient heat removal capability for the irradiated fuel in the core in case of an inadvertent draindown of the vessel. Requirements for ECCS OPERABILITY during MODES 1, 2, and 3 are discussed in the Applicability section of the Bases for LC0 3.5.1. ECCS subsystems are not required to be OPERABLE durino MODE 5 with the spent fuel storage pool gates removed at.. the water level maintained at , a 22 ft 1/8 inches above the RPV flange (equivalent to 21 ft i of water above the top of irradiated fuel assemblies seated ! in the spent fuel storage pool racks). This provides sufficient coolant inventory to allow operator action to terminate the inventory loss prior to fuel uncovery in case of an inadvertent draindown. The Automatic Depressurization System is not required to be OPERABLE during MODES 4 and 5 because the RPV pressure is s 150 psig, and the CS System and the LPCI subsystems can provide core cooling without any depressurization of the primary system. The High Pressure Coolant Injection System is not required to be OPERABLE during MODES 4 and 5 since the low pressure ECCS injection / spray subsystems can provide sufficient flow to the vessel. (continued) HATCH UNIT 2 B 3.5-18 REVISION A

ECCS - Shutdown B 3.5.2 BASES (continued) ACTIONS A.1 and 8.1 If any one required low pressure ECCS injection / spray subsystem is inoperable, the inoperable subsystem must.be restored to OPERABLE status in 4 hours. In this condition, the remaining OPERABLE subsystem can provide sufficient vessel flooding capability to recover from an inadvertent vessel draindown. However, overall system reliability is reduced because a single failure in the remaining OPERABLE subsystem concurrent with a vessel draindown could result in - the ECCS not being able tc, perform its intended function. The 4 hour Completion Time for restoring the required low pressure ECCS injection / spray subsystem to OPERABLE. status is based on engineering judgment that considered the remaining available subsystem and the low probability of a vessel draindown event. With the inoperable subsystem not restored to OPERABLE status in the required Completion Time, action must be immediately initiated to suspend operations with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown and the subsequent potential for fission product release. O continue until OPDRVs are suspended. Actions must C l. C.2. D.I. D.2. D.3. E.1. E.2, and E.3 ; With both of the required ECCS injection / spray subsystems inoperable, all coolant inventory makeup capability may be unavailable. Therefore, actions must immediately be initiated to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until 0PDRVs are suspended. One ECCS injection / spray subsystem must also be restored to OPERABLE status within 4 hours. The 4 hour Completion Time to restore at least one low pressure ECCS injection / spray subsystem to OPERABLE status ensures that prompt action will be taken to provide the required cooling capacity or to initiate actions to place the plant in a condition that minimizes any potential fission product release to the environment. (continued) HATCH UNIT 2 B 3.5-19 REVISION A

 ~ _ _ _            -__           _

ECCS - Shutdown B 3.5.2 BASES ACTIONS C.l. C.2. D.l. D.2. 0.3. E.1. E.2. and E.3 (continued) If at lea.st one low pressure ECCS injection / spray subsystem is not restored to OPERABLE status within the 4 hour Completion Time, additional actions are required to minimize any potential fission product release to the environment. This includes ensuring: I) secondary containment (at least including: the Unit 2 reactor building zone if in MODE 4; or the common refueling floor zone if in MODE 5) is OPERABLE; 2) sufficient standby gas treatment (SGT) subsystem (s) are OPERABLE to maintain the secondary containment at a negative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 2; single failure protection is not required while in this ACTION); and 3) secondary containment isolation capability is available (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability) in each associated secondary containment penetration flow path l not isolated that is assumed to be isolated to mitigate radioactivity releases. OPERABILITY may be verified by an administrative check, or by examining logs or other information, to determine whether the components are out of lg service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, the Surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE. l l (continued) HATCH UNIT 2 B 3.5-20 REVISION)ff

ECCS - Shutdown i B 3.5.2 i BASES SURVEILLANCE SR 3.5.2.1 and SR 3.5.2.2 (continued) REQUIREMENTS suppression pool water level less than the required limit,

all ECCS injection / spray subsystems are inoperable unless they are aligned to an OPERABLE CST. l When suppression pool level i:: < 146 inches, the CS System is considered OPERABLE only if it can take suction from the CST, and the CST water level is sufficient to provide the required NPSH for the CS pump. Therefore, a verification that either the suppression pool water level is 2 146 inches or that CS is aligned to take suction from the CST and the CST contains 2 150,000 gallons of water, equivalent to 12 ft, ensures that the CS System can supply at least 50,000 gallons of makeup water to the RPV. The CS suction is uncovered at the 100,000 gallon level. However, as noted, only one required CS subsystem may take credit for the CST option during OPDRVs. During OPDRVs, the volume in the CST may not provide adequate makeup if the RPV were completely drained. Therefore, only one CS subsystem is allowed to use the CST. This ensures the other required ECCS subsystem has adequate makeup volume, s The 12 hour Frequency of these SRs was developed considering operating experience related to suppression pool water level and CST water level variations and instrument drift 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 suppression pool or CST water level condition. SR 3.5.2.3. SR 3.5.2.5. and SR 3.5.2.6 The Bases provided for SR 3.5.1.1, SR 3.5.1.7, and SR 3.5.1.10 are applicable to SR 3.5.2.3, SR 3.5.2.5, and SR 3.5.2.6, respectively. However, the LPCI flow rate requirement for SR 3.5.2.5 is based on a single pump, not the two pump flow rate requirement of SR :. 5.1.7. SR 3.5.2.4 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are (continued) HATCH UNIT 2 8 3.5-21 REVISION A

ECCS - Shutdown B 3.5.2 BASES h SURVEILLANCE SR 3.5.2.4 (continued) REQUIREMENTS locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is appropriate because the valves are operated under procedural control and the probability of their being mispositioned during this time period is low. In MODES 4 and 5, the RHR System may operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor. Therefore, RHR valves that are required for LPCI subsystem operation may be aligned for decay heat removal. Therefore, this SR is modified by a Note that allows ou LPCI subsystem of the RHR System to be considered OPE.RABLE for the ECCS function if all the required valves in the LPCI flow path can be manually realigned (remote or socal) to allow injection into the RPV, and the system'is not otherwise inoperable. This will ensure adequate core cooling if an inadvertent RPV draindown should occur. REFERENCES 1. NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Analysis," December 1986.

2. Technical Requirements Manual. l

3. NRC No. 93-102, " Final Policy Statement on Technical g Specification Improvements," July 23, 1993.

O HATCH UNIT 2 B 3.5-22 REVISIONkG

Secondary Containment l B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment l BASES BACKGROUND The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA).- In conjunction with operation of the Standby Gas Treatment (SGT) System and closure of certain valves whose lines

       -                    penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place _inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment.

The secondary containment is a structure that completely encloses the primary containment and those components that may be postulated to contain primary system fluid. This ' structure forms a control volume that serves to hold up and dilute the fission products. It is possible for the

pressure in the control volume to rise relative to the l environmental pressure (e.g., due to pump and motor heat load additions). The secondary containment encompasses three separate zones: the Unit 1 reactor building (Zone I),

the Unit 2 reactor building (Zone II), and the common refueling floor (Zone III). The secondary containment can be modified to exclude the Unit I reactor building (Zone I) provided the following requirements are met:

a. Unit 1 Technical Specifications do not require !

OPERABILITY of Zone I; j i

b. All hatches separating Zone III from Zone I are closed and sealed; and l i i

c. At least one door in each access path separating Zone t III from Zone I is closed. l (continued)

HATCH UNIT 2 8 3.6-81 REVISION,IG

1 Secondary Containment l l B 3.6.4.1 i BASES Similarly, other zones can be excluded from the secondary containment OPERABILITY requirement during various plant j operating conditions with the appropriate controls. For I example, during Unit 2 shutdown operations, the secondary l containment can be modified to exclude the Unit 9. reactor building (Zone II) (either alone or in combination with excluding Zone I as described above) provided the following requirements are met:

a. Unit 2 is not conducting operations with a potential for draining the reactor vessel (0PORV);

b. All hatches separating Zone III from Zone II are closed and sealed; and

c. At least one door in each access path separating Zone III from Zone II is closed.

9 (continued) i HATCH UNIT 2 B 3.6-82 REVISIONM 1

Secondary Containment l B 3.6.4.1 BASES BACKGROUND To prevent ground level exfiltratior, while ailowing the (continued) secondary containment to be designed as a. conventional structure, the secondary containment requires support systems to maintain the control volume pressure at less than the external pressure. Requirements for these systems are specified separately in LC0 3.6.4.2, " Secondary Containment Isolation Valves (SCIVs)," and LC0 3.6.4.3, " Standby Gas Treatment (SGT) System." When one or more zones are excluded from secondary containment, the specific requirements for the support systems will also change (e.g., securing particular SGT or drain isolation valves). APPLICABLE There are two principal accidents for which credit is taken SAFETY ANALYSES #or secondary containment OPERABILITY. These are a loss of coolant accident (LPCA) (Ref.1) and a fuel handling accident inside secondary containment (Ref. 2). The secondary containment performs no active function-in response to either of these. limiting events; however, its l leak tightness is required to ensure that the release.of radioactive materials from the primary containment is O restricted to those leakage paths and associated leakage rates assumed in the accident analysis and that fission ' products entrapped within the secondary containment structure will be treated by the Unit 1 and Unit 2 SGT Systems prior to discharge to the environment. Postulated l LOCA leakage paths from the primary containment into , secondary containment include those into both the reactor ' building and refueling floor zones (e.g., drywell head leakage). .l l Secondary containment satisfies Criterion 3 of the NRC Policy Statement (Ref. 4). l I i LCO An OPERABLE secondary containment provides a control volume l into which fission products that bypass or leak from primary ! containment, or are released from the reactor coolant ' pressure boundary components located in the secondary l containment, can be diluted and processed prior to release to the environment. For the secondary containment to be ca.nsidered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and O (continued) HATCH UNIT 2 B 3.6-83 REVISION

Secondary Containment l B 3.6.4.1 BASES maintained. The secondary containment boundary required to be OPERABLE is dependent on the operating status of both units, as well as the configuration of doors, ratches, refueling floor plugs, SCIVs, and available flow paths to SGT Systems. The required boundary encompasses the zones which can be postulated to contain fission products from accidents required to be considered for the condition of each unit, and furthermore, must include zones not isolated from the SGT subsystems being credited for meeting LC0 3.6.4.3. Allowed configurations, associated SGT subsystem requirements, and associated SCIV requirements are detailed in the Technical Requirements Manual (Ref. 3). O (continued [ HATCH UNIT 2 B 3.6-84 REVISIONf

~                                                                                          .

Secondary Containment l l B 3.6.4.1 l BASES (continued) APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment (the reactor building zone and potentially the refueling floor zone). Therefore, secondary containment OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY. In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining secondary containment OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during 0PDRVs, during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment. (Note, moving irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3.) Since CORE ALTERATIONS and movement of irradiated fuel assemblies are only postulated to release radioactive material to the refueling floor zone, the secondary containment configuration may consist of only Zone III during these conditions. Similarly, during OPDRVs while in MODE 4 (vessel head O bolted) the release of radioactive materials is only postulated to the associated reactor building, the secondary containment configuration may consist of only Zone II. ACTIONS A.1 If secondary containment is inoperable, it must be restored l to OPERABLE status within 4 hours. The 4 hour Completion Time provides a period of time to correct the problem that is commensurate with the importance of maintaining secondary containment during MODES 1, 2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment OPERABILITY) occurring during periods l where secondary containment is inoperable is minimal. B.1 and B.2 If, secondary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must l l be brought to a MODE in which the LC0 does not apply. To (continuedJ HATCH UNIT 2 B 3.6-85 REVISION /G

Secondary Containment l B 3.6.4.1 BASES (continued) achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 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. C.I. C.2. and C.3

   ~                Movement of irradiated fuel assemblies in the secondary containment, CORE ALTERATIONS, and OPDRVs can be postulated to cause fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the 5

environment. CORE ALTERATIONS and movement of irradiated fuel assemblies must be immediately suspended if the secondary containment is inoperable. Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPORVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until 0PDRVs are suspended. Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LC0 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel asseinblies would not be a sufficient reason to require a reactor shutdown. (continued) HATCH UNIT 2 B 3.6-86 REVISIONg/j

Secondary Containment l . B 3.6.4.1 BASES (continued) -SURVEILLANCE SR 3.6.4.1.1 and SR 3.6.4.1.2 ' REQUIREMENTS Verifying that secondary containment equipment hatches and l access doors are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the secondary containments will not occur. SR 3.6.4.1.1 also requires equipment ' hatches to be sealed. In this application, the term

                   " sealed"-has no connotation of leak tightness. Maintaining secondary containment OPERABILITY requires verifying each         l door in the access opening is closed, except when the access opening is being used for normal transient entry and exit (then at least one door must remain closed). When the secondary containment configuration excludes Zone I and/or Zone II, these SRs also include verifying the hatches and doors separating the common refueling floor zone from the reactor building (s). The 31 day Frequency for these SRs has      l been shown to be adequate, based on operating experience, and is considered adequate in view of the other indications of door and hatch status that are available to the operator.

SR 3.6.4.1.3 and SR 3.6.4.1.4 The Unit 1 and Unit 2 SGT Systems exhausts the secondary e containment atmosphere i.c the environment through appropriate treatment equipment. To ensure that all fission products are treated, SR 3.6.4.1.3 verifies that the appropriate SGT System (s) will rapidly establish and maintain a pressure in the secondary containments that is less than the lowest postulated pressure external to the secondary containment boundary. This is confirmed by demonstrating that the required SGT subsystem (s) will draw down the secondary containment to a 0.25 inch of vacuum water gauge in s 120 seconds. This cannot be accomplished if the secondary containment boundary is not intact. SR 3.6.4.1.4 demonstrates that the required SGT subsystem (s) l , p (continued) HATCH UNIT 2 8 3.6-87 REVISIONJ( I

Secondary Containment l B 3.6.4.1 BASES (continued) h can maintain 2: 0.25 inch of vacuum water gauge for 1 hour at a flow rate s 4000 cfm for each SGT subsystem. The I O (continued) h HATCH UNIT 2 B 3.6-87A REVISION [y 4

4 N

                                                        -Secondary Containment    l B 3.6.4.1 g  BASES SURVEILLANCE SR   3.6.4.1.3 and SR   3.6.4.1.4 (continued)                      l            l REQUIREMENTS I hour test period' allows secondary containment to be in thermal equilibrium at steady state conditions. Therefore,                .

these two tests are used to ensure the secondary containment l i boundary integrity. Since these SRs are secondary 7 containment tests, they need not be performed with each SGT i subsystem. The SGT subsystems are tested on a STAGGERED TEST BASIS, however, to ensure thtt in addition to the j requirements of LC0 3.6.4.3, each SGT subsystem or / combination of subsystems will perform this test. The i > number of SGT subsystems and the required combinations are ' dependent on the configuration of the secondary containment and are detailed in the Technical Requirements Manual (Ref. I 3). The Note to SR 3.6.4.1.3 and to SR 3.6.4.1.4 specifies that the number of required SGT subsystems be one less than , the number required to meet LC0 3.6.4.3, " Standby Gas Treatment (SGT) System," for the given configuration. Operating experience has shown these components usually pass. the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. O REFERENCES 1. FSAR, Section 15.1.39.

2. FSAR, Section 15.1.41.

3. Technical Requirements Manual.

4. NRC No. 93-102, " Final Policy Statement on Technical l Specification Improvements," July 23, 1993.

i 1 O HATCH UNIT 2 B 3.6-88 REVIS10N g Q i

                                                                     -              -           l

SCIVs B 3.6.4.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) l BASES , BACKGROUND The function of the SCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs). Secondary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that fission products that leak from primary containment following a DBA, or that are released during certain operations when primary containment is not required to be OPERABLE or take place outside primary containment, are maintained within the secondary containment boundary. The OPERABILITY requirements for SCIVs help ensure that an adequate secondary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. These isolation devices consist of either passive devices or active (automatic) devices. Manual valves, de-activated automatic valves secured in their closed position, check valves with flow through the valve secured, and blind flanges are considered passive devices. Automatic SCIVs close on a secondary containment isolation signal to establish a boundary for untreated radioactive material within secondary containment following a DBA or other accidents. Other penetrations are isolated by the use of valves in the closed position or blind flanges. APPLICABLE The SCIVs must be OPERABLE to ensure the secondary SAFETY ANALYSES containment barrier to fission product releases is established. The principal accidents for which the secondary containment boundary is required are a loss of coo' ant accident (Ref. 1) and a fuel handling accident inside secondary containment (Ref. 2). The secondary containment performs no active function in response to ei.ther of these limiting events, but the boundary established by SCIVs is required to ensure that leakage from the primary containment is processed by the Standby Gas (continued) l HATCH UNIT 2 B 3.6-89 REVISION [/j i

SCIVs B 3.6.4.2 l l O \ Q BASES APPLICABLE Treatment (SGT) System before being released to the SAFETY ANALYSES environment. (continued) Maintaining SCIVs OPERABLE with isolation times within l limits ensures that fission products will remain trapped inside the secondary containment so that they can be treated by the SGT System prior to discharge to the environment. SCIVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 4). l LC0 SCIVs form a part of the secondary containment boundary. l The SCIV safety function is related to control of offsite radiation releases resulting from DBAs. The power operated isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves

            .      covered by this LCO, along with their associated' stroke times, are listed in Reference 3.

l The normally closed isolation valves or blind flanges are considered OPERABLE when manual valves are closed, or open in accordance with appropriate administrative controls, automatic SCIVs are de-activated and secured in their closed position, and blind flanges are in place. These passive isolation valves or devices are listed in Reference 3. l The SCIVs required to be OPERABLE are dependent on the l configuration of the secondary containment (which is dependent on the operating status of both units, as well as the configuration of doors, hatches, refueling floor plugs, and available flow paths to SGT Systems). The required boundary encompasses the zones which can be postulated to contain fission products from accidents required to be considered for the condition of each unit, and furthermore, must include zones not isolated from the SGT subsystems being credited for meeting LC0 3.6.4.3, " Standby Gas Treatment (SGT) System." The required SCIVs are those in l penetrations communicating with the zones required for secondary containment OPERABILITY and are detailed in Reference 3. (continued) 1 1 HATCH UNIT 2 B 3.6-90 REVISION /

SCIVs B 3.6.4.2 i BASES h APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore, the OPERABILITY of SCIVs l is required. O 1 l I l i (continued) HATCH UNIT 2 B 3.6-91 REVISION \(.

F SCIVs ! B 3.6.4.2

   \ BASES V

APPLICABILITY In MODES 4 and 5, the probability and consequences of a LOCA (continued) are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant radioactive releases can be postulated, such as during operations with a potential for draining the reactor vessel (0PDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment. (Note, moving irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3.) ACTIONS The ACTIONS are modified by three Notes. The first Note allows penetration flow paths to be unisolated intermittently under administrative controls. These i controls consist of stationing a dedicated operator, who is in continuous communication with the control room, at the ! controls of the isolation device. In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated.

  %                The second Note provides clarification that for the purpose of this LC0 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 SCIV. Complying with the Required Actions may allow for continued operation, and subsequent iro perable SCIVs are governed by subsequent Condition entry and application of associated Required Actions.

The third Note ensures appropriate remedial actions are taken, if necessary, if the affected system (s) are rendered ! inoperable by an inoperable SCIV. A.1 and A.2 In the event that there are one or more penetration flow paths with one SCIV inoperable, the affected penetration l 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. Is'olation barriers that meet this criterion are a closed and de-activated automatic SCIV, a closed manual valve, and a 1 (continued) HATCH UNIT 2 B 3.6-92 REVISION k(, l

i SCIVs B 3.6.4.2 l BASES l blind flange. For penetrations isolated in accordance with ! Required Action A.1, the device used to isolate the penetration should be the closest available device to secondary containment. The Required Action must be completed within the 8 hour Completion Time. The specified time period is reasonable considering the time required to I I O (continued) HATCH UNIT 2 8 3.6-93 REVISION 4(

I SCIVs B 3.6.4.2 BASES ACTIONS A.1 and A.2 (continued) isolate the penetration, and the probability of a DBA, which requires the SCIVs to close, occurring during this short time is very low. For affected penetrations that have been isolated in accordance with Required Action A.1, the affected penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that secondary containment penetrations required to be isolated following an accident, but no longer capable of being automatically isolated, will be in the isolation position should an event occur. The Completion Time of once per 31 days is appropriate because the isolation devices are operated under administrative controls and the probability of their misalignment is low. This Required Action does not require any testing or device manipulation. Rather, it involves verification that the affected penetration remains isolated. Required Action A.2 is modified by a Note that applies to devices located in high radiation areas and allows them to be verified closed by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment once they have been verified to be in the proper position, is low. Ed With two SCIVs in one or more penetration flow paths inoperable, the affected penetration flow path must be isolated within 4 hours. 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 4 hour Completion Time is reasonable ; considering the time required to isolate the penetration and the probability of a DBA, which requires the SCIVs to close, occurring during this short time, is very low. l (continued) 1 HATCH UNIT 2 B 3.6-94 REVISION %(, 1 i

SCIVs B 3.6.4.2 BASES l ACTIONS C.1 and C.2 I (continued) If any Required Action and associated Completion Time of Condition A or B cannot be met, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this l l status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 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 j challenging plant systems. D.I. D.2. and 0.3 l l If any Required Action and associated ' Completion Time of Condition A or B are not met, the plant must be placed in a ! condition in which the LC0 does not apply. If applicable, ' CORE ALTERATIONS and the movement of irradiated fuel assemblies in the secondary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, ',f applicable, actions must be immediately initiated to susr,end OPDRVs in order to minimize the probability of t. vessel draindown and the subsequent potential for fission product release. Actions must continue until 0PDRVs are suspended. Required Action D.1 has been modified by a Note stating that LC0 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LC0 3.0.3 would not specify any action. If moving fuel while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. SURVEILLANCE SR ' 6 4.2.1 l REQUIREMENTS This SR verifies that each secondary containment manual l isolation valve and blind flange that is required to be closed during accident conditions is closed. The SR helps to ensure that oost accident leakage of radioactive fluids or, gases outside of the secondary containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification (continued) HATCH UNIT 2 8 3.6-95 REVISION

SCIVs B 3.6.4.2 A V BASES that those isolation devices in secondary containment that are capable of being mispositioned are in the correct position. Since these isolation devices are readily accessible to personnel during normal operation and verification of their position is relatively easy, the 31 day Frequency was chosen to provide added assurance that the isolation devices are in the correct positions. Two Notes have been added to this SR. The first Note applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative controls. Allowing verification by administrative controls 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 isolation devices, once they have been verified to be in the proper position, is low. A second Note has been included to clarify that SCIVs that q are open under administrative controls are not required to meet the SR during the time the SCIVs are open. Q l l 1 (continued) HATCH UNIT 2 B 3.6-96 REVISIONgG i

SCIVs ) B 3.6.4.2 1 BASES h)l l SURVEILLANCE SR 3.6.4.2 2 l ! REQUIREMENTS (continued) Verifying that the isolation time of each power operated and each automatic SCIV is within limits is required to l demonstrate OPERABILITY. The isolation time test ensures that the SCIV will isolate in a time period less than or equal to that assumed in the safety analyses. The Frequency of this SR was developed based upon engineering judgment and the similarity to PCIVs. SR 3.6.4.2.3 l Verifying that each automatic SCIV closes on a secondary l containment isolation signal is required to prevent leakage of radioactive material from secondary containment following l a DBA or other accidents. This SR ensures that each automatic SCIV will actuate to the isolation position on a secondary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this SR to provide complete testing of the safety function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. FSAR, Section 15.1.39.

2. FSAR, Section 15.1.41.

3. Technical Requirements Manual.

4. NRC No. 93-102, " Final Policy Statement on Technical l Specification Improvements," July 23, 1993.

1 O HATCH UNIT 2 B 3.6-97 REVISION [b7 1 I

SGT System B 3.6.4.3 .O B 3.6 CONTAINMENT SYSTEMS B 3.6.4.3 Standby Gas Treatment (SGT) System l BASES BACKGROUND The SGT System is required by 10 CFR 50, Appendix A, GDC 41,

                      " Containment Atmosphere Cleanup" (Ref. 1). The function of the SGT System is to ensure that radioactive materials that leak from the primary containment into the secondary containment following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment.

The Unit I and Unit 2 SGT Systems each consists of two fully redundant subsystems, each with its own set of dampers, charcoal filter train, and controls. The Unit 1 SGT subsystems' ductwork is separate from the inlet to the filter train to the discharge of the fan. The rest of the ductwork is common. The Unit 2 SGT subsystems' ductwork is separate except for the suction from the drywell and torus, which is common (however, this suction path is not required l for subsystem OPERABILITY). Each charcoal filter train consists of (components listed in order of the direction of the air flow):

a. A demister or moisture separator;

b. An electric heater;

c. A prefilter;

d. A high efficiency particulate air (HEPA) filter;

e. Two charcoal adsorbers for Unit I subsystems and one charcoal adsorber for Unit 2 subsystems;

f. A second HEPA filter; and

g. An axial vane fan for Unit I subsystems and a centrifugal fan for Unit 2 subsystems.

The sizing of the SGT Systems equipment and components is based on the results of an infiltration analysis, as well as an' exfiltration analysis of the secondary containment. The l internal pressure of the SGT Systems p (continued) d HATCH UNIT 2 B 3.6-98 REVISIONf4

SGT System B 3.6.4.3 BASES BACKGROUND boundary region is maintained at a negative pressure of (continued) 0.25 inches water gauge when the system is in operation, which represents the internal pressure required to ensure zero exfiltration of air from the building when exposed to a 10 mph wind. The demister is provided to remove entrained water in the air, while the electric heater reduces the relative humidity of the airstream to < 70% (Refs. 2 and 3). The prefilter removes large particulate matter, while the HEPA filter removes fine particulate matter and protects the charcoal from fouling. The charcoal adsorbers remove gaseous elemental iodine and organic iodides, and the final HEPA filter collects any carbon fines exhausted from the charcoal adsorber. The Unit 1 and Unit 2 SGT Systems automatically start and operate in response to actuation signals indicative of conditions or an accident that could require operation of the system. Following initiation, all required charcoal filter train fans start. Upon verification that the required subsystems are operating, the redundant required subsystem is normally shut down. APPLICABLE The design basis for the Unit 1 and Unit 2 SGT Systems SAFETY ANALYSES is to mitigate the consequences of a loss of coolant accident and fuel handling accidents (Refs. 2, 3, 4, and 5). For all events analyzed, the SGT Systems are shown to be , automatically initiated to reduce, via filtration and ' adsorption, the radioactive material released to the environment. The SGT System satisfies Criterion 3 of the NRC Policy l Statement (Ref. 7). l 1 l l

i (continued) HATCH UNIT 2 B 3.6-99 REVISION M

SGT System B 3.6.4.3 p) t, BASES LCO Following a DBA, a minimum number of subsystems are required to maintain the secondary containment at a negative pressure with respect to the environment and to process gaseous releases. Meeting the LCO requirements for OPERABLE subsystems ensures operation of the minimum number of SGT subsystems in the event of a single active failure. The required number of SGT subsystems is dependent on the configuration required to meet LC0 3.6.4.1, " Secondary Containment." For secondary containment OPERABILITY consisting of all three zones, the required number of SGT subsystems is four. With secondary containment OPERABILITY consisting of one reactor building and the common refueling floor zones, the required number of SGT subsystem is three. Allowed configurations and associated SGT subsystem requirements are detailed in the Technical Requirements Manual (Ref. 6). O) m (continued) HATCH UNIT 2 B 3.6-99A REVISION M

1 SGT System B 3.6.4.3 BASES 1 LCO In addition, with secondary containment in modified (continued) configurations, the SGT System valves to excluded zone (s) are not included as part of SGT System OPERABILITY (i.e., the valves may be secured closed and are not required to open on an actuation signal). APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to secondary containment. Therefore, Unit I and Unit 2 SGT Systems OPERABILITY are required during these MODES. In MODES 4 and 5, the probability and consequences of a LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT Systems in OPERABLE status is not required in MODE 4 or 5, except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (0PDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment. O l ACTIONS The Actions are modified by a Note to indicate that when both Unit 1 SGT subsystems are placed in an inoperable status for inspection of the Unit 1 hardened vent rupture disk, entry into associated Conditions and Required Actions may be delayed for up to 24 hours, provided both Unit 2 SGT subsystems are OPERABLE. Upon completion of the inspection or expiration of the 24 hour allowance, the Unit 1 SGT subsystems must be returned to OPERABLE status or the applicable Conditions entered and Required Actions taken. The 24 hour allowance is based upon precluding a dual unit shutdown to perform the inspection, yet minimizing the time both Unit 1 SGT subsystems are inoperable. 1 A.1 and B.1 l With one required Unit 1 or Unit 2 SGT subsystem inoperable, th.e inoperable subsystem must be restored to OPERABLE status. In this condition, the remaining required OPERABLE SGT subsystems are adequate to perform the required l l (continued) HATCH UNIT 2 B 3.6-100 REVISION A (.,

s SGT System B 3.6.4.3 l BASES \_/ radioactivity release control function. However, the overall system reliability is reduced because a single failure in one of the remaining required OPERABLE subsystems l could result in the radioactivity release control function not being adequately performed. The 7 and 30 day Completion Times are based on consideration of such factors as the availability of the i w 1 i ( i (continued) HATCH UNIT 2 B 3.6-101 REVISION 1 l

SGT System B 3.6.4.3 BASES ACTIONS A.1 and B.1 (continued) l OPERABLE redundant SGT subsystems and the low probability of a DBA occurring during this period. Additionally, the 30 day Completion Time of Required Action A.1 is based on three remaining 0PERABLE SGT subsystems, of which two are Unit 2 subsystems, and the secondary containment volume in the Unit I reactor building being open to the common refueling floor where the two Unit 2 SGT subsystems can readily provide rapid drawdown of vacuum. Testing and analysis has shown l that in this configuration, even with an additional single failure (which is not necessary to assume while in ACTIONS) the secondary containment volume may be drawn to a vacuum in the time required to support assumptions of analyses. C.1 and C.2 l If the SGT subsystem cannot be restored to CPERABLE status within the required Completion Time in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, g'j based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In the event that a Unit 1 SGT subsystem is the one not restored to OPERABLE status as required by Required Action A.1 or B.1, and:

1. All three zones are required for secondary containment OPERABILITY; and

2. Unit 1 is shutdown with its Technical Specifications not requiring secondary containment OPERABILITY (i.e.,

not handling irradiated fuel, performing CORE ALTERATIONS, or conducting OPDRV), operation of Unit 2 can continue provided that the Unit I reactor building zone is isolated from the remainder of secondary containment and the St;T System. In this modified se.condary containment configuration, only three SGT subsystems are required to be OPERABLE to meet LC0 3.6.4.3, and no limitation is applied to the inoperable Unit 1 SGT (continued) HATCH UNIT 2 B 3.6-102 REVISION /Cf

SGT System B 3.6.4.3 BASES ACTIONS C.1 and C.2 (continued) subsystem. This in effect is an alternative to restoring the inoperable Unit 1 SGT subsystem, i.e., shut down Unit I and isolate its reactor building zone from secondary containment and SGT System. D.I. D.2.1. D.2.2. and 0.2.3 During movement of irradiated fuel assemblies in the secondary containment, during CORE ALTERATIONS, or during OPDRVs, when Required Action A.1 or B.1 cannot be completed within the required Completion Time, the remaining required OPERABLE SGT subsystems should immediately be placed in operation. This action ensures .that the remaining subsystems are OPERABLE, that no failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected. An alternative to Required Action D.1 is to immediately suspend activities that represent a potential for releasing radioactive material to the secondary containment, thus

 'O V

placing the plant in a condition that minimizes risk. If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies must immediately be suspended. Suspension of these activities must not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until DPDRVs are suspended. The Required Actions of Condition D have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a i reactor shutdown. 1 l l

 '                                                                                     I (continued) i HATCH UNIT 2                    B 3.6-103                         REVISIONfg

SGT System B 3.6.4.3 BASES I ACTIONS E.1 (continued) I If two or more required SGT subsystems are inoperable in MODE 1, 2 or 3, the Unit 1 and Unit 2 SGT ( stems may not be capable of supporting the required radioactivity release control function. Therefore, LC0 3.0.3 must be entered immediately. F.1. F.2. and F.3 When two or more required SGT subsystems are inoperable, if I applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in secondary containment must immediately be suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to m pend OPDRVs in order to minimize the probabi'.ity of a vessel draindown and subsequent potential for fission product release. Actions must continue until 0PDRVs are suspended. Required Action F.1 has been modified by a Note stating that LC0 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LC0 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.6.4.3.1 l REQUIREMENTS Operating each required Unit 1 and Unit 2 SGT subsystem for 2 10 continuous hours ensures that they 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. ' Operation with the heaters on for 2 10 continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in co.nsideration of the known reliability of fan motors and controls and the redundancy available in the system. (continued) HATCH UNIT 2 B 3.6-104 REVISION)(h

SGT System B 3.6.4.3 ( BASES SURVEILLANCE SR 3.6.4.3.2 ffQUIREMENTS (continued) This SR verifies that the required Unit 1 and Unit 2 SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies O (continued) HATCH UNIT 2 B 3.6-105 REVISIONf(

F^ SGT System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.2 (continued) l REQUIREMENTS and additional information are discussed in detail in the VFTP. SR 3.6.4.3.3 l This SR verifies that each required Unit 1 and Unit 2 SGT l subsystem starts on receipt of an actual or simulated initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps th's SR to provide complete testing of the safety function. While this Surveillance can be performed with the reactor at power, operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was found to be acceptable from a reliability standpoint. REFERENCES 1. 10 CFR 50, Appendix A, GDC 41.

2. Unit 1 FSAR, Section 5.3.

3. FSAR, Section 6.2.3.

4. FSAR, Section 15.1.39.

1

5. FSAR, Section 15.1.41

6. Technical Requirements Manual

7. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

l l l 1 O HATCH UNIT 2 8 3.6-106 REVISIONK(,

MCREC System-B 3.7.4 > BASES (continued) APPLICABLE The ability of the.MCREC System to maintain the SAFETY ANALYSES habitability of the control room is an explicit assumption for the safety analyses presented in the FSAR, Chapters 6 and'15 (Refs. 3 and 4, respectively).. The. pressurization mode of the MCREC System is assumed to operate following a loss of coolant accident, fuel-handling accident, main steam line break, and control rod drop accident, as discussed in the FSAR, Section 6.4.1.2.2 (Ref. 5). The radiological doses to control room personnel as a result of the various DBAs are summarized in Reference 6. No single active or passive failure will cause the loss of outside or recirculated air- from the control room. The MCREC System satisfies Criterion 3 of the NRC Policy Statement (Ref. 7). LC0 Two redundant subsystems of the MCREC System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem. Total system failure could result in exceeding a dose of 5 rem to the control room operators in the event of a DBA. The MCREC System is considered OPERABLE when the individual components necessary to control operator exposure are OPERABLE in both subsystems. A subsystem is considered , OPERABLE when its associated:

a. Fan is OPERABLE;

b. HEPA filter and charcoal adsorbers are not excessively restricting flow and are capable of performing their ,

filtration functions; and

c. Associated ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

In addition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, , ductwork, and access doors, such that the pressurization limit of SR 3.7.4.4 can be met. However, it is acceptable for access doors to be opened for normal control room entry 1 and exit and not consider it to be a failure to meet the l LCO. . l O (continued) i HATCH UNIT 2 8 3.7-19 REVISION A

MCREC System B 3.7.4 BASES (continued) APPLICABILITY In MODES 1, 2, and 3, the MCREC System must be OPERABLE to control operator exposure during and following a DBA, since the DBA could lead to a fission product release. In MODES 4 and 5, the probability and consequences of a DBA are reduced because of the pressure and temperature l limitations in these MODES. Therefore, maintaining the ' MCREC System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

a. During movement of irradiated fuel assemblies in the secondary containment. Moving irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3;

b. During CORE ALTERATIONS; and

c. During operations with potential for draining the reactor vessel (0PDRVs).

ACTIONS A.1 With one MCREC subsystem inoperable, the inoperable MCREC subsystem must be restored to OPERABLE status within 7 days. With the unit in this condition, the remaining OPERABLE MCREC subsystem is adequate to perform control room radiation protection. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced MCREC System capability. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and that the remaining subsystem can provide the required capabilities. Ibl and 8.2 In MODE 1, 2, or 3, if the inoperable MCREC subsystem cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be pl. aced in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the (continued) < HATCH UNIT 2 B 3.7-20 REVISIONk

g MCREC System B 3.7.4 O g BASES ACTIONS B.1 and B.2 (continued) required unit conditions from full power conditions in an orderly manner and without challenging unit systems. C.1. C.2.1. C.2.2. and C.2.3 r The Required Actions of Condition C are modified by a Note indicating that LC0 3.0.3 does not apply. If moving irradiatd fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown. During movement' of irradiated fuel assemblies in the l secondary containment, during CORE ALTERATIONS, or during OPDRVs, if the inoperable MCREC subsystem cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE MCREC subsystem may be placed in the pressurization mode. This action ensures that the remaining subsystem is , 4 OPERABLE, that no failures that would prevent automatic actuation have occurred, and that any active failure will be readily detected. An alternative to Required Action C.1 is to immediately

suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.

If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in the secondary containment must be j s suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission ' product release. Actions must continue until the OPDRVs are ; suspended, i [ (continued) HATCH UNIT 2 B 3.7-21 REVISIONk6

MCREC System B 3.7.4 BASES ACTIONS Rd (continued) If both MCREC subsystems are inoperable in MODE 1, 2, or 3, l the MCREC System may not be capable of performing the l intended function and the unit is in a condition outside the accident analyses. Therefore, LC0 3.0.3 must be entered immediately. l l E.1. E.2. and E.3 1 The Required Actions of Condition E are modified by a Note l indicating that LC0 3.0.3 does not apply. If moving l irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor , shutdown. I During movement of irradiated fuel assemblies in the l secondary containment, during CORE ALTERATIONS, or during ! OPDRVs, with two MCREC subsystems inoperable, action must be l l taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in the secondary containment must be l l suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. If applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of l a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended. SURVEILLANCE SR 3.7.4.1 i REQUIREMENTS l This SR verifies that a subsystem in a standby mode starts j on demand and continues to operate. Standby systems should be. checked periodically to ensure that they start and i function properly. As the environmental and normal l operating conditions of this system are not severe, testing 1 (continued) I HATCH UNIT 2 B 3.7-22 REVISION \(, l l l

Control Room AC System B 3.7.5 BASES 1 APPLICABILITY In MODES 4 and 5, the probability and consequences of a (continued) Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Control Room AC System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioact ) ..' releases can be postulated:

a. During movement of irradiated 'el assemblies in the secondary containment. Moving irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3;

b. During CORE ALTERATIONS; and

c. During operations with a potential for draining the reactor vessel (0PDRVs).

ACTIONS A.1 and A.2 A With one control room AC subsystem inoperable, the outside V air temperature must be verified to be s 65 F within I hour and the maximum outside air temperature in the previous 24 hours must be verified to be s 65 F within 1 hour. With temperature s 65 F, analysis has shown that only one control room AC . subsystem is needed to meet the design basis 105 F equipment limit. Thus, since there are still two OPERABLE control room AC subsystems, (and the single failure criterion is maintained), operation may continue provided outside air temperature remains s 65 F. The 1 hour Completion Times allow a reasonable period of time to verify the outside air temperature. The 12 hour periodic Completion Time ensures the operators are aware of the outside air temperature and any changes to the outside air temperature, which could change the number of control room AC subsystems needed to meet the design basis equipment limiting temperature. B.1 With one control room AC subsystem inoperable and the outside air temperature not within the limitations of Re' quired Actions A.1 and A.2, the inoperable control room AC subsystem must be restored to OPERABLE status within 30 days. With the unit in this condition, the remaining (continued) HATCH UNIT 2 B 3.7-27 REVISION \(, l l

Control Room AC System B 3.7.5 BASES ACTIONS fL1 (continued) OPERABLE control room AC subsystems are adequate to perform the control room air conditioning function. However, the overall reliability is reduced because a single failure in an OPERABLE subsystem could re: ult in loss of the control room air conditioning function. The 30 day Completion Time is based on the low probability of an event occurring requiring control room isolation, the consideration that the remaining subsystems can provide the required protection. C.l. C.2. and C.3 With two control room AC subsystems inoperable, the outside air temperature must be verified to be s 65 F within 1 hour and the maximum outside air temperature in the previous 24 hours must be verified to be s 65 F within 1 hour. With temperature s 65 F, analysis has shown that only one control room AC subsystem is needed to meet the design basis. Thus, with the unit in this condition, the remaining OPERABLE . control room AC subsystem is adequate to perform the control I room air conditioning function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in loss of the control room air conditioning function. Therefore, one inoperable control room AC subsystem must be restored to OPERABLE I status within 30 days. The 30 day Completion time is based 1 on the low probability of an event occurring requiring ' control room isolation, the consideration that the remaining subsystem can provide the requimd protection. D.1 and 0.2 In MODE 1, 2, or 3, with any Required Action and associated Completion Time of Condition B or C not met, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed in at least MODE 3 within ) 12 hours and in MODE 4 within 36 hours. The allowed l Completion Times are reasonable, based on operating I experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. (continued) HATCH UNIT 2 B 3.7-28 REVISION A

Control Room AC System B 3.7.5 l I (j BASES l ACTIONS E.1. E.2.1. E.2.2 and E.2.3 l (continued) The Required Actions of Condition E are modified by a Note indicating that LC0 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown. During movement of irradiated fuel assemblies in the l secondary containment, during CORE ALTERATIONS, or during OPDRVs, if any Required Action and associated Completion Time for Condition B or C is not met, the necessary OPERABLE control room AC subsystems may be placed immediately in operation. One operable control room AC subsystem is necessary if the outside air temperature is s 65 F and the maximum outside air temperature in the previous 24 hours has been s 65 F. If both of these conditions are not met, then two OPERABLE control room AC subsystems are necessary. This action ensures that the remaining subsystems are OPERABLE, that no failures that would prevent actuation will occur, and that any active failure will be readily detected. An alternative to Required Action E.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in the secondary containment must be l suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended. (continued) HATCH UNIT 2 B 3.7-29 REVISIONh

Control Room AC System B 3.7.5 BASES ACTIONS f_d (continued) If three control room AC subsystems are inoperable in MODE 1, 2, or 3, the Control Room AC System may not be capable of performing the intended function. Therefore, LC0 3.0.3 must be entered immediately. G.I. G 2. and G.3 The Required Actions of Condition G are modified by a Note indicating that LC0 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown. During movement of irradiated fuel assemblies in the j secondary containn.ent, during CORE ALTERATIONS, or during OPDRVs, with three control room AC subsystems inoperable, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. If applicable, CORE ALTERATIONS and handling of irradiated fuel in the secondary containment must be suspended l immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, action must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended. (continued) HATCH UNIT 2 B 3.7-30 REVISIONk(,

AC Sources - Operating B 3.8.1 ( BASES BACKGROUND b. 3100 kW - 2000 hours, (continued)

c. 3250 kW - 300 hours, and

d. 3500 kW - 30 minutes.

DG 1B has the following ratings:

a. 2850 kW - 1000 hours, and

b. 3250 kW - 168 hours.

A description of the Unit 1 onsite power sources is provided in the Bases for Unit 1 LC0 3.8.1. APPLICABLE The initial conditions of DBA and transient analyses in the SAFETY ANALYSES FSAR, Chapter 6 (Ref. 4) and Chapter 15 (Ref. 5), assume ESF systems are OPERABLE. The AC electrical power sources are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of nacessary power to ESF systems so that the fuel, Reactor Coolant System (RCS), and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for Section 3.2, Power Distribution Limits; Section 3.5, Emergency Core Cooling System (ECCS) and Reactor Core Isolation Cooling (RCIC) System; and Section 3.6, Containment Systems. The OPERABILITY of the AC electrical power sources is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the unit. This includes maintaining the onsite or offsite AC sources OPERABLE during accident conditions in the event of:

a. An assumed loss of all offsite power sources or all onsite AC power sources; and

b. A postulated worst case single failure.

AC sources satisfy Criterion 3 of the NRC Policy Statement (Ref. 13). l I ' h' (continued) HATCH UNIT 2 8 3.8-3 REVISIONh w

l 2 AC Sources - Operating ! B 3.8.1 ' BASES (continued) LC0 Two qualified circuits between the offsite transmission network and the onsite Unit 2 Class 1E Distribution System and three separate and independent DGs (2A, 2C, and 18) ensure availability of the required power to shut down the reactor and maintain it in a safe shutdown condition after an anticipated operational occurrence (A00) or a postulated DBA. In addition, since some components required by Unit 2 are powered from Unit 1 sources (i.e., Standby Gas Treatment (SGT) System, Main Control Room Environmental Control (MCREC) System, and Control Room Air Conditioning (AC) System), one qualified circuit between the offsite transmission network and the onsite Unit 1 Class lE Distribution System and one Unit 1 DG (1A or IC) must also be OPERABLE. Qualified offsite circuits are those that are described in the FSAR, and are part of the licensing basis for the unit. Each offsite circuit must be capable of maintaining rated frequency and voltage, and accepting required loads during an accident, while connected to the ESF buses. For the purpose of this LCO, each Unit 2 offsite circuit consists of incoming breaker and disconnect to the respective 2C and 2D SATs, the 2C and 2D transformers, and the respective circuit path including feeder breakers to 4.16 kV ESF buses. (However, for design purposes, the offsite circuit excludes the feeder breakers to each 4.16 kV ESF bus). Feeder breakers from each circuit to the 2F ESF bus are required to be OPERABLE; however, only one feeder breaker per bus to the 2E and 2G ESF bu:es is required to be OPERABLE, but they must be from different SATs (e.g., 2E feeder breaker from the 2C SAT and the 2G feeder breaker from the 2D SAT). With 2E and 2G ESF buses both fed from one SAT (normal line up is both buses fed from 2D SAT), both feeder breakers to each of these ESF buses are required to be OPERABLE. The Unit 1 offsite circuit also consists of the incoming breaker and disconnect to the 4.16 kV ESF buses required to be OPERABLE to provide power to the Unit 1 equipment required by LC0 3.6.4.3, LC0 3.7.4, and LC0 3.7.5. l Each DG must be capable of starting, accelerating to rated frequency and voltage, and connecting to its respective ESF bus on detection of bus undervoltage. This sequence must be accomplished within 12 seconds. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and must continue to operate until offsite power can be restored to the ESF buses. These (continued) HATCH UNIT 2 B 3.8-4 REVISION M

AC Sources - Operating B 3.8.1 ( BASES es REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.

2. FSAR, Sections 8.2 and 8.3.

( 3. Regulatory Guide 1.9, March 1971.

4. FSAR, Chapter 6.

5. FSAR, Chapter 15.

6. Regulatory Guide 1.93, December 1974.

7. Generic Letter 84-15.

8. 10 CFR 50, Appendix A, GDC 18.

9. Regulatory Guide 1.108, August 1977.

10. Regulatory Guide 1.137, October 1979.

11. IEEE Standard 387 - 1984.

12. IEEE Standard 308 - 1980.

13. NRC No. 93-102, " Final Policy Statement on Technical l Specification Improvements," July 23, 1993.

O I HATCH VNIT 2 B 3.8 ,41 8 / REVISION A

AC Sources - Shutdown B 3.8.2 ' 1 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources - Shutdown j i BASES l l BACKGROUND A description of the AC sources is provided in the Bases for . LC0 3.8.1, "AC Sources - Operating." l APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 SAFETY ANALYSES and 5 and during movement of irradiated fuel assemblies in the secondary containment ensures that: l

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate AC electrical power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident.

