ML20072V364
| ML20072V364 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/31/1994 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20072V356 | List: |
| References | |
| NUDOCS 9409190333 | |
| Download: ML20072V364 (869) | |
Text
{{#Wiki_filter:g LC0 Applicability j 3.0 l f i 3.0 LIMITING' CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LC0 3.0.2 and LCO 3.0.7. LC0 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as.
- provided in LC0 3.0.5 and LC0 3.0.6.
If the LC0 is met or is no longer applicable prior to expiration of the specified Completion Time (s), completion of the Required Action (s) is not required, unless otherwise stated. LCO 3.0.3 When an LC0 is not met and the associated ACTIONS are not met, an associated ACTI0ii is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within I hour to place the unit, as applicable, in: i
/~N a. MODE 2 within 7 hours; O b. MODE 3 within 13 hours; and
- c. MODE 4 within 37 hours.
Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit ! operation in accordance with the LC0 or ACTIONS, completion I of the actions required by LC0 3.0.3 is not required. LC0 3.0.3 is only applicable in MODES 1, 2, and 3. i LC0 3.0.4 When an LC0 is not met, entry into a MODE or other specified i condition in the Applicability shall not be made except when l the associated ACTIONS to be entered permit cont:nued i operation in the MODE or other specified condition in the l Applicability for an unlimited period of time. This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required 1 {
+ (continued)
HATCH UNIT 1 3.0-1 REVISION A 9409190333 940831 PDR ADOCK 05000321 P PDR
LC0 Applicability 3.0 3.0 LC0 APPLICABILITY LC0 3.0.4 to comply with ACTIONS or that are part of a shutdown of the (continued) unit. Exceptions to this Specification are stated in the individual Specifications. These exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered allow unit operation in the MODE or other specified condition in the Applicability only for a limited period of time. LC0 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned te service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LC0 3.0.2 for the system [ returned to service under administrative control to perform the required testing. LC0 3.0.6 When a supported system LC0 is not met solely due to a support system LC0 not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.5.10, " Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LC0 3.0.2. (continued) HATCH UNIT 1 3.0-2 REVISION D l l
SDM 3 1.1 t 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MAP, GIN (SDM) , LC0 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. B. Required Action and B.1 Be in MSDE 3. 12 hours associated Completion r Time of Condition A 1 not met. C. SDM not within limits C.1 Initiate action to- Immediately
- in MODE 3. fully insert all insertable control i
rods, i D. SDM not within limits D.1 Initiate action to Immediately in MODE 4. fully insert all , insertable control : rods. j l AND i O (continued) k. HATCH UNIT 1 3.1-1 REVISION A l
SDM 3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) D.2 Initiate action to I hour restore secondary containment to OPERABLE status. AND D.3 Initiate action to I hour ! restore two standby gas treatment (SGT) ; subsystems to OPERABLE status. AND D.4 Initiate action to I hour restore isolation capability in each lh ; j required secondary ; containment l penetration flow path l not isolated. j
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I 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 1 3.1-2 REVISION D
4 i SDM- ! 3.1.1: ; () ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
?
Initiate action to E. (continued) E.3 l' hour ' restore secondary containment to -l OPERABLE status. , t AND E.4 Initiate action to I hour restore two SGT subsystems to ! OPERABLE status. l AND ; E.5 Initiate action to I hour - restore isolation g ; capability in each . required secondary containment . penetration flow path g not isolated. :
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HATCH UNIT 1 3.1-3 REVISION D l
SDM 3.1.1 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 3.1-4 REVISION A HATCH UNIT 1 1
. Control Rod Scram Times 3.1.4
.(~$ SURVEILLANCE REQUIREMENTS () SURVEILLANCE FREQUENCY , t SR 3.1.4.1 (continued) Prior to , exceeding ,
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40% RTP after each reactor shutdown 2 120 days SR 3.1.4.2 Verify, for a representative sample, each 120 days - tested control rod scram time is within the cumulative limits of Table 3.1.4-1 with reactor steam operation in dome pressure 2 800 psig. MODE 1 SR 3.1.4.3 Verify each affected control rod scram time Prior to is within the limits of Table 3.1.4-1 with declaring any reactor steam dome pressure, control rod OPERABLE after . (%) work on control
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rod or CRD ' System that could affect > scram time SR 3.1.4.4 Verify each affected control rod scram time Prior to i is within the limits of Table 3.1.4-1 with exceeding l ' reactor steam dome pressure 2 800 psig. 40% RTP after work on control ! rod or CRD i System that could affect scram time i r k. HATCH UNIT 1 3.1-13 REVISION A
Control Rod Scram Times 3.1.4 Table 3.1.4-1 (page 1 of 1) Control Rod Scrim Times llll
-------------------------------------NOTES------------------------------------
- 1. OPERABLE control rods with scram times not within the limits of this Table are considered " slow."
- 2. Enter applicable Conditions and Required Actions of LC0 3.1.3, " Control Rod OPERABILITY," for control rods with scram times > 7 seconds to notch position 06. These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered " slow."
SCRAM TIMES WHEN REACTOR STEAM DOME PRESSURE 2: 800 psig (a)(b) NOTCH POSITION (seconds) 46 0.44 e#di 36 1.08 26 1.83 gg i 06 3.35 1 (a) Maximum scram time frof,. fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero. (b) When reactor steam dome pressure < 800 psig, established scram time limits apply. i . l t . r l l 9 HATCH UNIT 1 3.1-14 REVISION D
SRM Instrumentation l 3.3.1.2 l () SURVEILLANCE REQUIREMENTS
-------------------------------------NOTES------------------------------------
l
- 1. Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified conditions.
- 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 other required channel (s) is OPERABLE.
SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK. 12 hours SR 3.3.1.2.2 ------------------NOTES------------------
- 1. Only required to be met during CORE ALTERATIONS.
- 2. One SRM may be used to satisfy more 3
f4 than one of the following. \~_/ - - - - - - - - _ - - - - - _ . . . . _ - - - - _ - _ . - _ _ - - _ - - - Verify an OPERABLE SRM detector is 12 hours located in:
- a. The fueled region;
- b. The core quadrant where CORE ALTERATIONS are being performed, when -
the associated SRM is included in the fueled region; and i
- c. A core quadrant adjacent to where CORE ALTERATIONS are being performed, when the associated SRM is included in the fueled region.
i SR 3.3.1.2.3 Perform CHANNEL CHECK. 24 hours (continued) Q N- J i HATCH UNIT 1 3.3-11 REVISION A i l I
SRM Instrumentation 3.3.1.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.2.4 - - - -- ---- - - - - - - -- - N O T E S - - - - - - - -- - -- - -- - - -
- 1. Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.
- 2. Not required to be met during spiral unloading.
Verify count rate is 2 3.0 cps with a 12 hours during signal to noise ratio 2 2:1. CORE l ALTERATIONS AND 24 hours SR 3.3.1.2.5 Perform CHANNEL FUNCTIONAL TEST and determination of signal to noise ratio. 7 days G 4 SR 3.3.1.2.6 ------------------NOTE------------------- Not required to be performed until 12 hours after IRMs on Range ? or below. Perform CHANNEL FUNCTIONAL TEST and determination of signal to noise ratio. 31 days k SR 3.3.1.2.7 ------------------NOTES------------------
- 1. Neutron detectors are excluded.
- 2. Not required to be performed until 12 hours after IPJis on Range 2 or below.
Perform CHANNEL CALIBRATION. 18 months O HATCH UNIT 1 3.3-12 REVISION D
Control Rod Block Instrumentation 3.3.2.1 [) x_ - SURVEILLANCE REQUIREMENTS (continued) , SURVEILLANCE FREQUENCY SR 3.3.2.1.2 ------------------NOTE------------------- Not required to be performed until I hour i after any control rod is withdrawn at j{
< 10% RTP in MODE 2.
Perform CHANNEL FUNCTIONAL TEST. 92 days , SR 3.3.2.1.3 ------------------NOTE------------------- Not required to be performed until I hour after THERMAL POWER is < 10% RTP in MODE 1. Perform CHANNEL FUNCTIONAL TEST. 92 days b] SR 3.3.2.1.4 ------------------NOTE------------------- Neutron detectors are excluded. Verify the RBM: 18 months
- a. Low Power Range -- Upscale Function is not bypassed when THERMAL POWER is 2: 29% and < 64% RTP.
- b. Intermediate Power Range -- Upscale Function is not bypassed when THERMAL POWER is a 64% and < 84% RTP.
- c. High Power Range -- Upscale Function is not bypassed when THERMAL POWER is 2: 84% RTP.
(continued) O(% HATCH UNIT 1 3.3-17 REVISION D
Control Rod Block Instrumentation 3.3.2.1 (continued) SURVEILLANCE REQUIREMENTS llll SURVEILLANCE FREQUENCY SR 3.3.2.1.5 Verify the RWM is not bypassed when 18 months THERMAL POWER is < 10% RTP. SR 3.3.2.1.6 ------------------NOTE------------------- Not required to be performed until I hour after reactor mode switch is in the shutdown position. Perform CHANNEL FUNCTIONAL TEST. 18 months SR 3.3.2.1.7 ------------------NOTE------------------- Neutron detectors are excluded. Perform CHANNEL CALIBRATION. 18 months O SR 3.3.2.1.8 Verify control rod sequences input to the Prior to RWM are in conformance with BPWS. declaring RWM OPERAblF following loading of sequence into RWM O HATCH UNIT 1 3.3-18 REVISION A l l __m . _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . . _ - - _ _ _ _ _ _ _ _ _ _ . _ _ _ . . . _ . _ . - _ . . _ _ _ . . _ _ _ . _ . _ _ _ _ _ _ _
Remote Shutdown System 3.3.3.2 lg 3.3 INSTRUMENTATION
-3.3.3.2 Remote Shutdown System LCO 3.3.3.2 The Remote Shutdown System Functions shall be OPERABLE.
APPLICABILITY: MODES 1 and 2. ACTIONS
-------------------------------------NOTES------------------------------------
- 1. LC0 3.0.4 is not applicable.
- 2. Separate Condition entry is allowed for each Function.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required 30 days
'N Functions inoperable. Function to OPERABLE
{d status. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. i l r~%l V HATCH UNIT 1 3.3-25 REVISION A
Remote Shutdown System 3.3.3.2 SURVEILLANCE REQUIREMENTS ____________________-----------------NOTE------------------------------------- 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. SURVEILLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel that is normally energized. SR 3.3.3.2.2 Verify each required control circuit and 18 months Ithh transfer switch is capable of performing the intended function. SR 3.3.3.2.3 Perform CHANNEL CALIBRATION for each 18 months I required instrumentation channel. i
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l l l l 9 HATCH UNIT 1 3.3-26 REVISION D
15 Primary Co'ntainment Isolation Instrumentation-3.3.6.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME G. As required by G.1 Be in MODE 3. 12 hours _ Required Action C.1 l and referenced in AND , Table 3.3.6.1-1. G.2 Be in MODE 4. 36 hours E i Required Action and
- associated Completion
. Time of Condition F ,
not met. > H. As required by H.1 Declare Standby 1 hour Required Action C.1 Liquid Control (SLC) . and referenced in System inoperable. Table 3.3.6.1-1. E' H.2 Isolate the Reactor 1 hour
. (]- Water Cleanup (RWCU)
System. I. As required by I.1 Initiate action to Immediately Required Action C.1 restore channel to . and referenced in OPERABLE status. l Table 3.3.6.1-1. l M i I.2 Initiate action to' Immediately 1 isolete the. Residual Heat Removal (RHR) .l Shutdown Cooling. : System. ! e f O HATCH UNIT 1 3.3-51 REVISION A-
Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS _____--------------------------------NOTES------------------------------------
- 1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary 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. l /p\
SURVEILLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK. 12 hours SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.6.1.3 Perform CHANNEL CALIBRATION. 92 days l SR 3.3.6.1.4 Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.6.1.5 Perform CHANNEL CALIBRATION. 18 months SR 3.3.6.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 month,s 1 O HATCH UNIT 1 3.3-52 REVISION D l
.- = .. . - .. . . .- - -.
4 Secondary Containment Isolation Instrumentation - 3.3.6.2
) 3.3 INSTRUMENTATION
-3.3.6.2 Secondary Containment Isolation Instrumentation i
LCO 3.3.6.2 The secondary containment isolation instrumentation for each l 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 i
A. One or more channels A.1 Place channel in 12 hours for inoperable. trip. Function 2 AND 24 hours for l Functions other than Function 2 i i B. One or more automatic B.1 Restore isolation I hour { Functions with capability. ' isolation capability not maintained. C. ' Required Action and C.1.1 Isolate the 1 hour associated Completion associated zone (s). Time of Condition A . or B not met. OR i (continued)' l
./-~_
i. HATCH UNIT 1 3.3-57 REVISION A f i
l-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. E 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 j 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) HATCH UNIT 1 3.3-58 REVISION D
l ECCS - Shutdown 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM l 3.5.2 ECCS - Shutdown LCO 3.5.2 Two low pressure ECCS injection / spray subsystems shall be . OPERABLE. APPLICABILITY: MODE 4, MODE 5, except with the spent fuel storage pool gates j removed and water level 2: 22 ft 1/8 inches over the top of the reactor pressure vessel flange. 1 l ACTIONS j 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 I status. 1 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 OPDRVs. subsystems inoperable. AND C.2 Restore one ECCS 4 hours injection / spray subsystem to OPERABLE status. (continued)- A
. \.)
HATCH UNIT 1 3.5-7 REVISION A
i ECCS - Shutdown i 3.5.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action C.2 D.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 two standby gas treatment subsystems to OPERABLE status. AND D.3 Initiate action to Immediately A restore isolation capability in each g required secondary containment g penetration flow path w not isolated. l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
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SR 3.5.2.1 Verify, for each required low pressure 12 hours j coolant injection (LPCI) subsystem, the suppression pool water level is '> 2: 146 inches. 4 (continued) 1
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i 'w, HATCH UNIT 1 3.5-8 REVISION D
=
i Primary Containment Air Lock l 3.6.1.2 ) i i
! SURVEILLANCE REQUIREMENTS v
SURVEILLANCE FREQUENCY SR 3.6.1.2.1 ------------------NOTES------------------ l
- 1. An inoperable air lock door does not invalidate the previous successfe?,
perform?.nce of the overall air lock leakage test.
- 2. Results shall bt evaluated against acceptance criteria of SR 3.6.1.1.1 /\
in accordance witn 10 CFR 50, LD_\ l Appendix J, as modified by approved ! exemptions. Perform required primary containment air -----NOTE------ lock leakage rate testing in accordance SR 3.0.2 is not ! with 10 CFR 50, Appendix J, as modified applicable j by approved exemations. --------------- l The acceptance criteria for air lock In accordance l testing are: with 10 CFR 50, l ('i Appendix J, as
'w_) a. Overall air lock leakage rate is modified by s 0.05 L. .when tested at 2 P.. approved exemptions
- b. For each door, leakage rate is l
s 0.01 L. when the gap between the f door seals is pressurized to a 10 psig for at least 15 minutes. SR 3.6.1.2.2 ------------------NOTE------------------- Only required to be performed upon entry or exit through the primary containment air lock when the primary containment is de-inerted. l b rify only one door in the primary 184 days i containment air lock can be opened at a l 3 time.
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LJ HATCH UNIT 1 3.6-7 REVISION D g
PCIVs 3.6.1.3 3.6 CONTAINMENT SYSTEMS h 3.6.1.3 Primary Containment Isolation Valves (PCIVs) LC0 3.6.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE. lk APPLICABILITY: MODES 1, 2, and 3, When associated instrumentation is required to be OPERABLE per LC0 3.3.6.1, " Primary Containment Isolation Instrumentation." ACTIONS
NOTES------------------------------------
- 1. Penetration flow path:; except for 18 inch purge valve 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 PCIVs. g
- 4. Enter applicable Conditions and Required Actions of LC0 3.6.1.1, " Primary Containment," when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria.
CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE--------- A.1 Isolate the affected 4 hours except Only applicable to penetration flow path for main steam penetration flow paths by use of at least line - with two PCIVs. one closed and de-
---------------------- activated automatic AND valve, closed manual One or more valve, blind flange, 8 hours for main penetration flow paths or check valve with steam line with one PCIV flow through the inoperable except due valve secured, to leakage not within limit.
AND (continued) HATCH UNIT 1 3.6-8 REVISION D
p PCIVs 3.6.1.3 ( 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 the affected Once per 31 days penetration flow path for isolation is isolated. devices outside primary containment AND Prior to entering MODE 2 or 3 fron MODE 4 if primary (l V containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment (continued) V ; l HATCH UNIT 1 3.6-9 REVISION A i
PCIVs
'3.6.1.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. ---------NOTE--------- B.1 Isolate the affected I hour , Only applicable to penetration flow path { penetration flow paths by use of at least with two PCIVs. one closed and de- A , activated automatic & I valve, closed manual ! One or more valve, or blind I penetration flow paths fl ange. with two PCIVs ! inoperable except due l to leakage not within l limit. i C. ---------NOTE--------- C.1 Isolate the affected 4 hours except Only applicable to penetration flow path for excess flow penetration flow paths by use of at least check valve with only one PCIV. one closed and de- (EFCV) line l activated automatic l valve, closed manual AND One or more valve, or blind penetration flow paths flange. 12 hours for with one PCIV EFCV line inoperable except due AND A to leakage not within L& limits. C.2 --------NOTE--------- Valves and blind I flanges in high radiation areas may be verified by use of administrative means. l Verify the affected Once per 31 days penetration flow path is isolated. (continued) O HATCH UNIT 1 3.6-10 REVISION D
PCIVs 3.6.1.3 ', ) SURVEILLANCE REQUIREMENTS (continued) \J SURVEILLANCE FREQUENCY SR 3.6.1.3.3 ------------------NOTE-------------------
- 1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.
- 2. Not required to be met for PCIVs that are open under administrative controls.
Verify each primary containment manual Prior to isolation valve and blind flange that is entering MODE 2 located inside primary containment and is or 3 if primary required to be closed during accident containment was conditions is closed. de-inerted while in MODE 4, if not performed within the previous tN 92 days U SR 3.6.1.3.4 Verify continuity of the traversing 31 days incore probe (TIP) shear isolation valve explosive charge. _1 SR 3.6.1.3.5 Verify the isolation time of each power In accordance . operated and each automatic PCIV, except with the l for MSIVs, is within limits. Inservice _Il l Testing Program 1 (continued) l l 1 l l ,C% '%.) HATCH UNIT 1 3.6-13 REVISION D
PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify the isolation time of each MSIV is In accordance a: 3 seconds and s 5 seconds, with the Inservice Testing Program SR 3.6.1.3.7 Verify each automatic PCIV, excluding 18 months l EFCVs, actuates to the isolation position on an actual or simulated isolation signal. SR 3.6.1.3.8 Verify each reactor instrumentation line 18 months l EFCV actuates to restrict flow to within limits. SR 3.6.1.3.9 Remove and test the explosive squib from 18 months on a I each shear isolation valve of the TIP STAGGERED TEST system. BASIS SR 3.6.1.3.10 Verify leakage rate through each MSIV is -----NOTE------ 1 s 11.5 scfh when tested at 2 28.0 psig. SR 3.0.2 is not applicable. In accordance with 10 CFR 50, Appendix J, as modified by approved exemptions (continued) O1 l HATCH UNIT 1 3.6-14 REVISION D i l
PCIVs 3.6.1.3 ( SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.1.3.11 Replace the valve seat of each 18 inch 18 months Ib purge valve having a resilient material seat. SR 3.6.1.3.12 Cycle each 18 inch excess flow isolation damper to the fully closed and fully open 18 months Ik position. a v HATCH UNIT 1 3.6-15 REVISION D
Drywell Pressure 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell Pressure LC0 3.6.1.4 Drywell pressure shall be s 1.75 psig. I APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l A. Drywell pressure not A.1 Restore drywell 1 hour within limit. pressure to within limit. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.4.1 Verify drywell pressure is within limit. 12 hours 1 O HATCH UNIT 1 3.6-16 REVISION /hfh I
Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.8 I' D N_,/ 3.6 CONTAINMENT SYSTEMS 3.6.1.8 Suppression Chamber-to-Drywell Vacuum Breakers LC0 3.6.1.8 Ten suppression chamber-to-drywell vacuum breakers shall be OPERABLE for opening. AND Twelve suppression chamber-to-drywell vacuum breakers shall be closed, except when performing their intended function. APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required t. 1 Restore one vacuum 72 hours suppression chamber- breaker to OPERABLE / to-drywell vacuum status. '\ -) breaker inoperable for opening. B. One suppression B.1 Close the open vacuum 2 hours chamber-to-drywell breaker. vacuum breaker not closed. C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion i Time not met. AND C.2 Be in MODE 4. 36 hours 1 (r' i \_ / , i HATCH UNIT 1 3.6-23 REVISION A
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______________L'
Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.8.1 ------------------NOTE------------------- l Not required to be met for vacuum I breakers that are open during ) Surveillances. l Verify each vacuum breaker is closed. 14 days j l 1 l SR 3.6.1.8.2 Perform a functional test of each 31 days required vacuum breaker. AND Within 12 hours j after any I discharge of steam to the b ] suppression chamber from l the S/RVs , SR 3.6.1.8.3 Verify the opening setpoint of each 18 months l required vacuum breaker is :s 0.5 psid. O HATCH UNIT 1 3.6-24 REVISION D
Secondary Containment , 3.6.4.1 ACTIONS bd s ' CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 Suspend CORE Immediately , i ALTERATIONS. AND C.3 Initiate actior. to Imme'diately I suspend OPDRVs. f-SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY-i Verify all secondary containment SR 3.6.4.1.1 31 days equipment hatches are closed and sealed. I - V l 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------------------- ' During movement of irradiated fuel. ' assemblies in the-secondary containment, i CORE ALTERATIONS, and OPDRVs, the draw down time acceptance criteria is s 100 - seconds. Verify two standby gas treatment (SGT) 18 months on.a i subsystems will draw down the' secondary STAGGERED containment to 2: 0.25 inch of vacuum TEST BASIS water gauge in s 120 seconds. : (Continued) HATCH UNIT 1 3.6-37 REVISION D' i
Secondary Containment 3.6.4.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.4.1.4 Verify two SGT subsystems can maintain 18 months on a 2 0.25 inch of vacuum water gauge in the STAGGERED TEST secondary containment for 1 hour at a BASIS flow rate s 4000 cfm for each subsystem. O l l l l O' HATCH UNIT 1 3.6-38 REVISION A !
SCIVs 3.6.4.2 g tg 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) LCO 3.6.4.2 Fach 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 V 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 fl ange. AND g (continued) v HATCH UNIT 1 3.6-39 REVISION D
SCIVs 3.6.4.2 At.il0NS 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 31 days penetration flow path 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 ano 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) 1 i s l O l HATCH UNIT 1 3.6-40 REVISION D 1
AC Sources - Operating 3.8.1 [v SURVEILLANCE REQUIREMENTS
' -----------------------------------NOTE---------------------------------------
SR 3.8.1.1 through SR 3.8.1.18 are applicable only to the Unit 1 AC sources. O SR 3.8.1.19 is applicable only to the Unit 2 AC sources. SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and 7 days indicated power availability for each required offsite circuit. SR 3.8.1.2 -------------------NOTES-------------------
- 1. Performance of SR 3.8.1.5 satisfies this SR.
- 2. All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.
t' 3. A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR as recommended by the manufacturer. When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.5.a must be met.
- 4. For the swing DG, a single test will satisfy this Surveillance for both i units, using the starting circuitry of Unit 1 and synchronized to 4160 V bus IF for one periodic test, and the starting circuitry of Unit 2 and synchronized to 4160 V bus 2F during the next periodic test.