In general, when the unit is shut down the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or loss of all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs) that are analyzed in MODES 1, 2, and 3 have no specific analyses in MODES 4 and 5. Postulated worst case bounding events are deemed not credible in MODES 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and corresponding stresses result in the probabilities of occurrences significantly reduced or eliminated, and minimal consequences. These deviations from DBA analysis assumptions and design requirements during shutdown conditions are allowed by the LC0 for required systems. (continued) HATCH UNIT 2 B 3.8-40 REVISIONk(,

1 AC Sources - Shutdown B 3.8.2 /~T U BASES LC0 It is acceptable during shutdown conditions, for a single (continued) offsite power circuit to supply all 4.16 kV ESF buses on a unit. .No fast transfer capability is required for offsite circuits to be considered OPERABLE. APPLICABIL!TY The AC sources are required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment to provide assurance that: l

a. Systems providing adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel;

b. Systems needed to mitigate a fuel handling accident are available;

c. Systems necessary tc mitigate the effects of events that can lead to core damaje during shutdown are available; and h

(V d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition. AC power requirements for M DES 1, 2, and 3 are covered in LC0 3.8.1. ACTIONS A.1 An offsite circuit is considered inoperable if it is not available to one required ESF 4160 V bus. If two or more ESF 4.16 kV buses are required per LC0 3.8.8, the remaining buses with offsite power available may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, fuel movement, and operations with a potential for drainirg the reactor vessel. By the allowance of the option to declare required features inoperable with no offsite power availab:e, appropriate re.strictions can be implemented in accordance with the affected required feature (s) LCOs' ACTIONS. (continued) HATCH UNIT 2 B 3.8-43 REVISIONh(3

e AC Sources - Shutdown B 3.8.2 BASES ACTIONS A.2.1. A.2.2. A.2.3. A.2.4. B.1. B.2. B.3. and B.4 (continued) With one or more offsite circuits not available to all required 4160 V ESF buses, the option still exists to declare all required features inoperable (per Required Action A.1). Since this option may involve undesired administrative efforts, the allowance for sufficiently conservative actions is made. With one or more required DGs inoperable, the minimum required diversity of AC power sources is not available. It is, therefore, required to suspend CORE ALTERATIONS, movement of irt adiated fuel assemblies in the secondary containment, and activities that l could result in inadvertent draining of the reactor vessel. Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC sources and to continue this action until restoration is accomplished in order to provide the necessary AC power to the plant safety systems. The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The t restoration of the required AC electrical power sources should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power. Pursuant to LCO 3.0.6, the Distribution System ACTIONS would not be entered even if all AC sources to it are inoperable, resulting in de-energization. Therefore, the Required Actions of Condition A have been modified by a Note to indicate that when Condition A is entered with no AC power to any required ESF bus, ACTIONS for LC0 3.8.8 must be immediately entered. This Note allows Condition A to (tantinued) HATCH UNIT 2 B 3.8-44 REVISION k

DC Sources - Operating 8 3.8.4 ('g% BASES REFERENCES 12. IEEE Standard 485 - 1983. (continued)

13. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

( l l i i O , HATCH UNIT 2 3.8-67 REVIS*.0N A

DC Sources - Shutdown B 3.8.5 8 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC Sources - Shutdown BASES BACKGROUND A description of the DC sources is provided in the Bases for LC0 3.8.4, "DC Sources - Operating." APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the FSAR, Chapter 6 (Ref.1) and Chapter 15 (Ref. 2), assume that Engineered Safety Feature systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the diesel generators (DGs), emergency auxiliaries, and control and switching during all MODES of operation. The OPERABILITY of the DC subsystems is consistent with the initial assumptior>s of the accident analyses and the requirements for the supported systems' OPERABILITY. The OPERABILITY of the minimum DC electrical power sources during MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment ensures that: l

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and i

c. Adequate DC electrical power is provided to mitigate events postulated during shutdown, such as an l inadvertent draindown of the vessel or a fuel handling accident.

The DC sources satisfy Criterion 3 of the NRC Policy > Statement (Ref. 3). LCO The necessary Unit 2 DC electrical power subsystems - with:

1) each station service DC subsystem consisting of two 125 V batteries in series, two battery chargers, and the (continued)

HATCH VNIT 2 B 3.8-k REVISION M a

I DC Sources - Shutdown B 3.8.5 BASES LC0 corresponding control equipment and interconnecting cabling; (continued) and 2) each DG DC subsystem consisting of one battery bank, one battery charger, and the corresponding control equipment and interconnecting cabling - are required to be OPERABLE to support required DC distribution subsystems required OPERABLE by LCO 3.8.8, " Distribution Systems - Shutdown." In addition, some components that may be required by Unit 2 require power from Unit I sources (e.g., Standby Gas Treatment (SGT) System). Therefore, the Unit 1 DG DC electrical power subsystems needed to provide DC power to the required Unit 1 components are also required to be OPERABLE. This requirement ensures the availability of sufficient DC electrical power sources to operate the unit in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents and inadvertent reactor vessel draindown). 1 APPLICABILITY The DC electrical power sources required to be OPERABLE in f) v MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment provide assurance ' l that:

a. Required features to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel;

b. Required features needed to mitigate a fuel handling accident are available;

c. Required features necessary to mitigate the effects of ,

events that can lead to core damage during shutdown l are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

I i (continued) HATCH UNIT 2 B3.8-h REVISIONh(, 69

DC Sources - Shutdown B 3.8.5 BASES APPLICABILITY The DC electrical power requiremen 3 for MODES 1, 2, and 3 (continued) are covered in LC0 3.8.4. ACTIONS A.I. A.2.1. A.2.2. A.2.3. and A.2.4 If more than one DC distribution subsystem is required according to LC0 3.8.8, the DC subsystems remaining 0PERABLE with one or more DC power sources inoperable may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, fuel movement, and operations with a potential for draining the reactor vessel. By allowance of the opi. ion to declare required features inoperable with associated DC power sources inoperable, appropriate restrictions are implemented in accordance with the affected system LCOs' ACTIONS. In many instances, this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS, movement of irradiated fuel assemblies in the secondary containment, l and any activities that could result in inadvertent draining of the reactor vessel). Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of l postulated events. It is f'tr ther required to immediately initiate action to restore tr.e required DC electrical power subsystems and to continue this action until restoration is accomplished in order to provide the necessary DC electrical power to the plant safety systems. The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required DC electrical power subsystems should be completed as quickly as possible in order to minimize the time during which the plant safety systems may be without sufficient power. SURVEILLANCE SR 3.8.5.1 REQUIREMENTS SR 3.8.5.1 requires performance of all Surveillances required by SR 3.8.4.1 through SR 3.8.4.8. Therefore, see (continued) h HATCH UNIT 2 B3.8-k REVISION M

~10

Distribution Systems - Operating B 3.8.7 ~ O O O HATCH UNIT 2 B3.8,95'{'C,/ REVISION [

Distribution Systems - Shutdown B 3.8.8 8 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Distribution Systems - Shutdown BASES BACKGROUND A description of the AC and DC electrical power distribution system is provided in the Bases for LCO 3.8.7, " Distribution Systems - Operating." APPLICABLE The initial conditions of Design Basis Accident and SAFETY ANALYSES transient analyses in the FSAR, Chapter 6 (Ref.1) and Chapter 15 (Ref. 2), assume Engineered Safety Feature (ESF) systems are OPERABLE. The AC and DC electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System, and containment design limits are not exceeded. The OPERABILITY of the AC and DC electrical power distribution system is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY. The OPERABILITY of the minimum AC and DC electrical power sources and associated power distribution subsystems during MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment ensures that: l

a. The facility can be maintained in the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent l

draindown of the vessel or a fuel handling accident. The AC and DC electrical power distribution systems satisfy l Criterion 3 of the NRC Policy Statement (Ref. 3). i I (continued) ; HATCH UNIT 2 8 3.8-90 REVISION k(_,

i _ _ _ - . _ _ _____ - - _ __-_ _ _ - _ ---- - - 0

k Distribution Systems - Shutdown B 3.8.8 l BASES '(continued) LCO .Various combinations of subsystems, equipment, and j components are required OPERABLE by other LCOs, depending on i the specific plant condition. Implicit in those requirements is the required OPERABILITY of necessary support required features. This LC0 explicitly requires energization of the portions of the Unit 2 electrical distribution system necessary to support OPERABILITY of Technical Specifications required systems, equipment, and components'- both specifically addressed by their own LCO, j and implicitly required by the definition of OPERABILITY. In addition, some components that may be required by Unit 2 receive power through Unit 1 electrical power distribution subsystems (e.g., Standby Gas Treatment (SGT) System). Therefore, the Unit-1 AC and DC electrical power ^ distribution subsystems needed to support the required equipment must also be OPERABLE. Maintaining these portions of the distribution system energized ensures the availability of sufficient power to ; operate the plant in a safe manner to mitigate the consequences of postulated events during shutdown (e.g.,

fuel handling accidents and inadvertent reactor vessel draindown).

APPLICABILITY The AC and DC electrical power df.stribution subsystems required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment l provide assurance that:

a. Systems to provide adequate coolant inventory makeup
are available for the irradiated fuel in the core in

! case of an inadvertent draindown of the reactor

vessel; i: b. Systems needed to mitigate a fuel handling accident are available; i-; c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

, d .. Instrumentation and control capability is available i 1 for monitoring and maintaining the unit in a cold

shutdown condition or refueling condition.

1 (continued) i i HATCH UNIT 2 B 3.8-91 REVISIONKG

l Distribution Systems - Shutdown B 3.8.8 BASES h APPLICABILITY The AC and DC electrical power distribution subsystem (continued) requirements for MODES 1, 2, and 3 are covered in LC0 3.8.7. ACTIONS A.l. A.2.1. A.2.2. A.2.3. A.2.4. and A.2.5 Although redundant required features may require redundant electrical power distribution subsystems to be OPERABLE, one OPERABLE distribution subsystem may be capable of supporting sufficient required features to allow continuation of CORE ALTEP.ATIONS, fuel movement, and operations with a potential for craining the reactor vessel. By allowing the option to declare required features associated with an inoperable distribution subsystem inoperable, appropriate restrictions are implemented in accordance with the affected distribution subsystem LCO's Required Actions. In many instances this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made, (i.e., to suspend CORE ALTERATIONS, movement of irradiated fuel assemblies in the secondary l containment, and any activities that could result in inadvertent draining of the reactor vessel). Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and DC electrical power distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the plant safety systems. Notwithstanding performance of the above conservative Required Actions, a required residual heat removal-shutdown cooling (RHR-SDC) subsystem may be inoperable. In this , case, Required Actions A.2.1 through A.2.4 do not adequately 1 address the concerns relating to coolant circulation and heat removal. Pursuant to LC0 3.0.6, the RHR-SDC ACTIONS : would not be entered. Therefore, Required Action A.2.5 is j providec to direct declaring RHR-SDC inoperible, which 1 results in taking the appropriate RHR-SDC Af,TIONS. j l (continued) HATCH UNIT 2 B 3.8-92 REVISIONh(.,

RPV Water Level l B 3.9.6 O O BASES l l SURVEILLANCE SR 3.9.6.1 (continued) REQUIREMENTS The Frequency of 24 hours is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls on valve positions, , which make significant unplannad level changes unlikely. REFERENCES 1. Regulatory Guide 1.25, fiarch 23, 1972. I

2. FSAR, Section 15.1.41,

3. NUREG-0800, Section 15.7.4,

4. 10 CFR 100.11.

5. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

l l l 4 l l O HATCH UNIT 2 8 3.9-21 REVISION A l

RHR -High Water Level B 3.9.7 B 3.9 REFUELING OPERATIONS j B 3.9.7 Residual Heat Removal (RHR) -High Water Level BASES BACKGRLED The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34 (Ref. 1). Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal . Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recircuiation loop. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled. In addition to the RHR subsystems, the volume of water above the reactor pressure vessel (RPV) flange provides a heat j sink for decay heat removal. e i APPLICABLE With the unit in MODE 5, the RHR System is not required to SAFETY ANALYSES mitigate any events or accidents evaluated in the safety analyses. The RHR System is required for removing decay I heat to maintain the temperature of the reactor coolant. The RHR System satisfies criterion 4 of the NRC Policy Statement (Ref. 3). l LC0 Only one RHR shutdown cooling subsystem is required to be OPERABLE and in operation in MODE 5 with irradiated fuel in the RPV and the water level 2 22 ft 1/8 inches above the RPV flange (equivalent to 21 feet of water above the top of irradiated fuel assemblies seated within the spent fuel storage pool racks). Only one subsystem is required because the volume of water above the RPV flange provides backup decay heat removal capability. (continued) HATCH UNIT 2 8 3.9-22 \ REVISIONA,G,

I RHR - High Water Level B 3.9.7 BASES LC0 . An OPERABLE RHR shutdown cooling subsystem consists of an (continued) RHR pump, a heat exchanger, an RHRSW pump providing cooling-to the heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path. In MODE 5, the. RHR cross tie valve is not required to be closed; thus, the valve may be opened to allow RHR pumps in one loop to discharge through the opposite recirculation loop to make a complete subsystem. In addition, the RHRSW cross tie valves may be open to allow RHRSW pumps in one loop to provide cooling to a heat exchanger in the opposite loop to make a complete subsystem. Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in.the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. A Note is provided to allow a 2 hour exception to shut down the operating subsystem every 8 hours. O APPLICABILITY One RHR shutdown cooling subsystem must be OPERABLE and in operation in MODE 5, with irradiated fuel in the RPV and the water level 2: 22 ft 1/8 inches above the top of the RPV flange, to provide decay heat removal. RHR shutdown cooling subsystem requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS). RHR Shutdown Cooling subsystem requirements in MODE 5 with irradiated fuel in the RPV and the water level < 22 ft 1/8 inches above the RPV flange are given in LCO 3.9.8, " Residual Heat Removal (RHR) - Low Water Level ." ACTIONS A.1 With no RHR shutdown cooling subsystem OPERABLE, an alternate method of decay heat removal must be established within I hour. In this condition, the volume of water above the RPV flange provides adequate capability to remove decay heat from the reactor core. However, the overall reliability is reduced becaused loss of water level could (continued) HATCH UNIT 2 B 3.9-23 REVISION A

r RHR - High Water Level B 3.9.7 BASES ACTIONS A.1 (continued) result in reduced decay heat removal capability. The 1 hour Completion Time is based on decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of these alternate method (s) must be reconfirmed every 24 hours thereafter. This will ensure continued heat removal capability. Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures. For example, this may include the use of the Fuel Pool Cooling System, the Reactor Water Cleanup System, operating with the regenerative heat exchanger bypassed, or any other subsystem that can remove heat from the coolant. The method used to remove the decay heat should be the most prudent choice based on unit conditions. 8.1. B.2. B.3. and B.4 If no RHR shutdown cooling subsystem is OPERABLE and an alternate method of decay heat removal is not available in accordance with Required Action A.1, actions shall be taken immediately to suspend operations involving an increase in reactor decay heat load by suspending loading of irradiated fuel assemblies into the RPV. Additional actions are required to minimize any potential fission product release to the environment. This includes ensuring: 1) secondary containment (at least including the common refueling floor zone) is OPERABLE; 2) sufficient standby gas treatment subsystem (s) are OPERABLE to maintain the secondary containment at a negative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 2; single failure protection is not required while in this ACTION); and 3) secondary containment isolation capability is available (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable administrative controls to assure isolation capability) in each secondary containment penetration flow l path not isolated that is assumed to be isolated to mitigate 1 radioactive releases. This may be performed as an i administrative check, by examining logs or other information ' to determine whether the components are cet of service for maintenance or other reasons. It is not necessary to (continued) HATCH UNIT 2 8 3.9-24 REVISION %

l RHR - High Water Level B 3.9.7 I ( BASES perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to O i ( (continued) HATCH UNIT 2 B3.9-25'SiA- REVISION D l l 1 1

RHR - High Water Level B 3.9.7 BASES ACTIONS B.l. B.2. B.3. and B.4 (continued) OPERABLE status. In this case, a Surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are l OPERABLE. 1 C.1 and C.2 If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour. The Completion Time is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation. During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem), the reactor coolant temperature must be periodically monitored to ensure proper functionig of the alternate method. The once per hour Completion Time is deemed appropriate. I SURVEILLANCE SR 3.9.7.1 REQUIREMENTS This Surveillance demonstrates that the required RHR ! shutdown cooling subsystem is in operation and circulating ' reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room. i REFERENCES 1. 10 CFR 50, Appendix A, GDC 34.

2. Technical Requirements Manual. l

3. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

U HATCH UNIT 2 8 3.9-2[ [ REVISION'A 6

l I RHR - Low Water Level l B 3.9.8 l 1 8 3.9 REFUELING OPERATIONS l B 3.9.8 Residual Heat Removal (RHR) - Low Water Level BASES BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as required by GDC 34 (Ref. 1). Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suctioq f.om the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation loop. The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled. APPLICABLE With the unit in MODE 5, the RHR System is not required to SAFETY ANALYSES mitigate any events or accidents evaluated in the safety analyses. The RHR System is required for removing decay heat to maintain the temperature of the reactor coolant. The RHR System satisfies Criterion 4 of the NRC Policy Statement (Ref. 3). l LC0 In MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level < 22 ft 1/8 inches above the RPV flange, two RHR shutdown cooling subsystems must be OPERABLE. An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, an RHRSW pump providing cooling to the heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path. The two subsystems have a common suction source and are allowed to have a common heat exchanger and common discharge piping. (continued) HATCH UNIT 2 B 3. 9-27' .J h REVISION %

RHR - Low Water Level B 3.9.8 ( BASES LC0 Since the piping and heat exchangers are passive components (continued) that are assumed not to fail, they are allowed to be common to both subsystems. Thus, to meet the LCO, both pumps in one loop or one pump in each of the two loops must be OPERABLE. In MODE 5, the RHR cross tie valve is not required to be closed; thus, the valve may be opened to allow pumps in one loop to discharge through the opposite recirculation loop to make a complete- subsystem. In addition, the RHRSW cross tie valves may be open to allow RHRSW pumps in one loop to provide cooling to a heat' exchanger in the opposite loop to make a complete subsystem. Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. -However, to ensure adequate core > flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. A Note is provided to allow a 2 hour exception to shut down the operating subsystem every 8 hours. O APPLICABILITY Two RHR shutdown cooling subsystems are required to be OPERABLE, and one must be in operation in MODE 5, with irradiated fuel in the RPV and the water level < 22 ft 1/8 inches above the top of the RPV flange, to provide decay heat removal. RHR shutdown cooling subsystem requirements ! in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS). RHR shutdown cooling subsystem requirements in MODE 5 with irradiated fuel in the RPV and ) the water level 2 22 ft 1/8 inches above the RPV flange are l given in LC0 3.9.7, " Residual Heat Removal (RHR) - High ) Water Level." ' 1 4 ACTIONS L1 l With one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem is capable of providing th.e required decay heat removal. However, the overall reliability is. reduced. Therefore an alternate method of i decay heat removal must be provided. With both required RHR

(continued) HATCH UNIT 2 B3.9-2Gf REVISION A

RHR - Low Water Level B 3.9.8 0^ ACTIONS .A_d (continued) I shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the l LCO. The 1 hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of this i alternate method (s) must be recorifirmed every 24 hours I thereafter. This will ensure continued heat removal capability. Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating i Procedures. For example, this may include the use of the l Reactor Water Cleanup System, operating with the ; regenerative heat exchanger bypassed. The method used to l remove decay heat should be the most prudent choice based on unit conditions. B.l. B.2 and B.3 With the required RHR shutdown cooling subsystem (s) inoperable and the required alternate method (s) of decay heat removal not available in accordance with Required Action A.1, additional actions are required to minimize any potential fission product release to the environment. This includes ensuring: 1) secondary containment (at least including the common refueling floor zone) is OPERABLE; 2) sufficient standby gas treatment subsystem (s) are OPERABLE to maintain the secondary containment at a negative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 2; single failure protection is not required while in this ACTION); and 3) secondary containment isolation capability is available (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acce3 table administrative controls to assure isolation capa3ility) in each associated secondary containment l penetration flow path not isolated that is assumed to be isolated to mitigate radioactive releases. This may be performed as an administrative check, by examining logs or other information to determine whether the components are out of sarvice for maintenance or other reasons. It is not (continued) HATCH UNIT 2 B 3.9-29Y REVISION %(,

i RHR - Low Water Level l B 3.9.8 ( ,, BASES necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored f^s () l v (continued) HATCH UNIT 2 B 3.9-30_]h A REVISION D

RH - Low Water Level B 3.9.8 BASES ACTIONS B.1. B.2 and B.3 (continued) to OPERABLE status. In this case, the Surveillance may need to be performed to restore the. component to OPERABLE status. Actions must continue until all required components are OPERABLE. C.1 and C.2 If no RHR shutdown cooling subsystem is in operation, an alternate method of coolant circulation is required to be < established within 1 hour. The Completion Time is modified such that the I hour is applicable separately for each occurrence involving a loss of coolant circulation. During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour. Completion Time is deemed appropriate. SURVEILLANCE SR 3.9.8.1 REQUIREMENTS This Surveillance demonstrates that one required RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours is sufficient in view of other vi.sual and audible indications available to the operator for monitoring the RHR subsystems in the control room. REFERENCES 1. 10 CFR 50, Appendix A, GDC 34.

2. Technical Requirements Manual. l

3. NRC No. 93-102, " Final Policy Statement on Technical

                     . Specification Improvements," July 23, 1993.

O HATCH UNIT 2 B 3.9-31 / REVISION KG

l' l l O ux1r 2 =^axur or cuantur rtcaxicit SPECIFICATIONS AND DISCUSSION OF CHANGES i i e 4 0 . e 0 0

g DISCUSSION OF CHANGES j ITS: SECTION 3.1.1 - SHUTDOWN MARGIN ADMINISTRATIVE , A.1 In MODES 3 and 4, a single control rod may have been withdrawn under the provisions of the Special Operations LC0 3.10.3 and LC0 3.10.4, or some unanticipated event may have resulted in uninserted control rods. Therefore, rather than the passive "ve ri fy. . . i n s e rted , " the proposed action is active - " Initiate action to insert..." This wording provides the same intent in the event all insertable control rods are inserted. A.2 In MODES 3 and 4, the vessel head is bolted in place, and the only activity that can significantly reduce shutdown margin (SDM) is control rod withdrawal. Since an action which ensures control rods remain inserted is provided, any additional action to suspend activities that can reduce the SDM is repetitive and unnecessary. Similarly, in MODE 5, the only activities that can affect SHUTDOWN MARGIN are CORE ALTERATIONS and control rod withdrawal. Since Actions are provided to suspend CORE ALTERATIONS and ensure control rods remain inserted, any additional action to suspend other activities is also repetitive and unnecessary. A.3 This change replaces the use of the defined term SECONDARY CONTAINMENT INTCGRITY with the essential elements of that definition. The change is editorial in that the requirements are specifically addressed by the (O) proposed Required Actions D.2, D.3, D.4, E.3, E.4, and E.5. Therefore, the change is a presentation preference adopted by the BWR Standard Technical Specifications, NUREG 1433. Refer also to the Discussion of Changes in the Definitions section which addresses deletion of the various containment integrity definitions. In addition, the Bases include a discussion acknowledging the possibility of various configurations for the secondary containmeat boundary. Given the specific configuration, either one, two, or three SGT subsystems may be required to assure the necessary negative pressure when required. As discussed in the Bases, these details are provided in the Technical Requirements Manual (TRM). A.4 The existing action to " demonstrate SECONDARY CONTAINMENT INTEGRITY within ! I hour" appears to provide a period of time (1 hour) in which integrity could be violated even if capable of being maintained. Additionally, if the plant status is such that integrity is not capable of being demonstrated within 1 hour, the existing action results in "non-compliance with the Technical Specifications" and a requirement for an LER. The intent of the action is more appropriately presented in the proposed Required Actions D.2, D.3, D.4, E.3, E.4, and E.5. With the proposed actions, a significantly more conservati,e requirement to establish and maintain the secondary containment boundary is imposed. No longer would l the provision to violate the boundary for up to 1 hour exist. However, this conservatism comes the understanding that if best efforts to O.

demonstrate or establish the boundary exceeded I hour, no LER will be required. HATCH UNIT 2 1 REVISION /((~

 -                                DISCUSSION OF CHANGES                                 ,

', -), ITS: SECTION 3.1.1 - SHUTDOWN MARGIN l A.Q (ccn;r,ad) This interpretation of the action's intent is supported by the BWR l Standard Technical Specifications, NUREL 1433. Because this is an enhanced presentation of existing intent, the proposed change is considered administrative. O O HATCH UNIT 2 / /A REviSiONfc,

O O O bRt hrsc@$ tex cf _5 4.29 (gy (1,c ITs: 3 10.1 # TABLE tt? t (Continued) - >~

                $                                                                                                                      b Ap ruwe le.k r.d II3 drafA

ISOLATION ACTUATION INSTRUMENTATION $ C(* OJ % bkN c 3 -l0 3

a VALVE GROUPS tp.i MINIMUM NUMBER APPLICABLE y OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIG N AL(e) PER TRIP SYSTEM (b)(c) CONDITION ACTION Godbth4 I

2. SECOND ARY CONTAINMENT ISOLATION

                     ] w Reactor Buildiag Exhaust                                                                                                   **

Radiation - Hgh 6,10,12,* 2 1,2,3 and 24 C Q2Di1-K609 A,'s.,C_, D, Drywell Pressure - High 2 1,2,3 24 C L A. 2,6,7,10, h 1,4650 Q,3 12,* , j crf Reactor Vessel Water p(ot.95 \ Level - Low Low (Level 2) ! 5,

2 D= k
1, 2, **

24 ( (pB214482M, C 4 q A' Refueling Floct Exhaust 6 w Rjistion - High {Q 10,12,* 2 1,2,3,5,, and 24 b L Q-KD11 QC, D ,

3. REACTOR WAT ta CLEANUP SYSTEM ISOLATION: a. AFlow - High (2G31-N603 A, B) 5 1 1,2,3 25(1) fl l; b. Ares Temperature - High 5 1 1,2,3 25 (2G31-N662 A, D, E. H. J, M); c. Area Ventilation ATemp. - High 5 1 1,2,3 25 (2G31-N663 A D, E. H. J. M; e \fN g

2G31-N661 A, D, E. H. J, M; 2G31-N662 A, D, E. H. J. M)

)C,

b y'

B

d. SLCS Initiation (NA) 5W NA 1,2,3 25

                                                                                                                                                                                                  +
                $          e. Reactor Vessel Water Level - Low Low        5,

2 1,2,3 25 ['

{ ( (Level 2) (2821-N682 A B C, D)

,, F See Mscussua d %y> w

      .2T-q                                                            Q ITs: 1 u.i, is        '

b , u khew. ?'

              -p I              O

DISCUSSION OF CHANGES

 -(      ITS: SECTION 3.3.6.2 - SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE A.1   The response time for these Functions directly corresponds with the diesel generator start delay time.       Therefore, these response time tests are redundant    to   the diesel    generator    start  time tests in current Specification 3/4.8.1.1 (proposed LC0 3.8.1). NUREG 1366 and Generic Letter 93-05 both recommend deletion of these tests when they are redundant to the diesel generator tests. Therefore, these response time tests have been deleted, and their deletion is considered administrative.

A.2 These proposed changes provide more explicit instructions for proper application of the Actions for Technical Specification compliance. In conjunction with the proposed Specification 1.3, " Completion Times," the ACTIONS Note (" Separate Condition entry is allowed for each....") and the wording for ACTION B ("One or more automatic Functions") provides

direction consistent with the intent of the existing Action for an inoperable isolation instrumentation channel. Since this change only provides more explicit direction of the current interpretation of the existing specifications, this change is considered administrative.

A.3 This action has been deleted since it is redundant to the actions provided in current ACTIONS b and c. The action also provides no guidance as to how long is allowed to place the channel in trip. Thus, since the action O' to place a channel in trip within a certain time is already covered by another, more explicit ACTION, this change is considered administrative. The Specification 3.0.4 exception has been deleted since proposed LC0 A.4 3.0.4 contains this provision (allows continued operation once a channel is placed in the tripped condition). The Specification 3.0.3 exception has been deleted since proposed LCO 3.0.3 states it is only applicable in ! MODES 1, 2, and 3 (and this statement is applicable in MODE 5). A.5 This instrument detects radiation in the Unit 2 Reactor Building. Since fuel handling cannot occur in the Unit 2 Reactor Building, this applicability adds no requirements. Thus, it has been deleted and its deletion is considered administrative. A.6 This change replaces the use of the defined term SECONDARY CONTAINMENT INTEGRITY with the elements of that definition and clarifies the need to start the associated SGT subsystem (s). The change is editorial in that all the individual requirements are specifically addressed by the proposed Required Actions (Required Actions C.1.1 and C.2.1). Therefore the change is a presentation preference adopted by the BWR Standard Technical Specifications, NUREG 1433. Refer also to the Discussion of Changes associated with the Definitions Section which addresses deletion of the various containment integrity definitions. O HATCH UNIT 2 1 REVISION [(,-- l

                                                                                   $reci 9e,b 3 3,21 h        TABLE 3.3.6.7-1 (5'riEET-2-OF 21 MCRECS ACTUATION INSTRUMENTATION ACTION ON 52 - Take the ACTION requireb Specification 3.3 LACTI       x53 - Take t       CTION required by                 ecification 3.3.2.

ACTION 54 - h R? id W M. a Mthoneoftherehuiredradiationmonitorsino!hinthenext6 monitor to OPERABL status within 7 days or, wi erable, restore the hours, p M fuW ' initiate and maintain operation of the MCRECS in the pressurization mode of operation-R<pikh b. Vith no radiation monitors OPERABLE, within 1 hour initiate and e.a s n pdO6 maintain operation of the MCRECS in the pressurization mode of - dl Loperationm

c. (lhesprovision\of Specihotion 3.0'4sare not apMicabl _

NOTES IM jidthh ~\ When handling irradiated fuel in secondary containme t.

a H Daleted)- j Tb.With a design providing only one channel per t' rip system, an inoperable

thannel neea not be pi ed in the tripped conditi where this would ca'uqe the Trip Functio occur. In these cases, noperable channel shall be restored c OPERABLE status within 12 rs or the ACTION s required by Table 3.31.7-1 for that Trip Function a Q be taken. OActuates theMC5 in thET6Qrol room presturization modeQ d r (Deleted.) e,-(Deleted)., HATCH - UNIT 2 3/4 3-58b Amendraent No. M, 88, 96, MG,127 3ctlo

O O O y e TABLE 444,4 --4 S R AD1ATION MONITORING INSTRUMENTATION APPLICABLE y MINIMUM CH ANNELS OPERATIONAL AtAftfutTTR'P , [ ASUR h NT INSTRUMENTATION OPERABLE CONDITIONS SETPOtMT / ACTION { MANG

i. Oh -

Post-Trestrnent Moretore 2 1, 2 to) 10 ' to 10' epe 50 (2011-K615 A

2. Control Roorn Lto 3 3. r.1 3D (45 Inteke Monitors _ 2 1,2.3 -4 1 mr/hr 51 IQmr/hr Ano e qizeals=A s) 4. z 7 0 lic3.J.7.) L.I ~

                                                                                                             ._                                        f(*fo't b O ff rulu]ih: dar;p3 k
                                                                                                                                                         "'" * * "" LkJ * *l m - % .-

m g fcoeeuTaaYs

                                                                                                  ~h OPA/Us A
                                                                                                                                       ,-         t                      5 A br3 W% -1           -

l -

m .

              " Value not to exceed the equivalent of the eteck release lirnit indicated in
                                                                                                                                                   /                                                        .

Specification 8.18(7) e OEk Wrog & Q gy CTS : b4. 3 c. s, ,o ,q C be a. 1 it !. a #F e 4 s- .w N 3%e - am

l 1 i DISCUSSION OF CHANGES l O v ITS: SECTION 3.3.7.1 - MCREC SYSTEM INSTRUMENTATION ADMINISTRATIVE A.1 The only Function being retained is the Control Room Air Inlet Radiation-High Function (see comment R.1 for further discussion on the relocation of other Functions). Therefore, the LC0 statement has been simplified to only require this Function. Since only one Function is maintained, the Table format is not needed and has been deleted. The Applicability of the remaining Function has been written into the proposed Applicability and all references to Tables have been deleted. Since this change is a presentation preference only, it is considered administrative. A.2 This ACTION has been rewritten to require placing the channel in trip within the first 6 hours (proposed Required Action A.'2). Placing the channel in trip results in the system being initiated in the pressurization mode. If it is not desired to place the channel in trip, then ACTION B would require the system to be initiated (Required Action B.1) or the system to be declared inoperable (Required Action B.2). Once the system is declared inoperable, the ACTIONS provided in the system LC0 (proposed LCO 3.7.4) allows a 7 day restoration time, similar to the 7 day restoration time in this ACTION. Therefore, the proposed ACTION times are consistent with the current actions, thus, the change is considered administrative. O V A.3 The proposed Conditions and Required Actions will adequately cover all potential conditions for inoperable equipment in the system and as such, the indication that Specification 3.0.3 is not applicable is unnecessary. The Specification 3.0.4 exception has been deleted since proposed LC0 3.0.4 contains this provision (allows continued operation once the channel is tripped or the system initiated). This is considered to be a change in presentation only and therefore an administrative change. A.4 Comment number not used. l A.5 The Control Room Air Inlet Radiation-High Function is currently specified in two Technical Specifications; Specification 3/4.3.6.1 and 3/4.3.6.7. They have been combined into one proposed Specification, 3.3.7.1. Certain requirement:; are only specified in one of the two current specifications, thus, in the proposed Specification, the more limiting of the requirements is maintained, unless otherwise changed by another comment in this Discussion of Changes. The items affected are as follows: 1) The daily CHANNEL CHECK requirement of Specification 3/4.3.6.1 is maintained; 2) The allowance in current Specification 3/4.3.6.1, ACTION a, to not declare inoperable a channel with an improper trip setpoint for up to 4 hours, has been deleted since current Specification 3/4.3.6.7 does not provide for this allowance; 3) The Applicability of handling irradiated fuel in O HATCH UNIT 2 1 REVISIONfh

DISCUSSION OF CHANGES ITS: SECTION 3.3.7.1 - MCREC SYSTEM INSTRUMENTATION TECHNICAL CHANGE - LESS RESTRICTIVE LA.1 (continued) the required limitation for the parameter and this value is retained. Details relating to system design and operation (e.g., description of action of instrumentation) are also unnecessary in the LC0 and have been { relocated to the Bases and procedur;s. The design features and system operation are also described in the FSAR. Changes to the Bases. will be controlled by the provisions of the proposed Bases Control Process i described in Chapter 5 of the Technical Specifications. Changes to the ! FSAR and procedures will be controlled by the provisions of 10 CFR 50.59. LA.2 The MPL numbers are relocated to plant procedures. The numbers are controlled as part of the equipment location index and on plant drawings. Changes to the MPL numbers will be controlled by the provisions of 10 CFR i 50.59. i LA.3 Details of the methods for performing Surveillances are relocated to the Bases and procedures. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process described in Chapter 5 of the Technical Specifications. Changes to the procedures will be O controlled by the provisions of 10 CFR 50.59. LB.1 The allowed out of service time (A0T) is extended to 2 hours. This A0T has been shown to maintain an acceptable risk in accordance with previously conducted reliability analysis (GENE-770-06-1, February 1991). In addition, an optional allowance to declare the associated MCREC subsystems inoperable (proposed Pequired Action B.2) has also been added, since this is the effect of two inoperable channels.

 " Specific" L.1     This change limits the Applicability of the requirements for the system to during those operations which have potential to create a need for the system to operate.          The omitted conditions are not considered as initiators for events which require the system and therefore the change does not impact safety.           Thus, MODES 4 and 5 are deleted, while the conditions which could result in a potential for a radiation release in MODES 4 and 5, CORE ALTERATIONS, handling of irradiated fuel in the          l Secondary Containment, and operations with a potential for draining the reactor vessel, are maintained.

O HATCH UNIT 2 3 REVISION (__

i DISCUSSION OF CHANGES ITS: SECTION 3.5.2 - ECCS - SHUTDOWN [-)) ADMINISTRATIVE (continued) . A.7 This Surveillance requirement is now part of the Applicability. As such, it periodically verifies that the LC0 is still within that Applicability. Periodic verification that the unit condition remains within the Applicability is not used in the BWR Standard Technical Specifications, NUREG 1433 (and not typically found in current Technical Specifications). These types of surveillances are placed under plant specific control to assure control over Applicability changes are performed correctly and in compliance with LC0 3.0.4 and SR 3.0.4 rules. Therefore, this change is considered administrative. A.8 The suppression chamber water level measurement has been changed from feet and inches (i.e., 12' 2") to inches (i.e., 146"), since this is the parameter units provided in the control room. As such, this change is considered administrative. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 Actions have been added (proposed ACTIONS D and E), which require action to O be taken if an ECCS subsystem is not restored to operable status within 4 V hours. Proposed ACTION D requires action to be immediately initiated to restore to operable status the secondary containment, necessary SGT subsystem (s), and one isolation valve and associated instrumentation in each secondary containment penetration flow path not isolated. Similarly, proposed ACTION E requires action to be immediately initiated to restore to OPERABLE status the secondary containment, necessary SGT subsystem (s), and one isolation valve and associated instrumentation in each secondary containment penetration flow path not isolated. These requirements are additional restrictions on plant operation. In addition, the Bases include a discussion acknowledging the possibility of various configurations for the secondary containment boundary. Given the specific configuration, either one, two, or three SGT subsystems may be required to assure the necessary negative pressure when required. As discussed in the Bases, these details are provided in the Technical Requirements Manual (TRM). M.2 The frequency of this Surveillance has been increased from once per 24 hours to once per 12 hours. These requirements are additional restrictions on plant operation. M.3 A Note is proposed to be added to SR 3.5.2.2 such that when the Condensate Storage Tank is used as a water source for Core Spray, only one Core Spray subsystem can take credit for this alignment. During operations with a potential for draining the reactor vessel, the volume in the CST may not O provide adequate makeup if the RPV were completely drained. This Note helps HATCH UNIT 2 2 REVISION :

l q DISCUSSION OF CHANGES g ITS: SECTION 3.5.2 - ECCS - SHUTDOWN TECHNICAL CHANGE - MORE RESTRICTIVE N.3. (Continued) to ensure that the other ECCS subsystem provides the necessary makeup volume. .Since this provision is not in the current Technical Specifications, it is considered to be more restrictive. 1 l 1 i O l HATCH UNIT 2 2A REVISION G

CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT {t#<iGeEv 34 M.J SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION OJ l-

                                                                                                 /-l                                            1
                                                    .                                                   1                                       j po% %3.6.5.1 b h-Unit 2, rSECONDARY CONTAINMENT INTECRITY and.4,9                                                    (tlatch-Unit 1       l secondary containment integrity shall be maintained.cN7uuc.

APPLICABILITY: CONDITIONS 1, 2, 3, @- See h u"" d W w . . . , ,s , .~. .. ACTION: ' l_[g ta^.-J ug u s k

                                                                                                                 % % opn w, gg I^' " p A 'a.,o,
              .i.thoutJatch-Unit 2MDARY CONTAINMENT INTEGRITY (~and/or withotit-MN           atch-Unit 1 secondarlconfiihjent integrity restore }IatRi!UM t'2 SECONDARY CONTAINMENT INTEGRITY [and Hatch-Un (it T_secondar,y_Iontai.n ntegrity within 4 hours or be in at least HOT SHUTDOWN within the next p      D   12 hours and in COLD SHUTDOWN within the following 24 hours.                                                                      ,

SURVEILLANCE RE0VIREMENTS 4,6.5.1.1 Hatch-Unit 2 SECONDARY CONTAINMENT INTEGRITY shall be demon-strated by:

a. Verifying at least once per 31 days:

p.s,a u 1. All equipment hatches are closed and sealed, and p h s44.4112. At least M h access to the secondary containment is closed,

b. Verifying at h ast once per 92N ys that each seco arycontainR -

nt t ventilatioiNiystem automati olation damper i PERABLE /~ or cured in the Mosed position p Soecification 3. . 2 d

c. At least once per 18 months 6mNxubTwen[h

                        \fMaEDSR ncrc,                                ,     LA.\                                 Q u        1. vet"Tfying that @,, standby gas treatment subsystem will (fW 'a                    draw dcwn the secondary containment to 21/4 inch of vacuum water gauge in g l20 seconds, and                                          j d\

qw.m.: wrg q s 2'.~~DperatTng u,ng; standby gas treatment subsystem for one a hour and maintaining 2 1/4 inch of vacuum water gauge in S M .s.4 the secondary containment at a flow rate not exceeding 4000 CFM. F4 6.5.1.2 Hatch;UrittTTs~chndary containment integrity shall be demon , ' strated per Hatch-Unit 1 Technical Specifications. . w !

                                                                                                                                      /
   \      {*When t

ennlantperforming temoeratureinservice above 212hydrostatic F. or leak testing with thesreacTB&# HATCH - UNIT 2 3/4 6-36 Amendment No. 91 , lo! D l$

REFUELING OPERATIONS 3/4.9.5 SECONDARY CONTAINMENT REFUELING FLOOR m .

                                                                               -pa rRedw s-.L .'4.1     )

LIMITING CONDITION FOR OPERATION - 1 LtO 4 d l _ _ _ . O 1 3.9.5.1 (Hatch-Unit ~ li secondary containment integr4ty shall be maintaineth ]Q'/l. / . v'otnow ; APPLICABILITY: @DITIO}T3)and *. A_CTION: 4 ,.gg,_ _ Without Hatch-Unit I secondary containment integrity,;rettore_ Hatch 2 Sgf (Jn78CsecundarrtantMnment-1Mtogrir, witMa 4 nours29rJsuspend irradiated , fuel and p secondary containment, and, in CONDITION 5, suspend Hatch - Unit 2 CORE

                                                                                            -bl ALTERATIONS agEITUTttes tTiarrould reduththeJHUTDOWN MARG 10 The provisions of Spec _ification 3.0.3 are not . applicable, Q_)b Id SURVEILLANCE RE0VIREMENTS (T

1. Q khN'\ O 4.9.5.1Qlatch-lini~t1{secondarycontainmentshallbedemonstratedper N - ~

M M jj l O/ , i p '# east %nce irradiated fuelWeper_7 spenwuelday snimang#during Hatch casipis being handled- inUnit the 2 CORE A i L

                                                                  ~                                          '
            ' Qatch T UiiiT~l secondary containment.
                  /

L4 2

          -l r/L.1                             /g Ad (c4     69                                                                                             i 1

l f l'M

               *When irradiated _ fuel b he h i fuel sh pping dsD is being handled in the (fa~tch _ Unit 1 secondary containment.

HATCH - UNIT 2 b- /[7' " 3/4 9-7 g h,]-

I I q DISCUSSION OF CHANGES ig ITS: SECTION 3.6.4.1 - SECONDARY CONTAINMENT l ADMINISTRATIVE A.1 The definition of SECONDARY CONTAINMENT INTEGRITY h'as been deleted from the proposed Technical Specifications. It is replaced with the requirement for secondary containment to be OPERABLE. This was done because of the confusion associated with these definitions compared .to its use in the respective LCO. The change is editorial in that all the requirements are specifically addressed in the proposed LC0 for the secondary containment and in the Secondary Containment Isolation Valves and Standby Gas Treatment System Specifications. Therefore the change is a presentation preference adopted by the BWR Standard Technical Specifications, NUREG 1433. A.2 The requirements for secondary containment isolation dampers remain in the Technical Specifications. Providing a cross reference to them adds confusion when evaluating compliance with secondary containment OPERABILITY. Therefore, removal of these references is an administrative difference in presentation. A.3 The secondary containments for the two units at Plant Hatch can include one or more of three separable zones, i.e., the Unit I reactor building zone (Zone I), the Unit 2 reactor building zone (Zone II), and the common O C/ refueling floor zone (Zone III). Current Technical Specification requirements refer to various combinations of these three zones as either:

1) Hatch Unit-1 " normal" secondary containment (Zones I and III);

2) Hatch Unit-1 " modified" secondary containment (Zone III only); and

3) Hatch Unit-1 and Hatch Unit-2 secondary containuent (Here the Hatch Unit-1 secondary containment may be either " normal" or " modified";

i.e., Zones I, II, and III, or Zones II and III). The proposed revision to eliminate the specific terminology of " Hatch Unit-1" and " Hatch Unit-2" secondary containment will not alter these zones or the requirements regarding maintaining an appropriate secondary containment boundary when necessary. { Note however, that an additional requirement for secondary containment to be OPERABLE during operations with a potential for draining the reactor vessel (0PDRV) introduces a new secondary containment boundary requirement: in MODE 4, OPDRV will require only Zone II (Unit 1 has a similar change resulting in a requirement for only Zone I). Refer to more restrictive discussion M.4.} The proposed presentation will simply refer to the requirement for secondary containment to be OPERABLE. " Secondary containment" being defined in the Bases as those zone (s) that can be postulated to contain fission products from accidents required to be considered for the condition of each unit, and furthermore, must include zones not isolated from the SGT subsystems being credited for meeting LC0 3.6.4.3. Allowed configurations and associated SGT subsystem requirenents (refer to discussion LA.1) are fm proposed to be detailed in the Technical Requirements Manual. HATCH UNIT 2 1 REVISIOND'q

q l l DISCUSSION OF CHANGES (v .) ITS: SECTION 3.6.4.1 - SECONDARY CONTAINMENT l A.3 (Continued) This change does not alter current secondary containment boundary requirements, but only the associated terminology. The proposed approach provides a simplified, more user friendly, presentation of a single LC0 for secondary containment, rather than a separate LC0 for each operating configuration and combination of dual-unit applicability. Therefore, this change is considered administrative. A.4 This phrase has been deleted since it is redundant to suspending CORE ALTERATIONS. The only real activities that affect SHUTDOWN MARGIN are fuel movement in the RPV and control rod movement, both of which are considered CORE ALTERATIONS. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 The current Surveillance requires only one door to be closed. The proposed Surveillance requires both doors to be closed, except during entry and exit, and then only one door is required to be closed. This is an additional restriction on plant operation. O V M.2 The current Surveillance requires that the minimum required subsystem (s) (see also L.1) be tested every eighteen months. However, the same SGT subsystem (s) could be tested every eighteen months. The proposed Specification will now require a different combination of subsystems be tested every eighteen months - as represented by the Staggered Test Basis requirement of the Frequency. This is an additional restriction on plant operation. M.3 Since the Unit 1 Secondary Containment is now part of " Secondary Containment (refer to A.3), its Surveillances have been specifically written into this LCO. The secondary containment Surveillance will test the entire boundary as a combined test, rather than the current test of Unit-2 secondary containment separately from Unit-1 secondary containment. A change was made to proposed SR 3.6.4.1.3 adding the 120 second draw down time, which is not in the current Unit 1 Surveillance. Also, proposed SRs 3.6.4.1.1 and 3.6.4.1.2 apply to the required portions of all the secondary containment. These SRs are not required by the current Unit 1 Technical Specifications. These changes and additions, therefore, are considered additional restrictions on plant operation. M.4 An Applicability is being added requiring the Secondary Containment to be OPERABLE during OPDRVS. Appropriate ACTIONS and Surveillance Requirements are also added. This is consistent with the requirements of the BWR Standard Technical Specifications, NUREG 1433. O v HATCH UNIT 2 2 REVISION l

O V DISCUSSION OF CHANGES ITS: SECTION 3.6.4.1 - SECONDARY CONTAINMENT j TECHNICAL CHANGE - MORE RESTRICTIVE (continued) M.5 This out-of-service time has been deleted. With secondary containment inoperable, the applicable conditions must immediately be suspended, This is an additional restriction on plant operation. TECHNICAL CHANGE - LESS RESTRICTIVE

 " Generic" LA.1 SR 3.6.4.1.3 and SR 3.6.4.1.4' test the leak-tight ability of the secondary containment boundary. Since the boundary typically consists of more than just the current Hatch-Unit 2 Secondary Containment, the test will be required on the entire boundary (refer to discussion M.3 for this change).

The additional zones comprising the secondary containment will have additional SGT subsystems required to drawdown the required vacuum (see markup and discussion of changes for ITS 3.6.4.3). The number of SGT subsystems will vary depending on the specific zones included in the secondary containment boundary for the given operating status of both plants, and configuration of hatches and doors. Therefore, the required s number of SGT subsystems necessary to perform the leak-tight test of the secondary containment are provided in the Note to these SRs, and allow that number to vary depending on the current configuration of the secondary containment. Since the number is configuration dependent, it is proposed to be detailed in the Technical Requirements Manual, consistent with the required number of SGT subsystems (refer to L.4 in ITS 3.6.4.3 discussions). LA.2 If the spent fuel shipping cask has an irradiated fuel assembly in it, then the LC0 will be required because of the " movement of irradiated fuel assemblies" Applicability. Thus, for this case, the spent fuel shipping cask Applicability is redundant and has been deleted. For the case when the spent fuel shipping cask is empty, the requirement has been relocated to procedures. The licensing basis analysis is a dropped irradiated fuel assembly on other irradiated fuel assemblies, not a dropped piece of equipment (e.g., spent fuel shipping cask). The current Plant Hatch heavy loads analysis covers all loads considered heavy, which includes components much heavier than a spent fuel shipping cask. These loads will be handled under plant control as allowed by the NRC Policy Statement on Technical Specifications. (Refer to Plant Hatch Unit 2 Application of Selection Criteria, and current Specification 3/4:9.8 Discussion of Change). Therefore, handling of a spent fuel shipping cask will be handled in the same manner as other heavy loads which are not spent fuel assemblies. O HATCH UNIT 2 3 REVISION

1 l O - DISCUSSION OF CHANGES ITS: SECTION 3.6.4.1 - SECONDARY CONTAINMENT l TECHNICAL CHANGE - LESS RESTRICTIVE (Continued)

         " Specific" L1    The Applicability has been modified to require the secondary conta.inment during CORE ALTERATIONS, not all the time while in MODE 5. (The movement of irradiated fuel is unchanged).           CORE ALTERATIONS and movement of irradiated fuel are the only operations that are postulated to result in a f ~. sion product release requiring the secondary containment. This assertion is supported by the fact that the current ACTIONS only require these operations suspended (i.e., it does not require further actions to restore secondary containment after the operations are suspended.)