- 5. DG 1oadingsinay. include gradual loading as recommendad by the manufacturer.
- 6. Starting transients above the upper voltage limit do not invalidate this I
test. l lO d (continued) HATCH UNIT 1 3.8-7 REVISION D
l AC Sources - Operating 3.8.1 j SURVEILLANCE REQUIREMENTS (continued) h, SURVEILLANCE FREQUENCY SR 3.8.1.2 NOTES (continued)
- 7. Momentary transients outside the load range do not invalidate this test.
- 8. This Surveillance shall be conducted on only one DG at a time.
Verify each DG: 31 days
- a. Starts from standby conditions and achieves steady state voltage 2 3740 V and s 4243 V and frequency 2 58.8 Hz and s 61.2 Hz; and
- b. Operates for a 60 minutes at a load 21710 kW and s 2000 kW.
SR 3.8.1.3 Verify each day tank contains a 900 gallons 31 days of fuel oil. SR 3.8.1.4 Check for and remove accumulated water from 184 days each day tank. (continued) O HATCH UNIT 1 3.8-8 REVISION
# L 4 . + + , 4-. 4 AC Sources - Operating l 3.8.1 b
_V SURVEILLANCE REQUIREMENTS (continued) , SURVEILLANCE FREQUENCY I SR 3.8.1.5 -------------------NOTES------------------- ;
- 1. All DG starts.may be preceded by an :
engine prelube period. i
- 2. DG loadings may include gradual loading ,
as recommended by the manufacturer. .
- 3. Momentary load transients outside the load range do not invalidate this test.
- 4. This Surveillance shall be conducted on l 4 only one DG at a time. -
'i
- 5. For the swing DG, a single test will l satisfy this Surveillance for both ,
units, using the starting circuitry of Unit I and synchronized to 4160 V bus IF for one periodic test and the starting circuitry of Unit 2 and synchronized to 4160 V bus 2F during ; _______$__$___$_ $ ___$I_ _______________ Verify each DG: [
- a. Starts from standby conditions and ,
achieves, in s 12 seconds, voltage 2 3740 V and frequency a 58.8 Hz and , after steady state conditions are reached, maintains voltage a 3740 V and 184 days s 4243 V and frequency a 58.8 Hz and s 61.2 Hz; and >
- b. Operates for a 60 minutes at a load 2 2250 kW and s 2400 kW for DGs IA ;
and IC, and a 2360 kW and s 2425 kW for DG 1B. (continued) HATCH UNIT 1 3.8-9 REVISION [
=
AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.6 ---------------- NOTE--------------------- This Surveillance shall not be performed in MODE 1 or 2. However, credit may be taken for unplanned events that satisfy this SR. Verify automatic and manual transfer of 18 months unit power supply from the normal offsite circuit to the alternate offsite circuit. _ SR 3.8.1.7 ------------------NOTES--------------------
- 1. This Surveillance shall not be performed in MODE 1 or 2, except for the swing DG. For the swing DG, this Surveillance shall not be performed in MODE 1 or 2 using the Unit I controls.
Credit may be taken for unplanned events that satisfy this SR.
- 2. For the swing DG, a single test at the O
specified Frequency will satisfy this Surveillance for both units. Verify each DG rejects a load greater than 18 months or equal to its associated single largest iA post-accident load, and: /D\,
- a. Following load rejection, the frequency is s 65.5 Hz; and
- b. Within 3 seconds following load -
rejection, the voltage is 2: 3740 V and s 4580 V. (continued) O HATCH UNIT 1 3.8-10 REVISION D
AC Sources - Operating 3.8.1
)- SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.1.8 ------------------NOTES--------------------
- 1. This Surveillance shall not be performed in MODE 1 or 2, except for the swing DG. For the swing DG, this Surveillance shall not be performed in ,
MODE 1 or 2 using the Unit I controls. Credit may be taken for unplanned events that satisfy this SR.
- 2. If grid conditions do not permit, the l power factor limit is not required to D be met. Under this condition, the power factor shall be maintained as close to the limit as practicable.
- 3. For the swing DG, a single test at the i specified Frequency will satisfy this Surveillance for both units.
A __________________________________________ Verify each DG operating at a power factor s 0.88 does not trip and voltage is , maintained s 4800 V during and following a 18 months load rejection of a 2775 kW. (continued) (vD HATCH UNIT 1 3.8-11 REVISION D
I AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.9 -------------------NOTES-------------------
- 1. All DG starts may be preceded by an engine prelube period.
- 2. This Surveillance shall not be performed in MODE 1, 2, or 3.
However, credit may be taken for unplanned events that satisfy this SR. Verify on an actual or simulated loss of 18 months offsite power signal:
- a. De-energization of emergency buses;
- b. Load shedding from emergency buses; and
- c. DG auto-starts from standby condition and:
- 1. energizes permanently connected loads in s 12 seconds,
- 2. energizes auto-connected shutdown loads through automatic load sequence timing devices,
- 3. maintains steady state voltage e a 3740 V and s 4243 V,
- 4. maintains steady state frequency a 58.8 Hz and s 61.2 Hz, and
- 5. supplies permanently connected and auto-connected shutdown loads for a 5 minutes.
(continued) O HATCH UNIT I 3.8-12 REVISION A
AC_ Sources - Operating 3.8.1 (j SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.17 -------------------NOTES-------------------
- 1. All DG starts may be preceded by an engine prelube period.
- 2. This Surveillance shall not be performed in MODE 1, 2, or 3.
However, credit may be taken for unplanned events that satisfy this SR.
...--------- ...------------------ -=-----
Verify, on an actual or simulated loss of 18 months offsite power signal in conjunction with an actual or simulated ECCS initiation signal:
- a. De-energization of emergency buses;
- b. Load shedding from emergency buses; '
and .
- c. DG auto-starts from standby condition and:
- 1. energizes permanently connected '
loads in s 12 seconds,
- 2. energizes auto-connected emergency loads through automatic load sequence timing devices,
- 3. achieves steady state voltage 2 3740 V and s 4243 V,
- 4. achieves steady state frequency -
2 58.8 Hz and s 61.2 Hz, and
- 5. supplies permanently connected and auto-connected emergency loads for a 5 minutes.
(continued) i O i HATCH UNIT 1 3.8-17 REVISION A I i l
AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.18 -------------------NOTE-------------------- t All DG starts may be preceded by an engine prelube period. Verify, when started simultaneously from 10 years standby condition, the Unit 1 DGs and the swing DG achieve, in s 12 seconds, voltage 2 3740 V and frequency 2 58.8 Hz. SR 3.8.1.19 For required Unit 2 AC Sources, the SRs of In accordance Unit 2 Technical Specifications are with applicable, except SR 3.8.1.6 SR 3.8.1.10, applicable SRs O SR 3.8.1.11, SR 3.8.1.15, SR 3.8.1.17 and SR 3.8.1.18. O O HATCH UNIT 1 3.8-18 REVISION D
l AC Sources - Operating- l 3.8.1 : O : i i t 1 O ; I 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 IE AC electrical Id 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 IE AC electrical power distribution subsystem (s) required by LC0 3.8.8;
- c. One qualified circuit connected between the offsite 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 required by LC0 3.6.4.3, "SGT System;" and
- d. One Unit 2 DG capable of supplying the Unit 2 SGT subsystem required by LC0 3.6.4.3.
APPLICABILITY: MODES 4 and 5, O During movement of irradiated fuel assemblies in the ! secondary containment. 1 L l l HATCH UNIT 1 3.8-20 REVISION D 1 l
AC Sources - Shutdown 3.8.2 A ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required ------------NOTE------------- Id offsite circuit (s) Enter applicable Condition 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. j
.03 A.2.1 Suspend CORE Immediately ALTERATIONS.
AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies in the secondary containment. AND A.2.3 Initiate action to sus)end operations Immediately wit 1 a potential for draining the reactor vessel (0PDRVs). AND A.2.4 Initiate action to restore required Immediately offsite power circuit (s), to 2 OPERABLE status. (continued)
-s.
v HATCH UNIT I 3.8-21 REVISION D
AC Sources - Shutdown 3.8.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME
- 8. One or more required DG(s) inoperable.
B.1 Suspend CORE ALTERATIONS. Immediately d AND B.2 Suspend movement of Immediately irradiated fuel assemblies in secondary containment. AND B.3 Initiate action to Immediately suspend OPDRVs. AND B.4 Initiate action to Immediately restore required i DG(s) to OPERABLE status. 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 1 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 l (continued) O HATCH UNIT 1 3.8-22 REVISION D
Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.6.2 Verify battery cell parameters meet 92 days Table 3.8.6-1 Category B limits. AND Once within 24 hours after battery : overcharge l
> 150 V 1
SR 3.8.6.3 Verify average electrolyte temperature of 92 days l representative cells is 2 65 F for each station service battery, and 2 40 F for each DG battery, l i O 1 HATCH UNIT 1 3.8-37 REVISION A
Battery Cell Parameters 3.8.6 Table 3.8.6-1 (page 1 of 2) Battery Cell Parameter Requirements h CATEGORY A: CATEGORY B: CATEGORY C: LIMITS FOR EACH LIMITS FOR EACH LIMITS DESIGNATED PILOT 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 AND average of 111 connected cells Average of all connected cells MD ,
> 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. j i (b) Corrected for electrolyte temperature and level. Level correction is not required. However, when on float charge battery charging is < 1 amp for station service batteries and < 0.5 amp for DG batteries. l (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 9 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 .
measured prior to expiration of the 7 day allowance. I O HATCH UNIT 1 3.8-38 REVISION D
RHR -High Water Level I 3.9.7 ( ) ' ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3 Initiate action to Immediately restore two standby gas treatment subsystems to OPERABLE status. AND B.4 Initiate action to Immediately restore isolation capability in each lh required secondary containment penetration flow path not isolated. C. No RHR shutdown C.1 Verify reactor I hour from
, cooling subsystem in coolant circulation discovery of no
('N ,Js operation. by an alternate reactor coolant i method. circulation MQ Once per 12 hours thereafter AND C.2 Monitor reactor Once per hour coolant temperature. f']
\_
HATCH UNIT 1 3.9-11 REVISION D
.~
RHR -High Water Level 3.9.7 SURVEILLANCE REQUIREMENTS , SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify one RHR shutdown cooling subsystem 12 hours is operating. O O HATCH UNIT 1 3.9-12 REVISION A 1 l
RHR -- Low Water Level 3.9.8
/m It,-) 3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR) -- Low Water Level ;
1 i 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 (~'NI A. One or two required A.1 Verify an alternate 1 hour
'/ RHR shutdown cooling method of decay heat subsystems inoperable. removal is available ANQ for each inoperable required RHR shutdown Once per cooling subsystem. 24 hours thereafter
. Required Action and B.1 Initiate action to Immediately associated Completion restore secondary Time of Condition A containment to not met. OPERABLE status.
aHQ B.2 Initiate action to Immediately restore two standby gas treatment subsystems to OPERABLE status. AND (continued) l I c) HATCH UNIT 1 3.9-13 REVISION A l j l l
RHR - Low Water Level 3.9.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I B. (continued) B.3 Initiate action to Immediately A restore isolation /4L\ capability in each required secondary containment l 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 coolant temperature. l SURVEILLANCE REQUIREMENTS ! SURVEILLANCE FREQdENCY SR 3.9.8.1 Verify one RHR shutdown cooling subsystem 12 hours is operating. 3 O HATCH UNIT 1 3.9-14 REVISION D
p . Inservice Leak and Hydrostatic Testing Operation. > 3.10.1-( ( 3.10 SPECIAL OPERATIONS 13.10.1 Inservice Leak and Hydrostatic Testing Operation-LC0' 3.10.1 The average reactor coolant temperature specified in , j Table 1.1-1 for MODE 4 may be changed to "NA," and operation considercd not to be in MODE 3; and the requirements of A LC0 3.4.8, " Residual Heat Removal;(RHR) Shutdown Cooling- 1 OD l System -- Cold Shutdown," may be suspended, to allow ! performance of an inservice leak.or hydrostatic test provided the following MODE 3 LCOs are met: ;
- a. LCO 3.3.6.2, " Secondary Containment Isolation l Instrumentation," Functions 1, 3, and 4 of .
Table 3.3.6.2-1;
- b. LC0 3.6.4.1, " Secondary Containment";
i
- c. LCO 3.6.4.2, " Secondary Containment Isolation Valves (SCIVs)"; and .;
- d. LCO 3.6.4.3, " Standby Gas Treatment (SGT) System."
APPLICABILITY: MODE 4 with average reactor coolant temperature > 212*F. {} 1 b
)
F i 1 ('>');
- x. -
l HATCH UNIT 1 3.10-1 REVISION D ; a
l Inservice Leak and Hydrostatic Testing Operation 3.10.1 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 met. be in MODE 4 include reducing average reactor coolant temperature to s 212 F. Enter the applicable Immediately Condition of the affected LCO. 0.8 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. l HATCH UNIT 1 3.10-2 REVISION A
I Organization i 5.2 ' t 5.2 Organization 5.2.2 Unit Staff
- a. (continued) the required PE0s shall be assigned to each reactor- '
containing fuel.
- b. At least one licensed Reactor Operator (RO) shall be present
- in the control room for each unit that contains fuel in the '
reactor. In addition, while the unit is in MODE 1, 2, or 3, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.
- c. The minimum shift crew composition shall be in accordance with 10 CFR 50.54(m)(2)(1). Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements. ,
A d. An individual qualified to implement radiation protection _ ' b procedures shall be'on site when fuel is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate ' action is taken to fill the required position.
- e. Administrative procedures shall be developed and-implemented ;
to limit the working hours of unit staff who perform safety related functions (e.g., licensed and non-licensed : operations personnel, health physics technicians, key 4 maintenance personnel, etc.). M , Adequate shift coverage shall be maintained without routine , heavy use of overtime. The objective shall be to have ! operating personnel work a nominal 40 hour week while the i unit is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, . or during extended periods of shutdown for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:
- 1. An individual should not be permitted to work more than I 16 hours straignt, excluding shift turnover time; j i
'(continued) v ,
HATCH UNIT 1 5.0-3 REVISION D
Organization 5.2 5.2 Organization l 5.2.2 Unit Staff
- e. (continued)
- 2. An individual should not be permitted to work more than 16 hours in any 24 hour period, nor more than 24 hours in any 48 hour period, nor more than 72 hours in any 7 day period, all excluding shift turnover time;
~
- 3. A break of at least 8 hours should be allowed between j work periods, including shift turnover time;
- 4. Except during extended shutdown periods, the use of ,
overtime should be considered on an individual basis and not for the entire staff on a shift. Any deviation from the above guidelines shall be authorized by the AGM-PO, Assistant General Manager-Plant Support (AGM-PS), or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that & individual overtime shall be reviewed monthly by the AGM-PO, W AGM-PS, or designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.
- f. The Operations Manager shall hold an active or inactive SRO license.
- g. The Shift Technical Advisor (STA) shall provide advisory l
technical support to the Shift Supervisor (SS) in the areas l of thermal hydraulics, reactor engineering, and plant I analysis with regard to the safe operation of the unit. In l addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift. l 9 HATCH UNIT 1 5.0-4 REVISION A
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.4 Radioactive Effluent Controls Proaram This program conforms to 10 CFR 50.36a for the control of . i radioactive effluents and for maintaining.the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, . shall be l implemented by procedures, and shall include remedial actions to i be taken whenever the program limits are exceeded. The program l shall include the following elements: ;
- a. Limitations on the functional capability of radioactive :
liquid and gaseous monitoring instrumentation, including . surveillance tests and setpoint determination, in accordance ! with the methodology in the ODCM;
- b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 times the concentrations stated in 10 CFR 20, Appendix B (to paragraphs 20.1001-20.2401),
Table 2, Column 2; k .i
- c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with O. the methodology and parameters in the ODCM-l
- d. Limitations on the annual and quarterly. doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to. j unrestricted areas, conforming to 10 CFR 50, Appendix I; -
- e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter-and current calendar year, in accordance with the methodology and parameters in the ODCM, at least every 31 days;
- f. Limitations on the functional capability and use of the ,
liquid and gaseous effluent treatment systems to ensure that 1 appropriate portions of these systems are'used to reduce releases of radioactivity when the projected doses in'a ! period of 31 days would exceed 2% of the guidelines for the ! annual dose or dose commitment, conforming to 10 CFR 50, l Appendix I; i
? (continued) ]
HATCH UNIT 1 5.0-9 REVISION D J
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Proaram (continued) 9 Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary as follows:
- 1) For noble gases, less than or equal to a dose rate of 500 mrem / year to the total body and less than or equal to a dose rate of 3000 mrem / year to the skin, and
- 2) For Iodine-131, Iodine-133, tritium, and all _
radionuclides in particulate form with half-lives greater than 8 days, less than or equal to a dose rate of 1500 mrem / year to any organ;
- h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;
- i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and
- j. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.
5.5.5 Component Cyclic or Transient Limit This program provides controls to track FSAR Section 4.2, cyclic and transient occurrences, to ensure that reactor coolant pressure boundary components are maintained within the design limit.s. 5.5.6 Inservice Testina Proaram This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports.
- a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda are as follows:
(continued) HATCH UNIT 1 5.0-10 REVISION D t - - - - - - _ _
;1 Programs and Manuals. .!
5.5 l lj 5.5' ' Programs and Manuals-
-1 5.5.6 Inservice Testino Prooram (continued): a l
ASME Boiler and Pressure Vessel Code and Applicable Required Frequencies Addenda Terminology for- for Performing Inservice ; Inservice Testina Activities' Testina Activities ' ; Weekly. Monthly At least once per 7 days At least once per 31 days gi . Quarterly or every 3 months At least once per 92 days Semiannually' or every 6 months At least once per 184 days. , Yearly or annually At least once per 366 days j
- b. The provisions of SR 3.0.2 are applicable to the frequencies j for performing inservice testing activities; j
- c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and I .4 i l
- d. Nothing in the ASME Boiler and. Pressure' Vessel Cod'e shall be. IA construed to supersede the requirements of any Technical ,i Specification.
'u l
i I l i i (continued) HATCH UNIT.1- 5.0-10A REVISION D ! i i
, _ . . = . _
c d Reporting Requirements 5.6 i 5.6 Reporting Requirements (continued) ; 5.6.5 CORE OPERATING LIMITS REPORT (COLR) , t
- a. Core operating limits shall be established prior to each t reload cycle, or prior to any remaining portion of a reload :
cycle, and shall be documented in the COLR for the ' following: !
- 1) Control Rod Block Instrumentation - Rod Block Monitor r for Specification 3.3.2.1. ;
- 2) The Average Planar Linear Heat Generation Rate for :
Specification 3.2.1.
- 3) The Minimum Critical Power Ratio for Specifications [
3.2.2 and 3.3.2.1. -
- b. The analytical methods used to determine the core' operating limits shall be those previously reviewed and approved by l the NRC, specifically those described in the following documents: .
- 1) NEDE-24011-P-A, " General Electric' Standard Applic$ tion -
for Reactor Fuel," (applicable amendment specified in I the COLR). .
- 2) " Safety Evaluation by the Office of Nuclear Reactor :!
Regulation Supporting Amendment No.157 to Facility Operating License DPR-57," dated September 12, 1988. j
- c. The core operating limits'shall be' determined such that all ,
applicable limits (e.g., fuel thermal mechanical limits, ' core thermal hydraulic limits, Emergency Core Cooling. A - Systems (ECCS) limits, nuclear limits such as SDM, transient' * , analysis limits and accident analysis limits) of the* safety , analysis are met.
- d. The COLR, including any mid-cycle revisions or supplements, I shall be provided upon issuance for each reload cycle'to the j NRC. ,
~5.6.6 Reactor Coolant System (RCS) PRESSURE AND' TEMPERATURE LIMITS. ,
REPORT (PTLR) >
- a. RCS pressure and temperature limits for_heatup, cooldown, :
low temperature operation, criticality, and hydrostatic l testing as well as heatup and cooldown rates shall be (continued) U ! HATCH UNIT 1 5.0-19 REVISION _D
Reporting RequirGments 5.6 5.6 Reporting Requirements (continued) 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
- a. (continued) established and documented in the PTLR for LCO 3.4.9, "RCS Pressure and Temperature (P/T) Limits."
- b. The analytical methods used to determine the RCS pressure and temperature limits shall be determined in accordance with Regulatory Guide 1.99.
- c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluency period and for any revision or supplement thereto.
5.6.7 Post Accident Monitorina (PAMi Instrumentation Report When a report is required by LC0 3.3.3.1, " Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the < inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. I i 1 HATCH UNIT 1 5.0-20 REVISION /
/
Reporting Requirements 5.7 \n 'J 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area 5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601, in lieu of the requirements of 10 CFR 20.1601a, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is > 100 mrem /hr but < 1000 mrem /hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area. Entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Health Physics Technicians) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiatior, areas with exposure rates
< 1000 mrem /hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas. g Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device that continuously indicates l'g the radiation dose rate in the area.
V)
- b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.
- c. An individual qualified in radiation protecton procedures with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by-the facility Health Physics supervision in the RWP.