L.2 The normal periodic surveillance frequency for the Unit 1 Secondary Containment tests provides adequate assurance of OPERABILITY. As such, the requirement to perform the Surveillance Requirement "within 24 hours prior to and every 7 days thereafter" has been deleted. If the Surveillance has not been performed within the normal specified interval, use of the secondary containment is not allowed since proposed SR 3.0.4 (current Specification 4.0.4) requires a Surveillance be performed prior to entering the applicable MODE or condition and be current. Additionally, plant operational experience has shown the normal periodic O Surveillance Frequency to be adequate for maintaining OPERABILITY. I l O HATCH UNIT 2 4 REVISION ,A

l CONTAINMENT SYSTEMS SECONDARY CONTAINMENT AUTOMATIC ISOLATION DAMPERS b %i&;b M .41 d LIMITING CONDITION FOR OPERATION M .tA4 d 3g2 dampees)/(The secondary 6.h2-D shall becontainment (fientilat'ison system .au

nowrt inJable-L OPERABLE.

                                                    '                           LA,1 APPLICABILITY:           CONDITIONS 1, 2, 3, [

ACTION:

                                        '             b, p (r W o A e s 2 4 3 % :n.v g ,(es,;         gut.               c-s pp With one or more of tne secondary containment [$ENtu ation-system w toma.tib                                 l ifg      isolaW @yc411ed tisJable s.6>5_2-L inoperable, operatio_n may f+.2.                                          i (kl>66'niTlnue dwrttrnmTgroHrthsnorification 3 o ure_nnLafptt_cabiprovided.thatJt9Ta
                                                                                                     ~

(Ech'Tffected penetr{. ion that ___i.Lopen,, aid; no T&~ismper d is resto OPERABLE hatus within pA b. The affected pe.1etration is isolated by use of a closedIdainp?

       .rea Nf             within 8 hours, or bM                 (c .                                          Temonstrated wi           in 8 hours _        *
                                                                                                                 -'N

( SE50QUARY and.the damper is r& CONTAI stored to OP N NT INTEGRITY d ays_.(ERABLE_ statu Otherwise, be in at least H0T SHUTDOWN within the next 12 hours and W in COLD SHUTDOWN within the following'24 hours, or ,

    '55 SURVEILLANCE RE0VIREMENTS                                                                                   k (A

L/

                                                                                                       .i 4.6.5.2         Each secondary containmen,t_ftentWtionlyhem autbmatif i s o l a t i o n cdimfer; specWte d 1n:Tabl e1;fmi. 2;L s h a l llTHeFoiis t ra t ed OPERABLE:
          .m "

pr 'ia. At least once per 92 da K by cycling each' automatic dam r' w3.cA3 testab'le during plant ope on through at l'es(t one complete m cycle of%11 travel. k p bnrd.%g LIn '&tv l,3 % H L.sM t [ p~y p,3~

           *hhenperformingi             e  ice hydrostatic or leak testing with the reacd

_ coolant temperature above 212 F. - p'",w, HATCH - UNIT 2 3/4 6-37 Amendment No. 91 Icl6' '

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4 I

b. rior to returning he damper to service a er maintenan regair or replacemen ork is performed on t damper or its LA, assohted actuator, co trql or power circuit b erformance of the D clino test and verification of isolation time. _..

c. At least once per Q during@lHUTD0'Nr REFlAINC 1 by:

Cycling each automa ic damper through at least one

   %% $. 1.         complete cycle of full travel and measuring the isolation time, arid                                                 j
           ~ 3 2. Verifying that on a secondary containment isolation est 90 '.4 t         signal each automatic isolation damper actuates to its isolation position.

f~% . ( A HATCH - UNIT 2 3/4 6-38 / y gjfg ;

S d 3. L . V. 2. Qebuv TABLE 3.6.5.2-1 SECONDARY CONTAINM T_ VENTILATION SYSTEM AUTOMATIC OLATION DAMPERS ISO TION TIME DAMPE FUNCTION (Setonds)

1. Re or Building Normal (Sup ly) Ventilation 5 Damp s (2T41 - F011 A and B)

2. Reactor Building Normal (Exhaust) 5 Ventilat Dampers (2T41 - F044 A nd B)

O ! l O HATCH - U"'.' 2 3,'1 6-39 3d6

i REFUELING OPERATIONS SECONDARY CONTAINMENT AUTOMATIC ISOLATION DAMPERS _ SPccA' M M t2 d LIMITING CONDITION FOR OPERATION 1 Lc 0%%2- LA.\ 3.9.5.2 fThe secondary containment Feht43atiohy_ stem ammatic 1 solation (damnrieftshwtt in Iah 3.9.bM-D shall be OPERABLE. APPLICABILITY: @NDITION'S3and*. w NYh. _ ACTION: 9"ed 'MT% mwC Ip r

                       /P %, a@s zu ] um

a. With one or more of the secondary containment @lonsystehD GREoda'MgrTs o l a t i o nCa a'mser s_ s p e ett l ed in Mble 315.2-D inoperable, Q k"0 6 operation may continue provided thatlat gast one is'oi n damp 3 ;

O s mahuined OPERAutt in eachMfected penMration that <L coen(and

                                                                                                 ~
  ,g                       noperable per is restoreQ0PERABLE status %tthin QQ              t                          s pope)         2. The affected penetration is isolated by use of a closed fabef.,___

sch within 8 hours. - - ~

                                                                                                          'y       l/

pA. Otherwise, suspend handling of irradiated fuel in the Hatch - rp0 0 Unit I secondary containment, and suspend Hatch - Unit 2 CORE ( ALTERATIONS gmQ?tiv1Mes tnatbuld reduce the SHUTDOWN MA

i) he3lom D.s 46L H& TW

b. Tne provisions of Specifications 3.0.3 . are not applicable g NT.e 6 1 SURVEILLANCE REQUIREMENTS 4.9.5._2_.1 Each secondary containment M Hhystem%tomiDp isola-h- tion daeet speciNed in Table %S.5.2-V shall be demonstrated GPERABLE: m )

duringIOMHUTDDMLor REFUELING ((M.s a. At least once per damper through at least one complete

       ';W g .d t l . Cycling       each autom cycle of full travel and measuring the isolation time, and                              h et e.c h) 43 2.       Verifying that on a secondary containment isolation test                             L5 d 92%                 signal each automatic isolation damper actuates to its isolation position.

b M or to returh g the Q damper to se? 'ce after maintenance, pair or replace nt work is performe on the damper or its j p'3/ a iated actuator ontrol or power ci it by performance t of t cling test an verificationofisohtiontime. J

         *When irradiated fuel is being handled in the Hatch - Unit 1 secondary containment.

HATCH - UNIT 2 3/4 9-8 o4 c 1

Sm ;A,5 3.c .q. a s l o TABLE 3.9.5.2-1 s SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION AMPERS

                                         \
                                           \                                      '

ISOLA ON TIME DAMPER hjNCTION (Sec ds)

1. Refuel i' Floor Normal (Supply) Ventila on 4.2 '

Dampers T41 - F003 A and B)

2. Refueling fl r Normal (Exhaust) Ventilation 4.2 Dampers (2T41 F023 A and B)

L A, l O O HATCH - UNIT 2 3/4 9-9 gg g,

DISCUSSION OF CHANGES (eD ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENT ISOLATION VALVES l { %) l @MINISTRATIVE I l A.1 Proposed ACTIONS Note 2 (" Separate Condition entry'is allowed for each penetration flow path") provides explicit instructions for proper application of the ACTIONS for Technical Specification compliance. In conjunction with the proposed Specification 1.3, " Completion Times,." this ACTIONS Note provides direction consistent with the intent of the existing Actions for inoperable isolation valves. Similarly, proposed ACTIONS Note 3 facilitates the use and understanding of the intent to consider the affect of inoperable isolation valves on other systems. If a system is determined to be inoperable due to inoperable isolation valves, the affected systems Actions must be entered. With the proposed LC0 3.0.6, this intent would not necessarily apply. This clarification is consistent with the intent and interpretation of the existing Technical Specifications, and is therefore considered an administrative presentation preference. 1 A.2 The Specification 3.0.4 statement has been deleted since proposed LCO 3.0.4 provides this allowance (i.e., this allowance has been moved to LC0 3.0.4). A.3 Proposed Condition A only applies if one valve in a penetration is

 /~                         inoperable.                                This inherently ensures maintaining at least one isolation i                         valve OPERABLE.

A.4 The revised presentation of actions (based on the BWR Standard Technical l Specifications, NUREG 1433) is proposed to not explicitly detail options ! to " restore...to OPERABLE status." This action is always an option, and is implied in all Conditions. Omitting this action is editorial. A.5 This phrase has been deleted since it is redundant to suspending CORE 1 ALTERATIONS. The only activities that affect SHUTDOWN MARGIN are fuel l movement in the RPV and control rod movement, both of which are considered CORE ALTERATIONS. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 An additional Required Action is included (proposed Required Action A.2) to periodically verify that the isolated penetration remains isolated. This verification will assure that if the penetration were inadvertently I re-opened, it would be identified. Since no periodic reverification of isolation is currently required, this change is more restrictive. The l Required Action is also modified by a hote which allows isolation devices i in high radiation areas to be verified by use of administrative means. l M.2 This optional allowance has been deleted. With a valve inoperable, the penetration must be isolated. This is an additional restriction on plant operation. V HATCH UNIT 2 1 REVISIONp'q

t c DISCUSSION OF CHANGES ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENT ISOLATION VALVES TECHNICAL CHANGE - MORE RESTRICTIVE (continued) , M.3 An additional surveillance requirement is included to periodically verify that each secondary containment isolation manual valve and blind flange that is required to be closed is closed. These passive isolation devices have not previously been included in the verification of closure except through the ability of the Standby Gas Treatment System to develop and maintain a vacuum. Therefore, this periodic verification constitutes a more restrictive change. M.4 The isolation time test Frequency (proposed SR 3.6.4.4.2) has been reduced l to 92 days, consistent with the BWR Standard Technical Specifications, NUREG 1433. M.5 An Applicability is being added requiring the secondary containment l isolation valves to be OPERABLE during OPDRVS. Appropriate ACTIONS and Surveillance Requirements are also added. This is consistent with the l requirements of the BWR Standard Technical Specifications, NUREG 1433. M.6 An additional Surveillance Requirement is included (proposed SR 3.6.4.2.1) to periodically verify that each secondary containment isolation manual valve and blind flange that is required to be closed is closed. These s/ passive isolation devices have not previously been included in the verification of closure except through the ability of the standby gas treatment system to develop and maintain a vacuum. Therefore, this periodic verification constitutes a more restrictive change. TECHNICAL CHANGE - LESS RESTRICTIVE

    " Generic" LA.1 The list of secondary containment isolation dampers has been relocated to the Technical Requirements Manual consistent with Generic letter 91-08.

Any change to the Technical Requirements Manual will be controlled by the provisions of 10 CFR 50.59. In addition, due to the relocation, the name of the isolation dampers has been generically changed to secondary containment isolation valves (SCIVs). l LA.2 This surveillance has been deleted since it is duplicative of SR 3.6.4.2.2. The valve is still cycled every 92 days via proposed SR 3.6.4.2.2, which also measures the stroke time. O HATCH UNIT 2 2 REVISION)ff

1 l DISCUSSION OF CHANGES ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENY ISOLATION VALVES TECHNICAL CHANGE - LESS RESTRICTIVE

   " Generic" (continued)                                        ,

l LA.3 Any time the operability of a system or component has been affected by l repair, maintenance or replacement of a component, post maintenance l testing is required to demonstrate operability of the system or components l Explicit post maintenance Surveillance Requirements have therefore been ' deleted from the specifications. Entry into the applicable modes without performing this post maintenance testing also continues to be precluded as discussed in the Bases for SR 3.0.1. l

   " Specific"                                                                            !

L.1 An allowance is proposed for intermittently opening closed secondary containment isolation valves under administrative control as is allowed in the existing primary containment Technical Specifications. The allowance is presented in proposed ACTIONS Note 1 and SR 3.6.4.2.1 Note 2. Opening l of secondary containment penetrations on a intermittent basis is required for many of the same reasons as primary contcinment penetrations and the potential impact on consequences is less significant. L.2 Current ACTIONS list some, but not all, possible acceptable isolation A devices that may be used to satisfy the need to isolate a penetration with (,) an inoperable isolation valve. The proposed change provides a complete list of acceptable isolation devices. Since the result of the ACTION continues to be an acceptably isolated penetration for continued operation, the proposed change does not adversely affect safe operation. L.3 In the event both valves in a penetration are inoperable, the existing Specification, which requires maintaining one isolation valve operable, would not be met and an immediate shutdown is required. The proposed actions for the secondary containment penetrations provide 4 hours prior to commencing a required shutdown. This proposed 4 hour period is consistent with the existing time allowed for conditions when the secondary containment is inoperable. The proposed change will provide consistency in actions for these various secondary containment degradations. L.4 The proposed surveillance for a functional test of each secondary containment isolation valve does not include the restriction on plant conditions that requires the surveillance to be performed during Cold l Shutdown or Refueling. Some isolations could be adequately tested in i other than Cold Shutdowr. or Refueling, without jeopardizing plant operations. The control of the plant conditions appropriate to perform i the test is an issue for procedures and scheduling, and has been 1 determined by the NRC Staff to be unnecessary as a Technical Specification l restriction. As indicated in Generic Letter 91-04, allowing this control 1 7 is consistent with the vast majority of other Technical Specification ! surveillances that do not dictate plant conditions for the surveillance. l HATCH UNIT 2 3 REVISIONfh

l DISCUSSION OF CHANGES ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENT ISOLATION VALVES

                                                                           ~
   ' TECHNICAL CHANGE - LESS RESTRICTIVE
    " Specific" (continued).                                       ,

L.5 The phrase " actual or," in reference to the automatic isolation signal, has been added to the surveillance requirement for verifying that each , SCIV actuates on an automatic isolation signal. This allows satisfactory automatic SCIV isolations for other than surveillance purposes to be used to fulfill the surveillance requirements. Operability is adequately ! demonstrated in either case since the SCIV itself cannot discriminate between " actual" or " simulated" signals. L.6 The Applicability has been modified to require the SCIVs only during CORE ALTERATIONS, not all the time while in MODE 5. (The movement of irradiated fuel is unchanged). CORE ALTERATIONS and movement of irradiated fuel are the only operations that are postulated to result in i a fission product release requiring the Secondary Containment (hence the , need for the SCIVs). This assertion is supported by the fact that the current Actions only require these operations suspended (i.e., it does not require further actions to restore the SCIVs after the operations are , suspended.) O O : HATCH UNIT 2 4 REVISIONh(.

(2NIAINMENT SYSTEMS 3/4.6.6 CONTAINMENT ATMOSPHERE CONTROL A

                                 $J.ANDQY     GAS TREATMENT SYSTEM
                                                                                                                               ~     '

bG LIMITING CONDITION FOR OPERATION , \ ll o Hatch-Unit 2 independent standby gas treatment subsystems 3.L.43 l t h- atment Aubsystems shall ' j BLE tL

                                                           .O1 indep(endant4 APPLICABILITY: CONDITIONS 1, 2, 3 A00 DPDgR      mmhM. _y ga Q       ~

ACTION: r Ac.T!o g 4,,3 . wl s a. With of the above require s andby gas treatment subsystems pQW. 'd nok i it ays,or be in at least HOT SHU stem to OPERABLE status

                                                        ._r'estoretheinoperablesubsYDOWNwithinthenext12                        hours pp c kand in LULU MiUTDOWN within the following 24 hours.

b. With two or more of the above required standby gas treatment subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD

((cTied E (LA 3 3) SHUTDOWN within the next 24 hours,[Eq as siivML, L. dun c.;

c. With both of the Hatch-Uni gastreatmht subsystems inoperable for i 1indekendentstandbtalla ion of the Un t I torus har ned vent, Unit 2 operation may contin for a cumulative total of up to 7 ays provided all of the following quirements are met:

1. Prior to removing either Unit I standby gas treatment subsyste from service demonstrate that a Negative pressure can be maintained 'n '

theUnil2secondarycontainmentandtheUnit1modifiedseconda containment under the followin conditions:

The Unit I secondary contai ent is in the modified mode.
Both Unit 2 standby gas treat nt subsystems are aligned with suction from both of the subje t areas and are operating with each filter train flow rate not ore than 4000 cfm. i
Calm wind conditions (< 5 mph) ex t.

O 2. Main ain both Unit 2 standby gas treatme t subsystems OPERABLE.

3. Mainta n Unit 2 secondary containment inte rity, except for Unit 1 O 4 standb gas treatment system OPERABILITY re utrements.

Maintain nit 1 modified secondary containmen integrity, except for Unit I sta by gas treatment system OPERABILIT requirements.

5. Allow no Uni 1 CORE ALTERATIONS.

6. Allow no handl of irradiated fuel or spent fuel hipping casks in the modified Un I secondary containment.

If both Unit I standby gas eatment subsystems are not restore to OPERABLE status within the allowable c ulative time period of 7 days, or f any of the I above requirements cannot be m , be in at least HOT SHUTDOWN with'n the next 1 and in rni n SHUTDOWN w thin the following 24 hours. J SURVEILLANCE REOUIREMENTS 4.6.6.1.1 Each Hatch-Unit 2 standby gas treatment subsystem shall be demonstrated OPERABLE:

a. fronf he control Psam. finw throuah 'tp HEP f pfiltert y ingi,atinghrena%1' nd e l%d s andGerifving that the system o erates for atlTeasCa'gjert j (p2 g,3. )

                    /                           t se heaters on aumr.u n concroi fotil of    10 hTMb each 31 days with 2               l dt'                                            At least once per 18 months or (1) after any structural main-                                                ,
            "e g g' p 3,2,,                     tenance on the HEPA filter or charcoal adsorber housings, or (2) following painting fire or chemical release in any venti-
               ' ' ' '                          lationzonecommunicatingwiththesystemby:

1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures Nodd b. of Regulatory Positions C.5.a C.5.c and C.S.d of Regula-tory Guide 1.52 Revision I duly 1976, and the system flow rate is 40 0 + 0, -100 cfm

        $[% M 44a S . 5 ~1                            2.      Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory O                                                        Guide 1.52 Revision 1 July 1976, meets the laboratory testingcriteriaofRegulator Q ry Guide 1.52, Revision 1, y Position                  July 1976. C.6.a of Regula-        g bIsc W '" "p \

rk.~p O W \ hen performing inservice h polanttemperatureabove2{drostaticorleaktestingwiththere 2*F. 3.p.y lawne g g pf, A HACH - UNIT 2 3/4 6-40 Amendment No. M , 12 Th % 8eu.#, I r y ~ L IK in c L.pkr '3 n o l

CONTAINMENT SYSTEt's

                                                                      $ p g u ,q g f k.,        SURVEILLANCE REQUIREMENTS (Continued) g 3.G.% 3.L                                                                      3 4 ,3 Verifying a system flow rate of 4000 +0, -1000 cfm during system operation when tested in accordance with ANSI pod                     N510-1975.

ho 5peJJW c. After overy 720 hours of charcoal adsorber operation by g,5 3 verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accord- ., ance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 1, July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976.

d. It least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Water Gauge while operating the filter train at a flow rate of 4000 +0, -1000 cfm. -

,)

f2 . Verifying that the filter train starts and isolation g n dampers open on each of the following test, signal (t"pR 3.c . 4.13

a. Drywell pressure-high, A. 3

b. High radiation on the;

1) Refueling floor, f 2) Reactor building.

c. Reactor Vessel Water Level-Low Low (Level 2 V Veritying that the heaters dissipate 18.5 1 1.5 l

       ;i      3*gqy4{3.        when tested in ac.cordance with ANSI N910 1975. -

02 weA h> SRcdtc :ka f f. 7 HATCH - UNIT 2 3/4 6 al Amendment No. 39, 109 lef f ( '

CONTAINMENT SYSTEMS , SURVEILLANCE REQUIREMENTS (Continued)

1. Af ter each complete or partial replacement of a H:PA filter A.

bank by verifying that the HEPA filter banks remove 2 99's of ] the DOP when they are tested in place in accordance with ANSI McWg g N510-1975 while operating the system at a flow rate of 4000 5pe4.fu3 4 +0, -1000 cfm. (. f 'l f. After each complete or partial replacement of a charcoal adsorber bank by verif,ving that the charcoal adsorbers remove 2 99's of a halogenated nydrocarbon regrigerant test gas when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 + 0, -1 j cfm s.f M U l 4.676.1.2 Each Hatch-Unit 1 standby gas treatment subsystem shall be SR .M411 demonstrated OPERABLE per Hatch-Unit 1 Technical Specifications. 5x 3.b.4.3.3 - f% . T Ow A > O HATCH - UNIT 2 3/4 6-42 7 3d6 I'Y

rw REFUELING OPERATIONS STANDBY GAS TREATMENT SYSTEM 5pec Wedu 3 c4 3 d ( LIMITING CONDITION FOR OPERATION ii # 3.<,.4 3 3.9.5.3 M Hatch-Unit 2 independent standby gas treatment subsystems and (i

              ' Hatch-Unit 1 independent standby gas treatment subsystems shall be
    ,4     OPERABLE.
                                ~

APPLICABILITY: (C IiDITIONS 5')and *. ACTION:

a. With one of the above required standby gas treatment subsystems gcnoo inoperable, restore the inoperable subsystem to OPERABLE status within fj W.

fFh h Suspend all irradiated fuel snd hent fuel sNoping casYhandling

i u [i D in the Hatch - Unit I secondary containment, and ua InCONDITION5,suspendHatch-Unit 2_COREALTERATIONS4n'd Lopeqtiuns thirQuuid reduceW SHUTDOWN MAEI g pg (- hgg g.s g, 2.

b. (Withtwoormoreoftheaboverequiredstandbygastreatmentsubsystems Cm.,F o iaoperab e:

a: A ti 1. Suspend all irradiated fuelCankpent wel shioMncLeash handling in the Hatch - Unit I secondary containment, and i 2. In CONDITION 5, suspend Hatch - Unit 2 CORE ALTERATIO N (_'~iiWRioni 0 GiaRould reducMhe 5HUTDOWNMRGIN # Both Unit 2 indepe dent trains of standby gas treatmeh may be inoperable for 12 h rs during Unit I reactor operatio or surveillance N of the Unit 2 primary ontainment excess flow isolation mpers if the following conditions ar met: a Using Unit I standby as treatment system and normal ven ilation, maintain at least 1/4" ,0 vacuum in Unit I secondary con ainment

2. A sure operability of both nit 1 SGTS filter trains

3. Assu Unit 2 SGTS valves to t refueling floor cannot be ope d

( 4. Allow g fuel movement in Units 1 2 FA

           *When irradiated fuel (ob%pch fuel ship'bing ca                   ing handled in the Hatch - Unit I secondary containment.

HATCH - UNIT 2 3/4 9-10 Amendment No. 58 l ufr }/g

Lt o 3.6. 4.3 ,

                                                                       -               h g3         REFUELING OPERATIONS llMITING CONDITION FOR OPERATION (Continued)
                    . Unit 2secondha containment integrity is int ct except for Unit 2 standby gas treal ent system operability requi ments f

4 ,1 f any of the above c ditions cannot be met, an or rly shutdown i s 11 be initiated and e reactor shall be brought to ot Shutdown wit in 12 hours and shall in Cold Shutdown within the ollowing 24 h g .- _ t'g wf The provisions of Specification 3.0.3 are not applicable. W ,,,, At 4 d.

      'f SURVEILLANCE RE0VIREMENTS 9"-h.$

4.9.5.3.1 3 ch of the above required Hatch - Unit 2 standby gas treatment] t subsystems shall be demonstratort npFRABLE per Specification 4.6.6.1.1. / 4.9.5.3.2 Each of the above required Hatch - Unit 1 standby gas treat subsystems shall be demonstrated OPERABLE per Hatch -jUnit 1 Technic Specifications. k c/bsed sfb 3 c.4.3.1, 3 4 4 5 L d 3 ' 3 o v HATCH - UNIT 2 3/4 9-10a Amendment No. 58 54' h..

DISCUSSION OF CHANGES ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM ADMINISTRATIVE A.1 This allowance is being deleted since it is a one-time allowance only and the hardened vent has been installed. A.2 The technical content of As requirement is being moved to Section 5 of the proposed Technical Spn fications in accordance with the format of the BWR Standard Technical Specifications, NUREG 1433. Any technical changes to this requirement are addressed in the Discussion of Changes associated with proposed Specification 5.5.7. A surveillance requirement is added (proposed SR 3.6.4.3.2) to clarify that the tests of the Ventilation l Filter Testing Program must also be completed and passed for determining OPERABILITY of the SGT System. Since this is a presentation preference that maintains current requirements, this change is considered administrative. A.3 The technical content of this requirement is being divided into two Surtr i :l ances . The majority of the SurveiHance will be performed in LC0 3.3.6.2 requirements. The actual system functional test portion will be performed as SR 3.6.4.3.3. This ensures the entire system is tested with proper overlap. A.4 This phrase has been deleted since it is redundant to suspending CORE O, ALTERATIONS. The only activities that affect SHUTDOWN MARGIN are fuel movement in the RPV and control rod movement, both of which are considered CORE ALTERATIONS. A.5 This Action has been deleted since it is applicable to operation of Unit

1. Thus, any needed allowance will be located in the Unit 1 Technical Specifications. 1 A.6 These surveillances have been replaced with explicit Surveillance i' Requirements instead of references to other Specifications. Any changet to these Specifications are described in the Discussion of Changes for !

ITS: Section 3.6.4.3, SGT System, in this section. 1 TECHNICAL CHANGE - MORE RESTRICTIVE l M.1 The Unit 1 SGi System Surveillances have been specifically written into ! this LC0 instead of providing a cross reference. The current Hatch Unit I surveillances are written as proposed SRs 3.6.4.3.1 and SR 3.6.4.3.2. Also, proposed SR 3.6.4.3.3 now applies to the Unit 1 SGT System. This is not currently required by the Unit 1 Technical Specifications. These changes and additions, therefore, are considered an additional restriction on plant operation. j O l I HATCH UNIT ? 1 REVISION % l 4

i DISCUSSION OF CHANGES ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM TECHNICAL CHANGE - MORE RESTRICTIVE (continued) i M.2 SR 3.6.4.3.1 requires the SGT System to be run 10 con'inuous t hours each 31 l days, while the CTS state a total of 10 hours. This is an additional restriction on plant operations. M.3 With four SGT subsystems required OPERABLE (see L.4 for other re'axations), the allowed out of service time for an inoperable Unit 2 SGT suisystem has been reduced from 30 days to 7 days and additional restrictions placed on an inoperable Unit 1 SGT subsystem. These are additional restrictions on plant operation. M.4 An Applicability is being added requiring the SGT System to be OPERABLE l during OPDRVS. Appropriate ACTIONS and Surveillance Requirements are also added. This is consistent with the requirements of the BWR Standard l Technical Specifications, NUREG 1433. TECHNICAL CHANGE - LESS RESTRICTIVE

   " Generic" LA.1 Details of the methods for performing this surveillance are relocated to O        the Bases and procedures. Changes to the Bases will be controlled by the V        provisions of the proposed Bases Control Process described in Chapter 5 of the Technical Specifications.           Changes to the procedures will    be controlled by the provisions of 10 CFR 50.59.

LA.2 If the spent fuel shipping cask has an irradiated fuel assembly in it, then the LC0 will be required because of the " movement of irradiated fuel assemblies" Applicability. Thus, for this case, the spent fuel shipping cask Applicability is redundant and has been deleted. For the case when the spent fuel shipping cask is empty, the requirement has been relocated to procedures. The licensing basis analysis is a dropped irradiated fuel assembly on other irradiated fuel assemblies, not a dropped piece of equipment (e.g., spent fuel shipping cask). The current Plant Hatch heavy loads analysis covers all loads considered heavy, which includes components much heavier than a spent fuel shipping cask. These loads will be handled under plant control, as allowed by the NRC Policy Statement on Technical Specifications. (Refer to Plant Hatch Unit 2 Application of Selection Criteria, CTS: 3/4.9.8 discussion). Therefore, handling of a spent fuel shipping cask will be handled in the same manner as other heavy loads which are not spent fuel assemblies. .O HATCH UNIT 2 2 REVISION O l t

p) s. DISCUSSION OF CHANGES ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM TECHNICAL CHANGE - LESS RESTRICTIVE

   " Specific" L.1   The phrase " actual or," in reference to the automatic initiation signal, has been added to the surveillance requirement for verifying that each subsystem actuates on an automatic initiation signal.          This allows satisfactory automatic system initiations for other than surveillance purposes to be used to fulfill the surveillance requirements. Operability is adequately demonstrated in either case since the subsystem itself cannot discriminate between " actual" or " simulated" signals.

L.2 Comment number not used. L.3 An ACTION Note is proposed to allow inspection of the Unit I hardened vent rupture disk while Unit 2 is operating. This inspection will cause both the Unit 1 SGT subsystems to be inoperable and, thus the allowance to delay entry into associated Conditions and Required Actions is needed, provided both the Unit 2 SGT subsystems are operable. The 24 hour allowance allows Unit 2 to continue operation during the inspection and minimizes the time when the Unit 1 SGT subsystems are inoperable. L.4 (Refer also to ITS 3.6.4.1 discussions A.3 and LA.1.) The required number v of OPERABLE SGT subsystems is proposed to be dependent on the configuration of the secondary containment. The current requirement for four OPERABLE SGT subsystems is based on all three zones of secondary containment being required for secondary containment OPERABILITY (Zone I - Unit I reactor building; Zone II - Unit 2 reactor building; Zone III - common refueling floor). Thi:, requirement remains unchanged. However, based on the specific operational status of each of the two units, certain zones may be isolated from the secondary containment boundary; thereby reducing the volume of the secondary containment. With a reduced secondary containment volume, a reduced number of SGT subsystems would be required to drawdown to and maintain that volume at 0.25 inches vacuum in the required time. There are threa unique configurations allowed in CTS with only one or two zones required for the secondary containment boundary, but currently requiring all four SGT subsystems to be OPERABLE. (Note two other new and unique configurations, i.e., Zone I only and Zone II only, for MODE 4 OPDRV, are addressed as a more restrictive change; see M.4 for this Specification.) These three configurations each can be adequately supported, given the single failure of one subsystem, by less than the currently required four subsystems. A brief discussion of each follows: O HATCH UNIT 2 3 REVISION [---

i DISCUSSION OF CHANGES ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM l TECHNICAL CHANGE - LESS RESTRICTIVE L.4 (continued) A. With Unit 2 operating, and Unit 1 shutdown with its reactor building (Zone I) isolated from Zone III such that these volumes do not communicate (i.e. CTS " modified" Hatch-Unit 1 secondary containment), Zones II and III become the secondary containment. In this configuration only two SGT subsystems are required to adequately perform the required secondary containment drawdown function. This specific configuration, with two Unit 2 SGT subsystems was previously reviewed by the NRC in an SER dated March 10, 1993 for Unit 2 amendment 124 and found satisfactory. Testing and analysis has shown that one Unit 2 and one Unit I subsystem can also adequately perform the required secondary containment drawdown function. (Note that two Unit 1 subsystems has not been shown to be adequate.) Based on this, and to accommodate a single failure, three subsystems (two Unit 2 and one Unit 1) are adequate to support secondary containment OPERABILITY. B. With Unit 2 in MODE 5 and handling irradiated fuel or performing CORE ALTERATIONS only Zones I and III may be required (i.e., CTS requirement for Hatch-Unit 1 secondary containment). This configuration is that which is required by Unit I while Unit 1 is s operating; and Unit 1 CTS require only three SGT subsystems be OPERABLE. While the Unit 1 CTS specify that two of the three subsystems be Unit I subsystems, analysis has shown that two Unit 2 subsystems can adequately perform the required secondary containment drawdown function. Therefore any three subsystems can provide the required support for secondary containment OPERABILITY in the event of any single failure. C. With both Units shutdown, with refueling floor activities requiring secondary containment OPERABILITY (CORE ALTERATIONS or handling irradiated fuel), Zone I and II can be isolated from Zone III. In this configuration, only Zone III need be drawn down and maintained at a negative pressure. Testing and analysis has shown that any single SGT subsystem can perform the necessary function. The analyses performed in support of the above configurations and minimum SGT subsystems also confirmed no significant impact on probabilities or consequences of an accident, and no reduction in any margin of safety. Additionally, the proposed presentation of SGT requirements in ITS does not explicitly detail the minimum number of OPERABLE SGT subsystems. Rather it states, "The Unit 1 and Unit 2 SGT subsystems required to support LC0 3.6.4.1, " secondary Containment," shall be OPERABLE." The Bases provides an outline of the above discussed configurations and number of required SGT subsystems, with reference to the Technical Requirements ( Manual for complete details. Given the number of variations of secondary

/-

HATCH UNIT 2 4 REVISION [/

O DISCUSSION OF CHANGES ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM TECHNICAL CHANGE - LESS RESTRICTIVE , L.4 (continued) containment boundaries, dependent on variations in the operational status of both units, the resulta~nt complexity of providing the details of all options within the ITS is deemed detrimental to operator usefulness. The simplified presentation provides adequate requirements to assure proper implementation without unwarranted complexity. L.5 The Applicability has been modified to require the SGT System just during CORE ALTERATIONS, not all the time while in MODE 5. (The movement of irradiated fuel is unchanged). CORE ALTERATIONS and movement of-irradiated fuel are the only operations that are postulated to result in a fission product release requiring the Secondary Containment (henct the need for the SGT System). This assertion is supported by the fact that the current Actions only require these operations suspended (i.e., it does not require further actions to restore SGT system after the operations are suspended.) L.6 An alternative is proposed to suspending operations if a standby gas treatment subsystem cannot be returned to operable status within seven O days, and movement of irradiated fuel assemblies or CORE ALTERATIONS are being conducted. The alternative is to initiate two OPERABLE subsystems of standby gas treatment and continue to conduct the operations. Since two subsystems are sufficient for any accident, the risk of failure of the subsystems to perform their intended function is significantly reduced if they are running. L.7 In evaluating the minimum SGT subsystem requirements in support of changes addressed in L.4 above, one specific configuration was deemed justified of an additional allowed out-of-service time: one inoperable SGT subsystem while all three zones comprise the secondary containment boundary (and ! therefore all four SGT subsystems required OPERABLE). In this configuration, if the Unit I reactor building-to-refueling floor plug is not installed (open communication between Zone I and Zone III) a Unit 1 SGT subsystem can be inoperable, and tests have shown that a high level of confidence remains that even with an additional single failure of any SGT subsystem (which is not necessary to assume while in ACTIONS) the required drawdown function could still be performed. With this specific configuration, a 30-day Completion Time is proposed, as presented in ACTION A. (Note also, that this fourth SGT subsystem is being added to the minimum OPERABLE requirements for Unit 1, and a "more restrictive" 30-day Completion Time is added to Unit 1. This change for Unit 2 will allow for consistency between the two units.) HATCH UNIT 2 5 REVISION D

CONTAINMENT SYSTEMS MSIV LEAKAGE CONTROL SYSTEM 'IE'M 3N*W l LIMITING CONDITION FOR OPERATION - I

             .4   TwoMSIVLeakageControlSystemhlCS)subsystemsshallbe OPER B E.                                                                             ',) l l

APPLICABI TY: CONDITIONS 1, 2 and 3. ' ACTION:

a. With on MSIV leakage control system subsyste inoperable, restore t inoperable subsystem to OPERABLE stKtus within 30 days or in at least HOT SHUTDOWN within th next 12 hours and in COLD UTDOWN within the following 24 hours. The pro-visions of Spe fication 3.0.4 are not applicable.

b. With both MSIV lea ge control system subsystems inopera le, be in at least HOT UTDOWN within 12 hours and in COLD 5 T-DOWN within the next hours.

SURVEILLANCE REQUIREMENTS 4.6.1.4 Each MSIV Leakage Control System s system shall be demonstrated ERABLE:

a. At least once per 31 days by starting e blower from the control room and operating the blower for at lea 15 minutes.

b. ach COLD SHUTDOWN, if not performed within e previous 92 s, by cy. ling each bleeder valve through at least one

{ com lete cycle of full travel,

c. At lea t once per 18 months by performance of a fun tional test whi includes simulated actuation of the subsys em throughout its operating sequence and verifying that ea h auto-matic valve ctuates to its correct position and the blo r .I starts and de lops:

1. For inboard . IVs - 100 scfm at a vacuum of 60" H 0,2 and

2. For outboard MSIt - 240 scfm at a vacuum of 50" H20.

O

       ..ATCH - LNIT 2                             3/4 0 /
                                                                                       /c4 }

DISCUSSION OF CHANGES CTS: SECTION 3/4.6.1.4 - MSIV LEAKAGE CONTROL SYSTEM TECHNICAL CHANGE - LESS RESTRICTIVE

         " Specific" L.1   This Specification is being deleted.      The Discussion of this change is provided in GPC letters dated January 6,1994 and February 3,1994 from J.

T. Beckham, Jr. to the NRC. O l O l l HATCH UNIT 2 1 REVISION A

DISCUSSION OF CHANGES ITS: SECTION 3.6 - CONTAINMENT SYSTEMS BASES The Bases of the current Technical Specifications for this section (pages B 3/4 6-1 through 8 3/4 6-7) have been completely replaced by revised Bases that reflect the format and applicable content of the proposed Hatch Unit 2 Technical Specification Section 3.6, consistent with the BWR' Standard Tachnical 3pecifications, NUREG 1433. The revised Bases are shown in the proposed Hatch Unit 2 Technical Specification Bases. i O L O HATCH UNIT 2 1 REVISION A

DISCUSSION OF CHANGES ITS: SECTION 3.7.5 - CONTROL ROOM AIR CONDITIONING (AC) SYSTEM TECHNICAL CHANGE - MORE RESTRICTIVE M.1 A Specification is added which delineates speci f'ic requirements for operability of the Control Room Air Conditioning System. This system is necessary to assure the habitability of the control room in a post design basis accident environment. The new Specification requires .three subsystems to be OPERABLE in MODES 1, 2, and 3, during movement of

            ' irradiated fuel assemblies in the secondary containment, during CORE         l ALTERATIONS, and during OPDRVs.       Appropriate ACTIONS and a Surveillance Requirement have been added. This new Specification replaces the current Surveillance to verify control room temperature is s 105 F, which did not adequately ensure the design basis was met. This change is consistent with the BWR Standard Technical Specifications, NUREG 1433 and is considered more restrictive on plant operations.

O O HATCH UNIT 2 1 REVISIONfh

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ELECTRICAL POWER SYSTEMS A.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION Lco 3.B. L 3.8.1.2 As a minimum, the following A.C. electrical power sources shall- , be OPERABLE: - u,d ,, emma a4&4aoJ Lto rguirimob

                                                                                 << uad L o%4 sosrce W 38 M a. One circuit]between the offsite transmission network and thi

( onsiteCass IE distribution system,'and 2 _ {a6&+16apM uj W r. de m ec h *3 LLo 18 Lb b. OnedieselgeNeratorEh: Q M 'f L # (I. A day tank contain!ng a minimum of 900 gallons of fuel go M'D y 2. A fuel storage tank containing a minimum of 33,000 gallons' l

    ^3 go 14'                of fuel, and M                   A fuel transfer pump.                          -
                                                                                              ]

APPLICABILITY: CONDITIONS 4 and 5. urig s,%4 g b)'"g.[p4 O ACTION: 'b () & %anw+EU* % . y. gu.3 With less than the above required A.C. electrical pour sources OPERABLE, u r3 suspend all operations involving rcRE ALTERATIC;is, irradiated fuel F

r's e d handlino.,6&i4tiveNeactWitrerEnGP or operations that have the A* uind S ential 'of draining the ' reactor UTre-crovlucas of Wific yddM (fion32.4airnotqppltt@ _ 1.h' propse J /2p 86*

g I" N 1* 4%4 Q 42.%J a ;

             .jLURVEILLANCE RE0VIREMENTS 5g.3% 2 \

4.8.1.2 At least the above_ required A.C. electrical power sources shall be h strated OPERABLEfper Surveillance . qui rment e a_A 1.1A L,l QJ. .8. l . l .2.a.M (f8.1.1) ar,d

            @ eel.f .4except tor 7Fe requirement of          _

Q \(JA7A rrue 4 h Lc 3 3,pg 7 j {, p/ p 5(' D O HATCH - UNIT 2 3/4 8-9 Amendment No. 119 Iel l

f l DISCUSSION OF CHANGES ITS: SECTION 3.8.2 - AC SOURCES-SHUTDOWN v ADMINISTRATIVE A.1 The details relating to the required day tank level have been moved to a Surveillance Requirement (proposed SR 3.8.2.1, which' requires performance of SR 3.8.1.3) . No technical changes are being made; therefore, this change is considered administrative in nature. A.2 The words "in the secondary containment" have been added to clarify what l irradiated fuel handling operations need to be suspended. This change provides needed clarity and is administrative only. A.3 This exception has been deleted since it is no longer needed. The referenced Surveillance has been deleted from the proposed LC0 3.8.1, thus, refer 'ncing it here is not required. A.4 The requirement for when a DG special report is required (which is what Specification 4.8.1.1.4 encompasses) is governed by proposed Specification 5.6.7. This cross-reference, therefore, is unnecessary and has been deleted. A.5 This statement has been deleted since proposed LC0 3.0.3 states it is only applicable in MODES 1, 2, and 3 (and this LC0 is applicable in MODES 4 and 5 or defueled). A.6 The technical content of this requirement is being moved to LC0 3.8.5 of the proposed Technical Specifications in accordance with the format of the BWR Standard Technical Specifications, NUREG 1433. Any technical changes to this requirement will be addressed with the content of the proposed LC0. A.7 The technical content of this requirement is being moved to LC0 3.8.3 of the proposed Technical Specifications in accordance with the format of the BWR Standard Technical Specification, NUREG 1433. Any technical changes to this requirement are addressed with the content of the proposed LCO. A.8 The phrase " positive reactivity changes" has been deleted since it is redundant to CORE ALTERATIONS. The only activities that result in positive reactivity changes are fuel movement in the RPV and control rod movement, both of which are considered CORE ALTERATIONS. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 Certain equipment needed to v.et Unit 2 accident analysis is powered from Unit 1 AC Sources. Currently, the Unit 1 AC Sources are required since the Unit 2 definition of OPERABILITY requires both normal and alternate power supplies to be OPERABLE. In the proposed Technical Specifications, I the definition of OPERABILITY only requires one source, since proposed LC0 3.8.1 provides the proper ACTIONS to take if sources are inoperable. V l I i HATCH UNIT 2 1 REVISION (-

I DISCUSSION OF CHANGES ITS: SECTION 3.8.2 - AC SOURCES--SHUTDOWN

O ' V TECHNICAL CHANGE - MORE RESTRICTIVE M.1 ' (continued) Therefore, the Unit I requirea AC Sources have been added to this LCO. Since Unit I sources are now described, the current LC0 for Unit 2 sources has been modified to explicitly use the Unit designator, for clarity. In addition, the SRs are also applicable to the Unit 1 sources; the proposed SR 3.8.2.2 has been added to ensure Unit I sources are tested. M.2 The existing requirement for one offsite circuit to be OPERABLE during shutdown conditions is not specific as to what that circuit must be capable of powering. The proposed requirement specifies that the circuit must be available to supply power to all equipment required to be OPERABLE in the current plant condition. This added restriction conservatively assures the single OPERABLE circuit is performing a vital function. Since the circuit OPERABILITY requirements are proposed to require availability to all necessary loads, if one or more required load centers, MCC, buses, etc. are not capable of being powered via an offsite circuit, that circuit is inoperable. In this event it may not be necessary to suspend all CORE ALTERATIONS, irradiated fuel handling, and OPDRVs. Conservative ACTIONS can be assured if all required equipment without

  ,)       qualified offsite power availability is declared inoperable and the associated ACTIONS taken.

Therefore, along with the conservative additional requirements placed on the OPERABLE circuit, Required Action A.1 is al so proposed. These additions represent restrictions consintent with implicit assumptions for operation in shutdown conditions; restrictions which are not currently imposed via the Technical Specifications. l l M.3 Similar to the added restrictions for an OPERABLE offsite circuit, the single required OPERABLE DG during shutdown conditions is not specific as to what Division that DG must be associated with. The proposed LC0 requirement will ensure the OPERABLE DG is associated with one or more ! systems, subsystems, or components required to be OPERABLE. This added ; restriction enforces a level of Technical Specification control which ! currently is enforced only via administrative procedures. M.4 An additional Applicability has been added, requiring the AC Sources during movement of irradiated fuel assemblies in the Secondary l , Containment. Since this could occur when the reactor ic defueled (thus, i not in M0DE 4 or 5), this change is an additional restriction on plant 1 operation. O l l HATCH UNIT 2 2 REVISION f (--

ELECTRICAL POWER SYSTEMS Spm; ( 6 p,7 A.C. SOURCES - SHUTOOWN LIMITING CONDITION FOR OPERATION J r$oi-k ) m 8 a minimum, the following A.C. electrical power sources shall

a. One circuit between the offsite transmission network and the onsite Class lE distribution system, and

b. One diesel generator with:

_1 A dadnk containing a _ minimum of 900 gallons of fuel,

2. A fuel storage tank containing a minimum of 33,000 gallons l of fuel, and 4

3. A fuel transfer pump.

APPLICABILITY: CONDITIONS 4 and 5 W3**d atADdd g.1 Gd c65% b lie s lo tA t ACTION: 3 " * " % " '^* With less than the above required electrical DowD sources OPERABLE, ha suspend all operations involving CORE ALTERATIONS, irradiated fuel A handlin_q.{pbqtive reast1vitrcTiEW!D1 or opeXro_ns that have the potential of dr'aining the reactor vessel. /TheNrovisionuf spec 1Nca-J (t%tt 3.DN are nok.applic'alil f(* enca 41 Ryco % q,3

h. t.9 SVRVEILLANCE RE0UIREMENTS 6 .1.2 At least the above required A.C. electrical power sources shaTD f be demonstrated OPERABLE one surveillanco Requirements 4.8.1.1.1.

4.8.1.1 g35ept. ivi t he Tentrtreme n t%4__. 8.1.1. 2 . a . 5 / 8. 6831- ,, R

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                        .See Obewaa6 (LpQ 5 g . 3.g.z                                 Et Niscassonsf A t So ace)_ .5 L J L '                          b        M : 3 s .f,
                   \ im 4                                                bul 6 \ oi\ c-J T4 W C41, w 4 5 % war '

O in t % b , HATCH - UNIT 2 3/4 8-9 Amendment No. 119 2cA k

j. DISCUSSION OF-CHANGES ITS: SECTION 3.8.5 - DC SOURCES-SHUTDOWN TECHNICAL CHANGE - MORE RESTRICTIVE M.1 The existing requirement for " Division 1 or Division 2" DC sources to be OPERABLE during shutdown conditions is not specific a's to what that single source must be powering. The proposed requirement specifies that the sources necessary to supply DC power to all equipment required to be OPERABLE in the current plant condition must be OPERABLE. This added restriction conservatively assures the needed sources of power are OPERABLE, even if this results in both the Division 1 and Division 2 sources being required. The ACTION has been subsequently modified to be "one or more" instead of the current "one", to account for this potential addition. Since the proposed DC source OPERABILITY requirements require supplying power to all necessary loads, if one or more required DC loads are not being supplied the required DC power, the DC source is inoperable. In this event it may not be necessary to suspend all CORE ALTERATIONS, irradiated fuel handling, and OPDRVs. Conservative ACTIONS can be assured if all required equipment without the necessary DC power is declared inoperable and the associated ACTIONS taken. Therefore, along with the conservative additional requirements placed on the DC system, Required Action A.1 is also proposed. These additions represent restrictions consistent with implicit assumptions for operation O in shutdown conditions (required equipment receiving the necessary required power)--restrictions which are not currently- imposed via the Technical Specifications. M.2 An additional applicability has been added, requiring the DG DC sources during movement of irradiated fuel assemblies in the secondary l l containment. Since this could occur when the reactor is (efueled (thus, not in MODE 4 or 5), this change is an additional restrit. tion on plant operation. l M.3 In the event the necessary DG DC sources are not OPERABLE. plant , conditions are conservatively restricted by suspending CORE ALTERAi'ONS, ! irradiated fuel handling, and OPDRVs. However, continued operation ; without the necessary DG DC sources is not considered acceptable. Therefore, an ACTION to commence and continue attempts to restore the necessary DG DC sour ces is proposed. (Note that if ACTIONS are taken in ; accordance with the proposed Required Action A.1, sufficiently conservative measures are assured by the ACTIONS for the individual components declared inoperable without requiring the efforts to restore the inoperable source). O i HATCH UNIT 2 2 REVISION h -

I l

                                                                     .5p*c ik ka 3 8.s ELECTRICAL POWER SYSTEMS A.C. DISTRIBUTION - SHUTDOWN LIMITING CONDITION FOR OPERATION                      hj)--l=4%JPpimeA)                         '

leo M<.if.o } ,.-h A 3.8. .2 As a miiiimum, the([ollowi@ A.C. distribution Gstem busI, Q ny e r te r sgiqw r geqe ra Luw ( 6 ) w i.y shall be OPERABLE: I A.3 so u. . ~ ~Two 4160 volt Essentia uses, 2E, 2F and/or 2G, g .s t-

b. 600 volt Essential Bus, C or 2D, I

c. One 12 '08 volt Essential Cabine 2A or 2B,

d. One 120/208 v t Instrument Bus, 2A o B, ands (e. A.C. inverters 2R44-S002and2R44-5003*.3 % .,d b h.6 W 362.