I l o 1 HATCH UNIT 1 5.0-21 REVISION D
I . l l- - Reporting Requirements 5.7 5.7 High Radiation Area (continued) 5.7.2 In addition to the requirements of Specification 5.7.I, areas with radiation levels 2: 1000 mrem /hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, but less than 500 Rads in I hour measured at 1 meter from the radiation source or from any surface that the radiation penetrates, shall be provided with locked or continuously guarded d doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervision on duty or Health Physics supervision. 0 1 O HATCH UNIT 1 5.0-22 REVISION D
;,l J
! l
/ UNIT I IMPROVED BASES l P
i
'l f
I l l i 3 , i I I t 6 I I I i I
t RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs) B 2.1.2 Reactor Coolant System (RCS) Pressure SL j i BASES BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding-failure, fission products are released into the reactor , coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the ; atmosphere. Establishing an upper limit on reattor steam dome pressure ensures continued RCS integrity. Per 10 CFR 50, Appendix A, GDC 14, " Reactor Coolant Pressure Boundary," and GDC 15, " Reactor Coolant System Design" . t (Ref.1), the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and l anticipated operational. occurrences (A00s). ; 1 During normal operation and A00s, RCS pressure is limited from exceeding the design pressure by more than 10%, in ; accordance with Section III'of the ASME Code (Ref. 2). To ; ensure system integrity, all RCS components are i hydrostatically tested at 125% of design pressure, in : accordance with ASME Code requirements, prior to initial ' operation when there is no fuel in the core. Any further , hydrostatic testing with fuel in the core may be done under i LCO 3.10.1, " Inservice Leak and Hydrostatic Testing Operation." Follow ng inception of unit operation, RCS t components shall be pressure tested in accordance with.the i requirements of ASME Code, Section XI (Ref. 3). Overpressurization of the RCS could result.in a breach of the RCPB, reducing the number of protective barriers ! designed to prevent radioactive releases from exceeding the- A limits specified in 10 CFR 100, " Reactor Site Criteria" dd , (Ref. 4). If this occurred in conjunction with a fuel : cladding failure, fission products could enter the containment atmosphere. 1 I 6 I (continued) HATCH UNIT 1 B 2.0-7 REVISION D
RCS Pressure SL B 2.1.2 BASES (continued) h APPLICABLE The RCS safety / relief valves and the Reactor Protection SAFETY ANALYSES System Reactor Vessel Steam Dome Pressure - High Function have settings established to ensure that the RCS pressure SL will not be exceeded. The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel Code, 1965 Edition, including Addenda through the Winter of 1966 (Ref. 5), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig. The SL of 1325 psig, as measured in the reactor steam dome, is equivalent to 1375 psig at the lowest elevation of the RCS. The RCS is designed to the USAS Nuclear Power Piping Code, Section B31.1, 1967 Edition, including Addenda A, C, and D (Ref. 6), for the reactor recirculation piping, which permits a maximum pressure transient of 120% of design pressures of 1150 psig for suction piping and 1325 psig for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes. O SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code, Section III, is 110% of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 120% of design pressures of 1150 psig for suction piping and 1325 psig for discharge piping. The most limiting of these two allowances is the 110% of the reactor vessel design pressure; therefore, the SL on maximum allowable RCS pressure is established at 1325 psig as measured at the reactor steam dome. APPLICABILITY SL 2.1.2 applies in all MODES. l 1 i (continued) 1 HATCH UNIT 1 B 2.0-8 REVISION A
LCO Applicability B 3.0 [v] BASES LCO 3.0.3 assemblies in the spent fuel storage pool." Therefore, this (continued) LC0 can be applicable in any or all MODES. If tl.a LC0 and the Required Actions of LC0 3.7.8 are not met while in MODE 1, 2, or.3, there is no safety benefit.to be gained by placing the unit in a shutdown condition. The Required Action of LC0 3.7.8 of " Suspend movement of irradiated fuel assemblies in the spent fuel storage pool" is the appropriate Required Action to complete in lieu of the actions of LC0 3.0.3. These exceptions are addressed in the individual Specifications. LCO 3.0.4 LC0 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It precludes placing the unit in a MODE or other specified condition stated in that LC0's Applicability (e.g., Applicability desired to be entered) when the following exist:
- a. Plant conditions are such that the requirements of an LC0 would not be met in the Applicability desired to
/3 be entered; and O b. Continued noncompliance with these LC0 requirements, if that Applicability were entered, would result in the unit being required to exit the Applicability desired to be entered to comply with the Required Actions. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change. Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before unit startup. ; The provisions of LC0 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability i [D w) (continued) ) HATCH UNIT I B 3.0-5 REVISION A i
LC0 Applicability B 3.0 BASES LCO 3.0.4 that are required to comply with ACTIONS. In addition, the (continued) provisions of LC0 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. Exceptions to LCO 3.0.4 are stated in the individual Specifications. Exceptions may apply to all the ACTIONS or to a specific Required Action of a Specification. Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing MODES or other specified conditions while in an ACTIONS Condition, either in compliance with LC0 3.0.4 or where an exception to LC0 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO. LC0 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment O to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LC0 3.0.2 (e.g., to not comply with the applicable Required Action (s)) to allow the performance of SRs to demonstrate:
- a. The OPERABILITY of the equipment being returned to service; or lA
- b. The OPERABILITY of other equipment.
g The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the allowed SRs. This Specification does not provide time to perform any other preventive or corrective maintenance. (continued) I HATCH UNIT 1 B 3.0-6 REVISION D l
SDM B 3.1.1 i BASES (continued) ACTIONS M With SDM not within the limits of the LC0 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, and the low probability of an event occurring during this interval. M 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. () Cl 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.1. D.2. D.3. and D.4 Vith 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 1 hour to provide means for control of potential radioactive releases. This includes ensuring:
- 1) secondary containment is OPERABLE; 2) at least two Standby Gas Treatment (SGT) subsystems are OPERABLE (any combination of Unit 1 and Unit 2 subsystems); and g
- 3) secondary containment isolation capability is available (i.e., at least one secondary containment isolation valve j and associated instrumentation ]
(continued) HATCH UNIT 1 B 3.1-3 REVISION D I l
SDM i B 3.1.1 1 BASES h ACTIONS D.I. D.2. D.3. and D.4 (continued) are OPERABLE, or other acceptable administrative controls to assure isolation capability) in each associated secondary A 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 it; operable,.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. L1. 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 & W 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 I hour to provide means for control of potential radioactive releases. This includes ensuring: 1) secondary containment is OPERABLE; Id (continued) , HATCH UNIT 1 B 3.1-4 REVISION D
I l SDM" " l B 3.1.1' ( BASES 2).at least two SGT' subsystems are OPERABLE (any combination , of Unit I and Unit 2 subsystems); and 3) secondary- _ i containment. isolation capability is available'(i.e., at A-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 , i t r h r l () 1 1 l i i
- i L' (continued) '
HATCH UNIT 1- B 3.1-4A REVISION D
SDM B 3.1.1 ! {j BASES ACTIONS E.1. E.2. E.3. E.4. and E.5 (continued) isolated that is assumed to be isolated to mitigate radioactivity releases. This may be performed as an g 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 iA 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 (l u/ movement or 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 control 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 (B0C) test must also account for changes in core reactivity during the cycle. Therefore, to 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 BOC 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 B0C 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 acccent for uncertainties in ! the calculation of the highest worth control rod. 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 m :
. (continued)
HATCH UNIT 1 B 3.1-5 REVISION D i
SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued) REQUIREMENTS 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 the core) is required to ensure adequate SDM is maintained l 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 l reactive configurations during the refueling may be i performed to demonstrate acceptability of the entire fuel ! movement sequence. These bounding analyses include l additional margins to the SDM limit to account for the l associated uncertainties. Spiral offload / reload sequences l inherently satisfy the SR, provided the fuel assemblies are j reloaded in the same configuration analyzed for the new j cycle. Removing fuel from the core will always result in an ' increase in SDM. i REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.
- 2. FSAR, Section 14.4.2.
- 3. NEDE-24011-P-A-US, " General Electric Standard Application for Reactor Fuel," Supplement for United States, (revision specified in the COLR).
- 4. FSAR, Section 14.3.3.3.
(continued) HATCH UNIT 1 B 3.1-6 REVISION A
Control Rod Scram Times B 3.1.4 BASES i APPLICABLE The scram function of_the CRD System protects the MCPR SAFETY ANALYSES Safety Limit (SL) (see Bases for SL 2.1.1, " Reactor Core (continued) SLs" and LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)") and the 1% cladding plastic strain fuel design limit (see ! Bases for LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded. Above 800 psig, the scram function is designed to insert negative reactivity at a rate-fast enough to prevent the actual MCPR from becoming less than the MCPR SL, during the analyzed limiting power transient. Below 800 psig, the scram function is assumed'to perform during the control rod drop accident (Ref. 5) and, therefore, also provides protection against violating fuel _ damage 1imits during reactivity insertion accidents (see Bases for LC0 3.1.6, " Rod Pattern Control"). For the reactor vessel overpressure protection analysis, the scram , function, along with the safety / relief valves, ensure that' the peak vessel pressure is maintained within the applicable ASME Code limits. Control rod scram times satisfy Criterion 3 of the NRC A Policy Statement (Ref 8.) lA O-LC0 The scram times specified in Table 3.1.4-1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met (Ref. 6). To account for single failures and " slow" : scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis, The scram times have a margin that allows - up to approxim'ately 7% of the control rods (e.g., .137:x 7%
= 10) to have scram times exceeding the specified limits (i.e., " slow" control rods) assuming a single' stuck control rod (as allowed by LC0 3.1.3, " Control- Rod OPERABILITY") and ,
an additional: control rod failing to scram per the single failure criterion. The scram times are specified as a , function.of reactor steam dome pressure to account.for the. t pressure dependence of the scram times. The scram times are specified relative _ to measurements based ~ on reed switch positions, which provide the control rod position ; indication. The reed switch closes (" pickup") when the index tube passes a specific location and then opens 1 (" dropout") as the index tube travels upward. Verification
- f (continued)
HATCH UNIT 1 B 3.1-23 REVISION D
Control Rod Scram Times B 3.1.4 BASES LC0 of the specified scram times in Table 3.1.4-1 is (continued) accomplished through measurement of the " dropout" times. To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed
" slow" control rods may occupy adjacent locations.
Table 3.1.4-1 is modified by two Notes, which state that control rods with scram times not within the limits of the Table are considered " slow" and that control rods with scram times > 7 seconds are considered inoperable as required by SR 3.1.3.4. This LC0 applies only to OPERABLE control rods since inoperable control rods will be inserted and disarmed (LC0 3.1.3). Slow scramming control rods may be conservatively declared inoperable and not accounted for as " slow" control rods. APPLICABILITY In MODES 1 and 2, a scram is assumed to function during transients and accidents analyzed for these plant conditions. These events are assumed to occur during startup and power operation; therefore, the scram function of the control rods is required during these MODES. In MODES 3 and 4, with the mode switch in shutdown, control rod block prevents withdrawal of control rods. This provides adequate requirements for control rod scram capability during these conditions. Scram requirements in MODE 5 are contained in LC0 3.9.5, " Control Rod OPERABILITY - Refueling." ACTIONS L1_ When the requirements of this LC0 are not met, the rate of negative reactivity insertion during a scram may not be within the assumptions of the safety analysis. Therefore, 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 MODE 3 within 12 hours. 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. (continued) HATCH UNIT 1 B 3.1-24 REVISION A
1 l Control Rod Scram Times B 3.1.4 [w. BASES (continued) SURVEILLANCE The four SRs of this LC0 are modified by a Note stating that REQUIREMENTS during a single control rod scram time Surveillance, the CRD ' pumps shall be isolated from the associated scram accumulator. With the CRD pump isolated, (i.e., charging valve closed) the influence of the CRD pump head does not affect the single control rod scram times. During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times. SR 3.1.4.1 The scram reactivity used in DBA and transient analyses is based on an assumed control rod scram time. Measurement of the scram times with reactor steam dome pressure 2 800 psig demonstrates acceptable scram times for the transients analyzed in References 3 and 4. Maximum scram insertion times occur at a reactor steam dome pressure of approximately 800 psig because of the competing effects of reactor steam dome pressure and stored (~3 accumulator energy. Therefore, demonstration of adequate V scram times at reactor steam dome pressure 2 800 psig ensures that the measured scram times will be within the specified limits at higher pressures. Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed. To ensure that scram time testing is performed within a reasonable time following fuel movement within the reactor pressure vessel or after a shutdown 2120 days or longer, control rods are required to be tested before exceeding 40% RTP. In the event fuel movement is limited to selected core cells, it is the intent of this SR that only those CRDs associated with the core cells affected by the fuel movements are required to be scram time tested. This Frequency is acceptable considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by work on i control rods or the CRD System. ] l / \ (continued) V HATCH UNIT 1 B 3.1-25 REVISION A ! I
Control Rod Scram Times B 3.1.4 BASES h SURVEILLANCE SR 3.1.4.2 REQUIREMENTS (continued) Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods. The sample remains representative if no more than 20% of the control rods in the sample tested are determined to be " slow". With more than 20% of the sample declared to be " slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 20% criterion (i.e., 20% of the entire sample size) is satisfied, or until the total number of " slow" control rods (throughout the core, from all Surveillances) exceeds the LC0 limit. For )lanned testing, the control rods selected for the sample siould be different for each test. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample. The 120 day Frequency is based on operating experience that has shown control rod scram times do not significantly change over an operating cycle. This Frequency is also reasonable based on the additional Surveillances done on the CRDs at more frequent intervals in accordance with LC0 3.1.3 and LC0 3.1.5, " Control Rod Scram Accumulators." SR 3.1.4.3 O When work that could affect the scram insertion time is aerformed on a con: il rod or the CRD System, testing must 3e done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure. The scram testing must be performed once before declaring the control rod OPERABLE. The required scram time testing must demonstrate the affected control rod is still within acceptable limits. The limits for reactor pressures < 800 ')sig, required by footnote (b), A are included in the Technicai Requirements Manual (Ref. 7) ZM and are established based on a high probability of meeting the acceptance criteria at reactor pressures 2 800 psig. The limits for reactor pressures 2 800 psig are found in Table 3.1.4-1. If testing demonstrates the affected control rod does not meet these limits, but is within the 7 second j' limit of Table 3.1.4-1, Note 2, the control rod can be declared OPERABLE and " slow." Specific examples of work that could affect the scram times , are (but are not limited to) the following: removal of any ' CRD for maintenance or modification; replacement of a (continued) HATCH UNIT I B 3.1-26 REVISION D . l I
Control Rod Scram Times 1
-B 3.1.4 ,
BASES SURVEILLANCE SR 3.1.4.3 -(continued) . REQUIREMENTS control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator, isolation valve or check valve in the piping required for scram. l The Frequency of once prior to declaring the affected . control rod OPERABLE-is acceptable because of the capability- - to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of . control rod OPERABILITY. SR 3.1.4.4 ; When work that could affect the scram insertion time is performed on a control rod or CRD System, testing must be done to demonstrate each affected control rod is still within the limits of Table 3.1.4-1 with the reactor steam dome pressure 2 800 psig. Where work has been performed at-high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 can be satisfied with one test. However, for a control rod affected by work performed while shutdown, a . O- zero pressure test and a high pressure test may be required. This testing ensures that, prior to withdrawing the control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions. Alternatively, a control rod scram test during , hydrostatic pressure testing could also satisfy both criteria. The Frequency of once prior to exceeding 40% RTP is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY. This test is also used to demonstrate control rod OPERABILITY when a 40% RTP after work that could affect the . scram insertion time is performed on the CRD system. REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.
- 2. FSAR, Section 3.4.
- 3. FSAR, Appendix M. j l
d( (continued) . HATCH UNIT 1 B 3.1-27 REVISION A i
Control Rod Scram Times B 3.1.4 BASES REFERENCES 4. FSAR, Sections 14.3 and 14.4. (continued)
- 5. NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel," (revision specified in the COLR).
- 6. Letter from R. F. Janecek (BWROG) to R. W. Starostecki (NRC), "BWR Owners' Group Revised Reactivity Control Systems Technical Specifications", BWR0G-8754, September 17, 1987.
- 7. Technical Requirements Manual. Id
- 8. NRC No. 93-102, " Final Policy Statement on Technical IA Specification Improvements," July 23, 1993.
4 O i O HATCH UNIT'1 B 3.1-28 REVISION D
i j RPS Instrumentation B 3.3.1.1 p () BASES j l l SURVEILLANCE time required to perform channel Surveillance. Thatanalysis REQUIREMENTS demonstrated that the 6 hour testing allowance does (continued) not significantly reduce the probability that the RPS will trip when necessary. SR 3.3.1.1.1 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 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 determined by the plant staff based (V) 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 upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.1.1.2 To ensure that the APRMs are accurately indicating thE true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Frequency of once per 7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading between performances of SR 3.3.1.1.8. (continued) HATCH UNIT 1 B 3.3-25 REVISION A
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.2 (continued) REQUIREMENTS A restriction to satisfying this SR when < 25% RTP is provided that requires the SR to be met only at 2 25% RTP Ik because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when < 25% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At 2 25% RTP, the Surveillance is required to have been satisfactorily performed within the last 7 days, in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after reaching or exceeding 25% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. SR 3.3.1.1.3 The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function uses the recirculation loop drive flows to vary the trip setpoint. This SR ensures that the total loop drive flow signals from the flow units used to vary the setpoint is appropriately compared to an injection test flow signal to verify the flow signal trip setpoint and, therefore, the APRM Function accurately reflects the required setpoint as a function of flow. If the flow unit signal is not within the appropriate limit, one required APRM that receives an input from the inoperable flow unit must be declared inoperable. The Frequency of 7 days is based on engineering judgment, operating experience, and the reliability of this instrumentation. SR 3.3.1.1.4 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. (continued) HATCH UNIT 1 B 3.3-26 REVISION D
SRM Instrumentation B 3.3.1.2 m ( v
) BASES SURVEILLANCE SR 3.3.1.2.2 (continued)
REQUIREMENTS where CORE ALTERATIONS are being performed, and the other OPERABLE SRM must be in an adjacent quadrant containing fuel. Note I states that the SR is required to be met only during CORE ALTERATIONS. It is not required to be met at other times in MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a review of plant logs to ensure that SRMs required to be OPERABLE for given CORE ALTERATIONS are, in fact, OPERABLE. In the event that only one SRM is required to be OPERABLE (when the fueled region encompasses only one SRM), per Table 3.3.1.2-1, footnote (b), only the a. portion of this SR is required. Note 2 clarifies that more than one of the three regirements can be met by the same OPERABLE SRM. The 12 hour Frequency is based upon operating experience and supplements operational controls over refueling activities that include steps to ensure that the SRMs required by the LC0 are in the proper quadrant. SR 3.3.1.2.4 This Surveillance consists of a verification of the SRM instrument readout to ensure that the SRM reading is greater than a specified minimum count rate, which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. With few fuel assemblies loaded, the SRMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient " source" material, in the form of irradiated fuel assemblies, to establish the minimum count rate. To accomplish this, the SR is modified by a Note (Note 1) that states that the count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies are in the associated core quadrant. With four or less fuel assemblies loaded around each SRM and no other fuel assemblies in the associated core quadrant, even with a control rod withdrawn, the configuration will not be critical. In addition, Note 2 states that this requirement does not have to be met during spiral unloading. If the core is being unloaded in this manner, the various core configurations encountered will not be critical. g ( (continued) HATCH UNIT 1 B 3.3-39 REVISION A
SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.4 (continued) REQUIREMENTS The Frequency is based upon channel redundancy and other information available in the control room, and ensures that the required channels are frequently monitored while core o reactivity changes are occurring. When no reactivity changes are in progress, the Frequency is relaxed from 12 hours to 24 hours. SR 3.3.1.2.5 and SR 3.3.1.2.6 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated channel will function properly. SR 3.3.1.2.5 is required in MODE 5, and the 7 day Frequency ensures that the lA I channels are OPERABLE while core reactivity changes could be in progress. This Frequency is reasonable, based on operating experience and on other Surveillances (such as a CHANNEL CHECK), that ensure proper functioning between CHANNEL FUNCTIONAL TESTS. SR 3.3.1.2.6 is required in MODE 2 with IRMs on Range 2 or below, and in MODES 3 and 4. Since core reactivity changes do not normally take place in MODES 3 and 4 and core l reactivity changes are due only to control rod movement in l MODE 2, the Frequency has been extended from 7 days to 31 days. The 31 day Frequency is based on operating experience and on other Surveillances (such as CHANNEL CHECK) that ensure proper functioning between CHANNEL l FUNCTIONAL TESTS. Verification of the signal to noise ratio also ensures that the detectors are inserted to an acceptable operating level. In a fully withdrawn condition, the detectors are sufficiently removed from the fueled region of the core to essentially eliminate neutrons from reaching the detector. A Any count rate obtained while the detectors are fully withdrawn is assumed to be " noise" only. The Note to the SR 3.3.1.2.6 allows the Surveillance to be IA delayed until entry into the specified condition of the Applicability (THERMAL POWER decreased to IRM Range 2 or below). The SR must be performed within 12 hours after IRMs are on Range 2 or below. The allowance to enter the Applicability with the 31 day frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. (continued) HATCH UNIT 1 3.3 40 REVISION D
m Control Rod Block Instrumentation i B 3.3.2.1 BASES BACKGROUND The purpose of the RWM is to control rod patterns during . (continued) startup and shutdown, such that only specified control rod j sequences and relative positions are allowed over the i operating range from all control rods inserted to 10% RTP. : The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed , control rod sequences are stored in the RWM, which will I initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored i sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses 1 feedwater flow and steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed (Ref. 2). The RWM is a single channel system that provides input into both RMCS rod block circuits. l I With the reactor mode switch in the shutdown position, a l control rod withdrawal block is applied to all control rods 1 to ensure that the shutdown condition is maintained. This j Function prevents inadvertent criticality as the result of a i control rod withdrawal during MODE 3 or 4, or during MODE 5
/^ when the reactor mode switch is required to be in the !
(_)\ shutdown position. The reactor mode switch has two , channels, each inputting into a separate RMCS rod block : circuit. A rod block in either RMCS circuit will provide a ! control rod block to all control rods. l APPLICABLE 1. Rod Block Monitor SAFETY ANALYSES, LCO, and The RBM is designed to prevent violation of the MCPR APPLICABILITY SL and the cladding 1% plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event. The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 3. A statistical analysis of RWE events was performed to determine the RBM response for both channels for each event. From these responses, the fuel thermal performance as a function of RBM Allowable Value was determined. The Allowable Values are chosen as a function of power level. Based on the specified Allowable Values, operating limits are established. t
/' N .( ) (continued)
HATCH UNIT 1 B 3.3-43 REVISION A
Control Rod Block Instrumentation B 3.3.2.1 BASES h APPLICABLE 1 Rod Block Monitor (continued) ' SAFETY ANALYSES, LCO, and The RBM Function satisfies Criterion 3 of the NRC Policy APPLICABILITY Statement (Ref. 10). l Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Values, to ensure that no single instrument failure can preclude a rod block from this Function. The setpoints are calibrated consistent with applicable setpoint methodology (nominal trip setpoint). ! Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are thrse predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), 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 i from the limiting values of the process parameters obtained I 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. l The RBM is assumed to mitigate the consequences of an RWE event when operating 2 29% RTP. Below this power level, the ! consequences of an RWE event will not exceed the MCPR SL ! and, therefore, the RBM is not required to be OPERABLE l (Ref. 3) . When operating < 90% RTP, analyses (Ref. 3) have i shown that with an initial MCPR 21.70, no RWE event will result in exceeding the MCPR SL. Also, the analyses l demonstrate that when operating at E 90% RTP with l MCPR 21.40, no RWE event will result in exceeding the MCPR l (continued) HATCH UNIT 1 B 3.3-44 REVISION D
Control Rod Block Instrumentation B 3.3.2.1 l BASES 1 APPLICABLE 1. Rod Block Monitor (continued) l SAFETY ANALYSES, LCO, and SL (Ref. 3). Therefore, under these' conditions, the APPLICABILITY RBM is also not required to be OPERABLE.
- 2. Rod Worth Minimizer The RWM enforces the banked position withdrawal sequence (BPWS) to ensure that the initial conditions of the CRDA analysis are not violated. The analytical- methods and ,
assumptions used in evaluating the CRDA are summarized in References 4, 5, 6, and 7. In. addition, the Reference 6 analysis (Generic BPWS analysis) may be modified by plant specific evaluations. The BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked , positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LC0 3.1.6, " Rod Pattern Control." The RWM Function satisfies Criterion 3 of the NRC Policy
. Statement (Ref. 10). I A )
Since the RWM is a system designed to act as a backup to operator control of the rod sequences, only one channel of i the RWM is available and required to be OPERABLE (Re '. 7), 1 Special circumstances provided for in the Required A tion of l LCO 3.1.3, " Control Rod OPERABILITY," and LC0 3.1.6 may necessitata bypassing the RWM to allow continued operation with inoperable-control rods, or to allow correction of a ) control rod pattern not in compliance with the BPWS. The ! RWM may be bypassed as required by these conditions, but j then it must.be considered inoperable and the Required Actions of this LC0 followed. Compliance with the BPWS, and therefore OPERABILITY.of the RWM, is required in MODES 1 and 2 when THERMAL POWER is
< 10% RTP. When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal /gm fuel damage limit during a CRDA (Refs. 5 and 7). In MODES 3 and .4, all control rods are required to be inserted into the core;.
therefore, a CRDA cannot occur. In MODE 5, since only a i (continued) HATCH UNIT 1 B 3.3-45 REVISION D I l
Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 2. Rod Worth Minimizer (continued) SAFETY ANALYSES, LCO, and single control rod can be withdrawn from a core cell APPLICABILITY containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.
- 3. Reactor Mode Switch - Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch - Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.