{3.61.t .L,} hrig swmw4 c4 irru kle J u .t APPLICABILITY: CONDITIONS 4 and 5f L6adis W %< se+a rx g O' ACTION: hsed %=De A W h

                                                                                                   %< 4-%

thM

a. With

                          ' a ess than the above reauired A.C. distribution \syste#

ind inverterV 07ETABLE, suspend all operations invoiving [ 75' ] b h C0lE ALTERATIONS, irradiated fuel handling, puiwedeacuvit g Nor operations that have the potential of draining the reactor vessel. ffh M ovisiu q ui Spe Q{icetion 3.0. M re n % g

                        @NA                 .-fp rope:H Qerird M.~ 4.4$                                   l SURVEILLANCE REQUIREMENTS N P(*fd $*t"" ib '" M 1 h 'O
                                                                                        ~

A&b) l g kf> ^4.8.2.2 At least the above required A.C. distribution GEstem bussi and (nTertei'y shall be determined OPERABLE per Specification 4.8.2.1. a b

%d%

b %CL l l l l s.vd4,uu35.t O

Required to be operable when LPCI is required to be op D %

                                        ~                                                                 l 1

HATCH - UNIT 2 3/4 8-12 Amendment No. 23 J Ic4 2-

1 DISCUSSION OF CHANGES , ITS: SECTION 3.8.8 - DISTRIBUTION SYSTT.MS-SHUTDOWN ! O ADMINISTRATIVE l A.7 , (continued) This interpretation of the intent is supported by the BWR Standard l Technical Specifications, NUREG 1433. As an enhanced presentation of the l existing intent, the proposed change is deemed to'be administrative. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 The existing requirement for "following" AC Distribution buses, which only list one AC division and the DC " Division 1 or Division 2" Distribution systems to be OPERABLE during shutdown conditions, is not specific as to what that single system must be powering. The proposed requirement specifies that the distribution systems necessary to suppiy AC/DC power to all equipment required to be OPERABLE in the current plant condition must be OPERABLE. This added restriction conservatively assures the needed sources of power are OPERABLE, even if this results in both the Division 1 and Division 2 distribution systems being required. Since the distribution system OPERABILITY requirements are proposed to require supplying power to all necessary loads, if one or more required O loads are not being supplied the required power, that distribution subsystem is inoperable. In this event it may not be necessary to suspend all CORE ALTERATIONS, irradiated fuel handling, and OPDRVs. Conservative ACTIONS can be assured if all required equipment without the necessary power is declared inoperable and the associated ACTIONS taken. l Therefore, along with the conservative additional requirements placed on the distribution system, Required Action A.1 is also proposed. These additions represent restrictions consistent with implicit assumptions for operation in shutdown conditions (required equipment receiving the necessary required power) -- restrictions which are not currently imposed via the Technical Specifications. M.2 An additional applicability requiring the AC Distribution subsystems during movement of irradiated fuel assemblies in the secondary containment j has been added. Since this could occur when the reactor is defueled (thus, not in MODE 4 or 5), tnis change is an additional restriction on plant operation. O HATCH UNIT 2 2 REVISION f h

DISCUSSION OF CHANGES

 ;            ITS: SECTION 3.9.7 - RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL TECHNICAL CHANGE - MORE RESTRICTIVE M.1   The LC0 has been modified by requiring the shutdown" cooling subsystem to be in operation, except as allowed by the Note to the LC0 (which allows the subsystem to be removed from operation 2 hours every 8. hour period).

An appropriate ACTION (proposed ACTION C) has also been added to. provide compensatory measures when the subsystem is not in operation. This is an additional restriction on plant operation since currently the subsystem is not required to be operating. M.2 This action has been rewritten to include not only secondary containment and any penetrations, but to also include requiring necessary standby gas treatment (SGT) subsystem (s) to be OPERABLE. The actions now require: a) the secondary containment to be OPERABLE (Required Action B.2)

                            -- which is consistent with the current requirement to close all secondary containment penetrations; b)     at least one valve and associated instrumentation to be OPERABLE in each secondary containment penetration flow path not isolated (Required Action B.4) -- which is consistent with the current requirement since an OPERABLE, open valve will function to close, if required, if the associated O                        instrumentation is OPERABLE; and c)     OPERABLE SGT subsystem (s) Required Action B.3) -- which will    l maintain the secondary containment at a negative pressure, if required.      In addition, the Bases include a discussion acknowledging the possibility of various configurations for the secondary containment boundary.         Given the specific configuration, either one, two, or three SGT subsystems may be .

required to assure the necessary. negative pressure when required. As discussed in the Bases, these details are provided in the Technical Requirements Manual (TRM). M.3 The Frequency has been changed from once per 31 days to once per 12 hours. This is consistent with the BWR Standard Technical Specifications, NUREG 1433, and is an additional restriction on plant operation. TECHMICAL CHANGE - LESS RESTRICTIVE

        " Generic" LA.1 The details relating to system OPE TBILITY have been relocated to the Bases. The Bases will indicate that an OPERABLE RHR shutdown cooling system consists of at least an OPERABLE pump and heat exchanger. These are design features that are also described in the FSAR. Placing this                  :

detail in the Bases provides assurance that they will be maintained. Changes to the Bases will be controlled by the provisions of the Bases O- Control Process described in Chapter 5 of the Technical Specifications. HATCH UNIT 2 2 REVISIONfh

DISCUSSION OF CHANGES I- ITS: SECTION 3.9.8 - RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL l ADMINISTRATIVE l (continued) , A.5 This Surveillance has been rewritten to verify one RHR shutdown cooling subsystem is operating. This is the same as the current requirement, since if it is operating, 1) the RHR pump is not required to be started (current Surveillance 4.9.12.a) and 2) the system valves must be aligned for the subsystem to be operating properly. Thus, the proposed surveillance is equivalent to the current surveillance. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 The LCO has been modified by requiring two shutdown cooling subsystems instead of the current one subsystem. Thus, two RHR loops, each with a pump and heat exchanger, or one RHR loop with two pumps and a heat exchanger will be required. In addition, one subsystem must be in operation except as allowed by the Note to the LCO (which allows the subsystem to be . amoved from operation 2 hours every 8 hour period). An appropriate ACTION (proposed ACTION C) has also been added to provide compensatory measures when the subsystem is not in operation. This is an additional restriction on plant operation since currently only one pump is required to be OPERABLE and none are required to be operating. M.2 This action has been rewritten to include not only secondary containment and any penetrations, but to also include requiring necessary standby gas treatment (SGT) subsystem (s) to be OPERABLE. The actions now require: a) the secondary containment to be OPERABLE (Required Action B.1), -- which is consistent with the current requirement to close all secondary containment penetration; b) at least one valve and associated instrumentation to be OPERABLE in each secondary containment penetration flowpath not isolated (Required Action B.3), -- which is consistent with the current requirement since an OPERABLE, open valve will function to close, if required, if the associated instrumentation is OPERABLE; and c) OPERABLE SGT subsystem (s) (Required Action B.2), -- which will l maintain the secondary containment at a negative pressure, if required. In addition, the Bases include a discussion acknowledging the possibility of various configurations for the secondary containment boundary. Given the specific configuration, either one, two, or three SGT subsystems may be required to assure the necessary negative pressure when I required. As discussed in the Baser., these details are provided in the Technical Requirements Manual (TRM). .O HATCH UNIT 2 2 REVISION [ @ l

DISCUSSION 0F CHANGES ITS: SECTION 3.9.8 - RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL TECHNICAL CHANGE - MORE RESTRICTIVE (continued) M.3 The Frequency has been changed from once per 31 days to once per 12 hours. This is consistent with the BWR Standard Technical Specifications, NUREG 1433, and is an additional restriction on plant operation. - O l HATCH UNIT 2 ( ;, 4 REVISION %(_ 1 l

DISCUSSION OF CHANGES CTS: SECTION 3/4.9.5.2 - SECONDARY CONTAINMENT AUTOMATIC ISOLATION DAMPERS ADMINISTRATIVE A.1 The technical content of this requirement is being moyed to Section 3.6 of the proposed Technical Specifications. Any technical changes to this requirement are addressed with the content of proposed LC0 3.6.4.2. l O O HATCH UNIT 2 1 REVISION -

n DISCUSSION OF CHANGES (g CTS: SECTION 3/4.9.5.3 - STANDBY GAS TREATMENT SYSTEM ADMINISTRATIVE l A.1 The technical content of this requirement is being moyed to Section 3.6 of the proposed Technical Specifications. Any technical changes to this requirement are addressed with the content of proposed LC0 3.6.4.3. l V O HATCH UNIT 2 1 REVISION [ @

i I l O i l UNIT 2 NO SIGNIFICANT HAZARDS DETERMINATION : l O O , i 1

NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.1 - SECONDARY CONTAINMENT l l L.1 CHANGE l In accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

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

This change requires the secondary containment to be OPERABLE in MODE 5 only during CORE ALTERATIONS and movement of irradiated fuel. The secondary containment is not assumed to be an initiator of any analyzed accident. Therefore, the change does not significantly increase the probability of an accident previously evaluated. The consequences of any analyzed accident remain unchanged since secondary containment is still required during all conditions analyzed in the safety analysis. Therefore, the change does not significantly increase 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 any design changes, plant O modifications, or changes in plant operation. The secondary containment will continue to function in the same way as before the change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety is unchanged since the secondary containment is still required OPERABLE during CORE ALTERATIONS and movement of irradiated fuel assemblies. Since these are the only instances when a postulated fission product release is assumed to occur (with the exception of a vessel drain down event, which is covered by another Specification), this change does not involve a significant reduction in a margin of safety. O HATCH UNIT 2 1 / 4 REVISIONJ

 -                         NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.1 - SECONDARY CONTAINMENT                      l L.2 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, G'eorgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

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

The proposed change would remove an additional performance of a surveillance which has been performed within its normally required frequency. Not performing the surveillance would not affect any equipment which is assumed to be an initiator of any analyzed event. Since the surveillance continues to be performed on its normal frequency, there is no impact on the capability of the system to perform its required safety function. Therefore, the proposed change does not involve a significant increase in the probability or 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?

p The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed change does not introduce a new mode of plant operation does not involve physical modification to the plant.

3. Does this change involve a significant reduction in a margin of safety?

The normal surveillance frequency has been shown, based on operating experience, to be adequate for assuring the equipment is available and capable of performing its intended function. Additionally, the require-ments of SR 3.0.4 (current Specification 4.0.4) provide assurance the equipment is OPERABLE prior to beginning the functions for which it is required. Therefore, the proposed change does not involve a significant reduction in a margin of safety. O HATCH UNIT 2 2 REVISION %(,

NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENT ISOLATION VALVES l L.1 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, G'eorgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

1. Does the change involve a signi#icant increase in the probability or consequences of an accident previot. sly evaluated?

This change would allow an isolated secondary containment penetration to be opened under administrative controls similar to most other primary containment penetrations. Secondary containment isolation is not , considered as an initiator of any previously analyzed accident. Therefore, this change does not significantly increase the probability of such accidents. The proposed administrative controls provide an acceptable compensatory action to assure the penetration is isolated in the event of an accident. Therefore, the consequences of a previously analyzed event that may occur during the opening of the isolated line would not be significantly increased.

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

This change provides an additional acceptable compensatory action following failure of other equipment. The current requirements are based on providing a single active failure proof boundary to compensate for the loss of one of the two active boundaries. The proposed change provf des an alternative which essentially returns the system to its original configuration (i.e., configuration which can provide a single active failure proof boundary.) Therefore, this change does not create the possibility of a new or different kind of accident from any previously analyzed accident.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety considered in determining the required compensatory action is also based on providing the single active failure proof boundary. Since the proposed compensatory boundary essentially meets the original criteria and provides leakage characteristics essentially similar to currently approved compensatory boundaries, the change does not involve a significant reduction in the margin of safety. O HATCH UNIT 2 1 REVISIONh(_

NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENT ISOLATION VALVES l l L.2 CHANGE

In accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

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

The proposed change provides additional acceptable isolation devices for compliance with ACTIONS. Primary containment isolation is not considered as an initiator of any previously analyzed accident. Therefore, this change does not significantly increase the probability of such accidents. The p,^oposed additional isolation devices provide an acceptable compensatory action to assure the penetration is isolated in the event of an accident. Therefore, the consequences of a previously analyzed event is not significantly increased.

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

Since the result of the ACTION continues to be a acceptably isolated penetration for continued operation, the proposed change does not t adversely affect safe operation. Therefore, this change does not create the possibility f a new or different kind of accident from any previously analyzed accident.

3. Does this change involve a sigaificant reduction in a margin of safety?

l The margin of safety considered in determining the required compensat7ry action is based on providing the single active failure proof boundary. Since the result .of the ACTION continues to be an acceptably isola',ed penetration for continued operation, the proposed change does not adversely affect safe operation. Therefore, the change does not insolve a significant reduction in the margin of safety. U , l HATCH UNIT 2 2 REVISION h(- j

NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENT ISOLATION VALVES l l 1 L.3 CHANGE ! In accordance with the criteria set forth in 10 CFR 50.92, G'eorgia Power Company has evaluated this proposed Technical Specifications change and determined it does nut involve a significant hazards consideration based on the following:

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

This change would allow additional time to isolate a secondary containment penetration if both isolation devices are inoperable. Secondary containment isolation is not considered as an initiator of any previously , analyzed accident. Therefore, this change does not significantly increase the probability of such accidents. The proposed change allows additional limited operation with less than the required isolation capability. - However, the consequences of an event that may occur during the extended outage time would not be _ay different than during the currently allowed outage time for other loss of secondary containment integrity situations. Therefore, this change does not significantly increase the consequences of any previously analyzed accident. *

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

This change does not result in any changes to the equipment design or capabilities or to the operation of the plant. Further, since the change impacts only the required action completion time for the system and does not result in any change in the response of the equipment to an accident, the change does not create the possibility of a new or different kind of accident from any previously analyzed accident.

3. Doe it is change involve a significant reduction in a margin of safety?

This change impacts only the required action completion time for

inoperable valves that provide secondary containment isolation. The i methodology and limits of the accident analysis are not affected, and the secondary containment response in unaffected. Therefore, the change does not involve a significant reduction in the margin of safety.

5 O - HATCH UNIT 2 3 REVISION'y ;

1 N0 SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENT ISOLATION VALVES l L.4 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, G'eorgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

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

This change would remove a specific restriction to perform a surveillance of the secondary containment isolation valves during shutdown. Secondary containment isolation is not considered as an initiator of any previously analyzed accident. Therefore, this change does not significantly increase the probability of such accidents. The appropriate plant conditions for performance of the surveillance will continue to be controlled to assure the potential consequences are not significantly increased. This control method has been previously determined to be acceptable as indicated in Generic Letter 91-04. Therefore, this change does not significantly increase the consequences of any previously analyzed accident.

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

O V This change removes a specific restriction on the plant conditions for performing a surveillance, but does not change the method of performance. The appropriate plant conditions for performance of the surveillance will continue to be controlled to assure the possibility for a new or different kind of accident are not created. This control method has been previously determined to be acceptable as indicated in Generic Letter 91-04. Therefore, this change does not create the possibility of a new or different kind of accident from any previously analyzed accident.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety considered in determining the appropriate plant conditions for performing the surveillance will continue to be controlled to assure that there is no significant reduction. This control method has been previously determined to be acceptable as indicated in Generic Letter 91-04. Therefore, the change does not involve a significant reduction in the margin of safety. l 1 HATCH UNIT 2 4 REVISIONh(,

NO SIGNIFICANT HAZARDS DETERMINATION C) v ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENT ISOLATION VALVES l L.5 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, G'eorgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

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

The phrase " actual or," in reference to the automatic isolation signal, has been added to the system functional test surveillance test description. This does not impose a requirement to create an " actual" signal, nor does it eliminate any restriction on producing an " actual" signal. Creating an " actual" signal could increase the probability of an event, existing procedures and 10 CFR 50.59 control of revisions to them, dictate the acceptability of generating this signal. The proposed change does not affect the procedures governing plant operations and the acceptability of creating these signals; it simply would allow such a signal to be utilized in evaluating the acceptance criteria for the system functional test requirements. Therefore, the change does'not involvc a significant increase in the probability of an accident previously evaluated. O Q The system functional test remains unaffected and the secondary containment isolation valves continue to perform their safety function. Therefore, the 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 possibility of a new or different kind of accident from any accident previously evaluated is not created because the proposed change does not introduce a new mode of plant operation and does not involve physical modification to the plant.

3. Does this change involve a significant reduction in a margin of safety?

Use of an actual signal instead of the existing requirement which limits use to a simulated signal, will not affect the performance of the surveillance test. Operability is a.iequately demonstrated in either case ! since the system i tsel f can not discriminate between " actual" or

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

O HATCH UNIT 2 5 REVISION Q

l l N0 SIGNIFICANT HAZARDS DETERMINATION O U ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENT ISOLATION VALVES l L.6 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, G'eorgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following: I. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? The change requires the secondary containment isolation valves to be OPERABLE in MODE 5 only during CORE ALTERATIONS and movement of irradiated fuel. The secmdary containment isolation valves are not assumed to be an initiator of any analyzed accident. Therefore, the change does not involve a significant increase in the probability of an accident previously evaluated. The consequences of any analyzed accident remain unchanged since secondary containment isolation valves are still required during all conditions analyzed in the safety analysis. Therefore, the ! 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?

O Q The proposed change does not involve any design changes, plant modifications, or changes in plant operation. The valves will continue to function in the same way as before the change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety is unchanged since the secondary containment isolation valves are still required OPERASLE during CORE ALTERATIONS and movement of irradiated fuel assemblies. Since these are the only l instances when a postulated fission product release is assumed to occur (with the exception of a vessel drain down event, which is covered by another Specification), this change does not involve a significant reduction in a margin of safety. i

l 1 O HATCH UNIT 2 6 REVISION y j l i

l l q NO SIGNIFICANT HAZARDS DETERMINATION Q ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM l L.1 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, G'eorgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

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

The phrase " actual or," in reference to the automatic initiation signal, has been added to the system functional test surveillance test description. This does not impose a requirement to create an " actual" signal, nor does it eliminate any restriction on producing an " actual" signal. Creating an " actual" signal could increase the probability of an event, existing procedures and 10 CFR 50.59 control of revisions to them, dictate the acceptability of generating this signal. The proposed change does not affect the procedures governing plant operations and the acceptability of creating these signals; it simply would allow such a signal to be ctilized in evaluating the acceptance criteria for the system functional test requirements. Therefore, the change does not involve a significant increase in the probability of an accident previously evaluated. Since the function of the system functional test remains unaffected the 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 possibility of a new or different kind of accident from any accident previously evaluated is not created because the proposed change does not ) introduce a new mode of plant operation and does not involve physical modificaticn to the plant.

3. Does this change involve a significant reduction in a margin of safety?

J Use of an actual signal instead of the existing requirement which limits use to a simulated signal, will not affect the performance of the surveillance test. Operability is adequately demonstrated in either case since the system itself can not discriminate between " actual" or

        " simulated" signals. Therefore, the change does not involve a significant I

reduction in a margin of safety. O s HATCH UNIT 2 1 REVISIONk(_ l

NO SIGNifICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM l L.2 CHANGE

Not used. O O HATCH UNIT 2 2 REVISIONh(_

N0 SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM l L.3 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, G'eorgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

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

The proposed change will add a Note to the Actions to allow continued operation of Unit 2 while the Unit 1 inspection of the hardened vent rupture disc is taking place. In order to perform the inspection of the hardened vent, both Unit 1 SGT subsystems are rendered inoperable. Since there is no Action for both Unit 1 SGT subsystems being inoperable while Unit 2 is in Modes 1, 2, or 3, the plant would have to be shutdown per LC0 3.0.3. In order to avoid an unnecessary dual unit shutdown for this inspection and to maintain necessary plant safety, a 24 hour limit is placed on the time allowed for the inspection. At the end of the inspection or the 24 hour time limit, whichever comes first, the Unit 1 SGT subsystems must be returned to OPERABLE status or the Actions of LC0 3.6.4.7 must be entered. The SGT subsystems perform an accident mitigation function and, therefore, this change does not involve a significant increase in the probability of an accident previously ( evaluated. The 24 hour time limit for the inspection is long enough to allow completion of the work and still short enough to ensure that SGT subsystem unavailability is minimized. Therefore, the 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 introduce a new mode of plant operation and does not involve physical modifications to the plant. Therefore, it does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the change involve a significant reduction in a margin of safety?

The proposed change minimizes the time that both Unit 1 SGT subsystems are inoperable for hardened vent rupture disc inspection. The change will help to avoid a dual unit shutdown to perform this inspection and still alluw only a small window when the Unit 1 SGT subsystems will not be OPERABLE. The probability of an event occurring during the 24 hour period that wuuld require the use of the Unit 1 SGT subsystems is small. Therefore the change does not involve a significant reduction in a margin of safety. O b HATCH UNIT 2 3 REVISION h('

NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM l L.4 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, G'eorgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

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

The required number of OPERABLE SGT subsystems is proposed to be dependent on the configuration of the secondary containment. Based on the specific operational status of each of the two units, certain secondary containment 4 zones may be isolated from the secondary containment boundary; thereby reducing the volume of the secondary containment. With a reduced secondary containment volume, a reduced number of SGT subsystems would be required to drawdown to and maintain that volume at 0.25 inches vacuum in the required time. Since the SGT System is not assumed to be an initiator of any previously analyzed accident, the change does not significantly ' increase the probability of such accidents. The change will not increase the consequences of an accident previously analyzed since sufficient SGT subsystems remain OPERABLE to mitigate the previously evaluated accidents accounting for a single active failure. Additionally, the proposed change relocates the details of required number of SGT subsystems for various secondary containment configuration from the Technical Specifications to the Bases and Technical Requirements Manual (TRM). The Bases, and TRM containing the relocated information are , subject to the change control provisions in the Administrative Controls section of Technical Specifications, and will be maintained in accordance with 10 CFR 50.59. Since any changes to the Bases or TRM will be . evaluated per the requirements of 10 CFR 50.59, no increase (significant ' or insignificant) in the probability or consequences of an accident previously evaluated will be allowed. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. j i I O l HATCH UNIT 2 4 REVISIONh/-

NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM l L,4 CHANGE (continued)

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

The proposed change does involve new configurations of available SGT subsystems on various secondary containment boundaries; however, the secondary containment function to contain fission products released during accidents, and the necessary support provided by the SGT System to assure fission products released to the secondary containment are filtered prior to release, remains unaffected. The system will continue to function in the same way as before the change. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

With the minimum SGT subsystems OPERABLE, surricient SGT subsystems are available to mitigate the consequences of an accident and account for a single active failure of one of the subsystems. Therefore, since p sufficient SGT .;ubsystems are still required to be OPERABLE to meet the Q analysis assumptions, the change will not result in a significant reduction in a margin of safety. Since any future changes to these requirements in the Bases or TRM will be evaluated per the requirements of 10 CFR 50.59, no reduction (significant or insignificant) in a margin of safety will be allowed. l l l I l l l O HATCH UNIT 2 5 REVISION D

1 i NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM l 1 L.5 CHANGE i In accordance with the criteria set forth in 10 CFR 50.92, G'eorgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following: I. Dc?s the change involve a significant increase in the probability or co.> sequences of an accident previously evaluated? The change requires the SGT System to be OPERABLE in MODE 5 only during CORE ALTERATIONS and movement of irradiated fuel. The SGT System is not assumed to be an initiator of any analyzed accident. Therefore, the change does not significantly increase the probability of an accident

  ;        previously evaluated.           The consequences of any analyzed accident remain unchanged since SGT System is still required during all conditions analyzed in the safety analysis.               Therefore, the change does not significantly increase 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 any design changes, plant O modifications, or changes in plant operation. The system will continue to function in the same way as before the change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 1

3. Does this change involve a significant reduction in a margin of safety?

l The margin of safety is unchanged since the SGT System is still required ) l OPERABLE during CORE ALTERATIONS and movement of irradiated fuel l

 !         assemblies. Since these are the only instances when a postulated fission                               '

I product release is assumed to occur (with the exception of a vessel drain i down event, which is covered by another specification), this change does not involve a significant reduction in a margin of safety. ] l HATCH UNIT 2 6 REVISION h (-

i NO SIGNIFICANT HAZARDS DETERMINATION . ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM l L.6 CHANGE in accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

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

An alternative is proposed to suspending operations if a standby gas treatment subsystem cannot be returned to operable status that would allow continued movement of irradiated fuel assemblies, core alterations, or operations with the potential for draining the reactor vessel. The al ternative is to initiate two operable subsystems of Standby Gas Treatment (SGT) System and continue to conduct the operations. Operation of the SGT System is not considered as an initiator of a previously analyzed accident. Therefore, the operation does not significant'> increase the probability of an accident previously identified. Since two subsystems are sufficient to mitigate the consequences of previously evaluated accidents, the consequences of any previously evaluated accidents are not significantly increased. 2. O, Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? This change provides for continued performance of previously evaluated operations. Since these operations have been previously considered, their continued performance does not create the possibility of a new or different kind of .ccident from any previously analyzed accident.

3. Doe; this change involve a significant reduction in a margin of safety?

The margin of safety considered in performance of these operations is maintaired by starting and running the system that would be required to initiate should an accident occur. Operation of the system significantly reduces the risk that the system may not perform its intended function initiate when required. Therefore, the change does not involve a significant reduction in the margin of safety. I i HATCH UNIT 2 7 REVISIONh(,

NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM l L.7 CHANGE I In accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company has evaluated this proposed Technical Specifications change and determined it does not involve a significant hazards consideration based on the following:

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

This change would allow additional time to restore an inoperable Unit 1 SGT subsystem when three remaining subsystems are OPEPABLE. The SGT System is not considered an initiator of any previously analyzed accident. Therefore, this change does not significantly increase the probability of such accidents. The proposed change would allow additional temporary operation with less than the required SGT subsystems, however, since the change is in the allowed outage time, the consequences of an event that may occur during the extended outage time would not be any different than during the currently allowed outage time. Therefore, this change does not significantly increase the consequences of any previously analyzed accident.

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

This change does not result in any changes to the equipment design or capabilities, or plant operation. Therefore, this change does not create the possibility of a new or different kind of accident from any previously analyzed accident.

3. Does this change involve a significant reduction in a margin of safety?

The change increases the allowed outage time for an inoperable Unit 1 SGT subsystem when three remaining subsystems are OPERABLE, and only then if the reactor building-to-refueling floor plug is removed. The margin of i safety considered in determining the allowed outage time is based on i engineering judgement of the probability of occurrence of an event requiring the unavailable capabilities and the consequences of such an event. A Completion Time of 30 days is justified based on the capabilities of the remaining SGT subsystems in the required secondary containment configuration. In this condition specific plant testing has j demonstrated the ability of the SGT System to drawdown the secondary ' containment to 1 0.25 inches vacuum with any two of the three remaining subsystems. Therefore, the proposed ACTION allowing 30 days to restore an inoperable SGT subsystem still provides a high level of confidence remains that even with an additional single of any SGT subsystem (which is not necessary to assume while in ACTIONS) the required drawdown function could still be performed. Therefore the change is acceptable based on the minimal impact to the margin of safety and allows appropriate actions to O l HATCH UNIT 2 8 REVISION'D { l

NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.4.3 - STANDBY GAS TREATMENT SYSTEM l [~/T

s. ,

L7 CHANGE (Continued) be taken without undue haste and potentially prevents a shutdown. Therefore, the change does not involve a significant reduction in the margin of safety. g HATCH UNIT 2 9 REVISION d

NO SIGNIFICANT HAZARDS DETERMINATION CTS: SECTION 3/4.6.1.4 - MSIV LEAKAGE CONTROL SYSTEM L.1 CHANGE This specification is being deleted. The No Significant ffazards Determination for this change is provided in GPC letter dated January 6.1994, and February 3, 1994. The NRC issued this change as Amendment 132 to-the Unit 2 TS by letter dated March 17, 1994. t , 1 1 l l l l HATCH UNIT 2 1 REVISION A

1 NUREG 1433 COMPARISON DOCUMENT - SPECIFICATIONS O O

1 SDM 3.1.1 / 3 ; 3.1 REACTIVITY CONTROL SYSTEMS ! 3.1.1 SHUTDOWN MARGIN (SDM) l l LCO 3.1.1 SDM shall be:

a. t,TO.38j%M/k,withthehighestworthcontrolrod 3 analytically determined; or y.\ )

                  ~f       b. t de/0.28]%M/k,withthehighestworthcontrolrod tennined by test.

APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limits A.1 Restore SDM to within 6 hours in MODE 1 or 2. limits. O C/ B. Required Action and 8.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met.

                                             .i.       _!     . 1. . . 21       [T, , , , -'     ~~',

C. SDM not within limits C.1 ully insert all Ffiour in MODE 3. insertable control rods, s-fy s ,e*' Ac EfMr R_' r) D. SDM not within limits D.1 E dully insert all 1-hour-in MODE 4. insertable control ,.,,,,a.,,, ,i rods. - __1 ANp (continued) Ov BWR/4 STS 3.1-1 Rev. O, 09/28/92

r i SDM 3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) 0.2 Initiate action to . I hour restore '{ secondary 3' containment to OPERABLE status. AND Initiate action to Q D.3 I hour restore C standby gas treatment (SGT) subsystem to OPERABLE status. AND

                         -- -     D.4          Initiate action to        I hour
 ,D a    . _ .6 A. 2)    7,c         u __g       res h isolatia l
                  *K , '                 [, 's  a ve     d 1 so 'at'ad               i s um tat on                         f A Q                              P BLE tat       in                 P\

gh ' 'y#g k e each',(3econdaryl containment cf /- penetration flow path not isolated. E. SDM not within limits E.1 Suspend CORE Imediately in MODE 5. ALTERATIONS except for control rod insertion and fuel assembly removal. AND E.2 Initiate action to Imediately fully insert all insertable control rods in core cells containing one or more fuel assemblies. AND (continued) l BWR/4 STS 3.1-2 Rev. O, 09/28/92 1 1

7 SDM 3.1.1

                                                                           ~
  ,,                                                                                               l (v)           ACTIONS CONDITION         REQUIRED ACTION                 COMPLETION TIME l

E. (continued) E.3 Initiate action to I hour restore,{ secondary} containment to OPERABLE status. AND E. Initiate action to 1 hour restore.,3eSGT fl subsyste i

                                  ~.         status. gjto OPERABLE AND "Y'on Co pa blg
                   !                  E.5    Initiat t action to
           /\
         /D'x                                restore _ yon i::12t4 @

1 uN x. --g

                                          /~-vc k cr.d-:::: izted        7        _'~
                                                                                       ' . p 1-/

inn w... cat tien to b iOPEra"LE :t tu' n - k~~ es. eachA[ secondary}l~ c-,u 4

cont &1nment ('" ) 4\ penetration flow path not isolated. ) (v) BWR/4 STS 3.1-3 Rev. O, 09/28/92

                                                                                         .3 SDM 3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE                           FREQUENCY SR  3.1.1.1    Verify SDM is:                              Prior to each
                                   ..                             in vessel fuel           i

a. m 10.38J% M/k with the highest worth movement during control rod analytically determined; fuel loading :

or sequence l

b. m/0.287%M/kwiththehighestworth AND control rod determined by test.

Once within 4 hours after criticality following fu.el movement 6or control rod i replacement l TitMif OT's/\ g b .3 reactor \ pressurevessel] O l l i I O BWR/4 STS 3.1-4 Rev. O, 09/28/92

Primary Containment Isolation Instrumentation

                                   ,                                                                               3.3.6.1 h il CLup eyed.) Od:s wel(p, w,.h,)

, O Tabte 3.3.6.1 1 (page 6 of 6) i ( Primary Contairunent Isolation Instrumentation APPLICABLE CON 01TIONS MODES OR REQUIRED REFERENCED , OTHER CHANNELS FROM

                                                $PECIFIED     PER TRIP      REQUIRED       SURVEILLANCE        ALLOWABLE FUNCTION               CONDITIONS       SYSTEM     ACTION C.1      REQUIREMENTS          VALUE

6. Shutdown Cooting system Isolation f- a. Reector Steam Dome 1,2,3 f(1F F SR 3.3.6.1.1 5}145[psig

[ Pressure = Nign $R 3.3.6.1.2

                                                                                          ;; s:s:;;;:vo
         &                                                                                SR 3.3.6.1.4 4 -

b. Reactor vessel Water 3,4,5 -tf21) 3 st 3.3.6.1.1 1 Hel inches Levet - Low, Level 3 SR 3.3.6.1.2

, ffst i:i::i:fidD 3.3.6.i.rq (I) Only one trip system rew ired in MODES 4 and 5 when RHR Shutdown Cooling system integrity maintained. d O l BWR/4 STS 3.3-61 Rev. O, 09/28/92 <

l

Secondary Containment Isolation Instrumentation 3.3.6.2 3.3 INSTRUMENTATION 3.3.6.2 Secondary Containment Isolation Instrumentation LC0 3.3.6.2 The secondary containment isolation instrumentation for each Function in Table 3.3.6.2-1 shall be OPERABLE. APPLICABILITY: According tc Table 3.3.6.2-1. ACTIONS

                 -------.-----------------------------NOTE-------------------------------------

Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME l A. One or more channels A.1 Place channel in 12 hours for inoperable. trip. Function 2 AND 24 hours for Functions other than Function 2 B. One or more automatic B.1 Restore trew.idaiy- 1 hour Functions with ents mnen t- i sol ation ' !j [ U r'.- _.secondeiy w ioin cat-isolation capability capability. not maintained. I l C. Required Action and C.1.1 Isolate the v 1 hour associated Completion associated >{zondsW.' Time of Condition A or B not met. OR {

                                                      ~~~       'p* h b"              sO f4,,,y pa Y(C d    -'

(continued) O BWR/4 STS 3.3-62 Rev. O, 09/28/92 __ ___m_____ _ _ _ _ _

Secondary Containment Isolation InstrumerM tion S.3.6.2

                                                                     & sffYthH hU? cI p'

unN** bf'leTab3.3.6.21 sq (page 1 of 1) j

( pjf(f 5 C4r ',.- secondary Contatrynent Isolation Instrufentation APPLICABLE MODES OR REQUIRED OTHER CRANNELS

                                                           $PECIFIED                  PER                $URVEILLANCE               ALLOWABLE FUNCTION                 CONDITIONS            TRIP SYSTEM              REQUIREMENT $                VALUE

1. Reactor vessel Water 1,2,3, K st st 3.3.6.2.1 1/ 4g inches Levet - Low Low, Level 2 g(s)W m SR 3.3.6.2.2 n SR 3.3.6.2.8 W [

p :.3.;.:.x

2. Drywell Pressure - High 1,2,3 A2P sa 3.3.6.2.1 s J1.921*psig sR 3.3.6.2.2 4 3 n i:i:i:i:i W D

                                                                                           ,         st H4+m                   #          7,

3. Reactor Building Exhaust 1,2,3, p21P sa 33621 K6~ 0 pk R adi a t i on - Hi gh M( a),h an a.a.6.^.

st 3.3.6.2 7 st 3.3.6.2.F . 3dM , f.l/ -

4. Refueling Floor Exhaust 1,2,3, k(21P sR 3.3.6.2.1(p.2B s -420% nWt/hr

                                                                ),(b } }p                                  3.3.i.2.2 I                                /

Radi a t i on - H i gh 3;; s

         '                                              ')                                                               3                         /N
             /'                                                                           N
                                                                                                 -+{SR SR 3.3.6.2.8 3.3.6.2.*-     -

fw 9 L 3 "* i y =- -

                                                                    - - - . - -                                                                 .=-,

'v (7

        , g fW""                               ~.2 7;,2;,3g,7 - J i a'r s'aua) :_.SR'3;3.6.2.g NA
                                                                                                                                                      )

aq g I ca) grin, CORE Aumfig~ rat, ee wit a tentiai <or erainin, tse reacto, e .et.

                <b) jurin .ov ent o, irrsoiatee < i a...ne, in% cone.rvecontai, ,nt.

6 1 1 / k l m./ < BWR/4 STS 3.3-65 Rev. O, 09/28/92

i LLS Instrumentation 3.3.6.3 3.3 INSTRUMENTATION 3.3.6.3 Low-Low Set (LLS) Instrumentation j i LC0 3.3.6.3 The LLS valve instrumentation for each Function in l Table 3.3.6.3-1 shall be OPERABLE. I i l l APPLICAB'. _ITY : MODES 1, 2, and 3. ACTIONS 1 CONDITION REQUIRED ACTION COMPLETION TIME l A. One LLS valve u dA A.1 Restore channel':) to 24 hours CPEra0LL stetus__ g (jnoperdleduet-

i operable cher el(r)

  .\Q.       uv eia tion. capa s,ory3                ,z [$ val /i~M Heaf,or3J no t n,,,n un g,                         Qa?G6','9*/
             ~                             A

8. One or more safety / B.1 --------NOTE --------

relief valves (S/RVs) LCO 3.0.4 is not with one Function 3 applicable, channel inoperable. --------------------- Restore tailpipe Prior to pressure switches to entering MODE 2 OPERABLE status. or 3 from MODE 4 C. ---------NOTE--------- C.1 Restore one tailpipe (14fdays Separate Condition pressure switch to o entry is allowed for OPERABLE status. /' each S/RV. One or more S/RVs with two Function 3 channels inoperable. , (continued) O BWR/4 STS 3.3-66 Rev. O, 09/28/92 s

i ,

      > 'hff g y#t @@ o'wF, 1'

IMAGE EVAL.UATION SO c/'A<c,(4,/g

                                                                                        /
/o// l  th k                        TEST TARGET (MT-3) g/j/ p)'W                                                                     kj             )lff/$$
    +                                                                                     4 1.0      E "4 m if(llLM t .:: =

1.1 L " IWno l.8

===

1.25 1.4 l 1.6 i__ 150mm

6" >

4 _ _ .__ _._ _. . - - - -

  >>          %                                                                         + //p ef      y ,y 1                                      _
                                                                                               /h 4,,                                                                       -
                                                                                         .q ,

x 4

      %:fv,    '

4""p IMAGE EVALUATION 1 e

                                                                                                                 /;q,&
   \/ 'c

N)//7 'h' TEST TARGET (MT-3)

                                                                                                           /
                                                                                                               /l.,D[ 4g 9j                                                                         I////&              k!

l.0 l" 9 p$12,1 ll l _l. 8 l.25 1.4 1.6

                                                                    ==                   mm 4      ._  . - _ _ ..
                                           ._ _.- --- 1 5 0 m m                                              >

4 _ _ . . . _ . _ _ _ . - 6" - - - - - l O l . o[ 2Pg [4b /% 4

 *               >                                                                                ,. 4:,2         4l%

f p,,,,,} o Ky 0 4,4 j ,

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      >&        s'Qp
     *: h t 1 lp i.
                                                                                            ,t h, l' < '.' 3-$h' l MAGE EVAL.UATION TEST TARGET (MT-3)

((' / ffff* (g, [/o//f y, 7, l

                                                          '"         ?"

1.0 " p=.2.

                                                                    ~
                                                                      ?

2 1.25 lA l.6

==

4 - . _ . . _ . - - 150mm > w----- '---------6" -- l s - es x 0l g 1 I

>77,,,8p 3

49}//h

                                                                                %s  , '

T [?' o

     % , ,+
        +
                '4     @4
                   "'V                                IMAGE EVALUATION                     /,/j/g/j e.

J,//o/ ;. [ TEST TARGET (MT-3) 0 ef 4g,/gp

        #                                                                                             v I.0      ::^ 1" m m m p- (==

g gs l,l I " hN 11 1.25 'lA 1.6

==

150mm > 4-- -- - 6" > 4..._.._ _ . . _ . _ . _ . _ _ . - - - - - - - A A r>r4777 4++MsAsp g /ygNg o y,;

    , ; ; , y, //

o

                                                                                                  <g+y,  v af           ,
                                                                                         ..l

r INSERT A for orono ad TS 3.3.7.1 . c (> Two channels of the Control Room Air Inlet Radiation-High Function shall be ;

           . OPERABLE.

INSERT B for orocosed TS 3.3.7.1 MODES 1, 2, and 3, During novement of irradiated fuel as:a"nblies in the secondary containment, During CORE ALTERATIONS, During' operations with a potential for 61ning the reactor vessel (0PDRVs). P l [ O I t f l l l l 1 1 O

[MCREC] System Instrumentation \ 3.3.7.1 '

   . [$
                          \
         \                   -
                                                                                                                                                         /

Table 3.3.7.1 1 (pege 1 of 1) / kj IMain Control Room Envirorsnental control) system Instrunentation

                                                                                                                                                     /
                                    \                                                                                                           .
                                         \           APPLICA8LE                         CowDITIONS
                                                                                                                                     )#

3 MaDES OR REQUIRED REFERENCED , OTHER CHANNELS FROM

                                               \ $PECIFIED PER TRIP           REQUIRED              SURVEILLANCE '               ALLOWABLE FUNCTION               'C,0NDITIONS          STsTEM          ACTION A.1             REQUIREMENT 9'                 VALUE
                                                                                                                      ,,/                                --

1. Reactor Vossel Water 1,2,3;tal I2] B st 3'. 3. 7.1.1 : I-113) inches Level - Low Low Low, $st ' 3.3.7.1.2 Level 1 's (sR 3.3.7.1.3)

N / SR 3.3.7.1.4

                                                                                                      /     st 3.3.7.1.5 N                          j

2. Drywell Pressure - High 1,2,3 [2] 8/

st 3.3.7.1.1 5 II.921 psig l tR 3. 3. 7.1. 2 l -

                                                                                  \    f
                                                                                         /-                   A 3.3.7.1.33 SR   3.3.7.1.4 i

j' SR 3.3.7.1.5

'N

3. Main steem tine 1,2,3 [2.pe/ r 8 SR 3.3. 7.1.1 (1381% rated '

F low - High MSL3 s st 3.3.7.1.2 stems flow -

                                                                   ,/
                                                                      /                           s
                                                                                                     \'

[SR 3.3.7.1.31 SR 3.3.7.1.4 j SR 3.3.7.1.5 .

/.

4 Refueling Floor Ares 1,2,[ (13 C st 3.3.7.1.1 s (201 mat /hr Rediation - HIsh SR, 3.3.7.1.2 [( ),(b)] st 3.3.7.1.4 q _ SR 3 >3,.7.1.5 ,_.,

                               ~
                     'antrol'ao7om1r Inlet           e ,2,3,             '{ 1 }*             #A               sa   3.3.7.1.1         s til mR/hr Radiation - Hi gh (a),(b) st   3.3.7.1.2                             MS gg   3,3,7,3 g g                         4o g g
                 ,, o                   o                           % ed4o %                                  st 3.3.7.1 N                          3,3, , , f . 3 hk L(A s.-l'                            % k L , iek %Le.A f', \                                                                             \
  '-       ba) During CORE ALTERATIONS K operations with a potential for draining the reactor vessel.l i                                         i l     (b) Duri      movenent of irradiated fuel assemblies in the isecondaryt contairunent.

g o 16 3 g 1
tL n

                                                                                                                                .au Aes rad g y;aA                                                          g(

ss) i 4 BWR/4 STS 3.3-73 Rev. O, 09/28/92 l

l

                                                                                                                                          . LOP Instrumentation 3.3.8.1 3.3       INSTRUMENTATION 3.3.8.1           Loss of Power (LOP) Instrumentation LCO      3.3.8.1                 The LOP instrumentation for each Function in Table 3.3.8.1-1 shall be OPERABLE.

V (ps),}-]) c APPLICABILITY: MODES 1, 2, and 3, When the associated diesel generatorlis required to be OPERABLE by LCO 3.8.2, "AC Sources-Shutdown." ACTIONS

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

Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME

                                                                          ' (?c4gp tk w ee\ %

A. One or more channels A.1 Ph ee-chcnnci ni I hour

~              inoperable?' +]r                                                1r.iP - OfrA /4 6 V
                                                                   'd                        Sh (0) Ittoc4kri 16 e d 2.

,g Required Action and . Declareasdc ted Imediately

p associated Completion Time not met.

{ -dicni-ydantor (DGP inoperable. A p' y ' LC\ s.) 7

    .B . Om ci                  w ic. o n4 a .

d in p & ('c c v,1 w nc voilcy on 62 p.o c o c a r a 4. U d ku c b r V, 3. Em is > M2 5 i P.7o lj O

BWR/4 STS 3.3-74 Rev. O, 09/28/92

I ECCS-Shutdown ! 3.5.2 3.5

's)

EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATIO l COOLING (RCIC) SYSTEM 3.5.2 ECCS-Shutdown LCO 3.5.2 Two low pressure ECCS injection / spray subsystems shall be OPERABLE. APPLICABILITY: MODE 4, b MODE 5, except with the spent fuel storage pool gates removed and water level a '123=gf over the top of the reactor pressure vessel flange. ( . _. ACTIONS l M -t 'j/3 i e r h ol] CONDITION REQUIRED ACTION COMPLETION TIME A. One requireu ECCS A.1 Restore required ECCS 4 hours injection / spray injection / spray subsystem inoperable. subsystem to OPERABLE status. B. Required Action and associated Completion B.1 Initiate action to Imediately suspend operations Time of Condition A with a potential for not met. drainin the reactor vessel OPORVs). C. Two required ECCS injection / spray C.1 Initiate action to Immediately suspend OPDRVs. subsystems inoperable. AND C.2 Restore one ECCS 4 hours injection / spray subsystem to OPERABLE status. (continued) O V BWR/4 STS 3.5-7 Rev. O, 09/28/92

i ECCS-Shutdown 3.5.2 ACTIONS (continued)

                                                                                 -A                        e CONDITION                      REQUIRED ACTION             COMPLETION TIME D.               Required Action C.2          D.1       Initiate action to        Inynediately and associated                         restore {secondaryl Completion Time not                    containment to met.                                  OPERABLE status, h                 /

f I" - D.2 Immediately Ynitiate Alction to f restoreconejstandby s y

                                                /  \       gas treatment                               J x Q,)ly s
                                  /f     ,)                subsystemA to OPERABLE status.(ps)

/ AND I D.3 Initiate action to Immediately restore e r i =lation_ valve and ass wieted m'W26 4 instrumentatfon=to " ?f"'#N-bk'I / 2P.ERABf1 ste: 4 in / each;[ secondary] G-m,ceJ' contaThinent L A > penetration flow path not isolated. I

                                                                         - (,

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify, for each required low pressure 12 hours coolant injection (LPCI) subsystem, the suppression pool water level is 2 ,[ p L-2 inches)'.

4

                                                       't g

BWR/4 STS 3.5-8 Rev. O, 09/28/92

     ' YNIf s) o>k i                                                     CAD System y,q                                                                3.6.3.A e                                                                                        -g (s)                                                                                      ty SURVEILLANCE REQUIREMENTS v

SURVEILLANCE m FREQUEN V

                                        ~ /c h                                    -
                ;1) py)            / 1 3 9 0 2 )h p SR 3.6.3.\.1       Verify a f4350). gal of liquid nitrogen    31 days are contained in th: CAO Syster.
                                         &c4 All dzo3e f%k,      Pk9 SR 3.6.3 4.2       Verify each CAD subsystem manual, power    31 days A operated, and automatic valve in the flow l   path that is not locked, sealed, or P' ', ,#

otherwise secured in position is in the correct position or can be aligned to the correct position. O i e G BWR/4 STS 3.6-45 Rev. O, 09/28/92

(Secondary). Containment I j 3.6.4.1 j 3.6 CONTAINMENT SYSTEMS 3.6.4.1 { Secondary}; Containment i k. LCO 3.6.4.1 The { secondary} containment shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, Durina movement of irradiated fuel assemblies in the flB isecondary} containment, 1)uring CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs). ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (Secondary}" - 5.[ Restore (secondary)' 4 hours containment inoperable containment to in MODE 1, 2, or 3. OPERABLE status. O B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met. B.2 Be in MODE 4. 36 hours

C. { Secondary}. C.1 --------NOTE--------- containment inoperable ' LCO 3.0.3 is not during movement of ; applicable. irradiated fuel \ --------------------- assemblies in the { secondary}s L Suspend movement of Immediately p,G'

    . containment, during                    -

irradiated fuel CORE ALTERATIONS, or assemblies in the during OPDRVs. { secondary} containment. AND (continued) O BWR/4 STS 3.f-46 Rev. O, 09/28/92

s (Secondaryl Containment [jf ' - 3.6.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 Suspend CORE Immediately ALTERATIONS. AND C.3 Initiate action to Immediately suspend OPDRVs. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.i verify wcenda u 24 hours p?)> S.25HTicfo~s] f vacuum C00tainment' wateF gaugew- vacuuiiiis O SR 3.6. 4.1. / '. Verify all (secondary]' containment 31 days 7.b "p; equipment hatches are closed and sealed. d SR 3.6.4.1.3 7 Verify each efsecondaryJ containment 31 days v p access door is closed, except when the access opening is being used for entry and exit {, then at least one door shall

^h

(- be closed}'.

                                              $                                                           l r_                      A Veri fy pwa   standh
                                                     .2 as--treatment gn 418}:n iiio)ths on_

SR 3.6.4.1.49 a STAGGERED s ( . h >{(SGT) subshster(will secondary containment to draw down TESTthe BASIS X

                                 > $0.25) inch of vacuum water gauge                                      1 h

P hin s [120} seconds. (continued) , l O jggg y]f29 ' BWR/4 STS 3.6-47 Rev. O, 09/28/92

TSecondary}' Containment 3.6.4.1 c, SURVEILLANCE REQUIREMENTS (continued)

     --          ,s O39              SURVEILLANCE                                FREQUENCY
                                                -n
    /Q5ERT 48)-

hi"D Uj) is; w .$.6.4.1.5 Verify eace SGT subsyste n maintain f18}: months on f a f0.25}, inch of vacuum water gauge in a STAGGERED [n the isecondaryl containment for 1 hour at a flow rate 5 f4000}: cfm._ TEST BASIS (br eMA Suzw& ) f5^) O O BWR/4 STS 3.6-48 Rev. O, 09/28/92

                                                                                                             !l 1

[ INSERT 47 to SR 3.6.4.1.3 1 4 ___________--------------NOTE--------------------------------- The number of standby gas treatment (SGT) subsystem (s) ; required for this Surveillance is dependent on the secondary containment configuration, and shall be one less than the ; number required to n eet LCO 3.6.4.3, " Standby Gas Treatment (SGT) System," for the given configuration. INSERT 48 to SR 3.6.4.1.4 ______-------------------NOTE--------------------------------- ; The number of SGT subsystems required for this Surveillance is dependent on the secondary containment configuration, and shall be one less than the number required to meet LCO ; 3.6.4.3, " Standby Gas Treatment (SGT) System," for the given configuration. s i e 1 i e v i s

1 i SCIVs f-G 3.6.4.2 1 [- 1 3.6 CONTAINMENT SYSTEMS I % l 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) ) LC0 3.6.4.2 Each SCIV shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the Tsecondary}; containment, During 20RL' ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs). ACTIONS

  -----___-----------------------------NOTES------------------ -----------------

1. Penetration flow paths may be unisolated intennittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

D 3. Enter applicable Conditions and Required Actions for systems made 'd inoperable by SCIVs. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected 8 hours penetration flow paths penetration flow path with one SCIV by use of at least inoperable, one closed and In de-activated ID automatic valve, closed manual valve, or blind flange. AND (continued) O ; BWR/4 STS 3.6-49 Rev. O, 09/28/92 i l

                                                                       '              SCIVs i

3.6.4.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. 1 A.2 --------NOTE--------- ft (continued) j f -Yalves and bl4nd-ftanger in high g 715 . radiation areas may ('dem:0 be verified by use of administrative means. Verify the affected Once per 31 days l penetration flow path l is isolated.

8. ------ NOTE--------- B.1 Isolate the affected 4 hours OnJyahicableto penetration flow path

       -penetration flopaths                   by use of at least with'two isoittion                     one closed and (v     valves /

de-activated automatic valve, closed manual valve, g One or more or blind flange. penetration flow paths with two SCIVs inoperable. C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND or B not met in MODE 1, 2, or 3. C.2 Re in MODE 4. 36 hours (continued) O BWR/4 STS 3.6-50 Rev. O, 09/28/92 l 1

SCIVs ) 1[ 3.6.4.2 i f~% (continued) -Q ACTIONS COMPLETION TIME CONDITION REQUIRED ACTION D. Required Action and D.1 --------NOTE--------- associated Completion LCO 3.0.3 is not Time of Condition A applicable. or B not met during --------------------- movement of irradiated fuel assemblies in the Suspend movement of Immediately

         -fsecondaryP                  irradiated fuel
   .f     containment, during          assemblies in the CORE ALTERATIONS, or        (secondary}(

during OPDRVs. containment. AND D.2 Suspend CORE Immediately ALTERATIONS. AND D.3 Initiate action to Imediately suspend OPDRVs. O BWR/4 STS 3.6-51 Rev. 0, 09/28/92

SCIVs

                                                            ' (                                                            3.6.4.2 SURVEILLANCE REQUIREMENTS SilRVEILLANCE                                                                     FREQUENCY SR  3.6.4.2.1     ------------------NOTES------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for SCIVs that are open under administrative controls.

Verify each secondary containment 31 days isolation manual valve and blind flange that is required to be closed during accident conditions is closed. 1 1 l SR 3.6.4.2.2 Verify the isolation time of each power Ir, eucuidance J operated and each automatic SCIV is with the within limits. 4nten ice ys f,4 g 1 82 nregra a W 92 days SR 3.6.4.2.3 Verify each automatic SCIV actuates to E181 months the isolation position on an actual or simulated actuation signal. 9 BWR/4 STS 3.6-52 Rev. O, 09/28/92

SGT System I l[ 3.6.4.3 l 3.6 CONTAINMENT SYSTEMS l 3.6.4.3 Standby Gas Treatment (SGT) System hq j a, a UnfEfft Y h LCO 3.6.4.3 M SGT subsystems 4shall be OPERABLE. 3 frgwnd & Luffert kl0 344.\, Q &c w Jar 3 CmMn @ APPLICABILITY: MODES 1, 2, and 3, _puring movement of irradiated fuel assemblies in the y) fsecondary). containment,

                            'During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs).

ACTIONS / SErcT 5 %) umT 2. oM d CONDITION REQUIRED ACTION COMPLETION TIME ga pe ta i C l J. Required Action and Ef.1 Be in MODE 3. 12 hours c associated Completio Time of Condition A AND kM not met in MODE 1, 2, or 3. E.2 Be in MODE 4. 36 hours k -l /. Required Action and associated Completi

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

LCO 3.0.3 is not applicable,

  \p           Time of Condition not met during
                                                                   &~as0 6 n i

movement of irradiated fuel assemblies in the f.1 Place [0PERABL SGT subsystemp in Immediately o I isecondary) containment, during operation. @ CORE ALTERATIONS, or OR during 0PORVs. (continued) O BWR/4 STS 3.6-53 Rev. O, 09/28/92

SGT System 3.6.4.3 7 R ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME M (continued) g.2.1 Suspend movement of Imediately f g irradiated fuel assemblies 1 isecondary containment. AND p.2.2 Suspend CORE Imediate'v 0 ALTERATIONS. AND

0 f.2.3 Initiate action to Imediately suspend OPDRVs. c v> ]Two SGT,oc T

                 .               subsystems                 .1     --------NOTE---------

(g inoperable during (p LCO 3.0.3 is not movement of irradiated , applicable. I uel assemb ies in the / --------------------- l secondary 6C containment, during p/ Suspend movement of Imediately CORE ALTERATIONS, irradiated fuel during OPDRVs. '

                                               -                   assemblies in isecondary)fg
   }

containment. -- _s_ AND

                    .j                                     .2      Suspend CORE                Imediately w             ^

ALTERATIONS. e AND

                                                          !3       Initiate action to          Imediately suspend OPDRVs.

I _. . [, [tv 0 ch (Y)ofC F Y kJ S,l E r bt e l_ ( 9 3 ,'.1, 3 Tw

cc wys w~~I/1bebl/

( tt<f+iaz on Mo&E .

        \           1,   2 x 3.
                                                       ,w_-

p. . _ _ - -

Q BWR/4 STS 3.6-54 Rev. 0, 09/28/92

1 l l ( INSERT 53 a to LCO 3.6.4.3 ACTIONS { UNIT 2 ONLY} s

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

When two Unit 1 SGT subsystems are placed in an inoperable status solely for inspection of the Unit 1 hardened vent rupture disk, entry into associated Conditions and Required Actions may be delayed for up to 24 hours, provided both Unit 2 SGT subsystems are OPERABLE. INSERT 53 b to LCO 3.6.4.3 ACTIONS ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required Unit A.1 Restore required 30 days from 1 SGT subsystem Unit 1 SGT discovery of inoperable while: subsystem to failure to fove OPERABLE status. meet the LCO

1. 4 SGT subsystems required OPERABLE, and

2. Unit 1 reactor building-to-refuel floor plug not installed. l

i l l

1 l

    / khMT 63 U            M '74f4 (/

() ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. One required Unit B.1 Restore required 7 days 2 SGT subsystem SGT subsystem to inoperable. OPERABLE status. AND QE 30 days from discovery of One required Unit failure to 1 SGT subsystem meet the LCO inoperable for reasons other than Condition A. O

SGT System l{ 3.6.4.3 [9 U/ SURVEILLANCE REQUIREMENTS SURVEILLANCE [p. FREQUENCY ch5;:D ,

SR 3.6.4.3.1 Operate eachjdui sutifystem for a Of 31 days continuous hours with heaters f\

                                                              )d operatingll                       L[/                          \

SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP). fnga<eca)f,3D) SR 3.6.4.3.3 Verify eachASGT subsystem actuates on an f18}' months actual or simulated initiation signal. SR 3.6.4.3.4 Verify each SGT filtercooler bypasF -[18Lmonths Jmpee-canderofeTiid and the fan started. '-~ ,

          ,e .,,)

BWR/4 STS 3.6-55 Rev. O, 09/28/92

[MCRECSystem 3.7.4 j f i

3.7 PLANT SYSTEMS t 3.7.4 Main Control Room Environmental Control (MCREC) System l LCO 3.7.4 Two CREC subsystems shall be OPERABLE.

1 APPLICABILITY: MODES 1, 2, and 3, j[f 2') Durin movement of irradiated fuel assemblies in the A secondary containment, During CORE ALTERATIONS, Q6 During operations with a potential for draining the reactor vessel (0PDRVs). 1 ACTIONS _ CONDITION REQUIRED ACTION COMPLETION TIME A. One[CREC subsystem A.1 Restore CRE 7 days inoperable. subsystem to OPERABLE

  - ,i                                                status.

1 G' B. Required Action and 8.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours (continued) U BWR/4 STS 3.7-9 Rev. O, 09/28/92

[MCREC[ System 3.7.4 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and ------------NOTE------------- associated Completion LC0 3.0.3 is not applicable. Time of Condition A ----------------------------- not met during - movement of irradiated C.1 -------- TE------- fuel assemblies in the ace in t ic gas [ secondary @(f.Q p tection de if \ j containment, diiring CORE ALTERATIONS, or aut atic tra sfer ic gas

                                                                                 ' /. ]>I l                                         \ to         t protec ion mode 's during OPDRVs.

nopera e. Place OPERABLE Imediately i y'MCRECKsubsystem in y! pressurization}&

                                      ' '         mode.

l OE C.2.1 Suspend movement of Imediately &l irradiated fuel W 4 psemblies in the [containment. secondary}(I13 AND l C.2.2 Suspend CORE Imediately ALTERATIONS. AND C.2.3 Initiate action to Imediately suspend OPDRVs.

D. Two lN CREC[sub' systems D.1 Enter LCO 3.0.3. Imediately inoperable in MODE 1, 2, or 3. (continued) O BWR/4 STS 3.7-10 Rev. O, 09/28/92

MCREC System 3.7.4 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME _y E. Two CRE subsystems ------------NOTE------------- inoperable during LC0 3.0.3 is not applicable. movement of irradiated fuel assembli the fiecondaryl E.1 Suspend movement of Immediately containment, during irradiated fuel CORE ALTERATIONS, or a semblies in the during OPDRVs. condar containment f,1 AND E.2 Suspend CORE Immediately ALTERATIONS. AND E.3 Initiate action to Immediately suspend OPDRVs. O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 OperateeachhCRECfsubsystemfor-[1 10 31 days

                         ==te==:_.              w.a..:.=:= :;=tm
             ..m p \k/

er 'f:r :y;;;;. 2: 15 minute .

                                               .;i; ; h = t:r:}

V SR 3.7.4.2 Perform required hCRECf filter testing in In accordance accordance with the dVentilation Filter with the VFT Testing Program (VFTP)]& (continued) O BWR/4 STS 3.7-11 Rev. O, 09/28/92

CRECfSystem i 3.7.4 l % l l SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY l SR 3.7.4.3 Verify each hCREChsubsystem actuates on [8fmonths l /p,7 an actual or simulated initiation signal. ' l 3.7.4.4 (Verify each hCRECf subsystem can maintain [8fmnhs ' e a positive pressure of m y(0.lfinches on a l ifb L qbt water gau e relative to the t{' turbine building during thefpressurizationfmode STAGGERED TEST BASIS of operation at af ow rate of s fl cfm> (5vbsyskm ( zwo j

                                            \        --.
                                                                      /

A f ard %,vts)h aC

nu eam

                                        $Ju 0 O

O BWR/4 STS 3.7-12 Rev. O, 09/28/92

                    .      .=            . _ _ - .              ..         .                             .

[ControlRoomAC[ System 3.7.5

 ;           3.7 PLANT SYSTEMS 3.7.5 Nontrol Room Air Conditioning (AC[ System                                                i LCO 3.7.5                 kontro[roomAC[            s stems shall be OPERABLE.

ev APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the

         .                            isecondary3' containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLF. TION TIME l A On [ control room A.1 Rest a [contr room 0d s a 'sub stem ino rable. AC] su stem to lNSEllT B PERABLE tatus. x f 0

g. Required Action and g.1 Be in MODE 3. 12 hours p associated Completion ki) u Time of Condition A not met in MODE 1, 2, p AND l

i or 3. g.2 Be in MODE 4. 36 hours j (continued) l l BWR/4 STS 3.7-13 Rev. 0, 09/28/92 l l ( -

ControlRoomACfSystem 3 3.7.5 _V ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME

         /f. Required Action and               ------------NOTE-------------

g E associated Completi LCO 3.0.3 is not a

 . [,-             Time of Condition Bd }-----------------pplicable.-----------

not met during ((Ed ) movement of irradiated E.1 Place OPERABLE Imediately fyel assemblies in the E Rcontrol com A s Tsecondaryl* subsyste n y containment, during operation. \ CORE ALTERATIONS, or i', during 0PORVs. OR 2'.2.1 Suspend movement of Imediately g irradiated fuel h assemblies in the Ts'econdaryM containment. AND f.2.2 Suspend CORE Imediately g ALTERATIONS. AND R.2.3 Initiate action to Imediately E suspend OPDRVs. 61b U* l'pd] Ntt t \ l /. fweMontrol room AC8 .1 Enter LCO 3.0.3. Imediately F subsystems inoperable r g,$ in MODE 1, 2, or 3. l l (Continued) O BWR/4 STS 3.7-14 Rev. O, 09/28/92 c

l MontrolRoomACbystem g 3.7.5 ACTIONS (continued) I REQUIRED ACTION COMPLETION TIME Q.hCONDITION i

TNru Q g. 4*e fcontrol c room AC [ ------------NOTE------------- gv ' subsystems inoperable LCO 3.0.3 is not applicable.

   \f A during movement of              -----------------------------

irradiated fuel G

         \ assemblies in thef.1yp2                C Suspend movement of                Immediately
           -{ secondary.}A                             irradiated fuel containment, during                        assemblies in the CORE ALTERATIONS, or                    (TsecondaryM during OPDRVs.                             containment.

AND g.2 Suspend CORE Immediately ALTERATIONS. AND

                                     ,    .3           Initiate actions to             Immediately suspend OPDRVs.

O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each hntrol room ACfsubsystem has h8[ months ha, the capability to remove the assumed heat load. O , BWR/4 STS 3.7-15 Rev. O, 09/28/92

Main Condenser Offgas 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Main Condenser Offgas LCO 3.7.6 The gross gama activity rate of the noble gases measured at the main condenser evacuation system pretreatment monitor (p'Q%statio$shall s

                                                           -2 0 - ,,-] .

be7 s)240} mci /second [ ofter d: cay of h p (v.t APPLICABILITY: MODE 1, g MODES 2 and 3 with any kain steam line not isolated ana[

                                                         \        steam jet air ejector (SJAE) in operation.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Gross gama activity A.1 Restore gross gama 72 hours rate of the noble activity rate of the gases not within noble gases to within limit. limit. B. Required Action and I B.1 Isolate all main 12 hours

associated Completion steam lines. Time not met.

                                                                   @n:5     08 B.2       Isolate SJAE.           12 hours 0,,8 B.3.1     Be in MODI 3.           12 hours AND B.3.2     Be in MODE 4.           36 hours l

O BWR/4 STS 3.7-16 Rev. O, 09/28/92

I i Main Condenser Offgas 1 3.7.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 -------------------NOTE-------------------- Not required to be perfonned until 31 days 3 after any dmain steam line not isolated

l. / an S SJAE in operation.

Verify the gross gama'a 6vity rate of the 31 days noblegasesissd240PmCi/second i:f t:r & :; cf 2 -i ~. 4 AND Once within 4 hours after a a 50% increase in the nominal steady state fission gas release after factoring out increases due to changes in O' V THERMAL POWER level 1 BWR/4 STS 3.7-17 Rev. O, 09/28/92 ; 1

Main Turbine Bypass System 3.7.7 3.7 PLANT SYSTEMS 3.7.7 The Main Turbine Bypass System LCO 3.7.7 The Main Turbine Bypass System shall be OPERABLE. OR y LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," limits for N (1 (h[ an inoperable Main Turbine Bypass System, as specified in the COLRK are made applicable. APPLICABILITY: THERMAL POWER a 25's RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. hequirementsofthe A.1 /Satisfythe 2 hours LC0 not met er-Ment requirements of the

                           , p, -Tu rb i .m On e;; Smcm                  LCO cr rette e "2i-
                          / inepa able. h                       /7'9 To.u.a Cypr: Syrter Q ' -te OPEPf,SLE ;;;te.h B.      Required Action and          8.1        Reduce THERMAL POWER                      4 hours associated Completion                   to < 25's RTP.

Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify one complete cycle of each main 31 days turb.ine bypass valve. (continued) O BWR/4 STS 3.7-18 Rev. O, 09/28/92 _ _ _ _ _ _ _ _ _ _ _ _ _ ._.___m____ __________- D

(,,) M #) V INSERT LC0 3.8.1

c. The swing DG; OtLL

d. One Unit 2 DG capable of supplying power to Unit 2 Standby Gas Treatment (SGT) subsystem required by LC0 3.6.4.3, "SGT System;" and

e. One qualified circuit between the offsite transmission network and the Unit 2 onsite Class 1E AC Electrical Power Distribution subsystem (s) needed to support the Unit 2 SGT subsystem required by LC0 3.6.4.3. gf (5) c.

O HATCH UNIT 1

T- [)

  'a INSERT  LCO 3.8.1      41 e4@)

c. The swing DG;

d. One Unit 1 DG; and

e. One qualified circuit between the offsite transmission network and the Unit 1 onsite Class 1E AC Electrical Power Distribution subsystem (s) needed to support the O3 Unit 1 equipment required to be OPERABLE by LC0 3.6.4.Y, 'k

                          " Standby Gas Treatment (SGT) SystemtW4uNNtMs," LCO       i 3.7.4, " Main Control Room Environmental Control (MCREC)

System," and LCO 3.7.5, " Control Room Air Conditioning (AC) System."

  ,Q U

O HATCH UNIT 2

AC Sources-Operatinh 3.8.1 N ,

'         NN                                            Table 3.8.1-1

'( O ' kJ s s , Diesel Generator Test Schedule ' i

                                                                                            '                                  I
                     \IN'LAST NUMBER  25 VALID OFTESTS FAILURES (a)                           FREQUENCY
                     /        -

31 days

                /               xs3                   '

a4 j

                                                    '                    7 days (b) (but a 24 hours)

N ,/ /

                                            /
    /

f : 'N ,' Criteria for detemining number of failures and valid tests shall be in (a) accordance with Regulatory Position C.2.1 of Regulatory, Guide 1.9, Revision 3,.where the numbehof, tests and failures is' detemined on a , per DG basis.

                                                         /x '.

This test frequency shall be maintained until sev n consecutive failure /

                                                                                                                          '/ \ I (b) free starts from standby,donditions'and load and run tests have been perfomed. This is cons'istent with Reg.ulat,ory Position [ ],of                                    ,

Regulatory Guide 1.9,/R evision 3. If, subsequent to the 7 failure free tests,1 or more additional failures occur such that there are again 4 A or more failures in the last 25 tests,'the testing interval shall again V, be reduced as noted above and maintained untils7 consecutive failure free tests have' been perfomed. s ~ ' _ j , _ Note: If Revision 3 of Regulatory Guide 1.9 is not' approved, the above table wil,l' be modified to be,donsistent with the existi_ng' version of _ Regulatory Guide 1.108, GL,84-15, or other approved guidance. _

                    /                          '
                                                 /e
                                             /                                                       \

s

s %, t.- 3.8-17 Rev. O, 09/28/92 BWR/4 STS I I

AC Sources-Shutdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS ,- 3.8.2 AC Sources-Shutdown b w .+ 2. / u Y. LCO 3.8.2 The following AC electrica power sources shall be OPERABLE: a. l One qualified cir/cuit between the offsite transmission [-A network and the ensite Class lE AC electrical power distribution subsystem (s) required by LCO 3.8.1,

                                 " Distribution Systems-Shutdowq; 9                    g,   h

b. One diesel generator (DG) capable o supplying one I ision of the nsite Class 1E AC electrical power

distribution su system (s) required by LCO 3.8. R-APPLICABI'_ITY: MODES 4 and 5, _M . + t. During movement of irradiatt d fuel assemblies in the yEsecondary.P containment. ( ,=:: b..e

,y u
,      Lw    2.51
     %.           J 1

~~ Ol 3.8-18 Rev. O, 09/28/92 BWR/4 STS i

4 O INSERT LC0 3.8.2 LL I ve r Sion

c. One qualified circuit Wetween the offsite l transmission network and the onsite Unit 2 Class IE AC electrical power distribution subsystem (s) needed to support the Unit 2 Standby Gas Treatment (SGT) subsystem (s)

          & M              required by LCO 3.6:4.3, "SGT System;" and

d. One Unit 2 DG capable of supplying Unit 2 SGT subsystem required by LC0 3.6.4.3.

I h i O HATCli UNIT 1

INSERT LCO 3.8.2 6 "^i'#

c. One qualified circuit eennected between the offsite transmission network and the onsite Unit 1 Class 1E AC electrical power distribution subsystem (s) needed to support th.e Unit 1 equipment required to be OPERABLE by C0 3.6.4 7, " Standby Gas Treatment (SGT) System &

                       ,.RephtDHp," LCO 3.7.4, " Main Control Room Environmental Control (MCREC) System," and LCO 3.7.5, " Control Room Air Conditioning (AC) System;" and

d. One Unit 1 DG capable of supplying one subsystem of each of the Unit 1 equipment required to be OPERABLE by LCO 3.6.4. , LCO 3.7.4, and LCO 3.7.5. l 7 O- :

t O i HATCH UNIT 2 I

 ,         -                                                                         _  l

AC Sources-Shutdown 3.8.2 ACTIONS CONDITION' REQUIRED ACTION COMPLETION TIME A. for W D One required offsite f ---------- E5

                                                 -NOTE-------------                                        Ib circui    inoperable. Enterapp/icableCondition 92                             and RequTred Actions of O              LCO 3.8.!E, with one required J,ivisicfde-energizedasa                                                        '

r' result of Condition A.

                   'cu          A.1          Declare affected          Imediately required feature (s),

with no offsite power fJf available, inoperable. __ g 0R. A.2.1 SuspenkCORE Imediately ALTE)(ATONS. 9Y AND A.2.2 Suspend movement of Imediately irradiated fuel assemblies in the _ hp./ .1 secondary W containment. AND f A.2.3 Initiate action to Imediately suspend operations with a potential for draining the reactor vessel (OPDRVs). AND

       .                         A.2.4        Initiate action to        Imediately restore required                                                  i offsite power circuit-Q to OPERABLE status.                                               ,

1 (continued) O BWR/4 STS 3.8-19 Rev. O, 09/28/92 l l l

AC Sources-Shutdown 3.8.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME h OnerequiredDGg B. ormore B.1 Suspend CORE Imediately k0 inoperable. 4 ALTERATIONS. I AND B.2 Suspend movement of Imediately l irradiated fuel assemblies in fsecondaryP containment. AND B.3 Initiate action to Imediately suspend OPCRVs. AND B.4 Initiate action to p Imediately restorerequiredDGUV - to OPERABLE status. l I M l SURVEILLANCE REOUIREMENTS l SURVEILLANCE FREQUENCY SR 3.8.2.1

                                                                                                            .---------..-------40TE---                                                                             /       Q     E  f;'  ...      yricruc cMnn C M- n; c4 The following SRs arje not'requiredbo be                                                                                                       '*"'"

Opq f performed: SR 3.8.1.w, /$R 3.8.1.5 through h 3.8.131f,' SR 3.8.1.43 through SR 3.8.16J h {SR 3.8.1.7,andSR3.8.1. 7....--.... -----.....----- m ........ Qv) 1 u.6t

                                                                                             'W Ul i For/AC sourcfitequire Jto- bc OPEP@dhe SRs of Specification 3.8.1, except s In accordance with applicable I
                                                                                             ,us 1          SR 3.8.1.F and SR 3.8.1 70', are                                                                                                   SRs j

, T -

appl'icable. -MS f p m 7 l pn>g n =S BWR/4 ST 3.8-20 Rev. O, 09/28/92 o! l l

,n

  INSERT   LCO 3.8.4 0 va$ W The following DC electrical power subsystems shall be OPERABLE:

a. The Unit 2 Division 1 and Division 2 station service DC electrical power subsystems;

b. The Unit 2 and swing DGs DC electrical power subsystems;

c. The Unit 1 DG DC electrical power subsystems needed to su3 port the Unit 1 equipment required to be OPERABLE by A LCO 3.6J7, " Standby Gas Treatment (SGT) System A~

M," LC0 3.7.4, " Main Control Room Environmental d 06 Control (MCREC) System," LCO 3.7.5, " Control Room Air Conditioning (AC) System," and LCO 3.8.1, "AC Sources - Operating." f i ( l O HATCH UNIT 2

DC Sources-Operating 3.8.4 (5 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY v 5R 3.8.4.8 -------------------NOTFg------------------- '

                         # Thisp rJoSurveillance d irLN00E,b  shall anot n be_,,_

a * [ , dian1+themm&Pc events that satisfy this SR. kbh i Verify battery capacity is af80}f o t he 60 months manufacturer's rating when subjected to a perfomance discharge test AND Inst (T L* t

                                                             ~
                                                                              >-         O f.A -- \

Q --NOTp - ble \ /A 3 6.4.6 only a lica .

                                   /                g .eN.B                     when attery         i f M I4-                    "Y"1                     radation or 4                                  '%                                             as reached          l f           !

[85]%. of 6A 6 '.*.?.'" . ' . _'. . . I [ k12 mon s t _

    < L.,a                                                                                                      :

i s a 3.S.4. cf

                '                                                                                               f 1
              .                                                                                                 I
 .O BWR/4 STS                                 3.8-27                        Rev. O, 09/28/92

DC Sourcos-Shutdown l 3.8.5 l l

3.8 ELECTRICAL POWER SYSTEMS 3.8.5 DC Sources-Shutdown C LCO 3.8.5 'DC elec ical power su fystems shall be O ERABLE to su po%' _ the DC electrical pow distribution sub ystem(s) r j,& gL 3.8.10, "Dist ibution Systems- utdown."7 { 0 APPLICABILITY: MODES 4 and 5, Durin movement of irradiated fuel assemblies in the secondary}4ontainment. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare affected Imediately DC electrical power required feature (s) subsystems inoperable. inoperable. " E A.2.1 Suspend CORE Imediately ALTERATIONS. AND i A.2.2 Suspend movement of Imediately irradiated fuel assemblies in the _ 'gf p{secondaryK containment. A AND (continued) 1 O 3.8-28 Rev. O, 09/28/92 BWR/4 STS

( INSERT LC0 3.8.5 92v W The following DC electrical power subsystems shall be OPERABLE:

a. The Unit 2 DC electrical power subsystems needed to support the DC electrical power distribution subsystem (s) required by LC0 3.8.8, " Distribution Systems - Shutdown"; and

b. The Unit 1 DG DC electrical power subsystems needed to support the equipment required to be OPERABLE by LCO O3 3.6.47, " Standby Gas Treatment (SGT) System #t-A4ft@," and LC0 3.7.4, " Main Control Room Environmental Control (MCREC) System, "LC0 3.7.5,

                        " Control Room Air Conditioning (AC) System," and 3.8.2, "AC Sources - Shutdown."

n 1 ? v HATCH UNIT 2

INSERT LC0 3.8.7 (U2 VERSION) O The following AC and DC electrical power distribution subsystems shall be OPERABLE.

a. Unit 2 AC and DC-electric.al. power distribution subsystemsfcomprisedof: T ' '- -

1. 4160 V Essential Buses 2E, 2F, and 2G;

                        /                                                                      ;

j 2. 600 V Essential Buses 2C and 2D; y 4 l 3. 120/208 V Essential Cabinets 2A and 2B;

                    \
                     \        4. 120/208 V Instrument Buses 2A and 2B;                -

i

5. 125/250 V DC Station Service Buses 2A and 2B; \

t (y 6. DG DC Electrical Power Distribution Subsystems;- and '-

                                                                ,             ./    /

b. 'Uiift l'AC and DC electrical power distribution subsystems needed to support equipment required to be OrERABLE by i.t.u 4.b.4.Nt, " Standby Gas Treatment (SGT)

System," LC0 3.7.4, " Main Control Room Environmental O Control (MCREC) System," LCO 3.7.5, " Control Room Air Conditioning (AC) System," and LC0 3.8.1, "AC Sources - Operating." P 4 O HATCH UNIT %

i Distribution Systems-0perat b (~N ACTIONS (continued)

                                                                                                 @'@            i
\

R] CONDITION REQUIRED ACTION COMPLETION TIME

                      /        Required Action and        ,

()2

Be in MODE 3. 12 hours associated Completion Time of Condition A, AND Ohi B, +# not

,oc

[ i, h2 Be in MODE 4. 36 hours P Decla ~1 mediately

                               ~~iir~more-DG-0Q. E.1 E. One electrical power             p0Gfs)p4etC   inoperable.

gj distribution f ~- 2 subsystem Finoperable. %_

~~-

hsu c 3 s:, G47 S EILLAN E REQUIREMENTS SURVEILLANCE FREQUENCY p 9731 l Verify correct breaker alignments and 7 days SR . voltagetofrequired}^ACpDCf[and-ACvi+a@ 7 tu& electrical power distribution' fI : i subsystems. l O 3.8-39 Rev. O, 09/28/92 BWR/4 STS

l Distribution Systems-Shutdowe l 1 3.8(ly l 3.8 ,EI.ECTRICAL POWER SYSTEMS 3.8 0, istribution Systems-Shutdown P 'l 0 0 3. fhe- essary port ns of the AC,/DC, [and AC vitalWsM 1 rical power f stribution sy6 systems shall

           "                            support equiprhent required 70 be OPERABLE. pe OPER 3 m-,i J

Lco 3.to APPLICABILITY: MODES 4 and 5, , During movement of irradiated fuel assemblies in the ! i f{secondaryPcontainment. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. S Onefor more required A.1 Declare associated Imediately AC/ DC, der-AC7ttaf supported required fI JusPelectrical power feature (s) P7 distribution inoperable.

subsystems inoperable. l r OR A.2.1 Suspend CORE Imediately ALTERATIONS. j AND A.2.2 Suspend handling of Imediately irradiated fuel assemblies in the pecondary]^-

[l , containment. AND A.2.3 Initiate action to Imediately suspend ~ operations with a potential for draining the reactor vessel. AND (continued) O BWR/4 STS 3.8-40 Rev. O, 09/28/92

INSERT LC0 3.8.8 u 2 Nd b, The necessary portions of the following AC and DC electrical power distribution subsystems shall be OPERABLE:

a. The Unit 2 AC and DC electrical power distribution subsystems needed to support equipment required to be OPERABLE; and

b. The Unit 1 AC and DC electrical power distribution subsystems needed t support equipment required to be OPERABLE by LC0 3.6.4. , " Standby Gas Treatment (SGT)

System fMefdhitrt4," LC0 3.7.4, " Main Control Room Environmental Control (MCREC) System," LC0 3.7.5,

                       " Control Room Air Conditioning (AC) System," and LCO 3.8.2, "AC Sources - Shutdown."

,\ O HATCH UNIT 2

l l l RHR-High Water Level l 3.9.K l (3 $ V 3.9 REFUELING OPERATIONS / l 3.9.K Residual Heat Removal (RHR)-High Water Level ~ y w LC0 3.9.M One RHR shutdown cooling subsystem shall be OPERABLE and in

      %           (g        operation.                                                                l PO' #                 ----------------------------NOTE----------------------------

The required RHR shutdown cooling subsystem may be removed from operation for up to 2 hours per 8 hour period. , 1

                                       ) w ad.a+ea Cu< L w n,c << & n                         'llY ,7

[_ vessel (RFV)2nd. ;< ~ c o < <- b APPLICABILITY: MODE 5witNkhewaterlevela 23/=ft- above the top of the

                                 ) Reactor =prEssura=y.euel jRPV)' f1 angel.

r ACTIONS C' t/O^ ^ h[b f> CONDITION REQUIRED ACTION COMPLETION TIME

  ,_. A. Required RHR shutdown       A.1     Verify an alternate        1 hour                   I cooling subsystem                   method of decay heat (V)          inoperable.                         removal is available.      AND Once per                 1 24 hours thereafter               ,

1 l l B. Required Action and 8.1 Suspend loading Immediately ; associated Completion irradiated fuel l Time of Condition A assemblies into the I not met. RPV. l AND l B.2 Initiate action to Imediately restore [ssecondaryJ' containment to OPERABLE status. AND (continued) l

'w/

i BWR/4 STS 3.9-11 Rev. O, 09/28/92

RHR-High Water Level 3.95 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. kguirh5 L- g (continued) 8.3 Initiate action to Imediately O y restore ne- standby h/*gastreatment subsyste o OPERABLE

                                       'd         status. (6)

AND B.4 Initiate action to Imediately restore one: i geconda F, y /gontainmen isolation +- p' est c; nip i /\ gf,lJ' va+xe-andassociated

instrumealtatiaua-

                          -{      n

{'02f".ABLF-tatm in eac u ssociated l C c-- (r>uro1 O> penetration flow path not isolated. ,, _ k 0E[.\,)

        ;, b.d                                   y .t y 7 '

C. Required RHR shutdown d C.1 Est.dislish reactor 1 hour from cooling subsystem ast- coolant circulation discovery of no in operation. v -r by an alternate b reactor coolant

                         %,                      method.            ,.

ci.rcu,lation i

                      'G?. I          gno
                        ,~                                       'GP.l3
                                                                   ~f
                                                                                         >                         e        nei12v2)

ther eaHer J C.2 Monitor reactor Once per hour coolant temperature. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR Ver.ify one RHR shutdown cooling subsystem 12 hours r' 3.93).1 lg

\.,

1s operating. O BWR/4 STS 3.9-12 Rev. O, 09/28/92

I RHR-Low Water Level 3.9.K 3.9 REFUELING OPERATIONS [# 3.9., ' Residual Heat Removal (RHR)-Low Water Level

                   ~
                        " p/L LC0 3.9.)( #        Two RHR shutdown cooling subsystems shall be OPERABLE, and one RHR shutdown cooling subsystem shall be in operation.
                                 ------..--------------------NOTE---------------------------.

The required operating shutdown cooling subsystem may be ' removed from operation for up to 2 hours per 8 hour period. h',,

                                 ...----...-------.........---..-...-....--.....-..-..-...-.j.-

b yadia }eb. (je{ i n & Yf%cfat fxPrur

                                                  ,/            vessel C eFv) a e d              y APPLICABILITY:       MODE 5 with>the water level < J,231 ft above the top _of_the iteact@Jm-ess*+4hge}'. / f .P'/ J%r _                                     m-. .

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

                              -         m tr m u.cea) y v A. One or twotRHR         A        A.1        Verify an alternate       1 hour g                  shutdown cooling        \                 method of decay heat subsystems inoperable.:                    removal is available     AND
                                             /                for each inoperabl
                                      ~                       RHR shutdown cooling     Once per j                           subsystem.               24 hours

{ g,3 - thereafter u .J ^* ([ Yo rgw vuq $T.D

                                                  1         Estah1Mh reactor         1 hour from
           .Pf . No   RHRsubsystem cooling  shutdownin         ,
                                                      .       coolant circulation      discovery of no N

y).2 ]M operation. A by an alternate method. reactor coolant c circulation s. ( P-1 qW W Aj.A AND ).2 O m p.s 1:la@ fotecner

                                                 ' 8'. 2 g Monitor reactor             Once per hour
                                                        /     coolant temperature.

(continued) O BWR/4 STS 3.9-13 Rev. O, 09/28/92

1 l RHR-Loc Water level _ i 3.9.9's j k[<" & ACTIONS N W CONDITION REQUIRED ACTION COMPLETION TIME I B'. $3= Initiate action to Imediately ' p $,;ontinued)* C ~ ,,,.,J[ c.. ,

                                           @          restore ['secondaryl containment to x

7) 1 OPERABLE status.

l m e d a s u c;. &ie) ( z<,.ple + ve T rr e AND '

                                                         '                f                    f C --        -

n + > 57 M B./ ,/ Initiate /ctionto Imediately j ..L gl gas restore w s standby ( ( -

j'~

d treatment subsystemuto OPERABLE

status. @ AND B.%I Initiate action to diately restore u p ' ' vo, n y [ ( gt 1, O"o 3 secondary 3' j q j'O

                                                   ,{ containment)isolatione               'Jj -
          '#                                                                                                   /3 va+vc a-1 is-seeW                      -   v' 7

instrtimentation 40- , - - OPERABttTratus in eac W : ri ++d

'r

( 7's 3 'e 3F d] ' penetration flow path not isolated. __h SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.9.1 Verify one RHR shutdown cooling subsystem 12 hours is operating. 2@ l O1 ; BWR/4 STS 3.9-14 Rev. O, 09/28/92 ! l

l Inservice Leak and Hydrostatic Testing Operation l i 3.10.1 i 3.10 SPECIAL OPERATIONS 3.10.1 Inservice Leak and Hydrostatic Testing Operation LCO 3.10.1 The average reactor coolant temperature specified in Table 1.1-1 for MODE 4 may be changed to "NA," and operation m considered not to be in MODE 3; and the requirements of g [<p, ---tCO .3. 4./, " Residual Heat Removal (RHR) Shutdown Cooling

           .                 System-Cold Shutdown," may be suspended, to allow performance of an inservice leak or hydrostatic test provided the follcwing MODE 3 LCOs are met:

a. LCO 3.3.6.2, " Secondary Containment Isolation r. P.)Q Instrumentation," Functions)[1, 37 4 siid1ST of Table 3.3.6.2-1;

b. LCO 3.6.4.1, " Secondary Containment";

[7 c. - LCO 3.6.4.2, " Secondary Containment Isolation Valves ' (SCIVs)"; and

d. LCO 3.6.4.3, " Standby Gas Treatment (SGT) System."

O' APPLICABILITY: MODE 4 with average reactor coolant temperature > '{200FF. SR P.I I l l l l O BWR/4 STS 3.10-1 Rev. O, 09/28/92

i Inservice Leak and Hydrostatic Testing Operation 3.10.1 ACTIONS ________.______...______...._____.__ NOTE-------------_--------------..-.__--. Separate Condition entry is allowed for each requirement of tne LCO. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more of the A.1 --------NOTE--------- abo"e requirements not Required Actions to met. be in MODE 4 include reducing average reactor coolant temperature to s 200}'F. m ........____________. Enter the applicable Imediately Condition of the affected LCO. 03 A.2.1 Suspend activities Imediately that could increase the average reactor coolant temperature or pressure. AND A.2.2 Reduce average 24 hours reactor coolant temperature to s ;{200 }*F. b W i l O BWR/4 STS 3.10-2 Rev. O, 09/28/92

l C NUREG 1433 COMPARISON DOCUMENT - BASES l l O l 1 1 O

i SDM B 3.1.1 ) l B 3.1 REACTIVITY CONTROL SYSTEMS j B 3.1.1 SHUTDOWN MARGIN (SDM) BASES i i i BACKGROUND SDM requirements are specified to ensure: a. The reactor can be made subcritical from all operating conditions and transients and Design Basis Events; b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits; and c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. These requirements are satisfied by the control rods, as described in GDC 26 (Ref.1), which can compensate for the reactivity effects of the fuel and water temperature changes experienced during all operating conditions. O APPLICABLE SAFETY ANALYSES The control rod drop accident (CRDA) analysis (Refs. 2

                        -                 and 3) assumes the core is subcritical with the highest worth control rod withdrawn. Typically, the first control rHuTo m o M W                       rod withdrawn has a very high reactivity worth and, should S "' " # y; 4                         the core be critical during the withdrawal of the first m "F5'" I#                            control rod, the consequences of a CRDA could exceed the fe 6f*IO N #                          fuel damage limits for a CRDA (see Bases for LCO 3.1.6, " Rod e#le+ims ceabued                      Pattern Control"). Also, SDM is assumed as an initial is w s.pp                            condition for the control rod removal error during refueling C hopkr q"                             Ref. 4) and fuel assembly insertion error during refueling Ref. 5) accidents. The analysis of these reactivity C                                     insertion events assumes the refueling interlocks are OPERABLE operation. when the reactor is in the reiueling mode of
          )fd

( x Va'A These interlocks prevent the withdrawal of more than one control rod from the core during refueling. gger (Special consideration and requirements for multiple control rod withdrawal during refueling are covered in Special A. Operations LCO 3.10.6, " Multiple Control Rod Withdrawal-Refueling.") The analysis assumes +.his condition is acceptable since the core will be ahut down (continued) BWR/4 STS B 3.1-1 Rev. O, 09/28/92 1

SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES with the highest worth control rod withdrawn, if adequate SDM has been demonstrated. (continued) Prevention or mitigation of reactivity insertion events is necessary to limit energy deposition in the fuel to prevent significant fuel damage, which could result in undue release

                  /       of radioactivity (see Bases for LCO .l.7, "Stendby U yuid 0      -Control-(-SEC) Sys te .") . Adequate SDM ensures inadvertent criticalities and potential CRDAs involving high worth control rods (namely the first control rod withdrawn) will not cause significant fuel damage.
            @y            SDM satisfies Criterion 2 of the aNRC Polic I

LCO g 4u

The specified SDM limit accounts for the uncertainty in the demonstration of SDM by testing. Separate SDM limits are provided for testing where the highest worth control rod is

      /          o        detemined analytically or by measurement (Ref. 5) . This i g,op  a* A
               )

p c } highes y due to the reduced uncertainty in the SDM test when the' orth control rod is determined by measurement. .h

 \                       When SDM is demen5tieLed by calculations not associated wit f
 'o  %*cg\# m[.          a test 4 additional margin =st bcadded-to- the-ipeciGed-SD"r t

k #g"# -Hatt<to account for uncertainties in the calculation.

         '.1 -           ensure adequate SDM during the design process, a design To margin is included to account for uncertainties in the g     design calculations (Ref. 6).

APPLICABILITY In MODES 1 and 2, SDM must be provided because subcriticality with the highest worth control rod withdrawn is assumed in the CRDA analysis (Ref. 2). In MODES 3 and 4, SDM is required to ensure the reactor will be held subcritical with margin for a single withdrawn control rod. SDM is required in MODE 5 to prevent an open vessel, inadvertent criticality during the withdrawal of a single control rod from a core cell containing one or more fuel assembliegy - 3 (W- ) or sy

                              ;eh v                          ff.M v

(continued) BWR/4 STS 8 3.1-2 Rev. O, 09/28/92

ure ' _ , ,. __ ... - -- - - - SOM _B_3.1.1 BASES (continued) ACTIONS' _Ad With SDM not within the limits of the LCO in MODE 1 or 2, SDM must be restored within 6 hours. Failure to meet the specified SDM may be caused by a control rod that cannot be inserted. The allowed Completion Time of 6 hours is acceptable, considering that the reactor can still be shut down, assuming no failures of additional control rods to insert, this and the low probability of an event occurring during interval. B.1 If the SDM cannot be restored, the plant must be brought to MODE 3 in 12 hours, to prevent the potential for further reductions in available SOM (e.g., additional stuck control rods). The allowed Completion Time of 12 hours is

               \                    reasonable, based on operating experience, to reach MODE 3
       \                            from full power conditions in an orderly manner and without challenging plant systems, g,t M ir"g# 3.g C.1 F

With SDM not within limits in MODE 3, the operator must Pfully insert all insertable control rods. ie ' '.;=.t This ction res . (The iTioweults.ind Comple~ t the leist reactive condition for the core. ion Time _of~1 hour provides sufficient

                      > g,2 conhderi'ngs that the reactor can i is' acceptable, N "N time      textake  corrective-actiofr'and
                    ~

e be s N fs.\ & &e cwb A "O' *

eff' k"*

Y down,

f H WSOd'

                              - 0.1. 0.2. D.3. and 0.4 With SDM not within limits in MODE 4, the operator _mM
                      >           'nsert all insertable control rodsfir ' W vThis action          .

y n its_ in the least reactive condition for the core. R he 1Q1our s Completion 'Ti_me provides sufficient time to take \ correctixe action andis table accep\ sconsideri'ng assuming that there th are no react'o( ch stillsbe shQt down s . Cfalib(eAof _ ah be initiated dditional withincontrol 1 hourr$ds to lnser't.f Action to proviceWins must for control of potential radioactive releases. This includes ensuring '

. O (continued) )

BWR/4 STS B 3.1-3 Rev. O, 09/28/92 i l

                                                                                                                                 .          I

                                                                                                                     -h '. NM MP. S                                            Qunitlacd    in d   A,% hly

()f fQ J L C o vadno mac4e bd1 B 3.1.1 l ciworjs u D.1,D ? e Y # '7_one M {p],1%{K BASTS C 'S *> l 3r f ." , CTIONS 0.1. 0.2. 0.3. and 0.4 h.%3) 15condarx <eJo a ,d 1

        ', g                                                                          (continued)                               i                                                        kl; T)                                                                 l% Geo                                                 '
                 <~                    > secondary co     u innfe
                                                              %nt SGO-AG . 4 . ls '"ivec n d a ry c                                                                                    i p p g 4 4 00 3.5.4.3, " Standby Ces T.estment (SCT) Sy; tem" tis __0 D('CA.2               g oc,'L subsystemf ir OPERABL and 4t' TiRI one secondary containment isolation valve ft{0 3.5 A.2, "Seendary gg7 V                  cente n~ nt uniation valvx (tr!V:)")-and associated g                   instrumentation (4:G0 3.3.5.2, " Secondary COntai"me-nt 1+olatier !"str reat W ""}                                            are OPERABLE, sin each associated f                                       penetration flow path not isolatedq This may be perfomed b                                       as an administrative check, by examining l ogs or other infomation, to determin                                                                                                                         l!

service for maintenance p-i the comoonents are out of

                                                                                       ,c, ther reasons.

D] ( i It is not necessary to perfom th Ju veillancps nee [ded to demonstrate ; the OPERABILITY of the o onents. pf, h 3 wever, any W,P required component is inoperable, t n it must be restored to OPERABLE status. In this case, Rs ma / need to be l d perfomed to restore the component o OPE ABLE status. U g ^ ye Actions must continue until all r uired omponents are CA OPE BL ' A ,- ah a+ is a ssun,ed k h e

                                                          ',s ol.hJ +. m;H a te e n dioed                                         or o+her accep+ahle 3

a dm;rafr.Le cea,,

E.1. E.2. E.3. E.4. and E.5 *N'S += oss ure is ol 4, c,, m ca Q.h;I,+ With SDM not within limits in MODE 5, the operator' must imediately suspend CORE ALTERATIONS that could reduce SDM)( Thes-suspensions-wo-minsertion of fuel in the core or the

                  .4                     witfidrawaToflont f(rods). Suspension of these activities b'

p" . ) shall not preclude completion of movement of a component to

                    )                    a safe condition. Inserting control rods er remcving fuel frem the cerc-will reduce the total reactivity and we
                                       ,thereforegexcluded4 rom the suspended actions.,4
   @0 ',                  A.h
                                                    " i6 Action must alsob @e-i' mediately initiated to fully insert b         [g'                   all insertable control rods in core cells containing one or O   more fuel assemblies. Action must continue until all a4 ' gg/ use                     insertable control rods in core cells containing one or more ggM'"g0yC8gs                    fuel assemblies have been fully inserted. Control rods in W                               core cells containinc no fuel assemblies do not affect the reactivity of the core and therefore do not have to be Mg pc      p M'                 inserted.