The Reactor Mode Switch - Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10). IA Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when g required. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on reactor mode switch position. During shutdown conditions (MODE 3, 4, or 5), no positive reactivity insertion events are analyzed because assumptions are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, the control rod withdrawal block is required to be OPERABLE. During MODE 5 with the reactor mode switch in the refueling position, the refuel position one-rod-out interlock (LC0 3.9.2, " Refuel Position One-Rod-Out Interlock") provides the required control rod withdrawal blocks. ACTIONS Ad With one RBM channel inoperable, the remaining OPERABLE channel is adequate to oerform the control rod block function; however, overall reliability is reduced because a (continued) HATCH UNIT 1 B 3.3-46 REVISION D
l I Control Rod Block Instrumentation B 3.3.2;1 m) ( BASES l ACTIONS E.1 and E.2 (continued) required to be inserted. Action must continue until L11 insertable control rods in core cells containing one or more fuel assemblies are fully inserted. SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Control Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1. The Surveillances are modified by a second Note to indicate that when an RBM 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 control rod block 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. n This Note is based on the reliability analysis (Ref. 9) + i assumption of the average time required to perform channel O Surveillance. That analysis demonstrated that the 6 hour A m : testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary. I SR 3.3.2.1.1 , A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. It includes the Reactor Manual Control System j input. l Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on reliability analyses (Ref. B). SR 3.3.2.1.2 and SR 3.3.2.1.3 ! A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure ! that the entire system will perform the intended function. J
/^\
Q (continued) y l HATCH UNIT I B 3.3-49 REVISION D ; i l
Control Rod Block Instrumentation B 3.3.2.1 BASES The CHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block , occurs. This test is performed as soon as possible after l' the applicable conditions are entered. As noted in the SRs, 1 SR 3.3.2.1.2 is not required to be performed until I hour e i (continued) HATCH UNIT 1 B 3.3-49A REVISION D l'
l Control Rod Block Ir.strumentation 4 8 3.3.2.1 l G i b) BASES SURVEILLANCE SR 3.3.2.1.2 and SR 3.3.2.1.3 (continued) RE0UIREMENTS after any control rod is withdrawn at < 10% RTP in MODE 2, 1 1 and SR 3.3.2.1.3 is not required to be performed until ! I hour after THERMAL POWER is < 10% RTP in MODE 1. This allows entry into MODE 2 (and if entered during a shutdown, concurrent power reduction to < 10% RTP) for SR 3.3.2.1.2 lk . and THERMAL POWER reduction to < 10% RTP in MODE 1 for SR 3.3.2.1.3.to perform the required Surveillances if the 92 day Frequency is not met per SR 3.0.2. The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The 92 day Frequencies are based on reliability analysis (Ref. 8). SR 3.3.2.1.4 The RBM setpoints are automatically varied as a function of power. Three Allowable Values are specified in , Table 3.3.2.1-1, each within a specific power range. The j power at which the control rod block Allowable Values , l (3 automatically change are based on the APRM signal's input to l (/ each RBM channel. Below the minimum power setpoint, the RBM ; is automatically bypassed. These power Allowable Values i must be verified periodically to be less than or equal to ' the specified values. If any power range setpoint is nonconservative, then the affected RBM channel is considered inoperable. Alternatively, the power range channel can be placed in the conservative condition (i.e., enabling the proper RBM setpoint). If placed in this condition, the SR , is met and the RBM channel is not considered inoperable. As ! noted, neutron detectors are excluded from the Surveillance ! because they are passive devices, with minimal drift, and ; because of the difficulty of simulating a meaningful signal. 1 Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8. The 18 month Frequency is based on the actual trip setpoint methodology utilized for these channels. i l SR 3.3.2.1.5 l The RWM is automatically bypassed when power is above a specified value. The power level is determined from ; feedwater flow and steam flow signals. The automatic bypass setpoint must be verified periodically to be 2: 10% RTP. If n (v ) (continued) HATCH UNIT 1 B 3.3-50 REVISION D j
i Control Rod Block Instrumentation j B 3.3.2.1 BASES the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power O (continued) h]i i HATCH UNIT 1 B 3.3-50A REVISION D
n. Control Rod Block Instrumentation B 3.3.2.1
,) BASES REFERENCES 6. NED0-21231, " Banked Position Withdrawal Sequence,"
(continued) January 1977.
- 7. NRC SER, " Acceptance of Referencing of Licensing Topical Report NEDE-240ll-P-A," " General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.
- 8. NEDC-30851-P-A, " Technical Specification Improvement Analysis for BWR Control Rod Block In.;trumentation,"
October 1988.
- 9. GENE-770-06-1, " Bases For Changes To Surveillance Test Intervals and Allowed Out-0f-Service Times For Selected Instrumentation Technical Specifications," k February 1991.
- 10. NRC No. 93-102, " Final Policy Statement on Tech".ical Specification Improvements," July 23, 1993.
Ib A O v HATCH UNIT 1 B 3.3-53 REVISION D
Feedwater and Main Turbine High Waer Level Trip Instrumentation ; B 3.3.2.2 B 3.3 INSTRUMENTATION B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation BASES BACKGROUND The feedwater and main turbine high water level trip instrumentation is designed to detect a potential failure of the Feedwater Level Control System that causes excessive feedwater flow. With excessive feedwater flow, the water level in the reactor vessel rises toward the high water level setpoint, causing the trip of the two feedwater pump turbines and the main turbine. Reactor Vessel Water Level - High signals are provided by level sensors 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 in the reactor vessel (variable leg). Three channels of Reactor Vessel Water Level - High instrumentation are provided as input to a two-out-of-three initiation logic that trips the two feedwater pump turbines and the main turbine. The channels include electronic equipment (e.g., trip relays) that compare measured input signals with pre-established g setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a main feedwater and turbine trip signal to the trip logic. A trip of the feedwater pump turbines limits further increase in reactor vessel water level by limiting further addition of feedwater to the reactor vessel. A trip of the main turbine and closure of the stop valves protects the turbine from damage due to water entering the turbine. APPLICABLE The feedwater and main turbine high water level trip SAFETY ANALYSES instrumentation is assumed to be capable of providing a turbine trip in the design basis transient analysis fter a feedwater controller failure, maximum demand event (Ref.1). The high level trip indirectly initiates a reactor scram from the main turbine trip (above 50% RTP) and trips the feedwater pumps, thereby terminating the event. The reactor scram mitigates the reduction in MCPR. (continued) h HATCH UNIT I B 3.3-54 REVISION A
Remote Shutdown System B 3.3.3.2 .q l v) BASES ACTIONS M (continued) The Required Action is to restore the Function to OPERABLE status within 30 days. The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room. M If the Required Action and associated Completion Time of Condition A are not 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. The allowed Completion Time is reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when an instrument channel is placed in an inoperable status solely for performance of required Surveillances, entry into (n') associated Conditions and Required Actions may be delayed for up to 6 hours. 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. The Note is based upon a NRC Safety Evaluation Report (Reference 1) which concluded that the 6 hour testing allowance does not significantly reduce the probability of monitoring required parameters, when necessary. SR 3.3.3.2.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel against a similar parameter on A other channels. It is based on the assumption that (pn instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between 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 n g (continued) HATCH UNIT 1 B.3.3-75 REVISION A
Remote Shutdown System B 3.3.3.2 BASES h 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 sensor or the signal processing equipment has drifted outside its limit. As specified in the Surveillance, a CHANNEL CHECK is only required for those channels that are normally energized. The Frequency is based upon plant operating experience that demonstrates channel failure is rare. SR 3.3.3.2.2 lA SR 3.3.3.2.2 verifies each required Remote Shutdown System transfer switch and control circuit performs the intended II function. This verification is performed from the remote shutdown panel and locally, as appropriate. Operation of equipment from the remote shutdown panel is not necessary. ! The Surveillance can be satisfied by performance of a continuity check, or, in the case of the DG controls, the routine Surveillances of LCO 3.8.1 (since local control is utilized during the performance of some of the Surveillances (continued) HATCH UNIT 1 B 3.3-75A REVISION D b
PAM In trumentation
'B 3.3.3.1 1
Q f !- Q BASES SURVEILLANCE REQUIREMENTS SR 3.3.3.2.2 (continued) Ik! ' of LC0 3.8.1). This wil ensure that-if the control room becomes inaccessible, the plant can be placed and maintained l in MODE 3 from the remote shutdown panel and the local ; control stations. The 18 month Frequency is based on.the l need to perform this Surveillance under the conditions that ! apply during a plant outage and the potential for an 'i unplanned transient if the Surveillance were performed with i the reactor at power. Operating experience demonstrates- l that Remote Shutdown System controls usually pass the ; Surveillance when performed at the 18 month Frequency. : SR 3.3.3.2.3 lk l CHANNEL CALIBRATION is a complete check of the instrument i loop and the sensor. The test verifies the channel- responds ! to measured parameter values with the necessary range and j accuracy. ; The 18 month Frequency is based upon operating experience ! and consistency with the typical industry refueling cycle.
- 1. 10 CFR 50, Appendix A, GDC 19.
REFERENCES
- 2. Technical Requirements Manual. .l
- 3. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993. ;
- i
.I i
O i 1 HATCH UNIT 1 B 3.3-76 REVISION D
Primary Containment Isolation Instrumentation B 3.3.6.1 r% ! ) BASES v APPLICABLE 1.c. Main Steam Line Flow - Hiah (continued) SAFETY ANALYSES, LCO, and flow - High Function for each unisolated MSL (two channels APPLICABILITY per trip system) are available and are required to be OPERABLE so that no single instrument failure will preclude detecting a break in any individual MSL. The Allowable Value is chosen to ensure that offsite dose limits are not exceeded due to the break. The Allowable Value corresponds to :s 101 psid, which is the parameter i S monitored on control room instruments. This Function isolates the Group 1 valves. 1.d. Condenser Vacuum - Low The Condenser Vacuum - Low Function is provided to prevent overpressurization of the main condenser in the event of a loss of the main condenser vacuum. Since the integrity of the condenser is an assumption in offsite dose calculations, the Condenser Vacuum - Low Function is assumed to be
- OPERABLE and capable of initiating closure of the MSIVs.
The closure of the MSIVs is initiated to prevent the 'v)' ( addition of steam that would lead to additional condenser pressurization and possible rupture of the diaphragm installed to protect the turbine exhaust hood, thereby preventing a potential radiation leakage path following an accident. Condenser vacuum pressure signals are derived from four pressure transmitters that sense the pressure in the condenser. Four channels of Condenser Vacuum - Low Function : are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Allowable Value is chosen to prevent damage to the condenser due to pressurization, thereby ensuring its integrity for offsite dose analysis. As noted (footnote (a) to Table 3.3.6.1-1), the channels are not required to be OPERABLE in MODES 2 and 3 when all turbine stop valves (TSVs) are closed, since the potential for condenser , overpressurization is minimized. Switches are provided to manually bypass the channels when all TSVs are closed. This Function isolates the Group 1 valves, r) (continued) HATCH UNIT 1 B 3.3-153 REVISION D l
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 1.e. 1.f. Area Temperature - Hiah SAFETY ANALYSES, LCO, and Area temperature is provided to detect a leak in the RCPB APPLICABILITY and provides diversity to the high flow instrumentation. (continued) The isolation occurs when a very small leak has occurred. If the small leak is allowed to continue without isolation, offsite dose limits may be reached. However, credit for i these instruments is not taken in any transient or accident I analysis in the FSAR, since bounding analyses are performed for large breaks, such as MSLBs. j Area temperature signals are initiated from RTDs (for the i Main Steam Tunnel Temperature-High Function) or Temperature switches (for the Turbine Building Area Temperature-High Function) located in the area being monitored. While 16 channels of Main Steam Tunnel Temperature - High Function are available, only 12 channels (six per trip system) are j required to be OPERABLE. This will ensure that no single instrument failure can preclude the isolation function, assuming a line break on any line (the instruments assigned to monitor one line can still detect a leak on another line due to their close proximity to one another and the small confines of the area). While 64 channels of Turbine Building Area Temperature - High Function are available, only 32 channels are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. Each channel has one temperature element. The 32 channel requirement is further divided up, as noted in footnote (b), into 16 channels per trip system with 8 per trip string. Each trip string shall have 2 channels per main steam line, with no more than 40 feet separating any two OPERABLE channels. The ambient temperature monitoring Allowable Value is chosen to detect a leak equivalent to between 1% and 10% rated steam flow. These Functions isolate the Group 1 valves. (continued) HATCH UNIT 1 B 3.3-154 REVISION A
Prinary Containment Isolation Instrumentation B 3.3.6.1 uO BASES 1 ACTIONS f.J (continued) l For the RWCU Area and Area Ventilation Differential Temperature - High Functions, the affected penetration flow
- path (s) may be considered isolated by isolating'only that portion of the system in the associated room monitored by !
the inoperable chanr.el. That is, if the RWCU pump room A i area channel is inoperable, the pump room A area can be . isolated while allowing continued RWCU operation utilizing the B RWCU pump. Alternately, if it h ::ot desired to isolate the affected , penetration flow path (s) (e.g., as in the case where isolating the penetration flow path (s) could result in a , reactor scram), Condition G must be entered and its Required Actions taken. The I hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for personnel , to isolate the affected penetration flow path (s). , p) G.1 and G 2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, or any Required' - Action of Condition F is not met and the associated Completion Time has expired, the plant must be placed in a MODE or other specifled condition in which the LCO does not , apply. This is dont by placing the plant in at least MODE 3 , within 12 hours and in MODE 4 within 36 hours. The allowed i 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. I i (continued) HATCH UNIT I B 3.3-169 REVISION A
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES h ACTIONS H.1 and H.2 (continued) If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the SLC System is declared inoperable or the RWCU System is isolated. Since this Function is required to ensure that the SLC System performs its intended function, sufficient remedial measures are provided by declaring the SLC System inoperable or isolating the RWCU System. The 1 hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for personnel to isolate the RWCU System. I.1 and I.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated penetration flow path should be closed. However, if the shutdown cooling function is needed to provide core cooling, these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the channel to OPERABLE status or to isolate the RHR Shutdown Cooling System (i.e., provide alternate decay heat removal capabilities so the penetration flow path can be isolated). Actions must continue until the channel is restored to OPERABLE status or the RHR Shutdown Cooling System is isolated. SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Primary Containment Isolation instrumentation Function are found in the SRs co19mn of Table 3.3.6.1-1. 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 1D 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. 4 and 5) assumption of the
.(continued)
HATCH UNIT 1 B 3.3-170 REVISION D
Secondary Containment Isolation Instrumentation B 3.3.6.2 A U BASES I ACTIONS C.1.1. C.1.2. 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 iso M e the Unit I secondary containment and start the 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 , zones (closing the ventilation supply and exhaust automatic l isolation dampers) and starting the associated SGT i l subsystems (Required Actions C.1.1 and C.2.1) performs the ! I intended function of the instrumentation and allows I operation to continue. ) l ) 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 LC0 3.6.4.3) provide appropriate actions for the inoperable components. Since each trip system affects two SGT subsystems (one Unit I and 1 one Unit 2) Required Actions C.2.1 and C.2.2 can be I performed independently on each SGT subsystem. That is, one i SGT subsystem can be started (Required Action C.2.1) while (o) the other SGT subsystem 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 g Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the a3plicable Condition entered and Required Actions taken. T1is Note is based on the reliability analysis (Refs. 5 and 6) assumption of the average time required to perform channel surveillance. That analysis demonstrated I (continued) r. HATCH UNIT 1 B 3.3-181 REVISION D i
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES h 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 l 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 i gross channel failure; thus, it is key to verifying the l 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 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. (continued) HATCH UNIT I B 3.3-182 REVISION A
S/RVs B 3.4.3.
) B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.3 Safety / Relief Valves (S/RVs) BASES BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of S/RVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB). The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell . The S/RVs can actuate by either of two modes: the ' svety mode or the relief mode. In the safety mode (or 5.^ing mode of operation), the spring loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve. This - (7 satisfies the Code requirement. C/ Each S/RV discharges steam through a discharge line to a point belcw the minimum water level in the suppression pool. The S/RVs that provide the relief mode are the low-low set (LLS) valves and the Automatic Depressurization System (ADS) valves. The LLS requirements are specified in LC0 3.6.1.6,
" Low-low Set (LLS) Valves," and the ADS requirements are specified in LC0 3.5.1, "ECCS - Operating."
APPLICABLE The overpressure protection system must accommodate the most SAFETY ANALYSES severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main stea:a isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 1). For the ; purpose of the analyses,11 S/RVs are assumed to operate in the safety mode. The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure well below the ASME Code limit of 110% of vessel (continued) HATCH UNIT 1 B 3.4-13 REVISION A
S/RVs B 3.4.3 BASES - APPLICABLE design pressure (110% x 1250 psig = 1375 psig). Sensitivity SAFETY ANALYSES analyses have demonstrated that 8 or 9 S/RVs operating in (continued) the pressure relief mode will maintain the reactor vessel below 1375 psig. This LC0 helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event. From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above. Reference 2 discusses additional events that are g expected to actuate the S/RVs. S/RVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 4). LC0 The safety function of eleven S/RVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. I and 2), although margins to the ASME Vessel Overpressure Limit are substantial. The requirements of this LC0 are applicable only to the capability of the S/RVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety function). The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the FSAR are based on these setpoints, but also include the additional uncertainties of i 3% of the nominal setpoint drift to provide an added degree of conservatism. Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded. (continued) HATCH UNIT 1 B 3.4-14 REVISION D
S/RVs B 3.4.3 BASES (continued) APPLICABILITY In MODES 1, 2, and 3, all S/RVs must be OPERABLE, since considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat. In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents. In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The S/RV function is not needed during these conditions. < ACTIONS A_d With the safety function of one S/RV inoperable, the remaining OPERABLE S/RVs are capable of providing the (7 necessary overpressure protection. However, the overall Cl reliability of the pressure relief system is reduced because additional failures in the remaining OPERABLE S/RVs could result in failure to adequately relieve pressure during a limiting event. For this reason, continued operation is permitted for a limited time only. The 14 day Completion Time to restore the inoperable S/RV to OPERABLE status is based on the relief capability of the remaining S/RVs, the low probability of an event requiring S/RV actuation, and a reasonable time to complete the Required Action. B.1 and B.2 With more than one S/RV inoperable, a transient m'ay result in the violation of the ASME Code limit on reactor pressure. If the safety function of the inoperable S/RV cannot be restored to OPERABLE status within the associated Completion Time of Required Action A.1, or if the safety function of two or more S/RVs is inoperable, 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 MODE 3 within 12 hours h) v (continued) HATCH UNIT 1 B 3.4-15 REVISION A
S/RVs B 3.4.3 BASES h ACTIONS B.1 and B.2 (continued) and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillance requires that the S/RVs will open at the pressures assumed in the safety analysis of Reference 1. The demonstration of the S/RV safety lift settings must be performed during shutdown, since this is a bench test, to be done in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is i 3% for OPERABILITY; however, the valves are reset to 1% during the Surveillance to allow for drift. Performance of this SR in accordance with the Inservice Testing Program requires an 18 month Frequency. The W 18 month Frequency was selected because this Surveillance must be performed-during shutdown conditions and is based on the time between refuelings. SR 3.4.3.2 A manual actuation of each S/RV is performed to verify that, i mechanically, the valve is functioning properly and no i blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method' suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam 1 flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the S/RVs divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at , which this test is to be performed is 920 psig (the pressure i recommended by the valve manufacturer). Adequate steam flow (continued) h HATCH UNIT 1 B 3.4-16 REVISION D
S/RVs B 3.4.3-A 'y) L'ASES (continued) is represented by at least 1.25 turbine bypass valves open, or total steam flow 1 lE6 lb/hr. Plant d , l l l >O 'G) 1 l l i 1 1 I 1 I ()
\
f' (continued) l l HATCH UNIT 1 B 3.4-16A REVISION D !
RCS Operaticnal LEAKAGE B 3.4.4 . BASES (continued) ACTIONS M With RCS unidentified or total LEAKAGE greater than the limits, actions must be taken to reduce the leak. Because the LEAKAGE limits are conservatively below the LEAKAGE that would constitute a critical ' crack size, 4 hours is allowed to reduce the LEAKAGE rates before the reactor.must.be shut down. If an unidentified LEAKAGE has been identified and ! quantified, it may be reclassified and considered as l identified LEAKAGE; however, the total LEAKAGE would remain i unchanged. The total LEAKAGE must be averaged over the ! previous 24 hours for comparison to the 1:.Jt. M t An unidentified LEAKAGE increase of > 2 gpm within a 24 hour peried is an indication of a potential flaw in the RCPB and . ' must be quickly evaluated. Although the increase does not necessarily violate the absolute unidentified LEAKAGE limit, . certain susceptible components must be determined not to be i the source of the LEAKAGE increase within'the required Completion Time.
~
The 4 hour Completion Time is reasonable to properly reduce' the LEAKAGE increase before the reactor'must be shut down , without unduly jeopardizing plant safety. ; C.1 and C.2 [ If any Required Action and associated Completion Time of-Condition A or B is not met or if pressure boundary LEAKAGE exists, 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 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' safety systems. (continued) HATCH UNIT I B 3.4-21 REVISION A I
RCS Operational LEAKAGE B 3.4.4 BASES (continued) h SURVEILLANCE SR 3.4.4.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to provide alarms when LEAKAGE is indicated and to quantify the various types of LEAKAGE. Leakage detection instrumentation is discussed in more detail in the Bases for LC0 3.4.5, "RCS Leakage Detection Instrumentation." Sump level and flow rate are typically monitored to determine actual LEAKAGE rates; however, any method may be used to quantify LEAKAGE within the guidelines of Reference 7. In conjunction with alarms and other administrative controls, a 12 hour Frequency for this Surveillance is appropriate for identifying LEAKAGE and for tracking required trends (Ref. 8). The identified portion of the total LEAKAGE is usually determined by the drywell equipment drain sump monitoring system which collect expected leakage not indicative of a degraded RCS boundary. The system equipment and operation A is identical to that of the drywell floor drain monitoring /h.\ system described in the Bases for LC0 3.4.5, "RCS Leakage Detection Instrumentation." If a contributor to the unidentified LEAKAGE has been identified and quantified, it & may be reclassified and considered as identified LEAKAGE. W REFERENCES 1. 10 CFR 50.2.
- 2. 10 CFR 50.55a(c).
- 3. 10 CFR 50, Appendix A, GDC 55.
- 4. GEAP-5620, " Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flaws," April 1968.
- 5. NUREG-75/067, " Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactors," October 1975.
- 6. FSAR, Section 4.10.3.2.
(continued) HATCH UNIT 1 B 3.4-22 REVISION D
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- i
- l RCS Operational': LEAKAGE -
B 3.4.4: () BASES .(continued)
.l
- 7. Regulatory Guide 1.45, May 1973. .
- 8. GenericLetter88-01,SupplementL1,"NRCPosition:on i IGSCC in BWR Austenitic-Stainless Steel Piping," i February 1992.
- 9. NRC No. 93-102, " Final Policy Statement on Technical ,
Specification Improvements," July 23, 1993. l t . I 8 i r
-k i
r/'
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(continued) q HATCH UNIT 1 B 3.4-22A REVISION D
.I
)
RCS Leakage Detection Instrumentation B 3.4.5 ( i BASES s f v LC0 of other detection systems will be made to determine the ' (continued) extent of any corrective action that may be required. With the leakage detection systems inoperable, monitoring for LEAKAGE in the RCPB is degraded. , APPLICABILITY In MODES 1, 2, and 3, leakage detection systems are required to be OPERABLE to support LCO 3.4.4. This Applicability is consistent with that for LC0 3.4.4. ACTIONS A.1 With the drywell floor drain sump monitoring system inoperable, no other form of sampling can provide the equivalent information to quantify leakage. However, the primary containment atmospheric activity monitor will provide indication of changes in leakage. With the drywell floor drain sump monitoring system O inoperable, but with RCS unidentified and total LEAKArE () being determined every 12 hours (SR 3.4.4.1), operatio.i may continue for 30 days. The 30 day Completion Time of Required Action A.1 is acceptable, based on operating experience, considering the multiple forms of leakage detection that are still available. Required Action A.1 is modified by a Note that states that the provisions of LC0 3.0.4 are not applicable. As a result, a MODE change is allowed when the drywell floor dr;in sump monitoring system is inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage. Acceptable methods for quantifying both identified and , unidentified LEAKAGE include but are not limited to the ; following:
- 1) With a drifting sump monitoring system integrator, the A I sump can be manually pumped down with integrator /_4 readings taken before and after pumpdown. The difference in readings determines total gallons ,
pumped. Using time elapsed since last pumpdown, sump i inleakage rate can be calculated; and l I
.(v ) (continued) i HATCH UNIT 1 B 3.4-25 REVISION D i ~
RCS Leakage Detection Instrumentation B 3.4.5 BASES
- 2) With an inoperable sump monitoring system integrator, the sump can be manually pumped down and the time for pumpdown recorded. Utilizing pump flow rate, total gallons pumped is determined. Using time elapsed [ ,
since last pumpdown, sump inleakage rate can be calculated. B.1 and B.2 With both gaseous and particulate primary containment atmospheric monitoring channels inoperable (i.e., the required containment atmospheric monitoring system), grab samples of the primary containment atmosphere must be taken and analyzed to provide periodic leakage information. Provided a sample is obtained and analyzed once every 12 hours, the plant may be operated for up to 30 days to allow restoration of at least one of the required monitors. 9 (continued) HATCH UNIT 1 B 3.4-25A REVISION D ;
RCS Leakage Detection Instrumentation B 3.4.5 l 1 BASES U ACTIONS B.1 and B.2 (continued) The 12 hour interval provides periodic information that is adequate to detect LEAKAGE. The 30 day Completion Time for restoration recognizes that at least one other form of leakage detection is available. The Required Actions are modified by a Note that states that the provisions of LC0 3.0.4 are not applicable. As a result, a MODE change is allowed when both the gaseous and particulate primary containment atmospheric monitoring channels are inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage. 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 MODE 4 within 36 hours. The allowed Completion O Times are reasonable, based on operating experience, to C perform the actions in an orderly manner and without challenging plant systems. D_d With all required monitors inoperable, no required automatic means of monitoring LEAKAGE are available, and immediate plant shutdown in accordance with LC0 3.0.3 is required. SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS 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 other required instrumentation (either the drywell floor drain sump monitoring system or the primary containment atmospheric monitoring channel, as applicable) is OPERABLE. 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. m j () (continued) HATCH UNIT 1 B 3.4-26 REVISION A
RHR Shutdown Cooling System - Cold Shutdown B 3.4.8 (q) BASES l l APPLICABILITY In MODES 1 and 2, and in MODE 3 with reactor steam dome l (continued) pressure greater than or equal to the RHR low pressure l permissive pressure, this LC0 is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure mayl exceed the design pressure of the shutdown cooling piping. I Decay heat removal at reactor pressures greater than or i equal to the RHR low pressure permissive pressure is L typically accomplished by condensing the steam in the main 3 condenser. Additionally, in MODE 2 below this pressure, the i OPERABILITY requirements for the Emergency Core Cooling i Systems (ECCS) (LC0 3.5.1, "ECCS - Operating") do not allow I placing the RHR shutdown cooling subsystem into operation. Ib The requirements for decay heat removal in MODE 3 below the RHR low pressure permissive pressure and in MODE 5 are discussed in LC0 3.4.7, " Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown"; LC0 3.9.7,
" Residual Heat Removal (RHR) - High Water Level"; and LC0 3.9.8, " Residual Heat Removal (RHR) - Low Water Level ."
O V ACTIONS A Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems. Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent divisions, 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 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 shutdown cooling subsystems provide appropriate compensatory measures for separate inoperable shutdown cooling subsystems. As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem.
<m
) (continued)
HATCH UNIT 1 B 3.4-41 REVISION D i
RHR Shutdown Cooling System - Cold Shutdown B 3.4.8 I BASES ACTIONS L1 , (continued) l With one of the two required RHR shutdown cooling subsystems inoperable, except as permitted by LC0 Note 2, 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 RHR 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 i re-establishes backup decay heat removal capabilities, l similar to the requirements of the LCO. The I 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 these alternate method (s) must be reconfirmed every 24 hours thereafter. This will provide l assurance of continued heat removal capability. l The required cooling capacity of the alternate method should r' be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Decay heat removal by ambient losses can be considered as, or ! contributing to, the alternate method capability. Alternate methods that can be used include (but are not limited to) the Condensate / Main Steam Systems (feed and bleed) and the Reactor Water Cleanup System. B.1 and B.2 l With no RHR shutdown cooling subsystem and no recirculation l pump in operation, except as permitted by LC0 Note 1, and until RHR or recirculation pump operation is re-established, an alternate method of reactor coolant circulation must be placed into service. This will provide the necessary circulation for monitoring coolant temperature. The I hour Completion Time is based on the coolant circulation function and is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation. Furthermore, verification of the functioning of the alternate method must be reconfirmed every 12 hours thereafter. This will provide assurance of continued temperature monitoring capability. 1 (continued) l l HATCH UNIT 1 B 3.4-42 REVISION A 1
b RCS P/T L'imits B 3.4.9-i%
'Q BASES BACKGROUND as necessary, based on the evaluation' findings and the (continued) recommendations of Reference 5.
The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and-temperature rate of change, one location within the reactor' vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive. regions. The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The-thermal gradient reversal alters the location of the tensile stress between the outer and inner walls. The criticality limits include the Reference 1 requirement that they be at least 40 F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic O- testing. The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB componer.ts. ASME Code, Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits. APPLICABLE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed. The PTLR references the_ l D methodology for determining the P/T limits. Since the (continued) HATCH UNIT 1 B 3.4-45 REVISION D J
i RCS P/T Limits B 3.4.9 , 1 BASES APPLICABLE P/T limits are not derived from any DBA, there are no SAFETY ANALYSES acceptance limits related to the P/T limits. Rather, the (continued) P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition. RCS P/T limits satisfy Criterion 2 of the NRC Policy A Statement (Ref. 8). 1& LCO The elements of this LC0 are:
- a. RCS pressure, temperature, and heatup or cooldown rate are within the limits specified in the PTLR during RCS heatup, cooldown, and inservice leak and hydrostatic testing;
- b. The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is within the limit of the PTLR during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow;
- c. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel meets the limit of the PTLR during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow;
- d. RCS pressure and temperature are within the criticality limits specified in the PTLR, prior to achieving criticality; and
- e. The reactor vessel flange and the head flange temperatures are within the limits of the PTLR when tensioning the reactor vessel head bolting studs.
These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure. The rate of change of temperature limits controls the thermal gradient through the vessel wall and is used as input for calculating the heatup, cooldown, and inservice (continued) HATCH UMIT 1 B 3.4-46 REVISION D
-. . - -- - ~ ~ - . . - - . -
l RCS P/T Omits;- B 3.4.9 [ ~ BASES l C.1 and C.2 ACTIONS ] (continued) ,
' Operation.'outside the P/T limits'in other than MODES'1, 2, and 3-(including defueled conditions) must be corrected so ;
that the RCPB is returned to a condition that has been: ' verified by stress analyses. The Required Action must be' ! initiated without delay and continued until the limits are restored. i Besides restoring the P/T limit parameters to'within limits,. an evaluation is required to determine if RCS operation' is allowed. This evaluation must verify. that the RCPB l integrity is acceptable and must' be completed before - l approaching criticality or heating up to >'212*F. Several ! methods may be used, including comparison with pre-analyzed 4 transients, new analyses, or inspection of the components. . ASME Code, Section XI, Appendix E (Ref. 6), may be used.to support the evaluation; however, its use:is-restricted to evaluttiun of'the beltline. j Condition C is modified by a Note requiring Required Action !
'C.2 be completed whenever the Condition is entered. The ;
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D Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. - Restoration alone per Required Action C.1 is insufficient . ; b~ecause higher'than analyzed stresses may have occurred and- : may have affected the RCPB integrity.
~
' SURVEILLANCE SR 3.4.9.1 -
REQUIREMENTS . Verification that operation is within PTLR limits is required every 30 minutes when RCS pressure and temperature conditions.are undergoing planned changes. This'_ Frequency is considered reasonable in view of the control room indication available to monitor'RCS' status. Also, since temperature rate'of change limits are specified.in hourly increments, 30 minutes permits a reasonable time for i assessment and correction of minor deviations. ; Surveillance for heatup, cooldown, or inservice leakage and. I hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.' 4 (continued) HATCH UNIT 1 B 3.4-49 REVISION A j
RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.1 (continued) REQUIREMENTS This SR has been modified with a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and RCS inservice leakage and hydrostatic testing. SP 3.4.9.2 A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical. Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal. SR 3.4.9.3 and SR 3.4.9.4 Differential temperatures within the applicable PTLR limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 7) are satisfied. I D Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start. An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.4 is to compare the temperatures of the operating recirculation loop and the idle loop. (continued) HATCH UNIT 1 B 3.4-50 REVISION D
~ . .
RCS P/T Liaits [ B 3.4.9 l BASES :j i SURVEILLANCE SR 3.4.9.3 and SR 3.4.9.4'-(continued) l REQUIREMENTS , ! SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that. ! requires the Surveillance to be performed only in MODES 1, i 2, 3, and 4. In MODE 5, the overall stress on limiting .. components is lower. Therefore, AT limits are not required.. ] SR 3.4.9.5. SR 3.4.9.6. and SR 3.4.9.7 - ~ Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits ; during system heatup and cooldown. However, operations . approaching MODE 4 from MODE 5 and in MODE 4 with RCS i temperature less than or equal to certain specified values i require assurance that these temperatures meet the LCO limits. The flange temperatures must be verified to be above the l limits 30 minutes before and while tensioning the vessel j head bolting studs to ensure that once the head is tensioned j the limits are satisfied. When in MODE 4 with RCS ; temperature s 80*F, 30 minute checks of the flange O temperatures are required because of the reduced margin to 4 the limits. When in MODE 4 with RCS temperature s 100*F, j monitoring of the flange temperature is reouired every 1 12 hours to ensure the temperature is witlan the limits specified in the PTLR. The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time - e that the temperature limits could be exceeded. The 12 hour ! Frequency is reasonable based on the rate of temperature i change' possible at these temperatures. .) i SR 3.4.9.5 is modified by a Note that: requires the. l Surveillance to be performed only when tensioning the l reactor vessel-head bolting studs. SR 3.4.9.6 is modified "li by a Note that requires the Surveillance to be initiated 30 minutes after RCS temperature s 8_0*F in Mode 4. SR 3.4.9.7.is modified by a Note that. requires the Surve111ance' to be initiated 12 hours after RCS temperature 5 100'F in Mode.4. The Notes contained in these SRs are necessary to specify when the reactor vessel flange and head flange temperatures are required to be verifiel ? i be within the limits specified in the PTLR.
'(continued)
A I HATCH UNIT 1 B 3.4-51 REVISION A
, % - m.i c c- _ _ _ _
w
RCS P/T Limits B 3.4.9 BASES h REFERENCES 1. 10 CFR 50, Appendix G.
- 2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.
- 3. ASTM E 185-82, " Standard Practice for Conducting Surveillance Tests for Licht-Water Cooled Nuclear Power Reactor Vessels," July 1982.
- 4. 10 CFR 50, Appendix H.
- 5. Regulatory Guide 1.99, Revision 2, May 1988.
- 6. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E.
I k
- 7. FSAR, Section 14.3.6.2. Ib
- 8. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
lA 0 l l f-l l O HATCH UNIT 1 9 3.4-52 REVISION D
ECCS - Operating B 3.5.1 n BASES U SURVEILLANCE SR 3.5.1.7. SR 3.5.1.8. and SR 3.5.1.9 (continued) REQUIREMENTS The flow tests for the HPCI System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher and loweroperating ranges of the system. The pump flow rates are verified against a system head corresponding to the RPV pressure. The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. The reactor steam pressure must be 2 920 psig to perform SR 3.5.1.8 and a 150 psig to perform SR 3.5.1.9. Adequate steam flow for SR 3.5.1.8 is A represented by at least two turbine bypass valves open, or LDT 2 200 MWE from the main turbine generator; and for SR 3.5.1.9 adequate steam flow is represented by at least 1.25 turbine bypass valves open, or total steam flow 2 IE6 lb/hr. Therefore, sufficient time is allowed after adequate pressure and flow are achieved to perform these tests. Reactor startup is allowed prior to aerforming the low
,_N pressure Surveillance test because tie reactor pressure is
/ low and the time allowed to satisfactorily perform the d Surveillance test is short. The reactor pressure is allowed ; to be increased to normal operating pressure since it is ' assumed that the low pressure test has been satisfactorily completed and there is , no indication or reason to believe that HPCI is inoperable. Therefore, SR 3.5.1.8 and SR 3.5.1.9 are modified by Notes that state the Surveillances are not required to be performed until 12 hours after the reactor steam pressure and flow are adequate to perform the test. Therefore, implementation of these Notes requires these tests to be performed during reactor startup within 12 hours after adequate steam pressure and flow are achieved. The Frequency for SR 3.5.1.7 and SR 3.5.1.8 is consistent with the Inservice Testing Program pump testing requirements. The 18 month Frequency for SR 3.5.1.9 is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. (continued) HATCH UNIT 1 B 3.5-13 REVISION D
ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.10 REQUIREMENTS (continued) The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI, CS, and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions. This SR also ensures that the HPCI System will automatically restart on an RPV low water level (Level 2) signal received subsequent to an RPV high water level ' (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LC0 3.3.5.1 overlaps this Surveillance to provide complete testing of the assumed safety function. The 18 month Frequency is based on the need to perform the 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 that these components usually pass the SR when performed at the 18 month Frequency, which g is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. This SR is modified by a Note that excludes vessel injection / spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance. SR 3.5.1.11 The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e., ; solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.12 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LC0 3.3.5.1 (continued) HATCH UNIT 1 B 3.5-14 REVISION A
t ECCS - Operating B 3.5.1 p BASES U SURVEILLANCE SR 3.5.1.11 (continued) REQUIREMENTS overlap this ",urveillance to provide complete testing of the assumed safety function. The 18 month Frequency is based on the need to perform the 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 that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. This SR is modified by a Note that excludes valve actuation. This prevents an RPV pressure blowdown. SR 3.5.1.12 A manual actuation of each ADS valve is performed to verify that the valve and solenoid are functioning properly and that no blockage exists in the S/RV discharge lines. This is demonstrated by the response of the turbine control or (d-) bypass valve or by a change in the measured steam flow or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the ADS valves divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this SR. Adequate pressure at which this SR is to be performed is a 920 psig (the pressure recommended by the valve manufacturer). Adequate steam flow ' is represented by t least 1.25 turbine bypass valves open, or total steam flow a IE6 lb/hr. Reactor startup is allowed prior to performing this SR because valve OPERABILITY and the setpoints for overpressure protection are verified, par ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test. The 12 hours allowed for manual actuation after the required i pressure it reae.hed is sufficient to achieve stable conditions and provides adequate time to complete the Surveillance. SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LC0 3.3.5.1 b') V (continued) > HATCH UNIT 1 B 3.5-15 REVISION D
ECCS - Operating i B 3.5.1 BASES l SURVEILLANCE SR 3.5.1.12 (continued) REQUIREMENTS overlap this Surveillance to provide complete testing of the assumed safety function. The Frequency of 18 months is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the l SR when performed at the 18 month Frequency, which is based i on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. FSAR, Section 6.4.3.
- 2. FSAR, Section 6.4.4.
1
- 3. FSAR, Section 6.4.1.
- 4. FSAR, Section 6.4.2.
1
- 5. FSAR, Section 14.4.3.
- 6. FSAR, Section 14.4.5.
- 7. 10 CFR 50, Appendix K.
- 8. FSAR, Section 6.5.
- 9. NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Analysis," December 1986.
- 10. 10 CFR 50.46.
- 11. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. ;
(NRC), " Recommended Interim Revisions to LCOs for ECCS I Components," December 1, 1975.
- 12. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
O HATCH UNIT 1 B 3.5-16 REVISION A
l ECCS - Shutdown I B 3.5.2 ( f 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 l 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 l immediately initiated to suspend operations with a potential l for draining the reactor vessel (0PDRVs) to minimize the probability of a vessel draindown and the subsequent (3 potential for fission product release. Actions must C/ continue until OPDRVs are suspended. C.I. 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) HATCH UNIT 1 B 3.5-19 REVISION A l
ECCS - Shutdown B 3.5.2 BASES h ACTIONS C.1. C.2. D.1. 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 is OPERABLE; 2) two standby gas treatment subsystems (any combination of Unit 1 and Unit 2 subsystems) are OPERABLE; and 3) secondary containment isolation capability is A available (i.e., one secondary containment isolation valve la 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 A reasons. It is not necessary to perform the Surveillances I /d 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 W component to OPERABLE 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. ; 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 2: 146 inches i (continued) ; HATCH UNIT I B 3.5-20 REVISION D
P ECCS - Shutdown B 3.5.2 ; 1
~h ' - (Q BASES or that CS is aligned to take suction from the CST and the CST contains 2: 150,000 gallons of water, equivalent to /
( P 1 l
.. flL/ (continued)
J 1 HATCH UNIT 1 B 3.5-20A REVISION D I i
RCIC System B 3.5.3 r, () BASES SURVEILLANCE SR 3.5.3.2 (continued) REQUIREMENTS The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve position would affect only the RCIC System. This Frequency has been shown to be acceptable through operating experience. SR 3.5.3.3 and SR 3.5.3.4 The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated. The required flow rate (400 gpm) is the pump design flow rate. Analysis has demonstrated that RCIC can fulfill its design function at a system flow rate of 360 gpm (Reference 4). The pump flow rates are verified against a system head equivalent to the RPV pressure. The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the o vessel discharge, the piping friction losses, and RPV Q pressure. The flow tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow is tested both at the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Reactor steam pressure must be 2 920 psig to perform SR 3.5.3.3 and 2 150 psig to perform SR 3.5.3.4. Adequate steam flow is represented by at least one turbine bypass valve open, or [ for SR 3.5.3.3 2 200 MWE from the main turbine-generator and for SR 3.5.3.4 total steam flow 21E6 lb/hr. Therefore, sufficient time is allowed after adequate pressure and flow are achieved to perform these SRs. Reactor startup is allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance is short. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure Surveillance has been satisfactorily completed and there is no indication or reason to believe that RCIC is inoperable. Therefore, these SRs are modified by Notes that state the Surveillances are not required to be performed until 12 hours after the reactor steam pressure and flow are j n (continued) j ( ) HATCH UNIT 1 B 3.5-27 REVISION D , l 2
RCIC System B 3.5.3 BASES h adequate to perform the test. Therefore, implementation of these Notes require O l l 1 l 1 (continued) l HATCH UNIT 1 B 3.5-M27 A REVISION D
i i RCIC System B 3.5.3 j l m L (j BASES SURVEILLANCE SR 3.5.3.3 and SR 3.5.3.4 (continued) REQUIREMENTS these tests to be performed during reactor startup within 12 hours after the reactor steam pressure and flow are adequate to perform the test. A 92 day Frequency for SR 3.5.3.3 is consistent with the Inservice Testing Program requirements. The 18 month Frequency for SR 3.5.3.4 is-based on the need to perform the Surveillance under conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. SR 3.5.3.5 i The RCIC System is required to actuate automatically in order to verify its design function satisfactorily. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of the RCIC System will cause the system to tO, operate as designed, including actuation of the system throughout its emergency operating sequence; that is, automatic pump startup and actuation of all automatic valves , to their required positions. This test also ensures the RCIC System will automatically restart on an RPV low ater level (Level 2) signal received subsequent ~ to an RPV high , water level (Level 8) trip and that the suction is , automatically-transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LC0 3.3.5.2 overlaps this Surveillance to provide complete testing of the assumed safety function.
- The 18 month Frequency is based on the need to perform the ,
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. 1 Operating experience has shown that these components usually l pass the SR when performed at the 18 month Frequency, which ; is based on the refueling cycle. Therefore, the Frequency ; was concluded to be acceptable from a reliability standpoint. (continued) J l HATCH UNIT 1 B 3.5-28 REVISION A i
r 1 i i Primary Containment Air Lock B 3.6.1.2-l
/\
V, BASES' ACTIONS B.I. B.2. and B.3 (continued) J, typically restricted. Therefore, the probability of. misalignment of the door, once it has been' verified to be in i the proper position, is small. C.I. C.2. and C.3
~
If the air lock is inoperable.for reasons other than those ~ described in Condition A or B, Required Action'C.1 requires action to be-immediately-initiated to evaluate containment ! overall leakage rates using current air lock leakage test J results. An evaluation is acceptable since it is overly ; conservative to immediately declare-the primary containment. ; inoperable if both doors in the air lock have failed a seal test or if the overall air lock leakage is not' within - limits. 'In many instances (e.g., only one seal per door has. failed), primary containment remains OPERABLE, yet only I hour (according to LC0 3.6.1.1) would be provided to ' . restore the air lock door to OPERABLE status prior to i requiring a plant shutdown. In addition, even with both doors failing the seal test, the overall containment leakage : rate can still be within limits. 4 Required Action C.2 requires that one door in the primary I containment air lock must be verified closed. -This action must be completed within the 1 hour Completion' Time. This
~
i
.specified time period is consistent with the ACTIONS of ,
LC0 3.6.1.1, which require that primary containment'be' j restored to OPERABLE status within 1 hour. Additionally, the air lock 'must be restored to OPERABLE status within 24 hours. The 24 hour Completion Time is reasonable for restoring an inoperable air lock to OPERABLE status considering that-at least one door is maintained closed in the air lock. D.1 and D.2 - If the inoperable primary containment air: lock cannot be i
. restored to OPERABLE status within the assoc.iated Completion !
Time, the plant must be brought to a MODE in which_the LC0 l does not apply. To achieve this status, the plant must be brought to at-least MODE 3 within 12 hours and to MODE 4 I l (continued) HATCH UNIT 1. B 3.6-11 REVISION A j J 1
Primary Containment Air Lock B 3.6.1.2 BASES ACTIONS D.1 and D.2 (continued) within 36 hours. The allowed Completion Times are reasonable, based on operrting experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.1.2.1 REQUIREMENTS - Maintaining primary containment air locks OPERABLE requires compliance with the leakage rate test requirements of 10 CFR 50, Appendix J (Ref. 3), as modified by approved exemptions. This SR reflects the leakage rate testing requirements with respect to air lock leakage (Type B leakage tests). The acceptance criteria were established as a small fraction of the total allowable containment leakage. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall
' primary containment leakage rate. The Frequency is required by 10 CFR 50, Appendix J (Ref. 3), as modified by approved exemptions. Thus, SR 3.0.2 (which allows frequency extensicns) does not apply.