9W Action must also be initiated within 1 hour to provide means for control of potential radioactive releases. This includes ensurin secondary containmentg&GC J.G.4dt is b I M lost on AdQN . y &>+ wen eduts e e,{ u e) (continued) p

")

BWR/4 STS B 3.1-4 Rev. O, 09/28/92 i

/~'

INSERT B 4 to maintain the secondary containment at a negative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 8; single failure protection is not required while in this ACTION) 1 , O i i S 3. '-4 a lA

                                                                                                     -    -     --     -           ~^                     ~

g) Rsecon< lory Co&am**Q n b

   ,'         is ola k e
          .. mm W 6,c en               b; thy is             p f u & Llo +>u dcus                   ' ps                              GA      M      I LQ .
                                 .c. 3                                     _
                                                                                                                                          .1 h                         Meh*              - 1,   , F. 3, G V._J E T                                   ((e

-f" BASES s N, s

[- 2 .c o d aca tsfr.w e e ++ro t s H 1 g Assar* ;sel*La b ACTIONS E.2. E.3. E.4. and E.5 (continued) Db I

                                          "                                            s SGT-(LCO 2.5 f. f subsyste OPERABLE'l an at least one sec               ry containment ,

h on jk j})$GNr valve {LCO .0.4. & and ass lated instrumentation bn ' b5 4003.3.0.2-) are OPERABLE, in each associatedwpenetration [O I flow path not isolated. This may be performed as an administrative chec 7 by examining logs or other informat' , to determine if the components are out of %g g ser ' for maintenance or other reasons. It is not ecessary to perform the 5fts needed to demonstrate the@hQ (,A,1

OPERABILITY of the components. If, however, any required Od I5 a55u - d component is inoperable, t ien it must be restored to

                   ' (4. be isolahd b OPERABLE status. In this use, SRs may need to be performed I

ro d/e - to restore the component t OPERABLE status. Action must D \m;H decckh 2f y rehase continue until all require components are OPERABLE. vrveille,c& (h ; SURVEILLANCE SR 3.1.1.1 REQUIREMENTS Wr : h,cw ha act cap'.dej ) Adequate SDM musttrated be @=m$ed to ensure that the reactor can be made subcritical from any initial operating 9 C Nee *hak condition.+ Adequate SDM is demonstrated by testing before # rra cc @ c b d or during the first_ startup_af_t.er fuel moveme%, fontrol rod

           @- &o                                           replacemento6r shuffling within the reactor pressure)

(vessel,7 Control rod replacement refers to the decoupling and removal of a control rod from a core location, and

          -              ~                                 subsequent replacement with a new control rod or a control

{ y^ y# qp h u sh d m.

                              '                            rod from another core location. Since core reactivity will vary   during the cycle as a function of fuel depletion and                                              ,

poison burnup, the beginning of cycle (BOC) test must als account for i 4 4 sw ,,g & k[MTherefore,

                                        ;   A   cew                Ahangesto obtainin thecore SDM,     the initial during the cyclhas3[

reactivity c.A b4e must be ir'r::::d by & dd;iC "R", which is the difference va ue 4 %

     !      $* y [,              M re m                j/between     the     calculated  value   of  s imum     cor:     re::t m t7 sa lduring the operating cycle and the calculated BOC ee+e

( C,lMo Ak/ d *O m re ct M ty. If the value of R is ' dthat is, B be a & J % D the mest . m m c point in the cyc correction to the7

             ),,,,,+                                                                                                 ,p) h a c cow   of O,22%Ah  + G- 4                   BOC measured value is required (Ref.~ 7)g/                                                           :

1 g .r % c3 m e ; The SDM may be demonstrated during an in sequence x control '

         \ n icg A m ,c e j rod withdrawal, in which the highest worth control rod is j

F w;' e,h edo[ j analytically determined, or during local criticals, where j the highest worth control rod is determiined by testing. I r$cA- ' Local critical tests require the withdrawal of out of '

                                                                                                                            && MS                                  u c L,         ,    SG 3. / f,/ tu9+                         (continue h BWR/4 STS                                                                                    Rev. O, 09/28/92 e

SDM B 3.1.1 m BASES PM

                                                                      -                  ~

SURVEILLANCE SR 3.1.1.1 (continued) '/ M k '

                                                                                        " ' * ")

REQUIREMENTS sequence control rods This testing would therefore require bypassing of the Rod . atter " =^1 Syste'" to allow the out of sequence withdrawal, and therefore additional requirements must be met (see LC0 3.10.7, " Control Rod Testing-Operating") . The Frequency of 4 hours after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification. During MODE 5, adequate SDM is required to ensure that the reactor does not reach criticality during control rod 6b- withdrawals. An evaluation of each in yesselc fuel movement [9-i i : % es during fuel loading (including shuffling fuel within the

               '       D          coreh-sha' ' he verwyrmed to ensure adequate SDM is maintained during refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern. For example, Q,

f bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be g"" g"jl 9 performed to demonstrate acceptability of the entire fuel j, [ 3V movement sequence.v-For the SDti demonstet4en'; that rely

M w'wT'; ]a solely on calculationr addithnal mawin (0.10% sk/kt mus .

                   "9 ^3 ) la uded te- the-SDM-Hmit of-0:28bok/k-to secount for                       ' '3 '    &
 ' b N SDM l* * / uncertainties -i n-t he-ca l cul a tion . Spiral offload / reload                      f
     +' wgc %n                   sequences inherently satisfy the SR, provided the fuel 2   4          assemblies are reloaded in the same configuration analyzed hs,   a for the new cycle. Removing fuel from the core will always

_ _ result in an increase in SDM. REFERENCES 1. 10 CFR 50, Appendix A, GDC 26. f i 5 -

2. FSAR, Section , h [ d '" / y, L/, L-O i @ .g

3. NEDE-24011-P-A8-US, " General Electric Standard Application for' Reactor Fuel," Supplement for United

( States, & ction 1 2.2.3.z.,

c. rem,m e c m- es2

d. Sept emaet-MS[.oLR).

4. FSAR,Section45.1.1'

5. FSAR, Section 5.1.1 b t lis N L 3 7i )

s. _.i __

                                                               < W E I r o l(/. 3. 3. 'lm (continued)

BWR/4 STS B 3.1-6 Rev. O, 09/28/92

I i

[ INSERT B 5 to maintain the secondary containment at a negative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 8; single failure protection is not required while in this ACTION) 10 %) a B 31-5 9 la l

 . . . _ . .   .       .            . _ ~       . . _ .      -     -                                   .      -        .         .-. - .

SDM B 3.1.1 ! BASES t/' '

     .,a
       \                                                                                .

REFERENCES 6. FSAR, Section 4.3.2.4.1[ - (continued) < g,-L'J

7. NEDE-24011-P-Ag, " General Electric Standard Application for Reactor Fuel." Cer" -- ' 2. 4_.L-

                                                                     ~w,'    .-
                                                                           ^~ ' ' ~ ^ ^
                                                                                          -(Vg ),51 w Cp s . si e ) ; n % (OL R) .)

h, MC Mo. 93-fD7, " has l bo Iicq Shs.beM w Tec W sp ec, 9,c g w n , psue w ksl' h 32S rITS 5,

                                             %                    e. Techoia/ reyupenuu& Ahnwa/.

[( ) i t l O BWR/4 STS B 3.1-7 Rev. O, 09/28/9? ' i i

Reactivity Anomalies B 3.1.2 l B 3.1 REACTIVITY CONTROL SYSTEMS ;

                                                                 -                             1 B 3.1.2        Reactivity Anomalies ih/
                                                                           ~Q BASES
                                                                   /

BACKGROUND in accordance with GDC 26, GDC 28, an GDC 29 (Ref. 1), reactivity shall be controllable suc that subcriticality is maintained under cold conditions andf acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Therefore, reactivity anomaly is used as a measure of the predicted versus i mes**eed core reactivity during power operation. The 7; V continual confirmation of core reactivity is necessary to p ensure that the Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity anomaly could be the result of unanticipated changes in fuel reactivity or control rod worth or operation at conditions not consistent with those assumed in the predictions of core

#                        reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing
                   ' predicted versu h ' nd core reactivity validates the v,' ,jL
           ,             nuclear methods used in the safety analysis and supports the SDM demonstrations (LC0 3.1.1, " SHUTDOWN MARGIN (SDM)") in D                    assuring the reactor can be brought safely to cold, subcritical conditions.

When the reactor core is critical or in normal power - operation, a reactivity balance exists and the net p\ reactivity is zero. A comparison of predicted and aneastwed reactivity is convenient under such a balance, since \ , parameters are being maintained relatively stable under P - steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable cb;C , producing zero net reactivity. w p In order to achieve the required fuel cyTle energy output, the uranium enrichment in the new fuel loading and the fuel loaded in the previous cycles provide excess positive , reactivity beyond that required to sustain steady state !

               'O       operation at the beginning of cycle (BOC). When the reactor is critical at RTP and operating moderator temperature, the 1

w excess positive reactivity is compensated by burnable i absorbers (%), control rods, and whatever neutron M '5c c e , . . Vig ay, I Q900 **'A , v (continued) BWR/4 STS B 3.1-8 Rev. O, 09/28/92 i l

Secondary Containment Isolation Instrumentation B 3.3.6.2 DN 8 3.3 INSTRUMENTATION b B 3.3.6.2 Seconuary Containment Isolation Instrumentation BASES BACKGROUND The secondary containment isolation instrumentation automatically initiates closure of appropriate secondary containment Gas isolation Treatment (SGT) System. valves (SCIVs) The function and of these starts the systems, S in combination with other accident mitigation systems, is tof limit fission product release during and followin postulated Design Basis Accidents (DBAs) (Ref. . I und 2 Secondary containment isolation and establishmen o vacuum with the SGT System within the assumed time limits ensures that fission products that leak from primary containment following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be OPERABLE are maintained within applicable limits. The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of m secondary containment isolation. Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a secondary containment isolation signal pb to the isolation logic. Functional diversity is provided by monitoring a wide range of independent parameters. The dok input parameters to the isolation logic are (1) reactor vessel water level, (2) drywell pressure, (3) reactor k(r building exhaustl and (4) refueling floor exhaust high radiation. Redundant sensor input signals from each parameter are provided for initiation of isolation. &n. edditien, =rml i-itiatier Of the Icgic i: pree w beoutputsofthelogicqbannelsinatripsystemare h

                            ' "*"9'd i"t ** "*'*"f- f~t* t            5Y5f** I S i '5-      "*

gnEf A trip syst m initiate!s isolatiqn$ofonetheautomatic yalve (damp other x tNR system ini atid sstarts one iso a(ion of 5 rnTothereautom subsystem whjle\ati( isolaY n valve in t ('h subsystem. netration and s ts the other $GT achloggclo s,.one of the~ two ves on each penetration an s)4rts one SGT subsystem, so tha operation of either logic isolates the secondary containment and vides for the necessary filtration of fission products. ( - (continued) BWR/4 STS B 3.3-183 Rev. O, 09/28/92

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES (continued) APPLICABLE The isolation signals generated by the secondary containment SAFETY ANALYSES, isolation instrumentation are implicitly assumed in the LCO, and safety analyses of References 1 and 2 to initiate closure APPLICABILITY of valves and start the SGT System to limit offsite doses. Refer to LCO 3.6.4.2, " Secondary Containment Isolation valves (SCIVs)," and LCO 3.6.4.3, " Standby Gas Treatment (SGT) System," Applicable Safety Analyses Bases for more detail of the safety analyses. The secondary containment isolation instrumentatio (ReI' I satisfies Criterion 3 of the NRC Policy Statemen . Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion. The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have the required number of OPERABLE channels with their setpoints set within the specified Allowable Values, fM [asshowninTable3.3.6.2-1. The amese setpoint is calibrated consistent with applicable setpoint methodology L assumptiony A channel is inoperable if its actual trip [((LoM setpoint is not within itc required Allowable Value. Each channel must also respond within its assumed response time, 9,p ggy where appropriate. Allowable Values are specified for each Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calitiration, process, and some of the instrument errors. The trip setpoints are then determined accounting for the (continued) BWR/4 STS B 3.3-184 Rev. O, 09/28/92

l' INSERT A for orocosed BASES B 3.3.6.2 The outputs of the logic channels in a trip system are arranged into two two-out-of-two trip system logics. Any trip system initiates all SGT O subsystems and isolates the automatic isolation valves (dampers) in each secondary containment penetration. Each logic closes at least one of the two valves in each secondary containment penetration and starts the required SGT subsystems, so that operation of either logic isolates the secondary containment and provides for the necessary filtration of. fission products. O O

              -- .    ..                   __--__---__________________-___-______________-_-_-__-_--_-_-_-__A

Secondary Containment Isolation Instrumentation B 3.3.6.2 (qj v BASES APPLICABLE remaining instrument errors (e.g., drif t). The trip SAFETY ANALYSES, setpoints derived in this manner provide adeqttste protection LCO, and because instrumentation uncertainties, process effects, APPLICABILITY calibration tolerances, instrument drift, and severe (continued , environmen (for channels that must function in harsh 9, environments as define _d by 10 CFR 50.49) are accounted for. c'+4ec15) In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions when SCIVs and the SGT System are required. The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.

1. Reactor Vessel Water level-Low Low. Level 2 Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.

An isolation of the secondary containment and actuation of /* the SGT System are initiated in order to minimize the ( potential of an offsite dose release. The Reactor Vessel Water Level-Low Low, Level 2 Function is one of the Functions assumed to be OPERABLE and car 3ble of providing isolation and initiation signals. The isolation and initiation systems on Reactor Vessel Water Level-Low Low, Level 2 support actions to ensure that any offsite releases are within the limits calculated in the safety analysi . fHowev tne Rn~a or Vessel Wa Levei -(vw Luw, Levei C etion ciate isolatio ot dTre(tly%s.sumed

            <g\          ,

in ety ana e becaus ost limi ng DBA as a mnd

             @'       Qteam        ebreakoutsidesecondarycontainment[

Reactor Vessel Water Level-Low Low, Level 2 signals are initiated from level transmitters that sense the difference # between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low, Level 2 Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. (9 / (continued)

./

BWR/4 STS B 3.3-185 Rev. O, 09/28/92 l

l l Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 1. Reactor vessel Water Level-Low Low. Level 2 SAFETY ANALYSES, (continued) LCO, and 20 i' APPLICABILITY The Reactor Vessel Water Level-Low ow, Level 2 Allowable Value was chosen to be the same the High Pressure Coolant Injection / Reactor Core Isolation Cooling (HPCI/RCIC) Reactor Vessel Water Level-Low Low, Level 2 Allowable Value (LCO 3.3.5.1 and LCO 3.3.5.2), since this could indicate that the capability to cool the fuel is being threatened. The Reactor Vessel Water Level-Low Low, Level 2 Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and conrequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, this Function is not required. In addition, the Function is also required to be OPERABLE during N operations with a O(fl potential for draining the reactor vessel (OPDRVs) because the capability of isolating potential sources of leakage must be provided to ensure that offsite dose limits are not exceeded if core damage occurs.

2. Drywell Pressure-Hioh High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The isolation on high drywell pressure supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis.

However, the Drywell Pressure-High Function associated with isolation is not assumed in any FSAR accident or transient analyses. It is retained for the overall redundancy and diversity of the secondary containment isolation l instrumentation as required by the 'iRC approved licensing basis. l High drywell pressure signals are initiated from pressure tran,smitters that sense the pressure in the drywell. Four channels of Drywell Pressure-High Functions are available and are required to be OPERABLE to ensure that no single (continued) BWR/4 STS B 3.3-186 Rev. O, 09/28/92

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 2. Drywell Pressure-Hioh (continued) SAFETY ANALYSES, LCO, and instrument failure can preclude perforinance cf the isolation APPLICABILITY function. The Allowable Value was chosen to be the same as the ECCS Drywell Pressure-High Function Allowable Value (LCO 3.3.5.1) since this is indicative of a loss of coolant accident (LOCA). The Drywell Pressure-High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES. ._

3. 4. Reactor Buildina and Refuelina Floor Exhaust Radiation-High High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding.

The release may have originated from the primary containment due to a break in the RCPB or the refueling floor due to a i fuel handling accident. When Exhaust Radiation-High is detected, secondary containment isolation and actuation of the SGT System are initiated to limit the release of fission : products as assumed in the FSAR safety analyses (Ref. 4). l% The Exhaust Radiation-High signals arefinitiated from IWradiation 3G1 detecto exhaust s that are f W coming fromlocated int the the reactor ventilation building and the refueling floor zones, respectively. The signal from each k Tdetector is input to an individual monitor whose trip [cfg outputs are assigned to an isolation channel. Four channels of Reactor Building Exhaust Radiation-High Function and four cliannels of Refueling Floor Exhaust Radiation-High [ ..., Function are available and are required to be OPERABLE to b ensure that no single instrument failure can preclude the g, isolation function. ps jo CfL5 not e g,

                            ' The Allowable Values are chosen to pra ptly St=t m =~ =                          .

am" a $ fq$,tc 0*'c ; I (continued) t l BWR/4 STS B 3.3-187 Rev. O, 09/2B/92

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE 3. 4. Reactor Building and Refuelino Floor Exhaust SAFETY ANALYSES, Radiation-Hiah (continued) LCO, and APPLICABILITY The Reactor Building and Refueling Floor Exhaust Radiation-High Functions are required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these 7)I events are low due to the RCS pressure and temperature

       % 64 imitations of these MODES; thus, these Functions are not [@Es]
 /)p                  required.T        iii addition. tha F"nctiena out also required t d P                    be OPERABLE during CORE ALTERATIONS,(OPDNVs, ano movement of irradiated fuel assemblies in the secondary containment, because the capability of detecting radiation releases __due                                                           '7 to fuel failures ( ue to fuc! =ccvery-wrdropped f uel $                                                                       ,

must b rovided to ensure that offsite dose i I h assemblies) limits are not excee e 7 f 5.. Manual Initiation The Man Initiation push button channels introduce signals into the se dary containment isolation logic that are redundant to t automatic protective inftrumentation channels and provi manual isolationp apability. There is no specific FSAR safe analysistpttakescreditforthis Function. It is retaine for thgaverall redundancy and l diversity of the secondaryknt(inment isolation instrumentation as required 3Nthe NRC approved licensing basis. There are two push butt ns for the lo , one manual { initiation push buttorf per trip system. here is no ; AllowablevaluefoVthisFunction,sincet channels are mechanically act 'ted based solely on the pos ion of the push buttons. b Two channel of Manual Initiation Function are availaM e and I are requi ed to be OPERABLE in MODES 1, 2, and 3, and during I CORE AL RATIONS, OPDRVs, and movement of irradiated fuel assem ies in the secondary containment. These are the MODE and other specified conditions in which the Secondary Containment Isolation automatic Functions are required to be k OPERABLE. v - (continued) BWR/4 STS B 3.3-188 Rev. O, 09/28/92

l INSERT D for orocosed BASES B 3.3.6.2 lg . The Reactor Building Exhaust Radiation-High Function is also required to be

./

OPERABLE during OPDRVs (in MODE 4 and MODE 5) because the capability of lA detecting radiation releases due to fuel failures (due to fuel uncovery) must be provided to ensure that offsite dose limits are not exceeded. The Refueling Floor Exhaust Radiation-High Function is . I A O ; e p

l I

O INSEKT TO HAttA (A A(TS I AtJD 1 6 3. 3 -IST bis /w 4

Secondary Containment Isolation Instrumentation B 3.3.6.2 ( BASES (continued) ACTIONS i Renewer's Note: Certain LCO Com la+ Mimes are based on ""'" l

approved topical repm ? -_ r er for a licensee to use I g the times, th ee must justi e-Comp tion Times as rAq' require the staff Safety Evaluation Report for the i _ topica reonr+- _ l A Note has been provided to modify the ACTIONS related to l l secondary Section 1.3,containment Completionisolation instrumentation Times, specifies channels.g that once a d g @VI [3 ] Condition has been entered, subsequent-t & d subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel. (O V A.1 Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Function 2, and 24 hours for Functions other than Function 2, has been shown to be acceptable (Refs. 5 and 6) to pennit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where

                             . placing the inoperable channel in trip would result in an p,m      puhdre          isolation), Condition C must be entered and its Required
    ,                          Actions taken.

[ x (continued) i BWR/4 STS B 3.3-189 Rev. O, 09/28/92 l l 1

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ewkg CNmd ACTIONS B.1 , (continued) l l Required Action B.1 is intended to ensure that appropriate i actions are taken if multiple, inoperable, untripped e channels within the same Function result in a complete loss 'di ' of automatic isolation capability for the associated 4 - penetration flow path (s) or a lete loss of automatic lbM 1 A'c initiation caoability for the SGT SystemJ A Function is % 42. v' ~ considered to be maintaining secondary containment isolation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate a trip signal from the given_ Function on a valid signal. This ensures that one [1 . eachf of the two SCivs in'the a :::ciated-penetration flow path and f e- SGT subsystem 3can be initiated on an isolation signal a _ rom the given Function. JFor the' unctions witn two y lganvd u) 3

                           /one-out-of-two logic tri        ystems        nctions 1, 2, 3,           4 gg              and4)       his would  requir     a trin   y _ tam   to hava nna channel         RABIF nr in trin _J Londition            does not include etion 5), since 't is not }

Tne iianuai im u a' iun Function ,q assu d in any acc1 ent or transieit analysis. Thus, a l total ss of manual 'nitiation capa 'lity for 24 hou (as Lallowed Reauired Ac ' n A.1) is all ed. The Completion Time is intended to allow the operator time O to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. p,g C.1.1. C.I.2. C.2.1. and C.2.2 L b,.) If any Required Action and associated Completion Time of Condition A or B are not met, the ability to isolate the. secondary containment and start the tSGT System, cannot be 7 ensured. Therefore, further actions must be performed t ensure the ability to maintain the secondary containm g} f4 function. Isolating the associated %ene (closing the ventilation supply and exhaust automatic isolation dampers) g and starting the associated SGT subsysteme(Required (5) Actions C.1.1 and C.2.1) performs the intended function of the instrumentation and allows operation to continue. Alternately, declaring the associated SCIVs or SGT subsystem (s) inoperable (Required Actions C.1.2 and C.2.2) is also acceptable since the Required Actions of the (continued) BWR/4 STS B 3.3-190 Rev. O, 09/28/92 i

Secondary Containment Isolation Instrumentation B 3.3.6.2

 /   BASES ACTIONS             C.1.1. C . I . 2. C . 2.1. and C . 2.2   (continued)                8#

H respective LCOs (LC0 3.6.4.2 and LC0 3.6.4.3) pro 7ioe appropriate actions for the inoperable components. " j - One hour is sufficient for plant operations personnel to establish required plant conditions or to declare the associated components inoperable without unnecessarily challenging plant systems. SURVEILLANCE i Revil k ': "a+a* CertainFrequencjspa based on approved REQUIREMENTS topical reports. In o~ icensee to use these Frequencies ensee must justi e-Er.aq1tencies as p,qi _ requi r the staff SER for the topical report. -- As noted at the beginning of the SRs, the SRs for each Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1. The Surveillances are modified by a Note to indicate that b. D') when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 b ~ hours +"provided the associated

                                - -t isolation                  Function capability. Upon maintains completionsenaadgy of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.

This Note is based on the reliability analysis (Refs. 5 h3 and 6) assumationh : M ra .3 the average time required to perfom clannel surveillance. That analysis demonstrated the 6 hour testing allowance does not significantly reduce (o the probability that the SCIVs will isolate the associated penetration flow paths and that the SGT System will initiate when necessary. SR 3.3.6.2.1 Performance of the CHANNEL CHECK once every 12 hours ensures { that a gross failure of instrumentation has not occurred. A CHANNEL CHECK isa a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is fl0*(Lj based on the assumption that instrument channels monitoring n (continued) tv/ BWR/4 STS B 3.3-191 Rev. O, 09/28/92

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2.1 (continued) l REQUIREMENTS the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are detemined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the meed criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based on operating experience that ' [8g demonstrates channel failure is rare. Th . , ps t - - the C""=EL CHECK cu>ure:, that unueiectd ^"tH;ht ch=1 l fcils.e .3 limited te 12 huu,3. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during nomal operational use of the displays associated with channels required by the LCO. SR 3.3.6.2.2 A CHANNEL FUNCTIONAL TEST is perfomed on each required channel to ensure that the entire channel will perform the intended function. % I the as setp ' Ts stot withtFQs reg 3 i(ed Al blpM lant spe'cif ci. setpornbmethbdqlogy may be rev , as" ppro iate,-i'f the'htstory'aitth.all' other pertinent informatinn indicate a need fnr the revisio % g. f 's Att( setpoin shall be Efi ;;E consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on the reliability O h g ent analysis of References 5 and 6. r .

~~

SR 3 . 3 . 6. 2.1 N_ ', ovides a checE'trf-theactub Cali 'on of trip ( ( trip setpoi The channel mu be declared inoperable if 1 (Continued) 1 BWR/4 STS B 3.3-192 Rev. O, 09/28/92

l INSERT H for proposed BASES B 3.3.6.2 l Since each trip system affects multiple SGT subsystems, Required Actions C.2.1 g3 and C.2.2 can be performed independently on each SGT subsystem. That is, one f

 .g }   SGT subsystem can be started (Required Action C.2.1) while another SGT                                                    1/   )
  '     subsystem can be declared inoperable (Required Action C.2.2).                                                                  ;

i i 1 l O l i 4 l O w an , m an z w_sa 1

i Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE bR 3.3.6.2.3 (continued) REQUIREMENTS th trip setting is discovered to be less conse tive than the A le Value specified in Table 3.3. -1. If the J trip settin 'scovered to be less c rvative than accounted for in t ropriate s nt methodology, but is not beyond the Allowab , performance is still within the requirements of t safety analysis. Under these conditions, the oint must adjusted to be f4 equal to or more c appropriate se rvative than accoun nt methodology. or in the The Frgq ency of 92 days is based on the reliability analyliis of References 5 and 6. w / SR 3.3.6.2.4 d SR 3.3.6.2.{ A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive O , calibrations, "; ; a . . .. . . erd ;;;i;eint 77.,7 '.i:ter5! 4ete-in:ti;r.; ... . .. ,... . .. . consistent with the plant specific setpoint methodology. Th: ch:: e! & l' 5: 1:ft

litr;.t;d ces.eistir.t ith th; ear ;;;ti;n ; ;f th; ity. 7.t -

                            .i.; O,..'u iv3, . -

fI the as foun 'within its Jequ ed

             \              Al         le Valu         e point antis gnot,ifibsetpernt c                   met       logyemay}

b}6' be revis s appropr,ia f the. hist 6 d all r perti 15?o tion indicat 'eed for the sion. he se int shall ef.tsetconistentwithaneassumptionsj l Lofthecurrentplantspecificsetpointmethodology. hlThe Frequencies of SR 3.3.6.2.( SR 3.3.6.2.E' ased nn the assumption d 2 92 % cd .r. O muu m voiiu.. .... p 2 :t;r.;! , . oves t; . ;!, , ' ... .. . . ...__i;r of the

                       ' magnitude of equipment drift in the setpoint analysis.

hSR 3.3.6.2. The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific (continued) BWR/4 STS B 3.3-193 Rev. O, 09/28/92

I Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES SURVEILLANCE SR 3.3.6.2. o (continuedl l l REQUIREMENTS channel. The system functional testing perfomed on SCIVs and the SGT System in LCO 3.6.4.2 and LC0 3.6.4.3, respectively, overlaps this Surveillance to provide complete testing of the assumed safety function.

       '     7 e 18~r6cTdh F quenTyTb3Te              the need to~ pe fom~this '

Su veillance und. the conditions t apply durin a plant

  ,fa#g      out e and the pot tial for an unplan              transient        the T   Surve . 'ince era ner rmed with the reac               at power.__
            ,dperating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency.Ther<           keA frgueq as, bl 4 b+ <rcep4ble b* R            lh sbdpo,at

{ SR 3.3.6.2.7

This SR ensures at'the individual channel response times 1 are less than equal to the maximum value' assumed in the accident a ysis. 'The instrume(t response times must be added e SCIV clos'ure 's to obtain the ISOLATION SYSTpt ESPONSE TIME. I TION SYSTEM RESPONSE ll E l ages tance criteria ar inch ed in Referener7', A Note to e Sur eillance states he radiation detectors ma 6 excluded from 15% N SYSTEM RESP 0!pE . TIME testin

               'fficult of gen is Note is necessary ting an      ropriate ause of et     nput         .\

b I si al d because t prind ples of detectp operation virt lly ensure an ins 6taneous response' time Response i time' f radiation dete'ct channels Shill be mea gred from detector tput or,tfie inpu f the<first electroni g y ~ component the, channel. A

                             /                                                              '

ISOLATION SY,S - RESPONSE II E tes are condu.c ori an 18 month S7AGGE TEST BASIS. The A month 4 requency is consistent with th typ'ical industry re ling cycle and is l based on' plant opera

g experience, ich shows that random failures of instrumen (ion compone s causing serious response time degradatioh, but not channel failure, are '

regent occurrences. , _,__ j (continued) BWR/4 STS 8 3.3-194 Rev. O, 09/28/92

ECCS-Shutdown B 3.5.2 ('-) B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEN 8 3.5.2 ECCS-Shutdown BASES BACKGROUND A description of the Core Spray (CS) System and the low pressure coolant injection (<.PCI) mode of the Residual Heat Removal (RHR) System is provided in the Bases for LCO 3.5.1, "ECCS-Operating." APPLICABLE The ECCS performance is evaluated for the entire spectrum of SAFETY ANALYSES break sizes for a postulated loss of coolant accident (LOCA). The long tenn cooling analysis following a design basis LOCA (Ref.1) demonstrates that only one low pressure ECCS injection / spray subsystem is required, post LOCA, to maintaingne peak iaaaing teiiiperature ce(ow tne allowaule s g,'j 11 t. To p ovide edundancy, a minimum o twolowpressur$q I ECCS injectio spray ubsystems areN[equire tob\ OPERABLE in H0 S 4 and . Two PERABbE low p essure (CCS subsyst'ams (- so en vre adeq te ve el inv keup in the \ vent 6f / phy an,inadvgtentves 1 dr indown\ntory . y 0 The low pressure ECCS subsystems satisfy Criterion 3 of the NRC Policy Statemen g ( LC0 Two low pressure ECCS injection / spray subsystems are required to be OPERABLE. The low pressure ECCS injection /

     /
                      ' 7 spray subsystems consist of two CS subsystems and two LPCI
    /ihe ne~ m y gom' subsystems. Each CS subsystem consists of one motor driven g a v hadheroce_. pump, piping, and valves to transfer water from the kh 4. q em. en a d isuppression              pool or condensate storage tank (CST) to the
                                    ! reactor pressure vessel (RPV). Each LPCI subsystem consists a ls o r eaw re'A *           'of one motor driven pump, piping, and valves to transfer p,,y, J c0;ppretsddwaterfromthesuppressionpooltotheRPV. Only a single p ". , , I . ,,O b gy A
          ,                        lLPCI pump is required per subsystem because of the larger reauire4              c c *e  injection capacity in relation to a CS subsystem.      In HODES 4 and 5, the RHR System cross tie valve is not r dsys Fc m -                 required to be closed.+

u .j ' f One LPCI subsystem may be aligned for decay heat removal and W considered OPERABLE for the ECCS function, if it can be

             .b \

(continued) BWR/4 STS B 3.5-17 Rev. O, 09/28/92

ECCS-Shutdown B 3.5.2 BASES i LC0 ' manually realigned (remote or local) to the LPCI mode and is (continued) not otherwise inoperable. Because of low pressure and low ! temperature conditions in MODES 4 and 5, sufficient time ' will be available to manually align and initiate LPCI subsystem operation to provide core cooling prior to postulated fuel uncovery. 1 APPLICABILITY OPERABILITY of the low pressure ECCS injection / spray subsystems is required in MODES 4 and 5 to ensure adequate coolant inventory and sufficient heat removal capability for f,@ the irradiated fuel in the core in case of an inadvertent draindown of the vessel. Requirements for ECCS OPERABILITY F,

                 .            during MODLS 1, 2, and 3 are discussed in the Applicability section of the Bases for LCO 3.5.1. ECCS subsystems are not M,b !t~8.8rd6 required      to be OPERABLE during MODE 5 with the spent fuel i

storage pool gates removed and the water level maintained at yJ39ftabovetheRPVflange4 This provides sufficient 6 Tant invenTWy to7 Trow operator action to terminate the inventory loss prior to fuel uncovery in case of an T'.6 [, . , A+ + o inadvertent draindown. dI +o D3* he Automatic Depressurization System is not required to be a bo ve th e b g OPERABLE during MODES 4 and 5 because the RPV pressure is c>f f,racha4ch 5 150 psig, and the CS System and the LPCI subsystems can g}y provide core cooling without any depressurization of the primary system. S ep +e d i n h 2 9,J fu ~ eSb ' C he High Pressure Coolant Injection System is not required { ~ to be OPERABLE during MODES 4 and 5 since the low pressure ol rac #5) ECCS injection / spray subsystems can provide sufficient flow ( to the vessel. ACTIONS A.1 and B.1 If any one required low pressure ECCS injection / spray PS subrystem is inoperable, the inoperable subsystem must be restored to OPERABLE status in 4 hours. In thisd'ondition, the remaining OPERABLE subsystem can provida sufficient vessel flooding capability to recover from an inadvertent

                            , vessel draindown. However, overall system reliability is reduced because a single failure in the remaining OPERABLE subsystem concurrent with a vessel draindown could result in (continued)  g BWR/4 STS                             B 3.5-18                    Rev. O, 09/28/92

ECCS - Shutdown B 3.5.2 BASES ACTIONS A.1 and B.1 (continued) the ECCS not being able to perform its intended function. The 4 hour Completion Time for restoring the required low pressure ECCS injection / spray subsystem to OPERABLE status is based on engineering judgment that considered the remaining available subsystem and the low probability of a vessel draindown event. With the inoperable subsystem not restored to OPERABLE status in the required Completion Time, action must be immediately initiated to suspend operations with a potential for draining the reactor vessel (0PDRVs) to minimize the probability of a vessel drsindown and the subsequent potential for fission product release. Actions must continue until 0PDRVs are suspended. C.1. C.2. D.1. D.2. and D.3 With both of the required ECCS injection / spray subsystems inoperable, all coolant inventory makeup capability may be unavailable. Therefore, actions must immediately be

 ;4                                     initiated to suspend OPDRVs to minimize the probability of a
   '                                    vessel draindown and the subsequent potential for fission product release. Actions must continue until 0PDRVs are suspended. One ECCS injection / spray subsystem must a1so be estored to OPERABLE status within 4 hours.       Lsert 13 Go,n P.A Qt )                                            93Y 6% -10 t

at least one low pressure ECCS injection / spray suosystem is not restored to OPERABLE status within the 4 hour F Completion Time, additional ctions are required to minimize A h any potential fission product release to the environment. ; V This includes initiati g i=cdfe aet-fe to restorothen-meO , ,g lic;ing tc OPEP^SLE ',tatus: secondarycontainmentdonef ws 24 isolation va W and gig' gg, @ tandby gas treatment subsystem f @h asso associated instrumentatiorvin eac r r A, L flow path not isolatedi OPERABILITY may berverified by an Q/NSEWci-administrative check /or by examining logs or other %x information, to determine whether the components are ou o ch6%t) service for maintenance or other reasons. 4'"if M 6 'i=: F ItJSETg c7, not n 'i

e performug:the Surveillances needed to demonstrate ths OPERABILITY of the components. If, however, H is M any required' component is inoperable, then it must be ^*W3I;

~~

restored to OPERABLE status. In this case, the Surveillance A 1 may need to be performed to restore the component to (L1 h u ,md ; 3

                                          /                                                                   Is A )

Db fo i, soI&/ % // ' m . + , n d s' ra el o c e 1 fy <rica & w-BWR/4 STS B 3.5-19 Rev. O, 09/28/92 l

ECCS-Shutdown B 3.5.2 BASES ACTIONS C.1. C.2. D.1. D.2. and 0.3 (continued) _ OPERABLE status. Actions must continue until all required 7h,lk components are OPERABLE. w _ ..______ -- - m [ The 4 hour Completion Time to restore at least one low pressure ECCS injection / spray subsystem to OPERABLE Ng>f t] b - ensures that prompt action will be taken to provide the required cooling capacity or to initiate actions to plac the plant in a condition that minimizes any potential .fo " Yp 1,], si . 'p( ssion product release to the environment. SURVEILLANCE SR 3.5.2.1 and SR 3.5.2.2 /g REQUIREMENTS The minimum water level of c .. "-" inchesj required for the 8 suppression pool is periodically verified to ensure that the suppression pool will provide adequate net positive suction head (NPSH) for the CS System and LPCI subsystem pumps, recirculation volume, and vortex prevention. With the suppression pool water level less than the required limit, all ECCS injection / spray subsystems are inoperable unless they are aligned to an OPERABLE CST. g} When suppression pool level is <)t12 n 2 inche , the CS System is considered OPERABLE only if it can ta e suction from the CST, and the CST water level is sufficient to provide the required NPSH for the CS pump. Therefore, a

        'l av ification that either the suppression pool water level is

[ t $12 E 2 inche $ or that CS is aligned to take suction

               ./     ,fromtheCSTandtheCSTcontainsa$50,0001 gallons of G, O                water, equivalent to 12 ft, ensures that the CS System can supply The CS at     leastis150,0001 suction    uncoveredgallons  of makeup at the M00,000        w(ater gallon   level.to the RPV However, as noted, only one required CS sub) system may take credit for the CST option during OPDRVs. During OPDRVs, the f,\               volume in the CST may not provide adequate makeup if the RPV w were completely drained. Therefore, only one CS subsystem is allowed to use the CST. This ensures the other required ECCS subsystem has adequate makeup volume.

The 12 hour Frequency of these SRs was developed considering operating experience related to suppression pool water level and CST water level variations and instrument drift during the applicable MODES. Furthermore, the 12 hour Frequency is (continued) BWR/4 STS B 3.5-20 Rev. O, 09/28/92

i

i f3 Insert C2 l are OPERABLE or other acceptable administrative controls to assure isolation I capability) i i U Insert B19a . g ,'[ (at least including: the Unit' reactor building zone if in MODE 4; or the ! common refueling floor zone if in MODE 5) is OPERABLE; 2) sufficient i Insert Cl I l to maintain the secondary containment at a negative pressure with respect to I the environment (dependent on secondary containment configuration, refer to Reference 2; single failure protection is not required while in this ACTION); - and 3) secondary containment isolation capability is available (i.e., one secondary containment l O v I l 1 l I Insert to B 3.5-19

INSERT F for oroposed BASES 3.5.2

2. Technical Requiremenss iianual . f$

'l 3. NRC No. 93-102, " Final Policy Statement on Technical Specification lj Improvements," July 23, 1993. l l 1 O I l

i l

                                                                  $5econdaryf Containment  I g               B 3.6 .1

,m B 3.6 CONTAINMENT SYSTEMS (V i B 3.6.4.1 JSecondaryfContainment BASES BACKGROUND Thefunctionofthefsecondaryfcontainmentistocontain, dilute, and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA). g t In conjunction with operation of the Standby Gas Treatment (SGT) System and closure of certain valves whose lines

                 '.e penetrate the (secondaryl containment, the fsecondaryF containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment.

OThe tsecondary)' containment is a structure that completely bencloses the primary containment and those components that may be postulated to contain primary system fluid. This structure fonns a control volume that serves to hold up and dilute the fission products. It is possible for the /7 -q pressure in the control volume to rise relative to the V ',y,/ environmental 0 (e.g., due to pump and motor heat Coad additions)4 essure .A To prevent ground level exfiltration while pnEgr A ' allowing the tsecondaryy containment to be designed as a conventional structure, the (secondaryF containment requires pPsupportsystemstomaintainthecontrolvolumepressureat less than the external pressure. Requirements for these systems are specified separately in LCO 3.6.4.2, " Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3,

                         " Standby Gas Treatment (SGT) System."                    g APPLICABLE           There are two principal accidents for which credit is taken SAFETY ANALYSES      for fsecondaryl containment OPERABILITY.      These are a loss of coolant accident (LOCA) (Ref.1) and a fuel handling accident inside { secondary} containmer.t (Ref. 2). The g(d' )                containment performs no active function in v 1     secondary)4nE response  to        of these limiting events; however, its leak tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage paths and associated leakage
                        ' rates assumed in the accident analysis and that fission n                                                                               (continued)

G l BWR/4 STS B 3.6-97 Rev. 0, 09/28/92 6 i

                                                                     ;{S condaryY Containment (p                   B 3.6.4.1       4 Pos4al&d sW, l ea M ay h,n aa, %,,g in t t e w.r I-.otA                                          ,w    a [,

m . _ .- 3 BApESe_ .n '% a % e '* A e, cous (, g % 7

                                                                                   .. s f,, g ,y N g                      2-APPLICABLE             prodncts entrapped within the econdaryKcontainment SAFETY ANALYSES        structure will be treated'by, theQGT Systemsprior to                      )7    l (continued)       discharge to the environment >            @iJ TI wauN#TM              .

condary}; containment satisfies Criterion 3 of the NRC Policy Statemen . q [n)f6 LC0 AnOPERABLE(secondary)FD containment provides a control volume into which fission products that bypass or leak from primary containment, or are released from the reactor coolantpressureboundarycomponentslocatedinfsecondary)[ containment, can be diluted and processed prior to release to the environment. For the {secondaryk containment to be i

  '}, N                 considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum a n be established and M               maintainedj APPLICABILITY          In MODES I, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to fsecondaryk(61 containment _.4 Therefore, { secondary} containment OPERABILITY ~'

/3q is required during the same operating conditions that require primary containment OPERABILITY. Oh~scrtT C In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining p (~,Q){secondaryk containment OPERABLE is not required in MODE 4 d or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during ep:r:ti= ith a potential d, ' i e o.uin; th: rer.;ter : nel >(OPDRVsf, during CORE ALTERATIONS, or during movement of irradiated fuel g assembliesintheisecondary containment q g,(6) ACTIONS A.1 e [ Iffsecondarykcontainentisinoperable,itmustbe

                       . restored to OPERABLE status within 4 hours.          The 4 hour                '

Completion Time provides a period of time to correct the problem that is commensurate with the importance of i (continued) h i BWR/4 STS B 3.6-98 Rev. O, 09/28/92 l l

O INSERT A for orocosed BASES B 3.6.4 1 l (tJ a;4 l) The secondary containment encompasses three separate zones: the Unit I reactor building (Zone I), the Unit 2 reactor building (Zone II), and the common refueling floor (Zone III). The secondary containment can be modified to exclude the Unit 2 reactor building (Zone II) provided the following requirements tre met:

a. Unit 2 Technical Specifications do not require OPERABILITY of Zone II;

b. All hatches separating Zone III from Zone II are closed and sealed, l and; l l

c. At least one door in each access path separating Zone III from Zone !

II is closed. g l Similarly, other zones can be excluded from the secondary containment OPERABILITY requirement during various plant operating conditions with the appropriate controls. For example, during Unit I shutdown operations, the secondary containment can be modified to exclude the Unit I reactor building (Zone I) (either alone or in combination with excluding Zone II as described above) provided the following requirements are met:

a. Unit 1 is not conducting operations with a potential for draining the i reactor vessel (OPORV); 1

b. All hatches separating Zone III from Zone I are closed and sealed, .

and;

c. At least one door in each access path separating Zone III from Zone I is closed.

i V

I /% l (V) JRSERT A for proposed BASES B 3.6.4.1 (continued) l { UNIT 2} The secondary containment encompasses three separate zones: the Unit 1 ; reactor building (Zone I), the Unit 2 reactor building (Zone II), and the i common refueling floor (Zone III). The secondary containment can ha r.adified ! to exclude the Unit I reactor building (Zone I) provided the following requirements are met:

a. Unit 1 Technical Specifications do not require OPERABILITY of Zone I;

b. All hatches separating Zone III from Zone I are closed and sealed, and;

c. At least one door in each access path separating Zone III from Zone I is closed.

Similarly, other zones can be excluded from the secondary containment OPERABILITY requirement during various plant operating conditions with the appropriate controls. For example, during Unit 2 shutdown operations, the secondary containment can be modified to exclude the Unit 2 reactor building (Zone II) (either alone or in combination with excluding Zone I as described above) provided the following requirements are met: (V3

a. Unit 2 is not conducting operations with a potential for draining the reactor vessel (OPDRV);

b. All hatches separating Zone III from Zone II are closed and sealed, and;

c. At least one door in each access path separating Zone III from Zone II is closed.

INSERT A.1 gneor mon 2ones ab hMSG) When Zonerb er zcnc m is excluded from secondary containment the 4equirements for the support systems-map 1= '-^'"" 'i f fered .ruub --mis #^ r = =,% securing particular SGT or drain isolation valves). a(so c(qe(e

@ INSERT B for proposed BASES B 3.6.4.1 l The secondary containment boundary required to be OPERABLE is dependent on the operating status of both units, as well as the configuration of doors, hatches, refueling floor plugs, SCIVs, and available flow paths to SGT Systems. The required boundary encompasses the zones which can be postulated to contain fission products from accidents required to be considered for the condition of each unit, and furthermore, must include zones not isolated from the SGT subsystems being credited for meeting LC0 3.6.4.3. Allowed configurations and associated SGT subsystem requirements and associated SCIV requirements are detailed in the Technical Requirements Manual (Ref. 3). INSERT C for orocosed BASES B 3.6.4.1 (the reactor building zone and potentially the refueling floor zone). INSERT D for proposed BASES B 3.6.4.1 (Note, moving irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3.) Since CORE ALTERATIONS and movement of irradiated fuel assemblies are only postulated to release radioactive material

 ~) to the refueling floor zone, the secondary containment configuration may        ~ .

consist of only Zone III during these conditions. Similarly, during 0PDRVs while in MODE 4 (vessel head bolted) the release of radioactive materials is only postulated to the associated reactor building, the secondary containment configuration may consist of only Zone I (Unit 1) and Zone II (Unit 2). I 1 l U l

i

                                                                    )[SecondaryPContainment B 3.6.4.1 O  BASES ACTIONS                  M (continued) maintaining (secondaryP containment during MODES 1, 2, and 3. This time period also ensures that the probability 7

of an accident (requiring 4 secondary) containment OPERABILITY) occurring during periods where { secondary 3-containment is inoperable is minimal. B.1 and 8.2 If 4secondaryV tainment 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 12 hours and to MODE 4 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. C.I. C.2. and C.3

                         ' Movement of irradiated fuel assemblies in the {secondaryk containment, CORE ALTERATIONS, and OPDRVs can be postulated to cause fission product release to the fsecondaryh containment. In such cases, the {secondaryk containment is the only barrier to release of fission products to the a

environment. CORE ALTERATIONS and movement of irradiated

                         ; (fuel  assemblies must be immediately suspended if the secondary.Fcontainmentisinoperable.

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be imediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPORVs are suspended. Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in

                          ' MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend f

(continued) BWR/4 STS B 3.6-99 Rev. O, 09/28/92

                                                                              $SecondaryFContainment 8 3.6.4.1 BASES g

ACTIONS C.I. C.2. and C.3 (continued) movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.6.4.1.1 REQUIREMENTS This ensures that the [ secondary] containmenJt ary is sufficie leak tight to preclude exfilt itFn under expected win nditions. The 24 ho equency of this SR was developed ba on operati perience related to q1 [ secondary] containme . v m variations during the

p. applicable MODES an e robability of a DBA occurring between survei ces.

Fur ore, the 24 hour Frequency is co ered adequate in

                                      'ew of other indications available in the c          ol room, f        including                           operator to an ab o 1

[ secondary] alarms, to alert containment the condition. vacuum P - gg g.g .y. y alsh . 2-SR 3.6.4.1.IandSR 3.6.4.1. vfwresa MW H

            "     h                Verifying that {secondaryF containment equipment hatches and eea9d. in ih.s     b,c",7*

A** dd access doors are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaini ng no cwoobb. d the desired negative pressure does not occur. Verifying id W # that all such openings are closed provides adequate assurance that exfiltration from the { secondary) containment will not occur.* Maintaining (secondary]ccontainment OPERABILITY requires verifying each door in the access

             ,9                   opening is closed, except when the access opening is being used for nonnal transient entry and exit (then at least one C            d          door must remain close_db The 31 day Frequency for these pl$ fin C                                  ~

SRs has been shown'to be adequate, based on operating experience, and is considered adequate in view of the other indications of door and hatch status that are available to the operator.