The SR has been modified by two Notes. Note 1 states that Ib an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event : of a DBA. Note 2 has been added to this SR, requiring the results to be evaluated against the acceptance criteria of SR 3.6.1.1.1. This ensures that air lock leakage is properly accounted for in determining the overall primary k containment leakage rate. SR 3.6.1.2.2 The air lock -interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since-both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary containment pressure, closure of either door will support primary containment OPERABILITY. Thus, the interlock feature supports primary containment OPERABILITY while the (continued) HATCH UNIT I B 3.6-12 REVISION D u
Primary Containment Air Lock : B 3.6.1.2 i i BASES '\ /. air lock is being used for personnel transit in and out of the containment. Periodic testing of this. interlock demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not P P
's ,)k f'%
) (continued) l HATCH UNIT 1 B 3.6-12A REVISION D
PCIVs B 3.6.1.3-Q BASES P BACKGROUND The primary containment purge supply lines are 18 inches in (continued) diameter; exhaust lines are 18 inches in diameter. The 18 inch primary containment purge valves are normally maintained closed in MODES 1, 2, and 3 to ensure the primary containment boundary is maintained. However, the 18 inch valves are qualified for use and may be opened when used for inerting, de-inerting, pressure control, ALARA or. air quality considerations for personnel entry, or Surveillances that require the valves to be open. These valves are qualifiea to be open because two additional redundant excess flow holation dampers are provided on the vent line upstream of the Standby Gas Treatment (SGT) System filter trains. These isolation dampers, together with the PCIVs, will prevent high pressure from reaching the SGT System filter trains in the unlikely event of a loss of coolant accident (LOCA) during venting. Closure of the excess flow isolation dampers will not prevent the SGT System from performing its design function (that is, to maintain a negative pressure in the secondary containment). To ensure that a vent path is available, a 2 inch bypass line is provided around the dampers. The isolation valves on the 18 inch exhaust lines have 2 inch bypass lines around them for use during normal reactor operation or when the 18 inch .L valves cannot be opened. .; 4 APPLICABLE The PCIVs LC0 was derived from the assumptions related to SAFETY ANALYSES minimizing the loss of reactor coolant inventory, and establishing the primary containment boundary during major - , accidents. As part of the primary containment boundary, - PCIV OPERABILITY supports leak tightness of primary containment. Therefore, the safety analysis of any event requiring isolation of primary containment is applicable to this LCO. The DBAs that result in a release'of radioactive material ; for which the consequences are mitigated by PCIVs are a LOCA y and a main steam line break (MSLB). In the analysis for each of these accidents, it is assumed that PCIVs are either 4 closed or close within the required isolation times- j
~following event initiation. This ensures that potential paths to the environment through PCIVs (including primary containment purge valves) are minimized. Of the events k analyzed in Reference 1, the MSLB is the most limiting event ;
due to radiological consequences. The closure time of the lg main steam isolation valves (MSIVs) is a significant' (continued) HATCH UNIT 1 B 3.6-15 REVISION D !
PCIVs B 3.6.1.3 BASES h variable from a radiological standpoint. The MSIVs are required to close within 3 to 5 seconds since the 5 second closure time is assumed in the analysis. The safety analyses assume that the pruge valves were closed at event initiation. Likewise, it is assumed that the primary containment is isolated such that O (continued) HATCH UNIT I B 3.6-15A REVISION D
PCIVs B 3.6.1.3
/3 (y BASES APPLICABLE release of fission products to the environment is SAFETY ANALYSES controlled.
(continued) The single failure criterion required to be imposed in the conduct of unit safety analyses was considered in the original design of the primary containment purge valves. Two valves in series on each purge line provide assurance [ that both the supply and exhaust lines could be isolated even if a single failure occurred. PCIVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 5). LC0 PCIVs form a part of the primary containment boundary. The PCIV safety function is related to minimizing the loss of reactor coolant inventory and establishing the primary containment boundary during a DBA. The power operated and the automatic isolation valves are required to have isolation times within limits and the automatic isolation valves actuate on an automatic isolation (,)
/'7 signal. While the reactor building-to-suppression chamber vacuum breakers isolate primary containment penetrations, they are excluded from this Specification. Controls on their isolation function are adequately addressed in LCO 3.6.1.7, " Reactor Building-to-Suppression Chamber Vacuum Breakers." The valves covered by this LC0 are listed with their associated stroke times in Reference 2.
The normally closed PCIVs are considered OPERABLE when manual valves are closed, or open in accordance with appropriate administrative controls, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact. These passive isolation valves and devices are those listed in Reference 2. i MSIVs must meet additional leakage rate requirements. Other l l PCIV leakage rates are addressed by LC0 3.6.1.1, " Primary l Containment," as Type B or C testing. This LCO provides assurance that the PCIVs will perform l 1 b (continued) HATCH UNIT 1 B 3.6-16 REVISION D
. - _ - - i
I i PCIVs B 3.6.1.3 l BASES O their designed safety functions to minimize the loss of ; reactor coolant inventory and establish the primary containment boundary during accidents. l 9 l l l (continued)
' HATCH UNIT 1 B 3.6-Kl6 4 REVISION D
PCIVs B 3.6.1.3 BASES (continued) i APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 ) and 5, the probability and consequences of these events are. reduced due to the pressure and temperature limitations of , these MODES. Therefore, most PCIVs'are not required to be ' l OPERABLE and the primary containment purge valves are not L required to be sealed closed in. MODES 4 and 5. Certain valves, however, are required to be OPERABLE to prevent inadvertent reactor vessel draindown. These valves are. those whose associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, " Primary Containment Isolation i Instrumentation." (This does not include the valves that isolate the associated instrumentation.) ACTIONS The ACTIONS are modified by a Note allowing penetration flow path (s) except for 18 inch purge valve flow path (s) to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary x containment isolation is indicated. Due to the size of the primary containment purge supply and exhaust line penetrations and the fact that those penetrations-exhaust directly from the containment atmosphere to the environment (via the SGT Systems), the penetration flow path containing these valves is not allowed to be opened under administrative controls. A second Note has been added to provide clarification that, for the purpose of this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable PCIV. Complying with the Required Actions may allow for continued operation, and subsequent inoperable PCIVs are governed by subsequent Condition entry and application of associated Required Actions. The ACTIONS are modified by Notes 3 and 4. Note 3 ensures that appropriate remedial actions are taken, if necessary, if the affected system (s) are rendered inoperable by an inoperable PCIV (e.g., an Emergency Core Cooling System subsystem is inoperable due to a failed open test return (continued) HATCH UNIT 1 B 3.6-17 ret.SION A
PCIVs-B 3.6.1.3 BASES h ACTIONS valve) . Note 4 ensures appropriate remedial actions are continued) taken when the primary containment leakage limits are exceeded. Pursuant to LC0 3.0.6, these actions are not required even when the associated LCO is not met. Therefore, Notes 3 and 4 are added to require the proper actions be taken. jg A.1 and A.2 With one or more penetration flow paths with one PCIV inoperable except for inoperability due to leakage not within a limit specified in an SR to this LCO, the affected penetration flow paths must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, a blind flange, and a check valve with flow through the valve secured. For a penetration isolated in accordance with Required Action A.I, the device used to isolate the penetration should be the closest available valve to the primary containment. The Required Action must be completed within the 4 hour Completion Time (8 hours for main steam lines). The Completion Time of 4 hours is reasonable considering the time required to isolate the penetration and the relative importance of supporting primary containment OPERABILITY during MODES I, 2, and 3. For main steam lines, an 8 hour Completion Time is ellowed. The Completion Time of 8 hours for the main steam lines allows a period of time to restore the MSIVs to OPERABLE status given the fact that MSIV closure will result in isolation of the main steam line(s) and a potential for plant shutdown. For affected penetrations that have been isolated in accordance with Required Action A.1, the affected penetration flow path must be verified to be isolated on a periodic basis. This is necessary to ensure that primary containment penetrations required to be isolated following an accident, and no longer capaW e of being automatically isolated, will be in the isolation position should an event occur. This Required Action does not require any testing or device manipulation. Rather, it involves verification that those devices outside containment and capable of potentially being mispositioned are in the correct position. The (continued) HATCH UNIT 1 B 3.6-18 REVISION D
PCIVs B 3.6.1.3 BASES ACTIONS A.1 and A.2 (continued) , Completion Time of "Once per 31 days for isolation devices outside primary containment" is appropriate because the devices are operated under administrative controls and.the probability of their misalignment is low. For the devices inside primary containment, the time period specified " Prior to entering MODE 2 or 3 from MODE 4, if primary containment , was de-inerted while in MODE 4, if not performed within the previous 92 days" is based on engineering judgment and.is , considered reasonable in view of the inaccessibility of the devices and other administrative controls ensuring that device misalignment is an unlikely possibility. Condition A is modified by a Note indicating that this Condition is only applicable'to those penetration flow paths with two PCIVs. For penetration flow paths with one PCIV, Condition C provides the appropriate Required Actions. Required Action A.2 is modified by a Note that applies to isolation devices located in high radiation areas, and , allows them to be verified by use of administrative means.- ' Allowing verification by administrative means is considered acceptable, since access to these areas is typically
\ -
restricted. Therefore, the probability of misalignment, once they have been verified to be in the proper position, ' is low. M i With one or more penetration flow paths with two PCIVs , inoperable except due to leakage not within limits, either-the inoperable PCIVs must be restored to OPERABLE status or i the affected penetration flow path must be isolated within : I hour. The method of isolation must include the use of at ; least one isolation barrier that cannot be adversely , affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated . automatic valve, a closed manual valve, and a blind flange. ! A check valve may not be used to isolate the affected i penetration. The I hour Completion Time is consistent with- _ the ACTIONS of LC0 3.6.1.1. l l
-(g) - (continued)
HATCH UNIT 1 B 3.6-19 REVISION A l Y
PCIVs B 3.6.1.3 BASES ACTIONS JL.1 (continued) Condition B is modified by a Note indicating this Condition is only applicable to penetration flow paths with two PCIVs. For penetration flow paths with one DCIV, Condition C provides the appropriate Required Actions. C.1 and C.2 With one or more penetration flow paths with one PCIV A inoperable, except due to leakage not within limits, the I m inoperable valve must be restored to OPERABLE status or the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by-a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration. Required Action C.1 must be completed within 4 hours for lines other than excess flow check valve (EFCV) lines and 12 hours for EFCV lines. The Completion Time of 4 hours is reasonable considering the relative stability of the closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of support'ing primary containment OPERABILITY during MODES 1, 2, and 3. The Completion Time of 12 hours is reasonable considering the instrument to act as a penetration isolation boundary and the small pipe diameter of the affected penetrations. In the event the affected penetration flow path is isolated in accordance with Required Action C.1, the affected penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that primary, containment penetrations required to be isolated following an accident are isolated. The Completion Time of once per 31 days for verifying each affected penetration is isolated is appropriate because the valves are operated under administrative controls and the probability of their misalignment is low. Condition C is modified by a Note indicating that this Condition is only applicable to penetration flow paths with only one PCIV. For penetration flow paths with two PCIVs, Conditions A and B provide the appropriate Required Actions. i (continued) HATCH UNIT 1 B 3.6-20 REVISION D ) l
PCIVs B 3.6.1.3 i Q,m BASES ACTIONS C.1 and C.2 (continued) Required Action C.2 is modified by a Note that applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is low. D_d With the MSIV leakage rate not within limit, the assumptions of the safety analysis may not be met. Therefore, the leakage must be restored to within limit within 4 hours. Restoration can be accomplished by isolating the penetration that caused the limit to be exceeded by use of one closed e and de-activated automatic valve, closed manual valve, or blind flange. When a penetration is isolated, the leakage rate for the isolated penetration is assumed to be the (3_ actual pathway leakage through the isolation device. If two () isolation devices are used to isolate the penetration, the leakage rate is assumed to be the lesser actual pathway leakage of the two devices. The 4 hour Completion Time is reasonable considering the time required to restore the leakage by isolating the penetration and the relative importance to the overall containment function. E 1 and E.2 If any Required Action and associated Completion Time cannot i be met 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 De 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. i 1 (continued) ( ) HATCH UNIT 1 B 3.6-21 REVISION A
r
=
PCIVs B 3.6.1.3 BASES h ACTIONS F.1 and F.2 (continued) If any Required Action and associated Completion Time cannot be met, the unit must be placed in a condition in which the LC0 does not apply. 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 subsequent potential for fission product release. Actions must continue until 0PDRVs are suspended and the valve (s) are restored to OPERABLE status. If suspending an OPDRV would result in closing the residual - heat removal (RHR) shutdown cooling isolation valves, an alternative Required Action is provided to immediately initiate action to restore the valve (s) to OPERABLE status. This allows RHR shutdown cooling to remain in service while actions are being taken to restore the valve. SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS This SR ensures that the 18 inch primary containment purge valves are closed as required or, if open, are open for an & allowable reason. If a purge valve is open in violation of W this SR, the valve is considered inoperable (Condition A applies) . The SR is modified by a Note stating that the SR is not required to be met when the 18 inch purge valves are open for the stated reasons. The Note states that these valves may be opened for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open. The 18 inch purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The 31 day Frequency is consistent with other PCIV requirements discussed in SR 3.6.1.3.2. k SR 3.6.1.3.2 This SR verifies that each primary containment isolation manual valve and blind flange that is located outside primary containment and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits. (continued) HATCH UNIT 1 B 3.6-22 REVISION D m
PCIVs l B 3.6.1.3 m BASES (] , SURVEILLANCE SR 3.6.1.3.2 (continued) l REQUIREMENTS This SR does not require any testing or valve manipulation. Rather, it involves verification that those isolation devices outside primary containment, and capable of being mispositioned, are in the correct position. Since verification of valve position for isolation devices outside primary containment 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 allows valves and blind flanges located in high radiation areas 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 PCIVs that are open under administrative controls are not required to meet the SR during the time 7 that the PCIVs are open. J SR 3.6.1.3.3 This SR verifies that each primary containment manual isolation valve and blind flange that is located inside primary containment and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits. For these isolation devices inside primary containment, the Frequency defined as " Prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days" is appropriate since these isolation devices are operated under-administrative controls and the probability of their misalignment is low. Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered j l (continued) HATCH UNIT 1 B 3.6-23 REVISION A
. - - - _ . _ - _ - _ - . - _ - - _ _ _ - - - _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - . - _ - _ _ _ _ _ _ _ _ - . . _ _ . _ _ _ _ _ _ _ _ _ _ _ - - - _ . _ _ _ _ _ _ - - - _ - _ _ _ _ . _ _ _ - _ - - _ . - - _ _ _ - - - ~ _ __
PCIVs B 3.6.1.3 BASES h SURVEILLANCE SR 3.6.1.3.3 (continued) REQUIREMENTS acceptable since the primary containment is inerted and access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA and personnel safety reasons. Therefore, the probability of misalignment of these isolation devices, once they have been verified to be in their proper position, is low. A second Note has been included to clarify that PCIVs that are open under administrative controls are not required to meet the SR during the time that the PCIVs are open. SR 3.6.1.3.4 The traversing incore probe (TIP) shear isolation valves are actuated by explosive charges. Surveillance of explosive charge continuity provides assurance that TIP valves will actuate when required. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The 31 day Frequency is based on operating experience that has demonstrated the reliability of the explosive charge continuity. SR 3.6.1.3.5 Verifying the isolation time of each power operated and each automatic PCIV is within limits is required to demonstrate OPERABILITY. MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.6. k i The isolation time test ensures that each valve will isolate l in a time period less than or equal to that listed in the l FSAR and that no degradation affecting valve closure since the performance of the last Surveillance has occurred. (EFCVs are not required to be tested because they have no specified time limit). The Frequency of this SR is in accordance with the requirements of the Inservice Testing Program. i SR 3.6.1.3.6 Verifying that the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY. k; The isolation time test ensures that the MSIV will isolate (continued) h HATCH UNIT I B 3.6-24 REVISION D l
i PCIVs B 3.6.1.3 i
. BASES
_ Nltl' 1 . in a time period that does not exceed the times assumed in the DBA analyses. This ensures that the calculated .
.A-radiological consequences of these events remain _within 10 4: -
CFR 100 limits. The Frequency of this SR is in accordance: i with the requirements of'the Inservice Testing Program. SR 3.6.1.3.7 II 4 Automatic PCIVs close on a primary containment-isolation . signal to prevent leakage of radioactive material from l primary containment following a DBA. This SR ensures that -' each automatic PCIV will actuate.to its isolation position- , f 1 o 1 l, t r e e .. ?
- f, s <
]
i
- \
Q (continued) ;
- HATCH UNIT'l B 3.6-24A REVISION D-a
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- a PCIVs- l B 3.6.1.3 J q
- ry
'( g BASES . SURVEILLANCE REQUIREMENTS SR 3.6.1.3.7 (continued) 'I k on a primary containment isolation signal. The LOGIC SYSTEM ; FUNCTIONAL TEST in SR 3.3.6.1.6 overlaps this SR to provide- , complete testing of the safety function. The 18 month ) Frequency was developed considering it is prudent that this ' Surveillance be performed only during a unit outage since y isolation of penetrations would eliminate cooling water flow > and disrupt the normal operation of many critical components. Operating experience has shown that these - components usually pass this Surveillance when performed at . the 18 month Frequency. Therefore, the Frequency was - concluded to be acceptable from a reliability standpoint. SR 3.6.1.3.8 lk This SR requires a demonstration that each reactor i instrumentation line excess flow check valve (EFCV) is - OPERABLE by verifying that the valve reduces flow to within : limits on an actual or simulated instrument-line break ! condition. This SR provides assurance that the instrumentation line EFCVs will perform.as designed. The ' O' 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. i Operating experience has shown that these components usually pass.this Surveillance when performed at the 18 month l Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. SR 3.6.1.3.9 I i The TIP shear isolation valves are actuated by explosive . charges. An in place functional test is not possible with ; this design. The explosive squib is removed and tested to ' provide assurance that the valves will actuate when ' required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been. certified by having one of the batch successfully fired. The Frequency of 18 months' on a STAGGERED TEST BASIS is considered adequate given the , administrative controls on replacement charges and the frequent checks of circuit continuity (SR 3.6.1.3.4). , [ (continued) ! s 1 HATCH UNIT I B 3.6-25 REVISION D l
)
- 1. ( 1 j
PCIVs B 3.6.1.3 l BASES h SURVEILLANCE SR 3.6.1.3.10 l REQUIREMENTS (continued) The analyses in References 1 and 3 are based on leakage that is less than the specified leakage rate. Leakage through each MSIV must be s 11.5 scfh when tested at 2 28.0 psig. The MSIV leakage rate must be verified to be in accordance with the leakage test requirements of 10 CFR 50, Appendix J (Ref. 4), as modified by approved exemptions. This ensures that MSIV leakage is properly accounted for in determining the overall primary containment leakage rate. The Frequency is required by 10 CFR 50, Appendix J, as modified by approved exemptions; thus, SR 3.0.2 (which allows Frequency extensions) does not apply. SR 3.6.1.3.11 Ik The valve seats of each 18 inch purge valve (supply and exhaust) having resilient material seats must be replaced every 18 months. This will allow the opportunity for repair before gross leakage failure develops. The 18 month Frequency is based on engineering judgment and operational expe.rience which shows that gross leakage normally does not occur when the valve seats are replaced on an 18 month W Frequency. SR 3.6.1.3.12 Ik The Surveillance Requirement provides assurance that the excess flow isolation dampers can close following an isolation signal. The 18 month Frequency is based on vendor recommendations and engineering judgment. Operating experience has shown that these dampers usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from i a reliability standpoint. l REFERENCES 1. FSAR, Section 14.4.
- 2. Technical Requirements Manual
- 3. FSAR, Section 5.2.
l (continued) h l HATCH UNIT 1 B 3.6-26 REVISION D j 1 l
A LLS Valves 1 B 3.6.1.6 l 1 BASES _ 1 APPLICABLE assumption that simultaneous S/RV openings occur only on l SAFETY ANALYSES the initial actuation for DBAs. Even though four LLS S/RVs l (continued)' are specified, all four LLS S/RVs do not operate in any DBA- : analysis. ; LLS valves satisfy Criterion 3 of the NRC Policy Statement' l (Ref. 3). l l LC0 Four LLS valves are required to be OPERABLE to satisfy the e assumptions of the safety analyses (Ref.1). The requirements of this LC0 are applicable to the mechanical j and electrical / pneumatic capability of the LLS valves to ! function for controlling the opening and closing of the ' S/RVs. ; APPLICABILITY In MODES 1, 2, and 3, an event could cause pressurization of the reactor and opening of S/RVs. In MODES 4 and 5, the' l i probability and consequences of these events are reduced due ! to the pressure and temperature limitations in these MODES. O Therefore, maintaining the LLS valves OPERABLE is not required in MODE 4 or 5. l
.l l
ACTIONS Ad l With one LLS valve inoperable, the remaining OPERABLE LLS , valves are adequate to perform the designed function. , However, the overall reliability is reduced. The 14 day + Completion Time takes into account the redundant capability ! afforded by the remaining LLS valves and the low probability ! of an event in which the remaining LLS valve capability l would be inadequate, j j B.1 and B.2 ' If two or more LLS valves are inoperable or if the l inoperable LLS valve cannot be restored to OPERABLE status 1 within the required Completion Time, .the plant must be i brought to a MODE in which the LCO does not apply. To ; Q (continued) HATCH UNIT I B 3.6-35 REVISION A
LLS Valves B 3.6.1.6 BASES h ACTIONS B.1 and B.2 (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. SURVEILLANCE SR 3.6.1.6.1 REQUIREMENTS A manual actuation of each LLS valve is performed to verify that the valve and solenoids are functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control or bypass valve, by a change in the measured steam flow, or by any other method that is suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Adequate pressure at which this test is to be performed is 2: 920 psig (the pressure recommended by the valve manufacturer). Also, l adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the LLS valves divert steam flow upon opening. Adequate steam flow is represented by at least 1.25 turbine bypass valves .open, or total steam flow > IE6 lb/hr. The A 18 month Frequency was based on the S/RV tests required by the ASME Boiler and Pressure Vessel Code, Section XI l (Ref. 2). Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a relirbility standpoint. Since steam pressure is required to Nrform the Surveillance, however, and steam may not be available during a unit outage, the Surveillance may be performed during the startup following a unit outage. Unit startup is allowed prior to performing the test because valve OPERABILITY and the setpoints for overpressure protection are verified by ASME Section XI testing prior to valve installation. After adequate reactor steam pressure and flow are reached, 12 hours is allcwed to prepare for and perform the test. (continued) HATCH UNIT 1 B 3.6-36 REVISION D
LLS Valves i B 3.6.1.6 O \J BASES Adequate pressure at which this test is to be performed is ; consistent with the pressure recommended by the valve ' manufacturer. j t i i I g , v I 1 1 j I l 1 I (Continued) HATCH UNIT I B 3.6-36A REVISION D l
Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.8 ' /~T Q BASES (continued) ACTIONS U With one of the required vacuum breakers inoperable for opening (e.g., the vacuum breaker is not open and may be stuck closed or not within its opening setpoint limit, so that it would not function as designed during an event that depressurized the drywell), the remaining nine OPERABLE i vacuum breakers are capable of providing the vacuum relief function. However, overall system reliability is reduced ' because a single failure in one of the remaining vacuum " breakers could result in an excessive suppression chamber-to-drywell differential pressure during a DBA. Therefore, t with one of the 10 required vacuum breakers inoperable, 'j 72 hours is allowed to restore at least one of the inoperable vacuum breakers to OPERABLE status so that plant ! conditions are consistent with those assumed for the design- > basis analysis. The 72 hour Completion Time is considered , acceptable due to the low probability of-an event in which l the remaining vacuum breaker capability would not be adequate. ; i O ' : An open vacuum breaker allows communication between the drywell and suppression chamber airspace,- and, as a result, there is the potential for ' suppression chamber ~ overpressurization due to this bypass leakage if a LOCA were to occur. Therefore, the open vacuum breaker must be closed. A short time is allowed to close the vacuum breaker due to the low probability of an event that would pressurize primary containment. If vacuum breaker position indication is not reliable, an alternate method' of verifying that the vacuum breakers are closed is to verify that a differential ; pressure of > 0.5 psid between the drywell and suppression chamber is maintained.for 1 hour without makeup. The required 2 hour Completion Time is considered adequate to i perform this test. , C.1 and-C.2.