                           ',       SR   3.6.4.1.A       SR   3.6.4.1.         r I                    SGT Systebxhausts the (secondary c tainment f                         atmosphere to the environment through appropriate treatment (continued)

BWR/4 STS B 3.6-100 Rev. O, 09/28/92

i l O INSERT E for orocosed BASES B 3.6.4.1 l When the secondary containment configuration excludes Zone I and/or Zone II, these SRs also include verifying the hatches and doors separating the common l +k !

refueling floor zone from the reactor building (s). l/ O \

b{ Secondary 7 Containment B 3.6.4.1 BASES l SURVEILLANCE JSR 3.6.4.1. ndh SR 3.6.4.1. (continued) s REQUIREMENTS J Caiprorriate) (d \ g equipment. To ensure thaT alijtissiony tro ucts are teeated, h SR 3.6.4.1.A verifies that the 5GT v System &will rapidl ; establish and maintain a pressure in the (secondary j g> containment that is less than the lowest postulated pressure , P. external.To~the'{ secondary) containment boundary.f This.is~ -l confinned by demonstrating that one_SGI subsystem' will dra ~ e down theTfsecondary) corttainment

~~~

to"d(0.25} inche 3oundary issm p T0fi vacuum watdr gauge accomplishedifthefsecondary in s D S ) T6ntainment not intact. F

t fcan maintain eSR 3.6.4.las 10.25} inchsedemonstrates of vacuum water that.ane gaugeSGTfor subsystem Y 1 hour at a flow rate s (4000) cf The I hour test period , g1- allows ifecondaryk confliniiiiiit ~ o e in thernal equilibrium i

    '(2 7                          at steady state conditions. Therefore, these two tests are                         -

r' used to ensure fsecondarykcontainment boundary integrity. SincetheseSRsare(secondaryFcontainmenttests,theyneed, b ( b not be performed with each SGT subsystem. The SGT subsystems ; ea.ch gGr are tested on a STAGGERED TEST BASIS. however, to ensure ' subg h

that in addition to the requirements of LCO 3.6.4.3,.s hher g SGT subsysteigwill perfona this test. Th: {! $ rath :

TFr:p:::; t:7:ced ca t'e ::d te perfe- thi: S recill:::e - WhdA* haderth: :::ditien th.; epply d. ring . pl.d e.t&ge end > h ( 0 615 M ' & pd ntiel fer :: ;. ;;i :::d -tran:ient lf the S.rvei'::::

                                  ,wer: p:rf: .:d ith the Pr eter et per r. Operating 6            experience has shown these components usually pass the QF           Surveillance when performed at the (18h month Frequency.

Therefore, the Frequency was concluded to be acceptable from l a reliability standpoint. ' i REFERENCES . FSAR,Sectionf15.1.0

2. FSAR, Section {15.1. 4' ** d' '

Q}RE j

        'I'he number of SGT subsystems and the required ~ddmblii5tions arh dependent on the configuration of the secondary containment and are detailed in the Technical Requirements Manual (Ref.                                                                 i 3).h Riote / to SR 3.6.4.1.3 and W " m to SR speciff that the number of required SGT subsystems be one less 3.6.4.1.4 than the number required to meet LCO 3.6.4.3, " Standby Gas                                                                 :

O tment System," for the given configuration. BWR/4 STS B 3.6-101 Rev. O, 09/28/92 ;

SCfVs j B 3.6.4.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) BASES BACKGROUND The function of the SCIVs, in combination with other accident mitigation systems, is to 14 nit fission product

                              - release during and following postulated Design Basis h' Accidents                         Secondary containment isolation within the(DBAs)     (men).

time limits specified for those isolation valves designed to close automatically ensures that fission products that leak from primary containment following a DBA, or that are released during certain operations when primary

   /flid M"^dgyJ                 containment is not required to be OPERABLE or take place outside primary containment, are maintained within               -

corthaltune ttb bou'l0*'j mp! M: '; -;t5. p I h,1I The OPERABILITY requirements for SCIVs help adequate 4secondaryP containment hd_dgh=/ ensur ss is potential paths to the environment. These isolation devices [c [ maintained during consist of either passive devicesand after or active an accident (automatic) devices. Manual valves, de-activated automatic valves secured in their closed position with flow through the valve secur,f4mstutHwg check valvesed considered passive devices. e6fdla- Automati_c SCIVs close on a (secondaryl containment isolation signal to2prever.t-leakage of untreated radioactive material 6oorde"I bt' from {secondaryF containment following a DBA or other

f. accidents. g v6fd Wdha Other penetrations are isolated by the use of valves in the closed position or blind flanges.

                                                                              ~               _

g),e - - APPLICABLE The SCIVs must be OPERABLE to ensure ;{ secondary SAFETY ANALYSES containment 3: : 'eeE *ipt barrier to fissio T~releasesA. The principal accidents for whict}(}!Lproduct sec6ndar CI 5 esca4W' V containment /a -ist=z is required are a loss of coolant ' accident (Ref.1) and a fuel handling accident inside secondary) containment (Ref. 2). The (secondaryF #

    .bp,b - $wrJdary (p ' containment performs no active function in response to
        -                     ~ either of these limiting events, but ite 1;;L C 3 ne.        is bour.dy           required to ensure that leakage from the primary containment 46 Lis b d (continued)

BWR/4 STS B 3.6-102 Rev. O, 09/28/92 ()

INSERT REF.

3. Technical Requirements Manual

4. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

O l l

                                             ~
                  ~

O l

SCIVs B 3.6.4.2 [] v BASES APPLICABLE is processed by the Standby Gas Treatment (SGT) System SAFETY ANALYSES before being released to the environment. (continued) Maintaining SCIVs OPERABLE with isolation times within limits ensures that fission products will remain trapped (Ginside b /by the(secondary). SGT Systemcontainment so thattothey prior to discharge can be treated the environment. SCIVs satisfy Criterion 3 of the NRC Policy Statemen ( E d '] LCO SCIVs form a part of the [ secondary)b containment boundary. The SCIV safety function is related to control of offsite radiation releases resulting from DBAs. The power operated isolation valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an automatic isolation signal. The valves ['y covered by this LCO. along with their associated stroke times, are listed in Reference 3. @ gg 7-

              $CIO                                                                         are The nonnally closed isolation valves or blind flan utomati(
                            %consideredOPERABLEwhenmanua?valvesareclosed s are de-activated and secured in their closed position, and blind flanges are in place. These passive
         $66k7f              isolation valves or devices are listed in Reference 3.
    ]k'                                                    n           _

tocA )N APPLICABILITY In MODES 1, 2, and 3, a sBA could lead to a fission product release to the primary containment that leaks to the Therefore, the OPERABILITY of QJ)SCIVs is required.{ secondary) containment. In MODES 4 and 5, the probability and consequences of 4hese ) events are reduced due to pressure and temperature ! limitations in these MODES. Therefore, maintaining SCIVs i OPERABLE is not required in MODE 4 or 5, except for other , situations under which significant radioactive releases can l be postulated, such as during operations with a potential i for draining the reactor vessel (OPDRVs), during CORE ; ALTERATIONS, or during movement of irradiated gyct vina 1

                    }.k l' assemblies in the lsecondaryl containment. irradiated may also occur in H0 DES 1, 2, and g

V (continued) BWR/4 STS B 3.6-103 Rev. O, 09/28/92

SCIVs B 3.6.4.2 i BASES (continued) h ACTIONS The ACTIONS are modified by three Notes. The first Note allows penetration flow paths to be unisolated intermittently under administrative controls. These

  }                     controls consist of stationing a dedicated operator, who is in continuous comunication with the control room, at the 16g ,yoft.    ~ controls of theTe+ee. In this wa rapidlyisolatedwhenaneedfor(y,thepenetrationcanbe secondaryk containment de* Ce                isolation is indicated.

The second Note provides clarification that for the purpose of this LCO separate Condition entry is allowed for each penetrationflowpath.3 friS6hI C The third Note ensures appropriate remedial actions are taken, if necessary, if the affected system (s) are rendered inoperable by an inoperable SCIV. A.1 and A.2 In the event that there are one or more penetration flow paths with one SCIV inoperable, the affected penetration i> flow pathW!) 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 SCIV, a closed manual valve, and a blind flange. For penetrations isolated in accordance with Reauired Action A.1, the W used to isolate the ~ penetration sTouid be the' closest available ve+ vet bV'Ce

gqice secondary}<. containment. The Required Action must be completed within the 8 hour Completion Time. The specifie p2

                    / time period is reasonable considering the time required to A isolate    the penetration, and the probability of a DBA, which equires the SCIVs to close, occurring during this short time is very low.

For affected penetrations that have been isolated in accordance with Required Action A.1, the affected penetration must be verified to be isolated on a periodic : basis. This is necessary to ensure that tsecondary) (f.T l containmentpenetrationsrequiredtobeisolatedfollowing) i an accident, but no longer capable of being automatically l

                     , isolated, will be in the isolation position should an event occur. The Completion Time of once per 31 days is appropriate because thefvalues are operated under (continued)

BWR/4 STS B 3.6-104 Rev. O, 09/28/92

O INSERT A for oroposed BASES B 3.6.4.2 lg The SCIVs required to be OPERABLE are dependent on the configurations of the secondary containment (which is dependent on the operating status of both units, as well as the configuration of doors, hatches, refueling floor plugs, I and available flow paths to SGT Systems). The required boundary encompasses the zones which can be postulated to contain fission products from accidents A required to be considered for the condition of each unit, and furthermore, M must include zones not isolated from the SGT subsystems being credited for meeting LC0 3.6.4.3. The required SCIVs are those in penetrations communicating with the zones required for secondary containment OPERABILITY and are detailed in Reference 3. INSERT B for orocosed BASES B 3.6.4.2 A lM or open in accordance with appropriate administrative controls, INSERT C for proDosed BASES B 3.6.4.2 This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable SCIV. Complying with the Required Actions may allow for continued operation, and subsequent inoperable SCIVs are governed by subsequent Condition entry and application of associated Required Actions. i 4 O

j SCIVs [ B 3.6.4.2 f s BASES ACTIONS A.1 and A.2 (continued) administrative controls and the probability of their misalignment is low. This Required Action does not require an_y testing or refte manipulation. Rather, it involves verification that the affected penetration remains isolated. deme Required Action A.2 is modified by a Note that applies to v9:n ed blM fi:7s located in high radiation areas and allows them to be verified closed by use of administrative c/e gy controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment e' - -- once they have been verified to be in the r position, is low. With two SCIVs in one or more penetratfor flow paths I A inoperable, the affected penetration flow path must be isolated within 4 hours. The method of isolation must p include the use of at least one isolation barrier that V 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 4 hour Completion Time is reasonable considering the time required to isolate the penetration and the probability of a DBA, which requires the SCIVs to close, occurring during this short time, is very low. Th tion has been modified b M ng that Condition i i penetration flow path ]s with two isolatio es. a only Condition entered if one SCIV is inopera e ifreech of g rations. - C.1 and C.2 Cetiddmn A or 8} IfanyRequiredActionandassociatedCompletionTime)cannot be 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 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on ooerating experience, to reach the (continued) i BWR/4 STS B 3.6-105 Rev. O, 09/28/92

SCIVs l 8 3.6.4.2 BASES ACTIONS C.1 and C.2 (continued) required plant conditions from full power conditions in an : orderly manner and without challenging plant systems. j e D.I. D.2. and 0.3 ## # If any Required Action and associated Completion Timejare not met, the plant must be placed in a condition in which the LCO does not apply. If applicable, CORE ALTERATIONS and the movement of irradiated fuel assemblies in the 1 Tsecondary3c containment must be imediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be immediately initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until OPDRVs are suspended. Required Action 0.1 has been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving fuel while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. SURVEILLANCE SR 3.6.4.2.1 L REQUIREMENTS I This SR verifies that each secondary containment manual t isolation valve and blind flange that is required to be

                            .                 closed during accident conditions is closed. The SR helps f
                                     )        to ensure that post accident leakage of radioactive fluids /g h 9@g' es          or gases outside of the ;fsecondary) Unfainment boundary is W'             within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification
                              /
                                   \

that thoselha in '{secondaryk ontainment that are

                              !   N' 15       capable of being mispositioned are in the correct position.

N Since there ve he.s are readily accessible to personnel

                                             'during normal operation and verification of their position is relatively easy, the 31 day Frequency was chosen to (continued)

BWR/4 STS B 3.6-106 Rev. O, 09/28/92

l SCIVs B 3.6.4.2 i BASES L LR 3.6.4.2.1 (cont'inued) provide added assurance that the 3% h positions. areinthecorrect] Two Notes have been added to this SR. The first Note applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative controls. - Allowing verification by ; administrative controls is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for. ALARA reasons. Therefore, t , probability of misalignment of these ;;';;ronce they have

                        'been verified to be in the proper position', is low.                    ,

IV A second Note has been included to clarify that SCIVs t at are cpen under administrative controls are not re meet the SR during the time the uhes are open. quired to h' SR 3.6.4.2.2 l O h(f Verifying that the isolation time of each power operated and each automatic SCIV is within limits is required to demonstrate OPERABILITY. The isolation time test ensures OSC /V" equal to'that assumed in the safety analyses.that the_pe6me w The 4se!ation ti== rd Frequency of this SR arc lin ecc;rd;nce with th; p:>5

              )   -

Imer'; ice Testirig Tregier.;r 02 d:y:@. ) i ghye SR 3.6.4.2.3

 *$                     Verifying that each automatic SCIV closes on a secondary jo        4            containment isolation signal is required to prevent akage M 5],.P 3              of radioactive material from isecondaryF containmen following a DBA or other accidents. This SR ensures at yf dqu                 ea.citautomatic SCIV will actuate to the isolation position 1

rbn alsecondaryFcontainment isolation _ signal. The LOGI SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this SR to rovide complete testing of the safety function. The i 18} month Frequency is based on the need to perfom this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. ' Operating experience has shown these components usually pass (continued) i BWR/4 STS B 3.6-107 Rev. O, 09/28/92

f

                    --                                                                 SCIVs B 3.6.4.2 BASES SURVEILLANCE   SR     3.6.4.2.3 (continued)                     %

REQUIREMENTS the Surveillance when perfonned at the (18)(month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. I REFERENCES 1. FSAR, Section((15.1.r] . 1

2. ' 3-FSAR,Section((15.1.il].

           ,g9
           \     3.       Fm. S=tw < t +           % ,ca Repu,ameas
                ?                                  Q Muu>L. ,

M 8 O l l 9 ewR,4 515 ,3.e.1,, ,,,. ,, ,,,,,,,,

1 l l

  $ M5 N h EL 4,NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

I V

j

O

1 i SGT System j_ B 3.6.4.3 O B 3.6 CONTAINMENT SYSTEMS i V l B 3.6.4.3 Standby Gas Treatment (SGT) System l BASES I l l BACKGROUND The SGT System is required by 10 CFR 50, Appendix A. GDC 41,

                            " Containment Atmosphere Cleanup" (Ref.1). The function of the SGT S" stem is to ensure that radioactive materials th leak from the primary containment into the fsecondar containment following a Design Basis Accident (DBA)          are yk filtered and adsorbed prior to exhausting to the                        !

environment. - UgjlSM The SGT Syste onsists of two fully redundant subsystems, ) W dg each with its own set of d=t=rk, dampers, charcoal filter train, and controls y

                   'I     Each charcoal filter train consists of (components listed in pf                order of the direction of the air flow):

! a. A demister or moi"ure separator; (80

b. An electric heater; d

c. A prefilter; 1

d. A high efficiency particulate air (HEPA) filter; hU e.

A charcoal adsorber{f[JA/5E/27 5)

f. A LCond.HEPA filter: and namt van w Wu+ 1sdesv5Ms awD g centrifugal ~ fang , p, t (faw L wks%ns) '

ThesizingoftheSGTSyste[equipmentandcompon is based on the results of an infiltration analysis as well as an exfiltration analysis of the fseconda kcontainment. The internal pressure of the SGT Syste oundary region is maintained at a negative pressure of .25).incheswater gauge when the system is in operation, which represents the internal air pressure from the buildingrequired when ex to ensure zero exfiltration of )f g2) g q -blu ing at an w !e of [40 " posed to buil t: th: a 510]e i g<sph wind , l The demister is provided to remove entrained water in the air, while the electric heater reduces the relative humidity (continued) i

. g

SGT System B 3.6.4.3 BASES 1 (.O @ 'd 3d BACKGROUND oftheairstreamtolessthan[f/0}%(Re 2$. The prefilter removes large particulate matter, while he HEPA filter (continued) i removes fine particulate matter and protects the charcoal from fouling. The charcoal adsorber removey gaseous ! elemental iodine and organic iodides, and the final HEPA l filter collects any carbon fines exhaus e from the charcoal adsorber. 9, 5 g gruS I " The SGT Systed(automatically starty and operated in response h-) Og2 to actuation signals indicative of conditions or an accident that could require peration of the system. Following initiation, charcoal filter train fans start. Upon verification that subsystems are operating, the redundant4 subsystem is n ally shut down. f &uwed) W tc92+cd) APPLICABLE he design basis for the GT Syste s to mitigate the SAFETY ANALYSES consequences of a loss of coolant accident and fuel handling accidents (Ret. 21 For all events analyzed, the SGT System i 15 shown to be aubomatically initiated to reduce, via

    )
                 #       filtration and ads       tion, the radioactive material released
               .         to the environment.        g3]                  gut The SGT System satisfies Criterion 3 of the NRC Policy z

6) \, k h

/@

LCO ns%c)I Following a DBA, a minimum [of h SGTtssubsyste required py to majntagjthesecondaryl containment at a negative C ~ pressure with respect to the environment and to process (f. gaseous releases. Meeting the LCO requirements for t w , OPERABLE subsystems ensures operation of at Ica:t 01: SGT M5663 subsystemintheeventofasinglefactivefailure/ 2 ,

Th'

(II3SETtc c) [mimmw nou% d\ f LocA %yp APPLICABILITY In MODES 1, 2, and 3, a BRA could lea to a fission product release to primary containment that leaks to secondary SGT System OPERABILITY 4e required containment. during these MODES.Therefore,hodif / ud 2.

                       .In MODES 4 and 5, the probability and consequences of Mese                  ,

events are reduced due to the pressure and temperature 1 o, bOM limitations in these MODES. Therefore, maintaining the SGT

1 (continued) BWR/4 STS B 3.6-110 Rev. O, 09/28/92

[) V INSERT A for orocosed BASES B 3.6.4.3 lg for Unit 1 subsystems and one charcoal adsorber for Unit 2 subsystems INSERT Al for orocosed BASES B 3.6.4.3 lg The Unit 1 SGT subsystems' ductwork is separate from the inlet to the filter to the discharge of the fan. The rest of the ductwork is common. The Unit 2 SGT subsystems' ductwork is separate except for the suction from the drywell and torus, which is common (however, this suction path is not required for A subsystem OPERABILITY). INSERT B for proposed BASES B 3.6.4.3 In addition, with secondary containment in modified configurations, the SGT lg System valves to excluded zona's)are not included as part of SGT System OPERABILITY (i.e., the valves may be secured closed and are not required to open on an actuation signal). INSERT C for orocosed BASES B 3.6.4.3 1 The required number of SGT subsystems is dependent on the configuration s required to meet LC0 3.6.4.1, " Secondary Containment." For secondary containment OPERABILITY consisting of all three zones, the required number of SGT subsystems is four. With secondary containment OPERABILITY consisting of [ one reactor building and the common refueling floor zones, the required number of SGT subsystem is three. Allowed configurations and associated SGT subs

m. S). ystem requirements are detailed in the Technical Requirements Manual (R WL*

SGT System i B 3.6.4.3 o f BASES APPLICABILITY System in OPERABLE status is not required in MODE 4 or 5, (continued) except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (0PDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the {secondaryK containment.

              \    ~

2 ACTIONS f f #'# eg m air S With on SGT subsystem inoperabl , inoperable subsystem must be restored to OPERABL tatus, M 6yt. In this r ondition, the remaining BLE SGT subsystem3 equat 3 g- to perform the required radioactivity release control ,, function. However, the overall system reliability is one o reduced because a single failure inithet0PERABLE subsyste could result in the radioactivity release control function

                         !  not being adequately performed. The7dayCompletionTimO based on consideration of such fa ors as the                             i ailability of the OPERABLE redund nt SGT Sy + r and thererop'e      N # "d low                                                      period.

O v C@ probability of a DBA occurring

                                                                      *d 30 during thi SubsyS$eh lHSGTW gel and 9'.2                  ( 6 iHg C.             -

If the SGT subsystem cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, or 3, 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 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power co.ditions in an orderly manner and without challenging plexe systems. HWEM Y (BtiI6 I g.1. C.2.1. f.2.2. and 7.2.3 Y& N '" * ' ' ' ^'9 D D D .h ref aire d During movement of irradiated fuel assembliesp in the secondary) containment, durin CORE ALTERATIONS, or durin OPDRVs, when Required Action .ircannot be completed w ,i the required Completion Time, OPERABLE SGT subsystr.m.

      'h)                  should immediately be placed in operation. This action ensures that the remaining subsyste OPERABLE, that no failures that could prevent automatic ctuation have
                                                                   .          are f                                                                  Ib

{)/ (continued) BWR/4 STS B 3.6-111 Rev. O, 09/28/92

SGT System B 3.6.4.3 BASES D D D ACTIONS 3 E'.1. E'.2.1. f.2.2. and I 2.3 (continued) occurred, and that any other failure would be readily detected. An alternative to Required Action is to immediately suspend activities that represent a potential for releasing

                   / radioactive material to the >[ secondary}< containment, thus placing the plant in a condition that minimizes risk.

If applicable, CORE ALTERATIONS and moverrent of irradiated fuel assemblies must imediately be suspended. Suspension of these activities must not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs sus ed. 1 __ The Required Actions of Co it on ave been modified by a Note stating that LCO 3.0.3 is not applicable. If moving

         ,--            irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 f                     would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is
 @h'6/g'                independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel 7tKN7 p            assemblies would not be a sufficient reason to require a reactor shutdown.

O ( _p . A.1. 2. and ff. gg WhentwoMGTsubsystemsareinoperable,ifapplicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in

                    ;{secondaryF containment must imediately be suspended.

Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. tions must continue until OPDRVs are suspended. Required ion has been modified by a Note statin that LCO 3.0.3 is not applicable. If moving irradiated fue assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify iny action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor (continued) g BWR/4 STS B 3.6-112 Rev. O, 09/28/92 t

V INSERT Bllla for Required Action A.1 Additionally, the 30 day Completion Time of Required Action A.1 is based on three remaining OPERABLE SGT subsystems, of which two are Unit 2 subsystems, and the secondary containment volume in the Unit I reactor building being open to the common provide refueling rapid drawdown floor where of vacuum. the two/ Testing h as shown that in thisUnit 2 SGT subsystems configuration, even with an additional single failure (which is not necessary to assume while in ACTIONS) the secondary containment volume may be drawn to a vacuum in the time required to support assumptions of analyses. INSERT Billb for Required Action C.1 VI QV In the event that a Unit SGT subsystem is the one not restored to OPERABLE status as required by Required Action A.1 or B.1, and: 1. AllthreegegarerequiredforsecondarycontainmentOPERABILITY;and

2. UnithisshutdownwithitsTechnicalSpecificationsnotrequiring secondary containment OPERABILITY (i.e., not handling irradiated fuel, A performing CORE ALTERATIONS, or conducting OPDRV), vs U 2.

Ui s} operation of Unithgv2.can continue provided that the Unitbreac@ tor build zone is isolated from the remainder of secondary containment and the SGT System. In this modified secondary containment configuration, only three SGT subsystems are required to be OP BLE to meet LCO 3.6.4.3, and no limitation is applied to the inoperable Uni SGT subs tem. This in effect is an alter tive ' restoring the inop able Unit SGT subsystem, i.e., shut down Unit aqc , ate its reactor building rom secondary containment and SGT tem. s. e _

                                   ~
                            ')I
                             \

Q 02 INSERT D for proposed BASES B 3.6.4.3 L.1 If two or more required SGT subsystems are inoperable in MODE 1, 2 or 3, the Unit 1 and Unit 2 SGT Systems may not be capable of supporting the required radioa'ctivity release control function. Therefore, LC0 3.0.3 must be entered immediately. O

SGT System l B 3.6.4.3 BASES ACTIONS A.1 .3 (continued) b-6 operatior.s. Therefore, in either case, inability to suspend j movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. o $b re vired Und I aaJ Un3 &

3. 3.1 SURVEILLANCE REQUIREMENTS SR k.2 Operating each SGT su system for a {10] continuous hours y J-l ensuresthat-[heh. d.:y:P re 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. Operation seir,t;'

Op2 /+w:ith the heaters on (eute;. tic h;;ter cyclir.; t:;mtMKfor af eliminates ufoisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system. SR 3.6.4.3.2 This SR verifies that the required SGT filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). Th: SCT Sy:t;; filter tert: :re i- i eccord=ce with Regel:t:ry Cuid: 1.52 ("ef. 3). The VFTP l f ' y) includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical , properties of the activated charcoal (general use and ' following specific operations). Specific test frequencies i and additional information are discussed in detail in the ' VFTP.

p. -2 )

SR 3.6.4.3.3 f@ l//rc. Wj l l This SR verifies that eactp/ SGT subsystem starts on receipt I of an actual or simulated initiation signal.gWhile this 7 Surveillance can be performed with the reactor at power, operating experience has shown that these components usu 11y , mi pass the Surveillance when )erfomed at the T18Fmont M 1 94 ' Freque M The LOGIC 5YSTE1 FUNCTIONAL TE5T in SR 3.3.5.Z h (overlaps this SR to provide complete testing of the safety f (continue BWR/4 STS B 3.6-113 Rev. O, 09/28/92

SGT System 7 B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.3 (continued) REQUIREMENTS _ ___ dg 9 Nu ion Therefore, the Frequency was found to be r ge i t acceptable from a reliability standpoint. f(p.e __

                 %   ~ SR             3.6.4.3.4                                                          #

This SR veri at the filter cooler bypas a opened and the fan s . This ens the j 9 ventilation mode of SGT Sys ton is available, p,)7 While this Surveillanc power, operatin e per ith the reactor at rience has shown th e components usually e Surveillance when performed a [1 nth Frequency, which is based on the refueling le. I erefore, the Frequency was found to be acceptable from a i

                /    -

reliability standpoint. N l l REFERENCES 1 . 10 CFR 50, Appendix A, GDC 41. t uzonH S 2. 1 FSAR, Section((S.2.3] . 5*3 1 l

2. .:

[\s.%) -,

                                       ":g m t;.,

s

.sc, t . [2].

                   - 3, (U^h t 2] Fs A 2, se e hon. C.2 3.
                                          '~hlon/y                                                         l 1

i 9 O BWR/4 STS B 3.6-114 Rev. O, 09/28/92

,\ INSERT REF Ul d Technical Requirements Manual U2 6 U2g FSAR, Section 15.1.41.

       ' Y
      'l ,- g NRC No. 93-102, " Final Policy statement on Technical Specification Improvements," July 23, 1993.

O O

h/MCREC}LSystem B 3.7.4 BASES (continued) ..M 4 2) ' V / t APPLICABLE Theabilityofthe4MCRECKSystemtomaintainit,he f[un SeM er2d1> SAFETY ANALYSES habitability of the control room is an explicit' asumntinn (unit 2.) . for afety nalyses resented in the FSAR,(Chapters 46h)) _ ... En" J15h (Ref s and espectively). The pressurization gip 4J mode of the>{MCREC). System is assumed to operate followi 3 a loss of coolant accident, fuel handling accident, main steam i

                                                                  !                                                                                              line break, and control rod drop acci           t, as discussed in (u mr 1) ....'                                                                                                                                                    e FSAR, Sectionr {6.4.1.2.2P (Ref. ) 5 The radiological doses to control room personnel as a result of the various unit i ....

f0N DBAs are summarized in Reference QNo single active or passive failure will cause the loss of outside or y opbl recirculated ir from the control room. The/-[HCRECK System satisfies Criterion 3 of the NRC Policy Staten.an

                                                                                                                                                                                   .Cac.70 F4                                                              l l

, LC0 Tworedundantsubsystemsofthe#fMCRECFSystemarerequired I to be OPFRABLE to ensure that at least one is available, assuming a single failure disables the other subsystem. l l Total system failure could result in exceeding a dose of l 5 rem to the control room operators in the event of a DBA. [ TheRMCRECfSystemisconsideredOPERABLEwhenthe V individual components necessary to control operator exposure are OPERABLE in both subsystems. A subsystem is considered OPERABLE when.its associated: I

a. Fan is OPERABLE; l

b. HEPA filter and charcoal adsorbers are not excessively restricting flow and are capable of perfonning their filtrat_ ion functions; and

[h ssocial<d)" 4' - " + a r ductwork, valves, and dampers are [h ,y c. OPERABLE, and air circulation can be maintained. In addition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, j ductwork, and access doors l I D,userr8el t pl . O V (continued) BWR/4 STS B 3.7-19 Rev. O, 09/28/92

                                                                              /MCREC]LSystem B 3.7.4 BASES   (continued)

APPLICABILITY ' In MODES 1, 2, and 3, tha dMCRECK System must be OPERABLE to control operator exposu'e iuring and following a DBA, since the DBA could lead to a fission product release. In MODES 4 and 5, the probability and consequences of a DBA f ,k are reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaining the i 4MCRECh. System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated: C t. During operations with potential for draining the reactor vessel (OPDRVs); W'

b. During CORE ALTERATIONS; and e.* During movement of irradia_ted fuel assemblies in the 9 I r{ secondary contai nment 6Nns irradiaM fwi ssse aattu sa .

(-Ha stewwtuV emisinmf lJo ocarin I,1, insyj3 Moors an y ACTIONS A.1 h With one J[MCRECPsubsystem inoperable, the inoperable sfMCRECP subsystem must be restored to OPERABLE status within l 7 days. With the unit in this condition, the remaining

                      ; OPERABLE 4MCREC}t subsystem is adequate to perforn control
                      ! room radiation protection.         However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced MCRECk System capability.
             /f.             The 7 day Completion Time is base on the low probability of a DBA occurring during this time period, and that the                ,

f remaining subsystem can provide the required capabilities. ' B.1 end 8.2 p) f In MODE 1, 2, or 3, if the inoperable 4MCRECK subsystem i a cannot be restored to OPERABLE status within the associated 1 Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours and in MODE 4 i within 36 hours. The allowed Completion Times are reascnable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. (continued) O BWR/4 STS B 3.7-20 Rev. O, 09/28/92

! i i j{tlCRECP-System B 3.7.4 ' J s (

    ']      BASES ACTIONS                                                                           C.I. C.2.1. C.2.2. and C.2.3                                            ~

(continued) The Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown. zh_ During movement of irradiated fuel assemblies in the ifsecondary}4 containment, during CORE ALTERATIONS, or during l 0PDRVs, if the inoperable r{MCREC}t subsystem cannot be

                                                                                         -    restored to OPERABLE status within the required Completion i Time, the OPERABLEdMCRECP subsystem may be placed in the pressurization mode. This action ensures that the remaining l                                                                                              subsystem is OPERABLE, that no failures that would prevent l                                                                                              automatic actuation m will be readily detected. g r , and that any active failure l

Q oema Requ' red tion 1 is mo 'fied by Note al ting e [ place the syst m in the toxic ga protec ion O (_) (h\ N pera or t de i in erab e] . the oxic g automa 'c trans r capabi *ty is I An alternative to Required Action C.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control i room. This places the unit in a condition that minimizes risk. l Ok I If applicable, CORE ALTERATIONS an. movement of irradiated fuelassembliesintheTsecondaryfcontainmentmustbe k suspended immediately. Suspension of these activities shall not preclude completion of movement of a com onent tolsafe ' position. Also, if applicable, action must be initiated I immediately to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission 1 I product release. Actions must continue until the OPDRVs are l suspended. D.1 i If both (MCRECf subsystems are inoperable in MODE 1, 2, g l, . I or 3, the [MCREC) System may not be capable of performing , ( (continued) l BWR/4 STS B 3.7-21 Rev. O, 09/28/92 l l

l 4 JHCRECPSystem B 3.7.4 BASES ACTIONS Q.J (continued) the intended function and the unit is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be entered imediately. E.1. E.2. and E.3 The Required Actions of Condition E are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not suffi ient reason to require a reactor shutdown. p 2, During moveme3t o irradiated fuel assemblies in the fsecondarylA s tonta OPDRVs, with two((inment, during CORE MCREC)(subsystems inoperable, ALTERATIONS, action must or during be taken imediately to suspend activities that present a potential for releasing radioactivity that might require k isolation of the control room. This. places the unit in a condition that minimizes risk. g If applicable, CORE ALTERATIONS nd movement of irradiated fuel assemblies in theTsecondaryF containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe If applicable, actionffnust be initiated gV jposition. immediately to suspend OPDVRs to minimize the probability af a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended. SURVEILLANCE SR 3.7.4.1 REQUIREMENTS This SR verifies that a subsystem in a standby mode starts on demand and continues to operate. Standby systems should be checked periodically to ensure that they start and function properly. As the environmental and normal operating conditions of this system are not severe, testing each subsystem on this system. once M=tMy every h %eateepe-et provides 4 =an drier adequate check cut ry--- p,p ( _ (continued) BWR/4 STS B 3.7-22 Rev. O, 09/28/92

p{ Control Room ACP System B 3.7.5 O 8 3.7 PLANT SYSTEMS (G B 3.7.5 [ Control Room Air Conditioning (AC)[Syste gg BASES feochm s( SA MNs c4ol Room EnvirenunM [( herenRee referr44 fo as % Gnfol Reins 4C Sysiwij f ThehontrolRoomAC[*Systemlprovidestemperaturecontrol BACKGRCUND I for the control room following isolation of the control l fp 9\ %<a. E ThedontrolRoomAC System consists of M independent, 9% teraciNbWdrt subsystems that provide cooling and heating of recirculated control room air. Each subsystem consists of

wak<n cood O'S ha . u s w . i n coolin_gcoils, fans,fulia% compressors,h (unik (tfri*#8+d / ductwork, pr dampers, and instrumentation and con ol s to_ . / , ura enntral W c*5mf M'/5 g.t]ide for control room temperat(iiceWe coelo4walv ' frem nu flus / The ControlRoomACfSystemisdisignedtoproiu h) controlled environment under bo normal and accident 1 conditions. N i g'^ subsyste 5 rovidefthe required temperature control to maintain a suitable [ control room g -- N.g environment for a sustained occupancy of 4e persons. The 52.-WEO j design conditions for the control room environment are T@F N and 70's relative humidity. The t(Control Room ACF stem (Q Mt i operation in maintaining the control room temperatur 's discussed in thyFSAR, Sectiondfrr4} (Ref.1). OM ' '

                           -- @yLQ                . -

[5 4W and 6) APPLICABLE CB/ ThedesignbasisofthedontrolRoomACfS'ystemisto qg SAFETY ANALYSES maintain the control room temperature for a 30 day continuous occupancy. ystem components are arranged in

                        ~
    %* 60*l*CQf"'INq         3
                                 ' The    iTontrol n h i.nt    safet Roomed        AC               subsystems. During emergency operation,the$yrntrol Room AC}^ System maintains a habitable environment and ensures the OPERABILITY of components in the                                                                                                                  A single failure of a l

component of the $ control room.ontrol Room ACD System, assumi of offsite power, does not impair the ability of the system to perform its design function. Redundant detectors and controls are provided for control room temperature control. i TherfControl Room ACf System is designed in accordance with I hp Seismic Category I requirements. Thet{ Control Room ACK ! System is capable of removing sensible and latent heat loads from the control room, including consideration of equipment l . /O (continued) V BWR/4 STS B 3.7-25 Rev. O, 09/28/92 l i l i

                                                                 /ControlRoomACPSystem B 3.7.5 BASES APPLICABLE            heat loads and personnel occupancy requirements to' ensure SAFETY ANALYSES       equipment OPERABILITY.

(continued) he hontrol Room AC[ System satisfies Criterion 3 of the NRC Policy Statemen . (CRef.4 IE _ [% & jSb *lo capaciFV) LCO M independent and .. r t = subsystems of the Control System are required to be OPERABLE to ensure that 7 g _at leas;( 6 Room ACi .3 available, uming a single failure p(g disab g the eeef subsyste Total system failure could resuit in the equipment operating temperature exceeding limits. ThefontrolRoomAC[SystemisconsideredOPERABLEwhenthe Walen. cooled Codin$ut individual components necessary to maintain the control room yfub/thrieeaant s temperature are OPERABLE in both subsystems. These components include the cooling coils, fans, c""'et' Ucompressors, ductwork, dampers, and associated instrumentationandcontrolsf APPLICABILITY In MODE 1, 2, or 3, the a ontrol Room Aql System must be O OPERABLE to ensure that the control room temperature will not exceed equipment OPERABILITY limits following control room isolation. In MODES 4 and 5, the probability and consequences of a Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the JControl Room AC)( System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:

4. During operations with a potential for draining the q

C- reactor vessel (0PDRVs); l Qu

          $g             b. During CORE ALTERATIONS; and                                    i gI    During movement of irradiated fuel assemblies in the ysecondary3A containment.rpns,is ina/;d/M muk/ea                f ca., m >ews.n crnhinndm aho ocwe                     j f                                f e - m ,, ,, a n f)              -

i (continued) BWR/4 STS B 3.7-26 Rev. O, 09/28/92

V

                                                                                  ')ControlRoomAC}a. System B 3.7.5 M&t ov&sik air lwycalvvb BASES     (continued)                                              h.h n,&g m deUmiM b *f
                                    ,                                              p r a m 4.i a n.2                   )

p, ACTIONS J_rd p, p N 1J With one control room AC.P subsystem inoperable <, the

      ' W5 G LT B 2'l 4) inoperable control room ACh subsystem must be restored to                               j p.p                 OPERABLE status within 30 days. With the unit in thi                                    ;

_ _ condition, the remaining OPERABLEJcontrol room ACFgFM 1 ps ar(J subsystem 4 adequate to perform the control room air conditioning reduced because function. However, a single failure the overallsubsystem in t.WOPERABLE reliability is g could result in loss of the control room air conditioning function. The 30 day Completion Time is based on the low probability of an event occurring requiring control room isolation, the consideration that the remaining subsysteh

            .\S                   can provide the required protectione 3rd th: :nildi'ita                          f l t:rr.:t: ::f:+y =ad aaa w a+y "^'iS --+hadt r@

h95e%7 621 (

                                 #*I " 
                                                     .]               ithamAsqus*redArSak f In MODE 1, 2, or 3,                   ' t'e 'r;:rdl        [rentr:1 rn;; ,'C]

P seeyt t;... m,...m L r;;;;r:d te ^"EP'"LE :t:: ; ith', th: OmAdm associated Completion Timu the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit O 'B e dn drnef ~ i must be placed in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. 5 E E G f.1. f.2.1. f.2.2. and /.2.3 The Required Actions of Condition 2 are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown. ,

               }                 During movemeyM irradiated fuel assemblies in the (secondaryPcontainment, during CORE ALTERATIONS, or during .p,7 OPDRVs, i f "r "'4 *=d 4:ti:r .'..! := .:t be cr ?'eted ithi ;

4 fysst-t 821 c.} r tb: r:r'~d Cx;10ti:r 'i=, t5: OPERABLE fcontrol room AC subsysthay be placed imediately in operation. ifThis action ensures that the remaining subsystem <i.s 0PERABLE, , h lNSEET' 6Tl 1 [y5 Ar4D l (continued) BWR/4 STS B 3.7-27 Rev. O, 09/28/92

JControlRoomA@ System B 3.7.5 O4k < BASES e e e e ACTIONS Y gl . E2.1. E2.2. and /.2.3 (continued) f that no failures that would prevent actuation will occur, and that any active failure will be re 'ly detected. E P& An alternative to Required Action p.1 to imediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. Ifapplicable,COREALTyETIONS an movement of irradiated fuel assemblies in the 1 secondary containment must be suspended imediately. Suspension of these activities shall not preclude completion of movement of a component to a safp l7 position. Also, if applicable, actionghnust be initiated imediately to suspend OPDRVs to minimize the probability o (F a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are sus ended. 8.1 If fcontrol room AC ubsystems are inoperable in MODE 1, 2, or 3, the d ontrol Room AC) System may not be capable of performing the intended function. Therefore. LCO 3.0.3 must be entered imediately. G 4 G I.1. f.2. and E'.3 The Required Actions of Condition f are modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown. , During movemen iof irradiated fuel assemblies in the

                 >{ secondary     nthinment, during CORE ALTERATIONS, or during k                  OPDRVs, withetwWontrol room ACk subsystems inoperable, action must be taken imediately to suspend activities that b [ present a potential for releasing radioactivity that might M                                                                           ,

(continued) BWR/4 STS B 3.7-28 Rev. O, 09/28/92 i i

ControlRoomA[ System B 3.7.5 BASES 6 6 ACTIONS q.f.1.f2.andf.3 (continued) require isolation of the control room. This places the unit . in a condition g minimizes risk. I If applicable ORE ALTERATIONS and handling of irradiated A fuel in the econdaryf containment must be suspended S immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended. SURVEILLANCE SR 3.7.5.1 g [,,9 4 p g REQUIREMENTS This SR verifies that the heat remova capability of_the system is sufficient to remove thefssumad] heat load"in the Lcontrol roond. The SR consists of a combination of testing O [S and calculation. Ther{18 nth Frequency is appropriate since significant degradatio of the ontrol Room AC}<,

                                                                                           ,3 System is not expected over his time period.
                                        ,           .            hb Funi+ TJ    ..Q4n14 / oNW)

REFERENCES f 1. / FSAR, Section4M. [T G.y and 9 4.I) I 9' ( f2 INSEfLTB29 l BWR/4 STS B 3.7-29 Rev. O, 09/28/92

Main Condenser Offgas B 3.7.6 B 3.7 PLANT SYSTEMS B 3.7.6 Main Condenser Offgas BASES BACKGROUND During unit operation, steam from the low pressure turbine is exhausted directly into the condenser. Air and noncondensible gases are collected in the condenser, then exhausted through the steam jet air ejectors (SJAEs) to the Main Condenser Offgas System. The offgas from the main condenser normally includes radioactive gases. The Main Condenser Offgas System has been incorporated into the unit design to reduce the gaseous radwaste emission. This system uses a catalytic recombiner to recombine radiolytically dissociated hydrogen and oxygen. The gaseous mixture is cooled by the offgas condenser; the water and cendensibles are stripped out by the offgas condenser and moistu a separator. The radioactivity of the remaining gaseous mixture (i.e., the offgas recombiner effluent) is monitored downstream of the moisture separator prior to entering the holdup line.

                                                                      , ..pn3H)            -

19 4 and Armda E ] APPLICABLE The main condenser offgas gross gamma activity rate is an SAFETY ANALYSES initial condition of the Main Condenser Offga System Cs 11.3 aad failure (Ref.1).event, discussed The analysis in the FSAR, assumes a grossSection%15.1.35k failure in the Main ...J,gg g/ Condenser Offgas System that results in the rupture of the Main Condenser Offgas System pressure boundary. The gross gamma activity rate is controlled to ensure that, during the event. the calculated offsite doses will be well within the limits of 10 CFR 100 (Ref. 2),= th; =C 3 tuff ;ppnv;d-(hma.n3uom. The main condenser offgas limits satisfy Criterion 2 of the NRC Policy Statemen . -- (,Ref.*O hQ v LCO To ensure compliance with the assumptions of the Main Condenser Offgas System failure event (Ref. 1), the fission product release rate should be consistent with a noble gas release to the reactor coolant of 100 Ci/MWt-second after decay of 30 minutes. The LC0 is established consistent with (continued) BWR/4 STS B 3.7-30 Rev. O, 09/28/92

4 INSERT 1,G90_1 U1 Version In addition, since some components required by Unit 1 are powered from Unit 2 sources (i.e., Stancby Gas Treatment (SGT) System), one qualified circuit between the offsite transmission network and the onsite Unit 2 Class 1E Distribution System, and one Unit 2 DG (2A or 20), capable of supplying power to one required Unit 2 SGT subsystem, l/h must also be OPERABLE. C) v V HATCH UNIT 1 B 3.8

 -INSERT   LCO 2 U2 Version O                 to the 2F ESF bus are required to be OPERABLE; however, only one feeder breaker per bus to the 2E and 2G ESF buses is required to be OPERABLE, but they must be from different SATs (e.g., 2E feeder breaker from the 2C SAT and the 2G feeder breaker from the 20 SAT). With 2E and 2G ESF buses both fed from one SAT- (normal line up is both buses fed from      -

2D SAT), both feeder brer.kers to each of these ESF buses are required to be OPERABLE. The Unit I offsite circuit also consists of the incoming breaker and disconnect to the 4.16-kV ESF buses required to be OPERABLE to provide power to the Unit 1 equipment required by LC0 3.6.4.3, LC0 3.7.4, and LC0 l Jk 3.7.5. i e 4 e O O HATCH UNIT JL B 3.8

i i AC Sources-Shutdown B 3.8.2 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources-Shutdown BASES BACKGROUND A description of the AC sources is provided #n the Bases for LCO 3.8.1, "AC Sources-Operating." APPLICABLE The OPERABILITY of the minimum AC sources during MODES 4 SAFETY ANALYSES and5yensuresthat: e m .t

*[

e

a. The facility can be maintained in the shutdown or m bH refueling condition for extended periods; Q c. c < J <s b. Sufficient instrumentation and control capability is a b su " 1

          " A-u l       available for monitoring and maintaining the unit
                            )          status; and d

c. Adequate AC electrical power is provided to mitigate g~

events postulated during shutdown, such as an W inadvertent draindown of the vessel or a fuel handling accident. In general, when the unit is shut down the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or loss of all onsite power is not required. The rationale for this is based on the fact that many Design Basis Accidents (DBAs), which are analyzed in MODE , 2, and 3, have no specific analyses in

                 -               MODES 4 and 5        st case bounding events are deemed not Op qb ' g.p,3
                             ~

credible in1 0 4 and 5 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and corresponding stresses result in the probabilities of occurrences si eliminated, and minimal consequences. gnificantly These reduced deviations from or DBA analysis assumptions and design requirements during shutcown conditions are allowed by the LCO for required systems. During MODES 1, 2, and 3, various deviations from the analysis assumptions and design requirements are allowed within the ACTIONS. This allowance is in recognition that

 /3
 \)

(continued) i BWR/4 STS B 3.8-35 Rev. O, 09/28/92 ! i j

AC Sources-Shutdown B 3.8.2 s BASES APPLICABLE certain testing and maintenance activities must be SAFETY ANALYSES conducted, provided an acceptable level of risk is not (continued) exceeded. During MODES 4 and 5, performance of a significant number of required testing and maintenance activities is also required. In MODES 4 and 5, the activities are generally planned and administratively controlled. Relaxations from typical MODES 1, 2, and 3 LCO requirements are acceptable during shutdown MODES, based on:

a. The fact that time in an outage is limited. This is a risk prudent goal as well as a utility economic consideration.

l

b. Requiring appropriate compensatory measures for certain conditions. These may include administrative controls, reliance on systems that do not necessarily meet typical design requirements applied to systems credited in operation MODE analyses, or both,

c. Prudent utility consideration of the risk associated with multiple activities that could affect multiple g systems,

d. Maintaining, to the extent practical, the ability to perform required functions (even if not meeting MODES 1, 2, and 3 OPERABILITY requirements) with systems assumed to function during an event.

f In the event of an accident during shutdown, this LCO , ensures the capability of supporting systems necessary for ! avoiding immediate difficulty, assuming either a loss of all offsite power or a less of all onsite (diesel generator (DG)) power. The AC sources satisfy Criterion 3 of the NRC Policy Statemen Q ReS.l[ Q LCO ui Onefoffsite circuit capable of supplying th te Class 1E l us t( power distribution subsystem (s) of LCO 3.8.h, " Distribution Of2, go Systems-Shutdown," ensures that all required loads are S ' ,"

                                .                               powered from offsite power. An OPERABLErDG, associated with

} a Distribution System Engineered Safety Feature (ESF) bus required OPERABLE by LCO 3.8. , ensures power source is available for providing lthat apower lectrical diverse

                                                                                                                                                                                                                      , u. o          g 6                                                                                     W g                                                u... L (go_ inued)

BWR/4 STS B 3.8-36 Rev. O, 09/28/92

t INSERT LCO 3.8.2 do;41 version In addition, some components that may be required by Unit I ' are powered from Unit 2 sources _(i.e., Standby Gas Treatment (SGT) System). Therefore, one qualified circuit between the offsite transmission network and the onsite Unit 2 Class IE Distribution System, and one Unit 2 DG capable of supplying power to one required Unit 2 SGT subsystem, must also be / jh ; OPERABLE. p O 0 S HATCH UNIT 1 B 3.8

AC Sources-Shutdown B 3.8.2 BASES j l 1 ACTIONS A.2.1. A.2.2. A.2.3. A.2.4. 8.1. B.2. B.3. and B.4 (continued) sufficient required D conservative actions is made. With noperable, the minimum required diversity of AC l k power sources is not available. It is, therefore, required to suspend CORE ALTERATIONS, movement of irradiated fuel

             }                  assembliesinthepecondary}Nontainment,andactivities that could result in inadvertent draining of the reactor vessel.

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to imediately initiate action to restore the required AC sources and to continue this action until restoration is accomplished in order to provide the necessary AC power to the plant safety systems. .

                              / Notwith tanding perfomance f the above conserva ve
      #                          Requi d Actions, the plan is still without suf cient A s' ~