-l If any Required Action and associated Completion Time 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 hourc and to MODE 4
-(continued)
HATCH UNIT 1 B 3.6-47 REVISION A j
Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.8 BASES > ACTIONS C,1 ard C,2 (continued) L l within 36 hours. The allowed Completion Times are l reasonable, based on operating experience, to reach the l required plant conditions from full power conditions in an } t orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.1.8.1 REQUIREMENTS Each vacuum breaker is verified closed to ensure that this potential large bypass leakage path is not present. This Surveillance is performed by observing the vacuum breaker position indication or by verifying that a differential pressure of 0.5 psid between the drywell and suppression chamber is maintained for 1 hour without makeup. The 14 day Frequency is based on engineering judgment, is considered adequate in view of other indications of vacuum breaker status available to operations personnel, and has been shown to be acceptable through operating experience. A Note is added to this SR which allows suppression chamber-to-drywell vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers. SR 3.6.1.8.2 Each required vacuum breaker must be cycled to ensure that it opens adequately to perform its design function and returns to the fully closed position. This ensures that the safety analysis assumptions are valid. The 31 day Frequency of this SR was developed, based on Inservice Testing Program 4 requirements to perform valve testing at least once every j 92 days. A 31 day Frequency was chosen to provide j additional assurance that the vacuum breakers are OPERABLE, I since they are located in a harsh environment (the ) suppression chamber airspace). In addition, this functional a test is required within 12 hours after a discharge of steam im to be suppression chamber from the safety / relief valves. I (continued) HATCH UNIT 1 B 3.6-48 REVISION D
1 l Secondary Containment B 3.6.4.1 o () 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 d
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 modified Unit 1 secondary containment configuration is used, these SRs also include verifying the hatches and doors separating the common refueling floor zone from the Unit I reactor building. 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 otner indications of door and hatch status that are available to the operator. G V SR 3.6.4.1.3 and SR 3.6.4.1.4 The SGT System exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment. To ensure that all fission products are treated, SR 3.6.4.1.3 verifies that the SGT System will rapidly establish and maintain a pressure in the secondary containment that is less than the lowest postulated pressure external to the secondary containment boundary. This is confirmed by demonstrating that two SGT subsystems, one of which may be a Unit 2 subsystem, will draw down the secondary containment to 2 0.25 inch of vacuum water gauge in s 120 seconds, consistent with the LOCA analysis (Ref.1). As noted, the draw down acceptance criteria time is s 100 seconds during movement of irradiated fuel assemblies in the secondary containment, CORE ALTERATIONS, and OPDRVs, which is consistent with the fuel handling accident inside secondary containment analysis (Ref. 2). This cannot be accomplished if the secondary containment boundary is not intact. SR 3.6.4.1.4 demonstrates that two SGT subsystems can maintain 2 0.25 inch of vacuum water gauge for 1 hour at a flow rate s 8000 cfm (s 4000 cfm for each of two O g (continued) HATCH UNIT 1 B 3.6-79 REVISION D
1 Secondary Containment B 3.6.4.1 BASES (continued) subsystems). The 1 hour test period allows secondary containment to be in thermal equilibrium at steady state 9 (continued) h HATCH UNIT 1 B 3.6-79A REVISION D
^
'i Secondary Containment ~ :l B 3.6.4.1
() BASES SURVEILLANCE SR 3.6.4.1.3 and SR 3.6.4.1.4 (continued) I REQUIREMENTS conditions. Therefore, these two tests are used-to ensure i secondary containment-boundary integrity. Since these SRs .! are secondary containment tests, they need not be performed I 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, both Unit I and Unit 2 SGT subsystems will perform this test. Operating experien'ce has- , shown these components usually pass the Surveillance when .i i performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability. standpoint. ! I REFERENCES 1. FSAR, Section 14.4.3.
'I
- 2. FSAR, Section 14.4.4. ,
- 3. NRC No. 93-102, " Final Policy Statement on Technical :
Specification Improvements," July 23, 1993. l o i Y i
?
I HATCH UNIT 1 B 3.6-80 REVISION A' l Z
- ._ y , y
AC Sources-Operating B 3.8.1 ,m (_) BASES ACTIONS f_d (continued) The intent here is to avoid the risk associated with an
- immediate controlled shutdown and to minimize the risk associated with this level of degradation.
According to Regulatory Guide 1.93 (Ref. 6), with two or more DGs inoperable, operation may continue for a period - that should not exceed 2 hours. (Regulatory Guide 1.93 assumed the unit has two DGs. Thus, a loss of both DGs results in a total loss of onsite power. Therefore, a loss of more than two DGs, in the Plant Hatch design, results in degradation no worse than that assumed in Regulatory Guide 1.93. In addition, the loss of a required Unit 2 DG concurrent with the loss of a Unit 1 or swing DG, is analogous to the loss of a single DG in the Regulatory Guide 1.93 assumptions, thus, entry into this Condition is not required in this case). G.1 and G.2 (T If the inoperable AC electrical power sources cannot be (./ restored to OPERABLE status within the associated Completion Time, the unit must be brought to a MODE in which the LC0 does not apply. To achieve this status, the unit 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. lb.1 Condition H corresponds to a level of degradation in which all redundancy in the AC electrical power supplies has been lost. At this severely degraded level, any further losses in the AC electrical power system will cause a loss of function. Therefore, no additional time is justified for continued operation. The unit is required by LC0 3.0.3 to commence a controlled shutdown. m (continued) HATCH UNIT I B 3.8-17 REVISION A
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE The AC sources are designed to permit inspection and i REQUIREMENTS testing of all important areas and features, especially those that have a standby function, in accordance with 10 CFR 50, GDC 18 (Ref. 8). Periodic component tests are supplemented by extensive functional tests during refueling outages under simulated accident conditions. The SRs for demonstrating the OPERABILITY of the DGs are generally consistent with the recommendations of Regulatory Guide 1.9 (Ref. 9), Regulatory Guide 1.108 (Ref. 10), and Regulatory l Guide 1.137 (Ref. 11), although Plant Hatch Unit 1 is not i committed to these Regulatory Guides. Specific commitments relative to DG testing are described in FSAR section 8.4 (Ref. 2). Where the SRs discussed herein specify voltage and frequency tolerances, the following summary is applicable. The allowable values for achieving steady state voltage are specified within a range of minus 10 percent (3740V) and l plus 2 percent (4243V) of 4160 V. The Allowable Value of l 3740 V is consistent with Regulatory Guide 1.9 for demonstrating that the diesel generator is capable of l attaining the required voltage. A more limiting value of 4243 V is specified as the allowable value for overvoltage due to overvoltage limits on the 600 V buses. The plus 2 percent value maintains the required overvoltage limits. T The specified minimum and maximum frequencies of the DG are 58.8 Hz and 61.2 Hz, respectively. These values are equal to 2% of the 60 Hz nominal frequency and are derived from the recommendations found in Regulatory Guide 1.9 (Ref. 9). The SRs are modified by a NOTE to indicate that SR 3.8.1.1 A I l through SR 3.8.1.18 apply only to the Unit 1 AC sources, and that SR 3.8.1.19 applies only to the Unit 2 AC sources. g SR 3.8.1.1 This SR ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure that distribution buses and loads are connected to their preferred power source and that (continued) HATCH UNIT 1 B 3.8-18 REVISION D
AC Sources-Operating - B 3.8.1
)
BASES (continued)
'I appropriate independence of offsite' circuits is maintained. )
The 7. day Frequency is adequate since breaker position is i not likely to change without the operator.being aware of it i and because its status is displayed .in the control room.- ; l
- i
.1 0 1 l .f]
i l( (continued) HATCH UNIT 1 B 3.8-18A REVISION D ;
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i AC Sources-Operating l B 3.8.1 i BASES
)
SURVEILLANCE SR 3.8.1.5 (continued) REQUIREMENTS Note 3 modifies this Surveillance by stating that momentary voltage or load transients because of changing bus loads do not invalidate this test. Note 4 indicates that this Surveillance is required to be l conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations. To minimize testing of the swing DG, Note 5 allows a single l 1 test (instead of two tests, one for each unit) to satisfy the requirements for both units, with the DG started using l the starting circuitry of one unit and synchronized to the ESF bus of that unit for one periodic test and started using the starting circuitry of the other unit and synchronized to the ESF bus of that unit during the next periodic test. This is allowed since the main purpose of the Surveillance, to ensure DG OPERABILITY, is still being verified on the' ' proper frequency, and each unit's starting circuitry and breaker control circuitry, which is only being tested every second test (due to the staggering of the tests), historically have a very-low failure rate. If the swing DG J fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit. -J l SR 3.8.1.6 Transfer of each 4.16 kV ESF bus power supply' from the normal offsite circuit to the alternate offsite circuit J demonstrates the OPERABILITY of the alternate circuit distribution network to power the shutdown loads. The 18 month Frequency of the Surveillance is based on engineering judgment taking into consideration the plant conditions required to perform the Surveillance, and is 1 intended to be consistent with expected fuel cycle lengths. ]
^
1
- Operating experience has shown that these components usually pass the SR when-performed on the 18 month Frequency.
Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. (continued) HATCH UNIT 1 B3.8-2h REVISION _ _ _ -_ -_ _ i
AC Sources-Operating i B 3.8.1 l BASES h 1 SURVEILLANCE SR 3.8.1.6 (continued) l REQUIREMENTS l This SR is modified by a Note. The reason for the Note is ' that, during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that coulu challenge continued steady state operation and, as a result, plant safety systems. Credit may be taken for unplanned events that satisfy this SR. This Surveillance tests the applicable logic associated with the Unit I swing bus. The comparable test specified in the Unit 2 Technical Specifications tests the applicable logic associated with the Unit 2 swing bus. Consequently, a test must be performed within the specified Frequency for each unit. The Note specifying the restriction for not performing the test while the unit is in MODE 1 or 2 does not have applicability to Unit 2. As the Surveillance represents separate tests, the Unit 1 Surveillance should not be performed with Unit 1 in MODE 1 or 2 and the Unit 2 test should not be performed with Unit 2 in MODE 1 or 2. SR 3.8.1.7 ' Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This Surveillance demonstrates the DG load response characteristics and capability to reject the largest single load without exceeding predetermined voltage and frequency and while maintaining a specified margin to the overspeed trip. The largest single load for DGs IA and It is a core spray pump at rated flow (1275 bhp). For DG 1B, the largest single load is a residual heat removal service water pump at rated flow (1225 bhp). This a Surveillance may be accomplished by: a) tripping the DG l 2 output breaker with the DG carrying greater than or equal to its associated single largest post-accident load while 3aralleled to offsite power or while solely supplying the Id aus, or b) tripping its associated single largest post-accident load with the DG solely supplying the bus. g Although Plant Hatch Unit 1 is not committed to IEEE-387-1984 (Ref. 12), this SR is consistent with the IEEE-387-1984 requirement that states the load rejection test is acceptable if the increase in diesel speed does not exceed 75% of the difference between synchronous speed and the (continued) HATCH UNIT 1 B 3.8-24 REVISION D
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AC Sources-Operating B 3.8.1 BASES h SURVEILLANCE SR 3.8.1.8 (continued) REQUIREMENTS including reconnection to the bus if the trip initiator can be corrected or isolated. In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, testing must be performed using a power factor s 0.88. This power factor is chosen to be representative of the actual design basis inductive loading that the DG would experience. The 18 month Frequency is consistent with the recommendation of Regulatory Guide 1.108 (Ref.10) and is intended to be consistent with expected fuel cycle lengths. This SR is modified by three Notes. The reason for Note I ld is that during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that would challenge continued steady state operation and, as a result, plant safety systems. Credit may be taken for unplanned events that satisfy this SR. Note 2 is provided in recognition 1 that if the offsite electrical power distribution system is lightly loaded (i.e., system voltage is high), it may not be possible to raise voltage without creating an overvoltage condition on the ESF bus. Therefore, to ensure the bus voltage, supplied ESF loads, and DG are not placed in an unsafe condition during this test, the power factor limit does not have to be met if grid voltage or ESF bus loading does not permit the power factor limit to be met when the DG is tied to the grid. When this occurs, the power factor should be maintained as close to the limit as practicable. To minimize testing of the swing DG, Note 3 allows a single test (instead of two tests, one for each unit) to satisfy Id the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on either unit (no unit specific DG components are ; being tested). If the swing DG fails one of these ' Surveillances, the DG should be considered inoperable on I both units, unless the cause of the failure can be directly related to only one unit. (continued) HATCH UNIT 1 B 3.8-26 REVISION D l L
AC Sources-Operating B.3.8.1 BASES SURVEILLANCE SR 3.8.1.18 REQUIREMENTS l (continued) This Surveillance demonstrates that the DG starting independence has not been compromised. Also, this Surveillance demonstrates that each engine can achieve ' proper speed within the.specified time when the DGs are started simultaneously. For the purpose'of this testing, ' the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and ; temperature maintained consistent with manufacturer recommendations. It is permissible to place all three DGs ; in test simultaneously, for the performance of this Surveillance. The 10 year Frequency is consistent with the recommendations of Regulatory Guide 1.108 (Ref.10). This SR is modified by a Note. The reason for the Note is to minimize wear on the DG during testing. r SR 3.8.1.19 With the exception of this Surveillance, all other -
- O Surveillances of this Specification (SR 3.8.1.1 through SR 3.8.1.18) are applied only to the Unit 1 DG and offsite l circuits, and swing DG. This Surveillance is provided to !
direct that the appropriate Surveillances for the required-Unit 2 DG and offsite circuit are governed by the Unit 2- . Technical Specifications. Performance of the applicable Unit 2 Surveillances will satisfy both any Unit 2 requirements, as well as satisfying this Unit 1 Surveillance . requirement. Several exceptions are noted to the Unit 2 SRs: SR 3.8.1.6 is excepted since only one Unit 2 circuit d - is required by the Unit 1 Specification (therefore, there is , not necessarily a second circuit to transfer to); I SRs 3.8.1.10, 11, 15, and 17 are excepted since they relate to the DG response to a Unit 2 ECCS initiation signal, which k ; is not a necessary function for support of the Unit I requirement for an OPERABLE Unit 2 DG; and SR 3.8.1.18 is excepted since there is only one Unit 2 DG required by the . Unit 1 Specification (therefore, there are not necessarily multiple DGs for simultaneous start). 4 f- , j i The Frequency required by the applicable Unit 2 SR also ) governs performance of that SR for both Units. j (continued) HATCH UNIT 1 B 3.8-37 REVISION D e
AC Sources-Operating B 3.8.1 BASES h l i l l l 9 l I l l REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.
- 2. FSAR, Sections 8.3 and 8.4.
(continued) HATCH UNIT I REVISION M B 3.8-f( 38
AC Sources - Shutdown B 3.8.2 /% U BASES APPLICABLE During MODES 1, 2, and 3, various deviations from the SAFETY ANALYSES analysis assumptions and design requirements are allowed (continued) within the ACTIONS. This allowance is in recognition that certain testing and maintenance activities must be conducted, provided an acceptable level of risk is not 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 administrative 1y controlled. Relaxations from typical MODES 1, 2, and 3 LC0 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,
- 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.
O V
- c. Prudent utility consideration of the risk associated with multiple activities that could affect multiple 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.
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 loss of all onsite (diesel generator (DG)) power. The AC sources satisfy Criterion 3 of the NRC Policy Statement (Ref. 1). LCO One Unit 1 offsite circuit capable of supplying the onsite- Ib Class 1E power distribution subsystem (s) of LC0 3.8.8,
" Distribution Systems - Shutdown," ensures that all required (continued)
HATCH UNIT 1 B 3.8-41 REVISION D
AC Sources - Shutdown B 3.8.2 BASES g ntI ds are powered from offsite power. An OPERABLE I O (continued) h HATCH UNIT I B 3.8-41A REVISION D
AC Sources - Shutdown B 3.8.2 m BASES (] LC0 associated with a Distribution System Engineered Safety (continued) Feature (ESF) bus required OPERABLE by LC0 3.3.8, ensures that a diverse power source is available for previding electrical power support assuming a loss of the'offsite cirt.uit. 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 the required Unit 2 SGT subsystem, must also be OPERABLE. Together, OPERABILITY of the required 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 7 basis for the unit. The Unit I and Unit 2 offsite circuits (V consist of incoming breaker and disconnect to the IC or ID and the 2C or 2D startup auxiliary transformers (SATs), associated IC 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 sequeace inust 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 engine hot and DG in standby with engine at ambient conditions. Additional DG capabilities must be demonstrated to meet required A Surveillances, e.g., capability of the DG to revert to du standby status on an ECCS signal while operating in parallel test mode. (continued) HATCH UNIT I B 3.8-42 REVISION D , 1
AC Sources - Shutdown B 3.8.2 BASES g Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY. 9 (continued) HATCH UNIT 1 B 3.8-42A REVISION D
AC Sources - Shutdown 4 B 3.8.2 : 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. l APPLICABILITY 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:
- 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 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.
~
AC power requirements for MODES 1, 2, and 3 are covered in 'I LCO 3.8.1. ACTIONS Ad , 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 1 operations with a potential for draining the reactor vessel.
- By the allowance of the option to declare required features inoperable with no offsite power available,-appropriate .
restrictions can be implemented in accordance with the . k. ! affected required feature (s) LCOs' ACTIONS. , (continued)- HATCH UNIT I B 3.8-43 REVISION D
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 l (continued) l 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 Id 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 irradiated fuel assemblies in the secondary containment, and activities that Id 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 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 ) f systems. Th - (.w.aletion Time of immediately is consistent with the ) regt.:r.:d times for actions requiring prompt attention. The l l restoration of the required AC electrical power sources l should be completed as quickly as possible in order to l minimize the time during which the plant safety systems may j be without sufficient power. l ! Pursuant to LC0 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 LCO 3.8.8 must be immediately entered. This Note allows Condition A to 1 i (continued) l HATCH UNIT 1 B 3.8-44 REVISION D
Diesel Fuel Oil and Transfer, Lube Oil, and Starting Air-B 3.8.3 W(/ BASES LCO addressed in LC0 3.8.1, "AC Sources - Operating," and (continued) LCO 3.8.2, "AC Sources - Shutdown." The starting air system is required to have a minimum capacity for five successive DG start attempts without recharging the air start receivers. Only one air start receiver per DG is required, since each air start receiver has the required capacity. APPLICABILITY The AC sources (LC0 3.8.1 and LC0 3.8.2) are required to ensure the availability of the required power to shut down ' the reactor and maintain it in a safe shutdown condition after an A00 or a postulated DBA. Because stored diesel fuel oil and transfer, lube oil, and starting air subsystem support LC0 3.8.1 and LC0 3.8.2, stored diesel fuel oil and transfer, lube oil, and starting air are required to be within limits when the associated DG is required to be l OPERABLE. l ACTIONS The ACTIONS Table is modified by a Note indicating that IA separate Condition entry is allowed for each DG. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable DG subsystem. Complying with the Required Actions for one inoperable DG subsystem may allow for continued operation, I$ and subsequent inoperable DG subsystem (s) are governed by separate Condition entry and application of associated-Required Actions. Ad -; l-With one or more required DGs with one fuel oil transfer r pump inoperable, the inoperable pump must be restored to OPERABLE status within 30 days. With the unit in this i condition, the remaining OPERABLE fuel transfer pump is I adequate to perform the fuel transfer function. However, the overall reliability is reduced because a single failure l-in the OPERABLE pump could result in loss of the associated 1 \ (- ! L i 1 ( (continued)
.," HATCH. UNIT 1 B 3.8-49 REVISION D i
I - -- i -
Diesel Fuel Oil and Transfer, Lube Oil, and Starting Air B 3.8.3 BASES h ACTIONS Ad (continued) DG and loss of the fuel oil in the respective tank. The 30 day Completion Time is based on the remaining fuel oil transfer capability, and the low probability of the need for the DG concurrent with a worst case single failure. Ed In this condition, the 7 day fuel oil supply for a required DG is not available. However, the Condition is restricted to fuel oil level reductions that maintain at least a 6 day supply. These circumstances may be caused by events such as:
- a. Full load operation required for an inadvertent start while at minimum required level; or
- b. Feed and bleed operations that may be necessitated by increasing particulate levels or any number of other oil quality degradations.
This restriction allows sufficient time fnr obtaining the requisite replacement volume and performing the analyses required prior to addition of the fuel oil to the tank. A period of 48 hours is considered sufficient to complete restoration of the required level prior to declaring the DG inoperable. This period is acceptable based on the remaining capacity (> 6 days), the fact that procedures will be initiated to obtain replenishment, and the low probability of an event during this brief period.
.C d With a required DG lube oil inventory < 400 gal, sufficient lube oil to support 7 days of continuous DG operation at full load conditions may not be available. However, the Condition is restricted to lube oil volume reductions that maintain at least a 6 day supply. This restriction allows sufficient time for obtaining the requisite replacement i volume. A period of 48 hours is considered sufficient to f complete restoration of the required volume prior to (continued)
HATCH UNIT 1 B 3.8-50 REVISION A 1
. . . ~.. .
DC Sources - Shutdown B 3.8.5 t i BASES M. i 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 LC0 3.8.8, " Distribution Systems - Shutdown." In addition, some components that may be required by Unit I require power from Unit 2 sources (e.g., Standby Gas Treatment (SGT) System). Therefore, the Unit 2 DG DC electrical power subsystems needed to provide DC power to the required Unit 2 components are also required to be' l 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). , 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 assurance 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. ) , h v (continued) j
' HATCH UNIT 1 B 3.8-69 REVISION D l
1
DC Sources - Shutdown B 3.8.5 I BASES h APPLICABILITY The DC electrical power requirements 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 LCO 3.8.8, the DC subsystems remaining OPERABLE with one or more DC powor 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 option to declare required features inoperable with associated DC power sources inoperable, appropriate restrictions are implemented in accordance with a 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, and any activities that could result in inadvertent draining of the reactor vessel). Suspension of these activities shall not preclude completion g 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 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 Surveillance: required by SR 3.8.4.1 through SR 3.8.4.8. Therefore, see (continued) h HATCH UNIT 1 B 3.8-70 REVISION A
DC Sources - Shutdown B 3.8.5 BASES l SURVEILLANCE SR 3.8.5.1 (continued) REQUIREMENTS the corresponding Bases for LCO 3.8.4 for a discussion of J each SR. This SR is modified by a Note. The reason for the Note is to preclude requiring the OPERABLE DC sources from being discharged below their capability to provide the required power supply or otherwise rendered inoperable during the performance of SRs. It is the intent that these SRs must still be capable of being met, but actual performance is not required. SR 3.8.5.2 This Surveillance is provided to direct that the appropriate , Surveillances for the required Unit 2 DC sources are { governed by the Unit 2 Technical Specifications. I Performance of the applicable Unit 2 Surveillances will satisfy both any Unit 2 requirements, as well as satisfying this Unit 1 Surveillance Requirement. The Frequency required by the applicable Unit 2 SR also governs 9 performance of that SR for both Units. REFERENCES 1. FSAR, Chapters 5 and 6.
- 2. FSAR, Chapter 14.