pow sources to operate 'n a safe manner. Th efore, (7A 9 ac on must be initiatep to restore the minim required AC wer sources and cont /nue until the LCO re irements ar estored. -- The Completion Time of imediately is consistent with the required times for actions requiring prompt attention. The restoration of the required AC electrical power sources should be completed as quickly as possible in order to minimize the time during which the plant safety systems may 1 be without sufficient power. Pursuant to LCO 3.0.6, the Distribution System ACTIONS would not be entered even if all AC sources to it are inoperable, resulting in de-energization. Therefore, the Required ON3 C

                    -           Actions of Condition A have been modified by a Note to indicate that when Condition A is entered with no AC power 3 A_--

cy" td w e ESF bus, ACTIONS for LCO 3.8.}& must be immediately i L" entered. This Note allows Condition' A to provide l requirements for the loss of the offsite circuit whether or i

                               'not a%divisie='is de-energized. LCO 3.8.J6providesthe O6       appropriate restrictions for the situati involving a de-energized division?

A B ur (is) (continued) BWR/4 STS B 3.8-39 Rev. O, 09/28/92

l l l l l AC Sources-Shutdown i B 3.8.2 I I BASES (continued) l SURVEILLANCE SR 3.8.2.1 REQUIREMEN I6 .

                             .. P4           SR 3.8.2.1 requires the SRs from LCO 3.8.1/ha are                            l l                         My                  necessary for ensuring the OPERABILITY of/the AC sources in a tea co a J 7ther than MODES 1, 2, and D SR 3.8.1.57 is not required                                     l r<-< \ 6 bc.                 to be met because the required OPERABLE DG(s) is not y h     ;m        a          required to undergo periods of being synchronized to the-3 u . 9 1. y 5

offsite circuit. SR3.8.1.jpisexceptedbecausestarting Ogg m u .4 1 e l independence is not require withtheDG(s)thatisnot c ,,c. a m ,, srsa.4 / required to be OPERABLE. R fer to the corresponding Bases

.'

bc 0 MAS 4ha. for LCO 3.8.1 for a discuss'on f ea R.

M nb l @3 This SR is modified by a Note. The reason for the Note is l to preclude requiring the OPERABLE DG(s) from being paralleled with the offsite power network or otherwise l (nguorn0 6mcR4 - rendered inoperable during the performance of SRst With limited AC sources available, a sin b* g,9(s BwR 0 ;c7 event d the DcppItO is theG , l g,3. g w p 2.f,, compromise both the required circu intent that these SRs must still be ca able of being met. . m but actual performance is not requiregurin; periods "e *- p ,, t h the DC i: required te be OPEPABLER g se sa2.z b w REFERENCES

                                           &ne. l. 4367T ke r)
                                                                        , -d '               e- de-~p r~ C4t(,o re, wi<cd             V ELF %s 3

a or biccMecN* *- f* u dt C om,4 erca dur, p ,4- uc of SR-s.

~ ~

l l l BWR/4 STS B 3.8-40 Rev. O, 09/28/92 m

l l DC Sources-Shutdown B 3.8.5 - B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC Sources-Shutdown BASES BACKGROUND A description of the DC sources is provided in the Bases for LC0 3.8.4, "DC Sources-Operating."

                                                                 ~#9 h h os s . J c,h        "

APPLICABLE The initial conditions of Desig B i i dent _ ancL_.w.+ t SAFETY ANALYSES transien' nalyses in the FSAR,, hapter,{6 ~{Ref. 1) and e.ai

                     *d? '?hapte          (Ref. 2), assume th W eered Safety Feature systems      OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the diesel generators (DGs), emergency auxiliaries, and control and switching during all MODES of operation.

The OPERABILITY of the DC subsystems is consistent with the initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY. m w The OPERABILITY of the minimum DC electrical power sources M during MODES 4 andf5 ensures that: v a^ A d"'vDg a. The facility can be maintained'in the shutdown or e'M refueling condition for extended periods; w 's..h l Gd us < AM W b. Sufficient instrumentation and control capability is 9, A . . ) e d="M available for monitoring and maintaining the unit status; and

c. Adcquate DC electrical power 'is provided to mitigate events pastulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident.

The DC sources satisfy Criterion 3 of the NRC Policy ch, Zix u.5i h-LC0 Thep p"oNer subsystems-with: 1) each station g service DC subsystem consisting of two 125 V batteries in series, two battery chargers, and the corresponding control , equipment and interconnecting cabling; and 2) each DG DC l (continued) i BWR/4 STS B 3.8-59 Rev. O, 09/28/92 1

DC Sources-Shutdown i B 3.8.5 l r w BASES h 6 LCO subsystem consisting of e a ery bank, one battery (continued) charger, and the corres onding control equipment and interconnecting cabli -are required to be OPERABLE to L'd h support required DC stribution subsystems required OPERABLE by LCO 3.8. ,

                                                   " Distribution Systems-Shutdown."
                   ~

4This requirement ensures the availability of sufficient DC LMM electrical power sources to operate the unit in a safe manner and to mitigate the consequences of postulated events during shutdown reactor(e.g., fuel handling). accidents and inadvertent vessel draindown k APPLICABILITY The DC electrical power sources required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel assemblies in the secondary containment provide a surance that:

a. Required features to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core in case of an inadvertent draindown of the reactor vessel;

b. Required features needed to mitigate a fuel handling accident are available;

c. Required features necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The DC electrical power requirements for MODES 1, 2, and 3 are covered in LCO 3.8.4. ACTIONS A.1. A.2.1. A.2.2. A.2.3. and A.2.4 If more than one DC di n subsystem is required according to LCO 3.8.8I4r, the DC subsystems remaining OPERABLE with one or more DC power sources inoperable may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, fuel movement, and

~.-

(continued) BWR/4 STS B 3.8-60 Rev. O, 09/ES/92

t DC Sources-Shutdown B 3.8.5 ( ( +; BASES ACTIONS A.I. A.2.1. A.2.2. A.2.3. and A.2.4 (continued) operations with a potential for draining the reactor vessel. By allowance of the option to declare required features inoperable with associated DC power sources inoperable, appropriate restrictions are implemented in accordance with the affected system LCOs' ACTIONS. In many instances, this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative N- actions is made (i.e., to suspend CORE ALTERATIONS, movement _ of irradiated fuel assembliest, and any activities that could [i^% result in inadvertent draining of the reactor vessel). s ew&' (" # "" 1 Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of A postulated events. It is further required to imediately initiate action to restore the required DC electrical power subsystems and to continue this action until restoration is accomplished in order to provide the necessary DC electrical l power to the plant safety systems. I f M nding performance of the above conser avativ , Required Ac unit is still wi ut 1 Tficient DC j Q4 power sources to operate action must be in ji ' ate M o restore manner. Therefore, i

                                                                                                                                                               * ' mum required DC power sourcesc1rM continue until the LCO requi                                                                          ntstarej Qestored'.g The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The                                                                          ,

restoration of the required DC electrical power subsystems should be completed as quickly as possible~in order to minimize the time during which the plant safety systems may be without sufficient power. SURVEILLANCE SR 3.8.5.1 REQUIREMENTS SR 3.8.5.1 requires performance of all Surveillances

              /J     -

required by SR 3.8.4.1 through SR 3.8.4.8. Therefore, see the corresponding Bases for LCO 3.8.4 for a discussion of L << s each SR. TL T.S.5 l ) Or N f2, o C (continued) ! BWR/4 STS L" 4 B 3.8-61 Rev. O, 09/28/92 sg 3.6 5 1 U____--__-___--_-_-_-_--____---_----__---__---------------------------- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - --

I DC Sources-Shutdown B 3.8.5 REFERENCES 1. FSAR, 2.

                                         @                      l rS R. m ,1er gg...qq
               @) 5.@Er ec+

l O

   -                                                          O BWR/4 STS            B 3.8-62       Rev. O, 09/28/92 I

c,gnc72sc cwtrnE 1

                $W12-t$$ CG&

NOT SHOWH PAO Distribution Systems-Operatin g- B 3.8

                                                                                                                                                            'l i    ).                                                '
 '~'"                                                              y-                          ,~                               Ull
                                               /                JgbleB3.8.9,-1'(page1of17
                                   /                 AC ar#DC Electric &' Power DistritItion Systett#
                       /                             /                        /                     /               /                                     ,

TYPE [ VOLpE [IVISION jdVISION2]* AC',s(fety ,,[4160V] ,. ' [ESF Bus]'[NB01] ((ESF Bus] [NB02]

                                                      /           [480 V],/                    Load Centers                    Load C         rs t
          /                                        '
                                                                     /

[$80 V]

[NG01,NG03

                                                                                         / Motor Control Centers

[NG016('NG01I, [NGO , NG04] tor Control Centers [NG02A,NG02I, f I

                                                                              /,              NG018, NG03C,
                                                        /                 ,

NG031, NG03D NG028', NG04C, NG04I, NG04D] i,

                                                     /                  /                     /                                                    _
                                                 /
                                                   /                                        /Distrib ion                       Distributio'n                        i

[/120 V] / P is Panels

        /                                  ,/                 /                     f          [      ,NP03]                   [NP02, NPO4]

DC b'uses / [125 V] / Bus [NK01 rom Bus [NK02] from . j/ battery [NK11 d battery [NK12] and

                                                     -'                                      charger [NK21]

f charger [NK22] Bus [NK03] from . s [NK04]' from

                   /                                                                      battery [NK13] and
                 '                                                                                                      batt ry [NN14] and
                                                                                                                                                               <f charger [NK23]                  char          [NK24]

k,/ / ' Distribut' ion f

                                 '                                                                                           ,Distrib                    /

Panels , PanelsN/ (on [NK41, ,,NK43, NK51] <[NK42,NK44,NK,5

                      ,NACg tal.                                 [120 V]                                     from
                    ,          buse N                                                       Bus inverter[NN011,NN11]Bus
                                                                                                         , ,                inv

[NN02].

                                                                                                                                         ,   N12]

4

                /
                  /

connected to bu nnected to bus [ -[NK02]

          /                          /                         :                                      /
                                                            /                               Bus [, 03} 4 cm                 Bus /[NN04] from                              .

e ' inverter [NN13hQ nverter [NN14 l connected to bus ed to s I y- / p' ,/ [NK03] /,<coh 8 ([NKU4 l n- ,; -- 7 ,c f ~/ j

               '* / Eac)-{divisionk the AC And1C electrical                                       -

powet,distributjoMy' stem is

                 ,/ Vsubsysjem.                                -
               ~

J o Cgg) U ^AO3 y BWR/4 STS B 3.8-87 Rev. O, 09/28/92

Distribution Systems-Shutdor B 3.8 8 3.8 TRICAL POWER SYSTEMS B 3.8 D) Distribution Systems-Shutdown BASES BACKGROUND A description of the AC Cp and 8.C " ital b# electrical h power di ribution system is provided in the Bases for LCO 3.8 " Distribution Systems-Operating." APPLICABLE The initial conditions of Desi Basis cc a y! 4D arii SAFETY ANALYSES .transie . lyses in the FSAR,( ptefj[6 -(Ref.1) and u,.t 15 (Ref. 2), assume Engineere afety Feature u-.f i q F) systems are OPERABLE. The AC& DCrand AC vital buF "g

       /              ud?- electrical power distribution systems are designed to provide sufficient capacity, capability, redundancy, and               P7 reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System, and                 -

containment design limits are t exceeded. ,,m 4-4 The OPERABILITY of the AC DCp -d AC tital buf electrical % power distribution system is consistent with the initial ac.sumptions of the accident analyses and the requirements for the supported systems' OPERABILI The OPERABILITY of the minimum ACr' DC And ^C . ital busR_ electrical power scurces and associated power distribution subsystems during MODES 4 andf5 ensures that: e 'd a. The facility can be maintained in the shutdown or refueling condition for extended periods; Wd"b aswan.cs dm [(c + su I q rd b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent draindown of the vessel or a fuel handling accident.

The AC and DC electrical power distribution systems satisfy Criterion 3 of the NRC Policy Statement _S3A f31 O (continued) BWR/4 STS B 3.8-88 Rev. O, 09/28/92

l l l Distribution Systems-Shutdo B 3.8 , BASES (continued) LCO Various combinations of subsystems, equipment, and components are required OPERABLE by other LCOs, depending on the specific plant condition. Implicit in those requirements is the required OPERABILITY of necessary

        *'              support required features. This LCO explicitly requires W.A L          enercization of the cortions of thyelectrical distribution pgt            system necessary to support OPERABILITY of Technical                             ,

Specifications required systems, equipment, and

        " '             components-both specifically addressed by their own LCO, y                 pd implicitly required by the definition of OPERABILITY.

u.< A . , Leo %E Maintaining these portions of the distribution system

   \                    energized ensures the availability of sufficient power to operate the plant in a safe manner to mitigate the consequences of postulated events during shutdown (e.g.,

fuel handling accidents and inadvertent reactor vessel draindown). APPLICABILITY The AC and DC electrical power distribution subsystems required to be OPERABLE in MODES 4 and 5 and durji g movement of irradiated fuel assemblies in the pecondaryr containment A > provide assurance that: g

a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel in the core in case of an inadvertent draindown of the reactor vessel;

b. Systems needed to mitigate a fuel handling accident :

are available, j l

c. Systems necessary to mitigate the effects of events ;

that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdo condition or refueling condition.

4-4 The AC D , and A.C " ital bnS electrical power distribution OF7 'subsyste requirements for MODES 1, 2, and 3 are covered in LCO 3.8 . O (continued) BWR/4 STS B 3.8-89 Rev. O,09/28/92 j i i

Distribution Systems-Shutd B 3. . = BASES (continued) 9 ACTIONS A.I. A.2.1. A.2.2. A.2.3. A.2.4. and A.2.5 Although redundant required features may require redundant divisions of electrical power distribution subsystems to be OPERABLE, one OPERABLE distribution subsystem division may be capable of supporting sufficient required features to allow continuation of CORE ALTERATIONS, fuel movement, and operations with a potential for draining the reactor vessel. By allowing the option to declare required features associated with an inoperable distribution subsystem inoperable, appropriate restrictions are implemented in accordance with the affected distribution subsystem LCO's Required Actions. In many instances this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made. (i.e., to O suspend CORE ALTERATIONS, WWvement of irradiated fuel Al assemblies in thei{secondaryMontainment, and any *

     $]               activities that could result in inadvertent draining of the reactor vessel).
~

Suspension of these activities shall not preclude completion @$ of actions to establish a safe conservative condition. W These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC and DC electrical power distribution subsystems and to continue this action I until restoration is accomplished in order to provide the necessary power to the plant safety s i te3. .sh A.2.1 N 4 A.2 Notwithstanding performance of tie a e con G vative Required Actions, a required r sidual heat removal-shutdown cooling (RHR-SDC) subsystem y be inoperable. In this case, deseiequired Actions Of C=ditir V do not

            /y        adequately address the concerns relating to coolant circulation and heat removal. Pursuant to LCO 3.0.6, the               l Q@ll \             RHR-SDC ACTIONS would not be entered. Therefore, t4W Required Actiorf oLC;aditica A> direct declaring RHR-SDC 4

inoperable, whi results in taking the appropriate RHR-SDC ACTIONS. A2 ' ioed The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The r'estoration of the required distribution subsystems should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power.

O (continued) BWR/4 STS B 3.8-90 Rev. O, 09/28/92

RHR-High Water level B 3.9. (mj B 3.9 REFUELING OPERATIONS 1 5 f B 3.9.% Residual Heat Removal (RHR)-High Water Level p BASE h (Re6 l V BACKGROUND The purpos of the RHR System in MODE 5 is to remove decay heat and nsible heat from the reactor coolant, as required by GDC 34. Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat pp , exchangers, to the reactor via the associated recirculation loop, N i.hc readvi ,;a the ic pre::::re uvulo..; tnj e 4an p n . The RHR heat exchangers transfer heat to the RHR Service Water System S C^ 3.7.1;. The RHR shutdown 6I.I cooling mode is manually controlled. In addition to the RHR subsystems, the volume of water above the reactor pressure vessel (RPV) flange provides a heat sink for decay heat removal. (]J APPLICABLE With the unit in MODE 5, the RHR System is not required to SAFETY ANALYSES mitigate any events or accidents evaluated in the safety analyses. The RHR System is required for removing decay heat to maintain the temperature of the reactor coolant. h (ke f D ~AlthoughMeRHRSystem.doesnotm==Y $Ii#i /riterion 9 of the NRC Policy Statement 1 it vies identified h. iiie nut. Pnlicy $titamant at an imnn Hant contritUtus lu s$d j reductier. Therefore. the CWD Syste- 1: retsiiied as a l LSpeci fice Lierr. l Cel) L V rradse+ ct. + u e l s er LCO Only one RHR shutdown cooling subsystem sthe kV wnmL requirea ue [ OPERABLE and in operation in MODE 5 with the water level e d(3K:ft above the RPV flanier. Only one subsystem is i l required because the volumefof water above the RPV flange ' provides backup decay heat removal capability. NN B irci;<c j

&~

(& vsl e a f s o ;t I $+ c C wa+er a bove. \ .A +ba C , r r a d'e kd; fu e ( a SteMI MS :(continued) () s d a h A i n -f h o s- e fu ai ~ l (:5+o ra c3 e P oo l ra c k S),P- l BWR/4 STS j B 3.9-2 P ' Rev. O, 09/28/92 R5

h RHR-High Water Lovel

B 3.9. I 'R Sn HFGN mp frav d,5 Co chng +- y - BASES ( i h e h e3+ e dy nge q LCO An OPERABLE RHR shutdown c oling subsystem consists of an (continued) RHR pump, a heat exchangur, valves, piping, instruments, and controls to ensure an OPERABLE flow path. In MODE 5, the RHR cross tie valve is not required to be closed; thus, the ( gb r % f\ t valve may be opened to allo wpumps in one loop to discharge through tlle 6ppMRe leep' 5 hed cxchan to make ,a complete subsystem p 'ee WWIdrx P C

Additionally, each RHR shutdown cooling subsystem is i M* */ N' considered OPERABLE if it can be manually aligned (remote or liH RS vJ cre +i e, local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one valver rnq b FP subsystem can maintain and reduce the reactor coolant f o a lla u3 R H R S t d temperature as required. However, to ensure adequate core fl w to allow for accurate average reactor coolant

     +f u mF i -, crr e loop temperature monitoring, nearly continuous operation is                  ,

o provid e < ooli"j

 \ fo a he at enhwqeMrequired.         slut down A the Note  is provided operating         to allow subsystem        a 28 hour every    hours.exception to /fA}

i-f no+he o rva k opo si t e hop)/ a e w ole +A e J Su bn/c te rn , ' t r a d i B +d +d ' n "# RPVAMd APPLICAB One RHR shutdown cooling subsystem must be OPERABLE and in ff,1 operation in MODE 5, withwthe water level a T43h4at above i y the top of the RPV flange, to provide decay heat removal. g RHRjSyst== requirements in other MODES are covered by LCO

f ,in Section 3.4, Reactor Coolant System (RCS)XpEi44hs /gindn (eTh M.ourn Entgency~a- Ce wty re Coei in-Sy:

- innn w ,,, deme

                                                                             ,, _, (ECR)

S w agwd-Reactee)-Gorervervs Coo li o n - Cantai.1 nt_Sy>d-> RiiRdyM-5# requirements in MODE 5 with 6P. sd3p t emj he water level <) above the RPV flange are given in Ml

                          ^~~
                                                             #   S& 8-f Y8 i nches r         .-                  ,.                                       -

ACTIONS (I' A.1

Resi d u.a l Ne d kemava l (F<H R) - Lot ; MM L '

3L_ove.l.# { With no RHR shutdown cooling subsystem OPERABLE, an' l alternate method of decay heat removal must be established ' within 1 hour. In this condition, the volume of water above I Udy'Q~ Qe) the RPV flange provides adequate capability to remove decay All heat from the reactor core. However, the overall g So .d reliability is reduced because loss of water level could

            #g                        result in reduced decay heat removal capability. The 1 hour Completion Time is based on decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the (continued)

BWR/4 STS B 3.9-26 Rev. O, 09/28/92

RHR-High Water Level B 3.9.) f7 BASES l ACTIONS /L1 (continued)

                           '           functional availability of these alternate method (s) must be reconfirmed every 24 hours thereafter. This will ensure h o

h continued heat removal capability.

I got ,9f bd"S cs (ero uc k"h Alternate' decay heat removal methods are available to the

              % g<                     operators for review and preplanning in the unit's Operating Mj                   Procedures.         For example, this may include the use of the Reactor Water Cleanup System, operat_ing with the                                           o regenerative heat exchanger bypassed 4 The method used to                                   L' remove the decay heat should be the most prudent choice based on unit conditions.                                                         Fue l Q Cmlig B.1. B.2. B.3. and B.4                         s awlk    t 3 If no RHRIsubsystem is OPERABLE and an alternate method of decay heat removal is not available in accordance with Required Action A.1, actions shall be taken immediately to f                           suspend operations involving an increase in reactor decay Md               p bN                heat load by suspending loading of irradi neombliac intn the RD L                                   P fuel (a                                                                                              ,

econ daq { h ((al b54 indudMk cewm ufudedloortsd is oNh* @ n5un6cy. 01_ l %'f i Additional actions are require to minim ze any potential fission product release to the nvironment. This include q gen L . . . . . . ,3 . . . . . ~ , . . . . _ ica te Pesta,m tliefolh;in; tariff,umf) 61 1 OpWABLE st:t=: -secondary containmen- me(standby gas i/

                                        .reatment subsyste                    an* one secondary containment isolation r\!                  )[                ve and associate instrumentation <in each -as: ci:te'!

h an.oPMg F'

l. penetration.not isolated. This may performed as an aaministrative check, y examining ogs or other information art m #

cr

     '3) s eccnduj        @.s          to determineorwhether                   the components It is not are  out of service     for du!w*aa'r' i

paib maintenance other reasons. necessary to ahx.

 %ggi-g                                 perform the surveillances needed to demonstrate the of the components. If, however, any required ad$

a blW A U.c.,si OPERABILI component s inoperable, then it m t be restored to cag4;liisi' OPERABLE status. In this case, a urveillance may need to h -Hut E M5*d be performed to restore the component to OPERABLE status. he isolaW h Actions must continue until all required components are mi4344c OPERABLE. (5[to albirl-lusr 5 -

(continued) BWR/4 STS B 3.9-27 Rev. O, 09/28/92

RHR-High Water Level B 3.9 %. BASES g ACTIONS C.1 and C.2 # j (continued) < (c) e If no RHR . utdown' Cooling 4 < stem is in operation, an ; alternate method of coolant circulation is required to be ! established within 1 hour. The Completion Time is modified i such that the 1 hour is applicable separately for each ! occurrence involving a loss of cool ci rcul ation. { j <M y O#} l t During the period when the reac or ant is being circulated by an alterAate me od (other than by the required RH@utdown@gooling ystem), the reactor coolant l temperature must be periodically monitored to ensure proper l functioning of the alternate method. The once per hour ' Completion Time is deemed appropriate. SURVEILLANCE SR 3.9. _l 6 gg&do[ ; REQUIREMENTS co This Surveillance demonstrates that th RHR ubsystem is in operationandcirculatingreactorcoolantg dhe required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours 's sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room. t REFERENCE None. TMSEk T D, I P4 I I O BWR/4 STS B 3.9-28 Rev. O, 09/28/92 l

- INSERT B27 C to maintain the secondary containment at a negative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 2; single failure protection is not required while in this ACTION) T O

v INSERT D for Dronosed BASES 3.9.7

1. 10 CFR 50, Appendix A, GDC 34.

2. ' Technical Requirements Manual f

3. NRC No. 93-102, " Final Policy Statement on Technical l l Specification Improvements," July 23, 1993. !

l i -

RHR-Low Water Level B 3.9. s o 8 8 3.9 REFUELING OPERATIONS B 3. 9.' Residual Heat Removal (RHR)-Low Water Level

                                                       )N BASES Q}  7 ie4. O)

BACKGROUND The purpos of the RHR System in MODE 5 is to remove decay heat and' 9 nsible heat from the reactor coolant, as required by GDC 34/. Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a comon suction from the same recirculation loop. Each pump discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via the associated recirculation hh'1 loop.crtothere::t:rsi;t6- Inw prau ura ecetent iajectier ;;th. The RHR heat exchangers transfer heat to the RHR Service Water System (t C 3 4 The RHR shutdown cooling mode is manually controlled. GT.l f) L' APPLICABLE With the unit in MODE 5, the RHR System is not required to SAFETY ANALYSES mitigate any events or accidents evaluated in the safety analyses. The RHR System is required for removing decay heat to maintain the temperatura nf the reactor coo nt.

Alth; ugh die RHR System does net sper et$hf.3. 4titerion 0 _SaQ: '/.,3 of the NRC Policy Statemen , it es f i a;;iiod ... Uic HRC 0g Da14 cy 92+amant me == i ;:...... ....... ...,. Lu ris

                       'O       r44uetien. Therefore, the "ll" M., i; rejibdas a Speci#i:;tiv&.        Qg fy( w +% reudn pressa r c)
                                                    /   vme L (ePV%A LCO                    In MODE 5 with      e water level <
                               'prc::u n w v1 XRPV)' f1ange above the T::" w shutdown cooling Q

b subsystems must be OPERA . p g f n gg c3 An OPERABLE RHR shutdown oli ubsystem consists of an R Pump, a heat exchanger,4 valves, piping, instruments, and

u. m d"' controls to ensure an OPERABLE flow path o meet the LCO, m /; fn f f # dto YAc $ pumps in one loop or one pump ingeach f the two loops both gV 8 In MODE 5, the RHRicross tie valve is not M p# T ^> must be OPERABLE. required to be closed; thus, the valve may

( (continued) BWR/4 STS B 3.9-29 Rev. O, 09/28/92 l

                /"-                                                                                                                                             1 a cf/M n , % MHCN cro:s4'e-y     ko /ve 5 e                           c, //c u.) M/'M'M
                                                                                  .RHR-Low Water Level f w,c: g,se y * / 3 i . ,  ,,g B 3.9.)(

bo a h @ er /s e ofjw.m tofiddpc~00/,}- m W afpoc t & BASES # 5' b ""' #

                      /                          7/"'I N f               f                                                                   ,

LCO allow pumps in one loop to discharge through the opposite (continued) exchanger to make a complete subsystem.(

                        /                      W Additionally, each RHR shutdown cooling subsystem is

[&f fbj) considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature tJ required. However, to ensure adequate core flow to allow i-r accurate average reactor coolant temperature moaltoring, nearly continuous operation is

                   ,         requi red. A Note is provided to allow a 2 hour exception to

[y shut down the operating subsystem every 8 hours. g.?

                 %                           7 BM+ /e u & ps.D (irreJ R M +" d                                                               1 M APPLICABILITY             TwoRHRshutdowncooiingsuosystemsarerequTFedtobe)

OPERABLE, and one. must be in operation in H0DE 5, with nhe gp> water level < '{23 ft- above the top of the RPV flange, to , provide decay heat removal. RHRjSystem requirements in 76tfreTMOD[S are covered ~by LCOs in Section 3.4, Reactor 3g"

  ,-juld0       #

Coolant' FCf Sve+ r System (RCS)XyJedh U. E-rgency u u l i n u 4 e 5n-(RM [T

 't00 po9 #

r>4-a ndda a c to r: Gore =isciatiEC Se k , end 5::tica 3.5, Cent:i m at Sy C .. RHR .q - r g a &5 #+- requirements in MODE 5-with the water level e P y ft bove the RPV flange are g ven in LCO 3.9.')l', " Residual Heat #' p,q. q R h emoval (RHR)-High Water Level." g - Grr&M L\ , & RPVaD g &- 8 "I1 - ACTIONS A.d h 3IO With one of the two required RHR shutdown cooling subsystems N 1 inoperable,theremainingsubsystemiscapableofproviding)- the required decay heat removal. However, the overall reliability is reduced. Therefore an alternate method of / decay heat removal must be provided. With both'liHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO. The 1 hour Completion Time is based on the decay heat removal function and the probability of a loss of the i available decay heat removal capabilities. Furthermore, verification of the functional availability of this (continued) BWR/4 STS 8 3.9-30 Rev. O, 09/28/92

1 i RHR-Low Water Level B 3.9 # 8 7.

/N BASES V i ACTIONS /L1 (continued) ) alternate method (s) must be reconfirmed every 24 hours I thereafter. This will ensure continued heat removal capability. l Alternate decay heat removal methods are available to the operators for review and preplanning in the unit's Operating Procedures. For example, this may include the use of the Reactor Water Cleanup System, operating with the regenerative heat exchanger bypassed. The method used to remove decay heat should be the most prudent choice based on unit conditions, n , nY N k')o _ , 6A 1 Ww e cArnf . If no RHR(subsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour. The Completion Time is modified such that the I hour is applicable separately for each occurrence g fust inc6bg involving a loss of coolant circulation. e

       +   & cmm e                                                                     k    l

( ) ngy ft,y During the period when the reactor coo an i ing goQ ( circulatedby/analternatemeto (other than by the required RHR hutdowngooling ystem), the reactor coolant temperature must be periodically monitored to ensure proper pJ , functioning of the alternate method. The once per hour 9 (3 3,y31 Completion Time is deemed appropriate. 4 C, A .1

                                         -If t ie n t erc ""R     subsyst a is not restered to CPERACLE status nc.,cdiete4y, additional actions are required to minimize any potential fission product release to the
     /s{ahgg           -

conladwmf

                                      ,3  environment. This includes et4t4et4ng-i=cdiate actiun to @segr83h                '

gi/W g restace t6 M e., ins Lv GPERABLE d d es; secondar ~gR W_ l 3M. ((i. e.., g ~43< % Q ontainment % standby gas treatment subsystem 1 an

                                 ' u secondary containment isolation valve and associate                  4/Q\    4 instrumentation *in each associated,penetrationdnot isolate

[g ggg g w M ,'5

                     . . g#    .

This may be performed 4s an administrative check, by [, examining logs or other information to determine whether the [ y "[ M components are out of service for maintenance or other ,

       @ la b, W # N-f)                   reasons. It is not necessary to perform the surveillances           *Og          I l

needed to demonstrate the OPERABILITY of the components. " l

                           " #            If, however, any required component is inoperable, then it          MS (cacolu1 *b'.                       mu'st be restored to OPERABLE status.       In this case, the l
                                         /urveillancemayneedtobeperformedtorestorethe f

(continued) i BWR/4 STS B 3.9-31 Rev. 0, 09/28/92

RHR-Low Water Level

       '                                                                   B 3.9.%/

i I

                                                                                @         l BASES g
                                   .Q                                                     '

ACTIONS 7_ B.l. B.2.*B.3v 4.1L M (continued) component to OPERABLE status. Actions must continue until all required components are OPERABLE. gp] SURVEILLANCE

@~

SR 3.9 I.1 Gv3

                                                             '---                   r REQUIREHENTS Mch uldouT This Surveillance demonstrates that one HR$ubsystem        6 is in ,;, v l operation and circulating reactor coolant. The required flow rate is detemined by the flow rate necessary to provide sufficient decay heat removal capabilityf

( ly CTh'e Frequency of 12 hours is sufficient in view of other p (v,j )J visual and audible indications available to the operator for j

          ..       monitoring the RHR subsystems in the control room.

REFERENCES 4HRE. 1NSE(LT O

F} By l l O' BWR/4 STS B 3.9-32 Rev. O, 09/28/92 l l

INSERT B 3.9-31 89 0) O

                                                      /    -

B.1, B.2, and B.3 gogshovh With the required RHR shutdown ooling subsystem (s) inoperable and the required alternate method (s) of decay heat removal not available in accordance with Required Action A.1, INSERT B31 to maintain the secondary containment at a negative pressure with respect to the environment (dependent on secondary containment configuration, refer to Reference 2; single j failure protection is not required while in this ACTION) f O O

INSERT F for Droposed BASES 3.9.8 Dr M 1. 10 CFR 50, Appendix A, GDC 34.

2. Technical Requirements Manual 9 f

3. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

I l 1 s v l i l l l U ,

Q NUREG 1433 COMPARISON DOCUMENT - JUSTIFICATION FOR DEVIATION -

                                                           -           l
                                                                       )

J 4 i i O l O

JUSTIFICATION FOR DEVIATION FROM NUREG 1433 ITS: SECTION 3.1 - REACTIVITY CONTROL [/) PLANT SPECIFIC DIFFERENCES P.1 The brackets have been removed, and the proper values /words used for each of the two units. Bases changes are made consistent with the Specifications. P.2 The number of SGT subsystems and the required comblo1tions are dependent on the configuration of the secondary containment and are detailed in the Technical Requirements Manual, consistent with the presentation for these requirements in Section 3.6. The required number of SGT subsystems is that necessary to maintain the secondary containment at a negative pressure with respect to the environment (single failure protection is not required while in this ACTION). P.3 The words in Condition D, "and not separated by two or more OPERABLE control rods" have been deleted since a plant specific BPWS evaluation allows this condition. Also, these words are not needed since the separation criteria will be addressed in the BPWS evaluation. Thus, the words "not in compliance with...(BPWS)" encompasses the words "and not separated..." The Bases have been (- modified to reflect this deletion by referencing both the generic licensing basis analysis, as well as any plant specific analysis. The Bases still state that in the generic licensing basis analysis, inoperable control rods are assumed to be separated by at least two OPERABLE control rods. NUREG ACTION E and the associated Bases have been deleted since the plant specific BPWS analysis does not require this limit (no more than 4 inoperable control rods in a group). The remaining ACTIONS have been renumbered to reflect the deletion. P.4 The words have been modified to provide a clear understanding of when to enter the Condition. Boron concentration is not the only variable; temperature and volume are also part of the overall requirement. The intent of this Condition has not changed; it will be entered when the sodium pentaborate in solution is not within the LCO limits but still meets the original licensing basis. l HATCH UNITS 1 AND 2 1 REVISION f (fr~

eg JUSTIFICATION FOR DEVIATION FROM NUREG 1433 i (Y ITS: SECTION 3.3 - INSTRUMENTATION PLANT SPECIFIC DIFFERENCES (continued) P.25 This Function is not included in the Plant Hatch design and has been deleted. P.26 Changes were made to provide additional information or. clarity, or were made to use plant specific terminology. P.27 The words " primary containment" and " secondary containment" are unnecessary. The LCO title addresses primary containment or secondary containment instrumentation, thus no confusion should result from the deletion of these modifiers. P.28 NUREG SR 3.3.6.2.2 for Function 4, which requires a CHANNEL FUNCTIONAL TEST every 92 days, has been deleted. The CHANNEL CALIBRATION (NUREG SR 3.3.6.2.4), which is also required every 92 days, includes the requirement to perform a CHANNEL FUNCTIONAL TEST. Therefore, it is not necessary to repeat the CHANNEL FUNCTIONAL TEST. For Function 3, the current Plant Hatch Frequency for a CHANNEL CALIBRATION is 92 days, and the SR number has been changed to SR 3.3.6.2.3

^g        to refer to an SR frequency that reflects the current 92
)         day test frequency. The CHANNEL FUNCTIONAL TEST SR has been deleted for the same reason as Function 4.

I P.29 The number of SGT subsystems (Unit 1 and Unit 2 SGT subsystems) and the required combinations are dependent on I the configuration of the secondary containment and are detailed in the Technical Requirements Manual, consistent with the presentation for these requirements in Section 3.6. The required number of SGT subsystems is that ) necessary to maintain the secondary containment at a negative pressure with respect to the environment (single 1 failure protection is not required while in this ACTION). The Bases cahnges are made to more accurately reflect these possibilities. P.30 These words have been modified for clarity, and to be consistent with other Conditions and Required Actions where the requirements are similar (e.g., NUREG LCO 3.3.6.2, ACTION B). The instrumentation section should address the condition where a component is inoperable due to an inoperable channel and also where initiation capability has been lost. The Note in the SRs has been modified for consistency with the change to the ACTIONS. Appropriate Bases changes have been made. ("' l V) HATCH UNITS 1 AND 2 6 REVISION f (by-

l f- JUSTIFICATION FOR DEVIATION FROM NUREG 1433 j ITS: SECTION 3.5 - ECCS AND RCIC SYSTEM i l l PLANT-SPECIFIC DIFFERENCES P.1 Brackets have been removed and the proper value/words have been used for each of the two units. Also, Bases changes were made to be consistent with the Specifications. P.2 The number of SGT subsystems and the required combinations are dependent on the configuration of the secondary containment and are detailed in the Technical Requirements Manual, consistant with the presentation for these 1 requirements in Section 3.6. The required number of SGT H subsystems is that necessary to maintain the secondary containment at a negative pressure with respect to the environment (single failure protection is not required while in tiais ACTION) . P.3 Typographical / grammatical errors have been corrected. l P.4 The proper references have been provided. P.5 This allowance has been added to both Units since it is currently licensed for Hatch Unit 2. , P.6 Plant Hatch has no recirculation pump discharge valve bypass valves. P.7 Plant Hatch has only one solenoid for each ADS valve. l P.8 This value is equal to the licensed value at which the spent fuel pool must be maintained. Twenty-one feet in the spent fuel pool is equivalent to 22 ft 1/8 inches over the top of the reactor pressure vessel flange. P.9 These changes have been made since the actions discussed are not certainties, but "could" or "may be" allowed. P.10 This discussion has been deleted since it discusses RCIC, which is not part of this LCO. O HATCH UNITS 1 AND 2 1 REVISION 'A (- l I

4 JUSTIFICATION FOR DEVIATION FROM NUREG 1433 ( ITS: SECTION 3.6 - CONTAINMENT SYSTEMS PLANT SPECIFIC DIFFERENCES (continued) j P.25 This Specification has not been added for Natch Unit 2 since the Unit 2 safety analysis does not assume the operation of this system to control Hydrogen. The Unit 2 safety analysis assumes the H Recombiners (LCO 3.6.3.1.) 2 operation. I P.26 Changes were made to provide additional information or clarity, or were made to use plant specific terminology. P.27 The number of SGT subsystems and the required combinations are dependent on the configuration of the secondary containment and are detailed in the Technical Requirements Manual. The required number of SGT subsystems is that necessary to maintain the secondary containment at a negative pressure with respect to the environment (single failure protection is not provided in the required number for SRs 3.6.4.1.3 and 3.6.4.1.4). One specific configuration is deemed justified of an additional allowed out-of-service time: one inoperable SGT subsystem while all three zones comprise the secondary O containment boundary (therefore all four SGT subsystems required OPERABLE) . In this configuration, if the Unit 1 reactor building-to-refueling floor plug is not installed (open communication between Zone I and Zone III) a Unit 1 SGT subsystem can be inoperable, and tests have shown that a high level of confidence remains that even with an additional single of any SGT subsystem (which is not necessary to assume while in ACTIONS) the required drawdown function could still be performed. With this specific configuration, a 30-day Completion Time is proposed, as presented in ACTION A. P.28 This SR has been deleted since the secondary containment analysis assumes no differential pressure at the start of the accident. That is, the SGT subsystems are assumed to drawdown the secondary containment from 0 inches to .25 inches vacuum when they are actuated. This SR is not currently in the Hatch Unit 1 or 2 Technical Specifications. L O HATCH UNITS 1 AND 2 5 REVISION d(3 - l .- . .~

I l JUSTIFICATION FOR DEVIATION FROM NUREG 1433 Ih ITS: SECTION 3.6 - CONTAINMENT SYSTEMS V PLANT SPECIFIC DIFFERENCES (continued) P.29 All secondary containment penetration flow' paths have two l isolation valves, therefore this Note is unnecessary and i has been deleted. P.30 Since there are more installed RHR pumps than are regulred to meet LCO 3.6.2.3, the word " required" has been added to l the applicable places, consistent with its use throughout the NUREG. P.31 Comment number not used. P.32 This SR has been deleted since the Hatch design does not include a filter cooler bypass damper and fan. P.33 A Note has been added to the ACTIONS to allow inspection of the Unit 1 hardened vent rupture disk while Unit 2 is operating. This inspection will cause both the Unit 1 SGT subsystems to be inoperable, thus the allowance is needed to continue operating Unit 2 while this inspection is being performed. Without this allowance, a dual unit shutdown would be required. () It is expected that the hardened vent inspection from hanging of the first tag of the clearance through completion of the inspection and restoration to operability will take aproximately 12 hours. Should, during the course of the inspection, there be an emergent need to return one of the tagged-out SGT trains to service, this action (bolting the rupture disk flange, racking the breaker, and removing clearance tags) should, at most, take approximately 1 hour. P.34 The proposed ACTION D added per NUREG change package BWR-04, Item C.8 has not been added into the Unit 2 ITS since it is not needed. The change was made to the NUREG because a Condition allowed two SGT subsystems to be inoperable (NUREG Condition D), and confusion existed as to which requirement applied if the Unit was in MODE 1, 2, or 3 at the same time fuel was being moved. Since the Unit 2 LCOs have been split up with the MODES 1, 2, and 3 requirements in a different LCO than the handling of irradiated fuel assemblies in the secondary containment, the confusion does not exist. In the Unit 2 ITS, no ACTIONS exist in the Operating LCO (MODES 1, 2, and 3) to allow two SGT subsystems to be inoperable; therefore, LCO 3.0.3 will apply, just as the proposed NUREG Condition requires. In addition, this proposed NUREG ACTION only applies for MODES 1, 2, and 3. Therefore, it is not needed in the other two Unit 2 SGT System LCOs. HATCH UNITS 1 AND 2 6 REVISION 1((,

l JUSTIFICATION FOR DEVIATION FROM NUREG 1433 () ITS: SECTION 3.6 - CONTAINMENT SYSTEMS - PLANT SPECIFIC DIFFERENCES (continued) P.55 The proper description of the system has b'een provided. : P.56 Unit 1 uses a CAD System, not H2 Recombiners and the Drywell cooling System. Thus the reference to the other systems has been changed. P.57 The safety analysis for Plant Hatch Unit 2 does not assume an initial oxygen concentration. However, the specification does meet criterion 4 of the NRC Policy Statement, thus it is retained for this reason. P.58 As described in comment No. P.27, the secondary containment can include: the common refueling floor, 2) the Unit i reactor building, and 3) the Unit 2 reactor building. Due to the design of the buildings, one or both reactor building (s) can be separated from the common refueling floor secondary containment and also separated from the required secondary containment boundary. Descriptions outlining these options are added to the Bases. (" P.59 This phrase has been deleted since Plant Hatch safety analysis does not assume a wind angle. P.60 The filter tests are not all in accordance with RG 1.52, Rev.2. However they are in accordance with the VFTP. Thus, this is the reference mentioned in the Bases. P.61 The MSIV leakage limit for Plant Hatch Unit 2 was proposed to be changed per GPC letter dated January 6, 1994, and subsequently, the NRC issued this change as Amendment 132 to the Unit 2 TS by letter dated March 17, 1994. Amendment l 132 includes the 250 scfh limit and the requirement to ' reduce leakage to < 11.5 scfh if an MSIV exceeds the 100 scfh limit. Further technical discussion is provided in 1 the GPC January 6, 1994, letter and the corresponding NRC SER dated March 17, 1994. " l P.62 Comment number not used. l P.63 The Unit 1 safety analysis does not have the same level of detail as the Unit 2 analysis. Therefore, the Unit 1 Bases ! has been modified to reflect that the Unit 2 FSAR analysis { is appropriate for Unit 1. See comment No. P.46. I 1 () HATCH UNITS 1 AND 2 9 REVISIONf(h-i

JUSTIFICATION FOR DEVIATION FROM NUREG 1433

' O                        ITS: SECTION 3.7 - PLANT SYSTEMS PLANT SPECIFIC DIFFERENCES P.6 (continued) each PSW subsystem is inoperable, consistent with the. time provided for an inoperable PSW subsystem.

t P.7 This SR has been modified to reflect current Technical Specification requirements. The UHS is the Altamaha River, which does not charge level rapidly. Therefore, it is not necessary to monitor river level frequently when river level 1 61.7 ft MSL. When the river level is lower, it must be monitored more frequently (every 12 hours). P.8 The river temperature has been shown, by a review of the USGS data, not to exceed the river temperature assumed in the safety analysis. The current Technical Specifications do not include a temperature limit. The NRC did not require this SR when the two units were originally licensed. A review of more recent data has continued to show that the maximum temperature limit has not been N exceeded. Therefore, it is unnecessary to perform this SR and has not been included in the Plant Hatch ITS submittal. Appropriate Bases changes have been made. P.9 Changes were made to provide additional information or clarity, or were made to use plant specific terminology. P.10 Comment number not used. l P.11 This Note has been deleted since Plant Hatch does not have ' a toxic gas protection mode. (Removed in Unit 1 Amendment ; 156 and Unit 2 Amendment 96, dated September 12, 1988.) 4 P.12 The MCREC System does not have any heaters installed; thus, I the appropriate run time is 15 minutes. This is consistent with the current Plant Hatch Technical Specifications. O HATCH UNITS 1 AND 2 2 REVISION ,I(37'

1 JUSTIFICATION FOR DEVIATION FROM NUREG 1433 (,.s) ITS: SECTION 3.9 - REFUELING OPERATIONS PLANT SPECIFIC DIFFERENCES P.1 Brackets have been removed and the proper value/words have been used for each of the two units. Also, Bases changes were made to be consistent with the Specifications. P.2 Plant Hatch's licensing basis analysis for a dropped fuel bundle assumes the bundle drops over the fuel in the reactor pressure vessel. It does not assume a drop on the flange. the current Technical Specifications supports this since water level, during fuel movement within the RPV, is only required to be 23 feet above the top of irradiated fuel assemblies seated within the RPV. Therefore, ITS LCOs 3.9.6 and 3.9.7 have been combined into one LCO (ITS 3.9.6) covering both irradiated fuel movement and new fuel and control rod movement, with the water level requirement the same as the current Specifications. Appropriate Bases changes have also been made. The subsequent LCOs have been renumbered due to this change. In addition, certain changes to the Bases provided in NUREG change package BWR-18 (Items C.2, C.4, C.75, and C.76) were either not adopted or were modified to be consistent with the Hatch licensing f) V basis. P.3 Typographical / grammatical errors have been corrected. P.4 The proper references have been provided. P.S This value is equal to the licensed value at which the spent fuel pool must be maintained. Twenty-one feet in the spent fuel pool is equivalent to 22 feet 1/8 inches over the top of the reactor pressure vessel flange. P.6 The number of SGT subsystems and the required combinations are dependent on the configuration of the secondary containment and are detailed in the Technical Requirements Manual, consistent with the presentation for these requirements in Section 3.6. The required number of SGT subsystems is that necessary to maintain the secondary containment at a negative pressure with respect to the environment (single failure protection is not required while in this ACTION). HATCH UNITS 1 AND 2 1 REVISION ( r-1 __ 1

,3
,                JUSTIFICATION FOR DEVIATION FROM NUREG 1433 (V  )              ITS: SECTION 3.9 - REFUELING OPERATIONS PLANT SPECIFIC DIFFERENCES P.7   These words have been deleted since, in some cases, there is more than one channel. Starting the sentence with the word " Instrumentation" is sufficient to convey the proper meaning.

P.8 This information is not needed since this LCO does not include the one-rod-out interlock. P.9 Changes were made to provide additional information or clarity, or were made to use plant specific terminology. P.10 The APRM neutron flux scram is not required while in MODE 5, thus reference to it has been deleted. P.11 Changes were made for consistency with other (~ Specifications. D} P.12 These words have been deleted since there is only one injection flow path; via the associated recirculation loop. P.13 The proper criterion from the final policy statement has been used. The current wording was developed prior to the issuance of the final policy statement, which now uses criterion 4 as meaning the current words of the NUREG. P.14 This LCO is dealing with RHR shutdown cooling requirements. Thus it has been phrased this way to prevent confusion. This LCO has nothing to do with the LPCI mode of RHR (ECCS function) or the suppression pool cooling mode of RHR. Thus the references to these LCOs has been deleted. P.15 Since each RHR pump (along with the associated piping, valves, heat exchanger, and cooling water) constitutes a subsystem, and there are four RHR pumps, these words have been changel to "two". P.16 The RHR crosstie valve is downstream of the heat l exchangers. Therefore, the words have been changed to be l

            " recirculation loop" to maintain the intent of the LCO words and for consistency with other LCO Bases.

A 'N) HATCH UNITS 1 AND 2 2 REVISION)q[_ L

l JUSTIFICATION FOR DEVIATION FROM NUREG 1433 () ITS: SECTION 3.10 - SPECIAL OPERATIONS PLANT SPECIFIC DIFFERENCES P.1 Brackets have been removed and the proper'value/words have been used for each of the two units. Also, Bases changes were made to be consistent with the Specifications. P.2 Comment number not used. l P.3 Typographical / grammatical errors have been corrected. P.4 The allowances provided by these Specifications are not needed at Plant Hatch; consequently, they have been deleted. P.5 The proper references have been provided. P.6 The Startup Tast Program has been completed at Plant Hatch; thus, a reference is not needed. P.7 Changes were made to provide for consistency with other Specifications. ' P.8 Changes were made to either provide additional information ( or clarity, or to incorporate plant-specific terminology.

P.9 The proposed hydrostatic testing requirement for RCS temperature is > 212*F (plant specific value). Therefore a sentence saying that the requirement may eventually be j greater than 200'F is unnecessary. The last sentence in the fourth paragraph has been deleted since this LCO is not exempting the Safety Limit from being met during a hydrostatic test. Therefore the Safety Limit is required to be met in accordance with SL 2.1.2. The temperature requirements are included in LCO 3.4.9. Additionally, a sentence regarding the system test pressure is added. P.10 Since Plant Hatch does not have a reactor high water level scram or a suppression pool makeup system; these references have been deleted. P.11 The previous sentence states that, the rod patterns assumed in the safety analysis may not be preserved. This sentence has been changed to state that a special CRDA analysis "may be" required. P.12 The correct power level (corresponding to the analysis value) is 10% RTP. As written, the power level corresponds to the low power setpoint. O HATCH UNITS 1 AND 2 1 REVISIONf(; _---}}

ML20078G716

Text

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