- 3. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
O HATCH UNIT 1 B 3.8-71 REVISION A
)- Battery Cell Parameters B 3.8.6 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.6 Battery Cell Parameters BASES BACKGROUND This LC0 delineates the limits on electrolyte temperature, A level, float voltage, and specific gravity for the DC I tm electrical power subsystems batteries. A discussion of these batteries and their OPERABILITY requirements is provided in the Bases for LC0 3.8.4, "DC Sources - Operating," and LC0 3.8.5, "DC Sources - Shutdown." APPLICABLE The initial conditions of Design Basis Accident (DBA) =.r.d SAFETY ANALYSES transient analyses in FSAR, Chapters 5 and 6 (Ref.1) and Chapter 14 (Ref. 2), assume Engineered Safety Feature systems are OPERABLE. The DC electrical power subsystems provide 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 is based upon meeting the design basis of the unit. This includes maintaining at least one division of DC sources OPERABLE during accident conditions, in the event of:
- a. An assumed loss of all offsite AC or all onsite AC power; and
- b. A postulated worst case single failure.
Since battery cell parameters support the operation of the DC electrical power subsystems, they satisfy Criterion 3 of the NRC Policy Statement (Ref. 4). I LCO Battery cell parameters must remain within acceptable limits to ensure availability of the required DC power to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence or a postulated DBA. Cell parameter limits are established to allow continued DC electrical system function even with Category A and B limits not met. (continued) HATCH UNIT 1 B 3.8-72 REVISION D
)
l Battery Cell Parameters B 3.8.6 g Q BASES (continued) i APPLICABILITY The battery cell parameters are required solely for the support of the associated DC electrical power subsystem. Therefore, these cell paramcters are only required when the DC power source is required to be OPERABLE. Refer to the Applicability discussions in Bases for LCO 3.8.4 and LCO 3.8.5. ACTIONS A Note has been added providing that, for this LCO, separate Condition entry is allowed for each battery. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable i battery. Complying with the Required Actions for battery l cell parameters allows for restoration and continued i operation, and subsequent out of limit battery cell l parameters may be governed by separate Condition entry and application of associated Required Actions. A.I. A.2. and A.3 With parameters of one or more cells in one or more w/ batteries not within limits (i.e., Category A limits not met or Category B limits not met, or Category A and B limits not ! met) but within the Category C limits specified in I Table 3.8.6-1, the battery is degraded but there is still sufficient capacity to perform the intended function. Therefore, the affected battery is not required to be i considered inoperable solely as a result of Category A or B l limits not met, and continued operation is permitted for a limited period. The pilot cell electrolyte level and float voltage are required to be verified to meet the Category C limits within 1 hour (Required Action A.1). This check provides a quick indication of the status of the remainder of the battery cells. One hour provides time to inspect the electrolyte level and to confirm the float voltage of the pilot cells. One hour is considered a reasonable amount of time to perform the required verification. Verification that the Category C limits are met (Required Action A.2) provides assurance that during the time needed to restore the parameters to the Category A and B limits, ( (continued) HATCH UNIT 1 B 3.8-73 REVISION A l l J
Battery Cell Parameters B 3.8.6 BASES ACTIONS A.I. A.2. and A.3 (continued) the battery is still capable of performing its intended function. A period of 24 hours is allowed to complete the initial verification because specific gravity measurements I must be obtained for each connected cell. Taking into consideration both the time required to perform the required verification and the assurance that the battery cell parameters are not severely degraded, this time is considered reasonable. The verification is repeated at 7 day intervals until the parameters are restored to Category-A and B limits. This periodic verification is consistent with the normal Frequency of pilot cell surveillances. Continued operation is only permitted for 31 days before battery cell parameters must be restored to within Category A and B limits. Taking into consideration that, while battery capacity is degraded, sufficient capacity exists to perform the intended function and to allow time to fully restore the battery cell parameters to normal limits, this time is acceptable for operation prior to declaring the associated DC battery inoperable. u O When any battery parameter is outside the Category C limit for any connected cell, sufficient capacity to supply the maximum expected load requirement is not ensured and the corresponding DC electrical power subsystem must be declared inoperable. Additionally, other potentially extreme conditions, such as not completing the Required Actions of Condition A within the required Completion Time or average electrolyte temperature of representative cells falling below the appropriate limit (65 F for station service and 40 F for DG batteries), also are cause for immediately declaring the associated DC electrical power subsystem inoperable. SURVEILLANCE SR 3.8.6.1 REQUIREMENTS This SR verifies that Category A battery cell parameters are consistent with IEEE-450 (Ref. 3), which recommends regular battery inspections (at least one per month) including (continued) 4 1 HATCH UNIT I B 3.8-74 REVISION D
Battery Cell Parameters B 3.8.6 [)J L BASES SURVEILLANCE SR 3.8.6.1 (continued) REQUIREMENTS voltage, specific gravity, and electrolyte level of pilot Id cells. SR 3.8.6.2 The 92 day inspection of specific gravity, cell voltage, and 1A l 1evel is consistent with IEEE-450 (Ref. 3). In addition, within 24 hours of a battery overcharge > 150 V, the battery must be demonstrated to meet Category B limits. This inspection is also consistent with IEEE-450 (Ref. 3), which recommends special inspections following a severe overcharge, to ensure that no significant degradation of the battery occurs as a consequence of such overcharge. 1 SR 3.8.6.3 This Surveillance verification that the average temperature of representative cells is within limits is consistent with
/~ N a recommendation of IEEE-450 (Ref. 3) that states that the temperature of electrolyte in representative cells should be ;
determined on a quarterly basis. Lower than normal temperatures act to inhibit or reduce battery capacity. This SR ensures that the operating temperatures remain within an acceptable operating range. This limit is based on IEEE-450 or the manufacturer's recommendations when provided. Table 3. L5-1 This table delineates the limits on electrolyte level, float voltage, and specific gravity for three different IA categories.. The meaning of each category is discussed below. Category A defines the normal parameter limit for each designated pilot cell in each battery. The cells selected as pilot cells are those whose temperature, voltage, and electrolyte specific gravity approximate the condition of the entire battery. d (continued) HATCH UNIT 1 B 3.8-75 REVISION D
Battery Cell Parameters B 3.8.6 BASES h SURVEILLANCE Table 3.8.6-1 (continued) REQUIREMENTS The Category A limits specified for electrolyte level are based on manufacturer's recommendations and are consistent with the guidance in IEEE-450 (Ref. 3), with the extra
% inch allowance above the high water level indication for operating margin to account for temperature and charge
)'
effects. In addition to this allowance, footnote a to )j Table 3.8.6-1 permits the electrolyte level to be above the ! specified maximum level during equalizing charge, provided l it is not overflowing. These limits ensure that the plates ) suffer no physical damage, and that adequate electron l transfer capability is maintained in the event of transient l conditions. IEEE-450 (Ref. 3) recommends that electrolyte I level readings should be made only after the battery has been at float charge for at least 72 hours. The Category A limit specified for fica voltage is 2 2.13 V per cell. This value is based on the recommendation of IEEE-450 (Ref. 3), which states that prolonged operation of cells below 2.13 V can reduce the life expectancy of cells. The Category A limit specified for specific gravity for each pilot cell is 2 1.200 (0.015 below the manufacturer's fully ' charged nominal specific gravity) or a battery charging current that had stabilized at a low value. This value is characteristic of a charged cell with adequate capacity. According to IEEE-450 (Ref. 3), the specific gravity readings are based on a temperature of 77 F (25 C). The specific gravity readings are corrected for actual electrolyte temperature and level. For each 3 F (1.67 C) [ above 77 F (25 C), 1 point (0.001) is added to the reading; 1 point is subtracted for each 3*F below 77 F. The specific gravity of the electrolyte in a cell increases with a loss I of water due to electrolysis or evaporation. Level l correction will be in accordance with manufacturer's recommendations. l Category B defines the normal parameter limits for each connected cell. The term " connected cell" excludes any battery cell that may be jumpered out. (continued) HATCH UNIT 1 B 3.8-76 REVISION D
Battery Cell Parameters
.B-3.8.6-
./ y Q BASES The Category B limits specified for electrolyte level and float voltage are.the same as those specified for Category A and have been discussed above. The Category B limit specified for specific gravity for each connected cell is 2: 1.195 (0.020 below the manufacturer's fully charged, . ' nominal specific gravity) with the average of all connected cells 1.205 (0.010 below the manufacturer's fully charged, nominal specific gravity). These. values are based on
..g-manufacturer's recommendations. The minimum specific gravity value required for each cell ensures that the-effects'of a highly charged or newly installed cell-do not mask overall degradation of the battery. ,
Category C defines the limits for each connected cell. These values, although reduced, provide assurance that . sufficient capacity exists to perform the intended function and maintain-a margin of safety. When any battery parameter is outside the Category C limit, the assurance of sufficient capacity described above no longer exists, and the battery must be declared inoperable. The Category C limits specified for electrolyte level (above O the top of the plates and not. overflowing) ensure that the ; \ plates suffer no physical damage and maintain adequate r I (continued) i HATCH UNIT 1 B 3.8-76A REVISION D {
Battery Cell Parameters B 3.8.6 BASES SURVEILLANCE Table 3.8.6-1 (continued) REQUIREMENTS electron transfer capability. The Category C limit for voltage is based on IEEE-450 (Ref. 3), which states that a cell voltage of 2.07 V or below, under float conditions and not caused by elevated temperature of the cell, indicates internal cell problems and may require cell replacement. The Category C Allowable Value of average specific gravity 2: 1.195, is based on manufacturer's recommendations (0.020 below the manufacturer's recommended fully charged, nominal specific gravity). In addition to that_ limit, it is required that the specific gravity for each connected cell must be no less than 0.020 below the average of all connected cells. This limit ensures that the effect of a highly charged or new cell does not mask overall degradation of the battery. The footnotes to Table 3.8.6-1 that apply to specific gravity are applicable to Category A, B, and C specific gravity. Footnote b of Table 3.8.6-1 requires the above mentioned correction for electrolyte level and temperature,
\, with the exception that level correction is not required A
(~'/ w when battery charging current, while on float charge, is $_\
< 1 amp for station service batteries and < 0.5 amp for DG batteries. This current provides, in general, an indication of overall battery condition.
Because of specific gravity gradients that are produced during the recharging process, delays of several days may occur while waiting for the specific gravity to stabilize. A stabilized charger current is an acceptable alternative to specific gravity measurement for determining the state of charge of the designated pilot cell. This phenomenon is discussed in IEEE-450 (Ref. 3). Footnote c to Table 3.8.6-1 allows the float charge current to be used as an alternate to specific gravity for up to 7 days following a battery recharge. REFERENCES 1. FSAR, Chapters 5 and 6.
- 2. FSAR, Chapter 14.
( ( (continued) HATCH UNIT 1 B 3.8-77 REVISION D
Battery Cell Parameters B 3.8.6
- f. BASES (continued)
- 3. IEEE Standard 450 - 1987.
IA
- 4. NRC No. 93-102, "Finai Policy Statement on Technical Ik Specification Improvements," July 23, 1993.
O l i l 1 f l l i O' HATCH UNIT 1 B 3.8-77A REVISION D
Distribution Systems - Operating B 3.8.7 1 m U) t B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.7 Distribution Systems - Operating BASES BACKGROUND The onsite Class IE AC and DC electrical power distribution system is divided into redundant and independent AC and DC electrical power distribution subsystems. The primary AC distribution system consists of three 4.16 kV Engineered Safety Feature (ESF) buses each having an offsite source of power as well as a dedicated onsite diesel generator (DG) source. Each 4.16 kV ESF bus is normally connected to a normal source startup auxiliary transformer (SAT) (10). During a loss of the normal offsite power source to the 4.16 kV ESF buses, the alternate supply breaker from SAT IC attempts to close. If all offsite sources are unavailable, the onsite emergency DGs supply power to the 4.16 kV ESF buses. The secondary plant distribution system includes 600 VAC emergency buses 1C and ID and associated load centers, and O transformers. O There are two independent 125/250 VDC station service electrical power distribution subsystems and three independent 125 VDC DG electrical power distribution subsystems that support the necessary power for ESF functions. ; A description of the Unit 2 AC and DC electrical power distribution system is provided in the Bases for Unit 2 LC0 3.8.7, " Distribution System-Operating." The list of required Unit I distribution buses is presented in LC0 3.8.7. l APPLICABLE The initial conditions of Design Basis Accident (DBA) and SAFETY ANALYSES transient analyses in the FSAR, Chapters 5 and 6 (Ref.1) and Chapter 14 (Ref. 2), assume 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 (continued) HATCH UNIT 1 B 3.8-J/% REVISION /( 1
Refuel Position One-Rod-Out Interlock B 3.9.2 BASES ACTIONS A.1 and A.2 (continued) containing no fuel assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted. SURVEILLdNCE SR 3.9.2.1 REQUIREMENTS Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel.
- During control rod withdrawal in MODE 5, improper.
positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks. Therefore, this Surveillance imposes an additional level of assurance that the refueling position one-rod-out interlock - will be OPERABLE when required. By " locking" the reactor mode switch in the proper position (i.e., removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel), an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation. rs C The Frequency of 12 hours is sufficient in view of other administrative controls utilized during refueling operations to ensure safe operation. SR 3.9.2.2 : Performance of a CHANNEL FUNCTIONAL TEST on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual signal indicative of a required condition is injected.into the logic. The-CHANNEL FUNCTIONAL TEST may be performed by any ' series of sequential, overlapping, or total channel steps so that the entire channel is tested.' The 7 day. Frequency is considered adequate because of demonstrated' , circuit reliability, procedural controls on control rod ' withdrawals, and visual and audible indications available in the control room to alert the operator to control rods not ! fully inserted. To perform the required testing, the ; applicable condition must be entered '(i.e., a control rod must be withdrawn from its full-in position). Therefore, 1 SR 3.9.2.2 has been modified by a Note that states-the I dh (continued). V(G ' HATCH UNIT 1 B 3.9-7 REVISION D
Refuel Position One-Rod-Out Interlock B 3.9.2 i BASES h l SURVEILLANCE SR 3.9.2.2 (continued) REQUIREMENTS CHANNEL FUNCTIONAL TEST is not required to be performed until I hour after any control rod is withdrawn. REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.
- 2. FSAR, Section 7.6.3.
- 3. FSAR, Section 14.3.3.3.
- 4. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
O 1 l l l O; HATCH UNIT 1 B 3.9-8 REVISION A l
1 RHR - High Water Level B 3.9.7 m () 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 orie 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. 3 t
%J APPLICABILITY One RHR shutdown cooling subsystem must be OPERABLE and in operation in MODE 5, with irradiated fuel in the RPV and the l
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 A_d 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 because loss of water level could n () (continued) HATCH UNIT 1 B 3.9-23 REVISION A
RHR - High Water Level B 3.9.7 BASES h ACTIONS A J (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 j removal capabilities. Furthermore, verification of the i functional availability of these alternate method (s) must be reconfirmed every 24 hours thereafter. This will ensure continued heat removal capability. l 1 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.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 is OPERABLE; 2) two standby gas treatment subsystems (any combination of Unit 1 and Unit 2 subsystems) are OPERABLE; and 3) secondary containment isolation capability is available (i.e., one secondary containment isolation valve and associated d 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 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 l (continued) ) HATCH UNIT 1 B 3.9-24 REVISION D t - ---- ------- -------_-- _
RHR - Low Water Level B 3.9.8 o t v) 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 q 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 j 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 l Water Level." i ACTIONS _Ad ; With one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem is capable of providing i 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 g) i (continued) HATCH UNIT I B 3.9-27 REVISION A I _)
RHR - Low Water Level B 3.9.8 BASES h ACTIONS L1 (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 I 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.1. 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 Action A.1, additional actions are required to minimize any potential fission product release to the environment. This includes ensuring: 1) secondary containment is OPERABLE;
- 2) two standby gas treatment subsystems (any combination of Unit I and Unit 2 subsystems) are OPERABLE; 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 other information to determine whether the components are out of service for i maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required 1 (continued) HATCH UNIT 1 B 3.9-28 REVISION D
RHR - Low Water Level B 3.9.8
,m '(j BASES I 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. I i m i
,.~.
( (continued)
)
HATCH UNIT I B 3.9-28A REVISION D
l Inservice Leak ~and Hydrostatic Testing Operation J B 3.10.1 ; l p) ( B 3.10 SPECIAL-OPERATIONS B 3.10.1 Inservice Leak and Hydrostatic Testing Operation BASES BACKGROUND The purpose of this Special Operations LCO is to allow certain reactor coolant pressure tests to be performed in MODE 4 when the metallurgical characteristics of the. reactor r pressure vessel (RPV) require the pressure testing at , temperatures > 212 F (normally corresponding to . MODE 3). . System hydrostat-ic testing and system leakage (same as inservice leakage tests) pressure tests required by Section XI of the American Society of Mechanical Engineers t (ASME) Boiler and Pressure Vessel Code (Ref. 1) are-performed prior to the reactor going critical after a : refueling outage. Inservice system leakage tests are ; performed at the end of each refueling outage with the ' system set for normal power operation. Some parts of the Class 1 boundary are not pressurized during these system : tests. System hydrostatic tests are required once per interval and include all the Class 1 boundary unless the O test is broken into smaller portions. Recirculation pump ..! O operation and a water solid RPV (except for an air bubble for pressure control) are used to achieve the necessary temperatures and pressures required for these tests. The minimum temperatures (at the required pressures) allowed for these tests are determined from the RPV pressure and , temperature (P/T) limits required by LCD .i.4.9, " Reactor Coolant System (RCS) Pressure and . Temperature (P/T) Limits." These limits are conservatively based on the-fracture toughness of the reactor vessel, taking into account anticipated vessel neutron fluence. The hydostatic. test requires increasing pressure to approximately 1106 psig. The system leakage test requires increasing pressure to > approximately 1005 psig. , With increased reactor vessel fluence over time, the minimum 1 allowable vessel temperature increases at a given pressure. Periodic updates to the RCS P/T limit curves are performed e as necessary, based upon the results of analyses of : irradiated surveillance specimens removed from the vessel. (continued) i HATCH UNIT 1 B 3.10-1 REVISION D
Inservice Leak and Hydrostatic Testing Operation B 3.10.1 BASES APPLICABLE Allowing the reactor to be considered in MODE 4 during SAFETY ANALYSES hydrostatic or leak testing, when the reactor coolant temperature is > 212 F, effectively provides an exception to MODE 3 requirements, including OPERABILITY of primary O l (continued) HATCH UNIT 1 B 3.10-1A REVISION D
Inservice Leak and Hydrostatic Testing Operation B 3.10.1 q
) BASES containment and the full complement of redundant Emergency APPLICABLE SAFETY ANALYSES Core Cooling Systems. Since the hydrostatic or leak tests l (continued) are performed nearly water solid (except for an air bubble j for pressure control), at low decay heat values, and near MODE 4 conditions, the stored energy in the reactor core
~ will be very low. Under these conditions, the potential for failed fuel and a subsequent increase in coolant activity above the LC0 3.4.6, "RCS Specific Activity," limits are minimized. In addition, the secondary containment will be OPERABLE, in accordance with this Special Operations LCO, ! and will be capable of handling any airborne radioactivity or steam leaks that could occur during the performance of hydrostatic or leak testing. The required pressure testing conditions provide adequate assurance that the consequences of a steam leak will be conservatively bounded by the consequences of the postulated main steam line break outside of primary containment described in Reference 2. Therefore, these requirements will conservatively limit radiation releases to the environment. In the event of a large primary system leak, the reactor vessel would rapidly depressurize, allowing the low pressure ('S core cooling systems to operate. The capability of the low ! ( pressure coolant injection and core spray subsystems, as required in MODE 4 by LC0 3.5.2, "ECCS - Shutdown," would be more than adequate to keep the core flooded under this low decay heat load condition. Small system leaks would be detected by leakage inspections before significant inventory loss occurred. l For the purposes of this test, the protection provided by normally required MODE 4 applicable LCOs, in addition to the secondary containment requirements required to be met by this Special Operations LCO, will ensure acceptable i consequences during normal hydrostatic test conditions and l during postulated accident conditions. 1 i As described in LCO 3.0.7, compliance with Special . Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs . provide flexibility to perform certain operations by { appropriately modifying requirements of other LCOs. A { discussion of the criteria satisfied for the other LCOs is i' provided in their respective Bases. p) t* (continued) HATCH UNIT 1 B 3.10-2 REVISION A
1 I Inservice Leak and Hydrostatic Testing Operation i B 3.10.1 i BASES (continued) LCO As described in LC0 3.0.7, compliance with this Special Operations LC0 is optional. Operation at reactor coolant temperatures > 212*F can be in accordance with Table 1.1-1 for MODE 3 operation without meeting this Special Operations LC0 or its ACTIONS. This option may be required due to P/T limits, however, which require testing at temperatures
> 212 F, while the ASME system hydrostatic test itself requires the safety / relief valves to be gagged, preventing their OPERABILITY.
If it is desired to perform these tests while complying with this Special Operations LCO, then the MODE 4 applicable LCOs and specified MODE 3 LCOs must be met. This Special Operations LC0 allows changing Table 1.1-1 temperature limits for MODE 4 to "NA" and suspending the requirements of A LC0 3.4.8, " Residual Heat Removal (RHR) Shutdown Cooling I $ System - Cold Shutdown." The additional requirements for secondary containment LCOs to be met will provide sufficient protection for operations at reactor coolant temperatures
> 212 F for the purpose of performing either an inservice leak or hydrostatic test.
This LC0 allows primary containment to be open for frequent
/] 'v unobstructed access to perform inspections, and for outage activities on various systems to continue consistent with the MODE 4 applicable requirements that are in effect immediately prior to and immediately after this operation.
APPLICABILITY The MODE 4. requirements may only be modified for the performance of inservice leak or hydrostatic tests so that these operations can be considered as in MODE 4, even though the reactor coolant temperature is > 212 F. The additional requirement for secondary containment OPERABILITY according to the imposed MODE 3 requirements provides conservatism in the response of the. unit to any event that may occur. Operations in all other MODES are unaffected by this LCO. 7 V (continued) HATCH UNIT 1 B 3.10-3 REVISION D i
Inservice Leak and Hydrostatic Testing Operation l B 3.10.1 BASES (continued) I l ACTIONS A Note has been provided to modify the ACTIONS related to inservice leak and hydrostatic testing operation. 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, 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 each requirement of the LC0 not met provide appropriate compensatory measures for separate requirements that are not met. As such, a Note has been provided that allows separate Condition entry for each requirement of the LCO. l
!L.1 If an LC0 specified in LC0 3.10.1 is not met, the ACTIONS applicable to the stated requirements are entered immediately and complied with. Required Action A.1 has been modified by a Note that clarifies the intent of another LC0's Required Action to be in MODE 4 includes reducing the average reactor coolant temperature to s 212 F.
A.2.1 and A.2.2 Required Action A.2.1 and Required Action A.2.2 are alternate Required Actions that can be taken instead of Required Action A.1 to restore compliance with the normal MODE 4 requirements, and thereby exit this Special Operation LC0's Applicability. Activities that could further increase reactor coolant temperature or pressure are suspended immediately, in accordance with Required Action A.2.1, and the reactor coolant temperature is reduced to establish normal MODE 4 requirements. The allowed Completion Time of 24 hours for Required Action A.2.2 is based on engineering judgment and provides sufficient time to reduce the average reactor coolant temperature from the highest expected value to s 212 F with normal cooldown procedures. The Completion Time is also consistent with the time provided in LC0 3.0.3 to reach MODE 4 from MODE 3. (continued) HATCH UNIT 1 B 3.10-4 REVISION A
Single Control Rod Withdrawal - Cold Shutdown B 3.10.4 m
) BASES APPLICABLE on the withdrawn control rod, are inserted and incapable of SAFETY ANALYSES withdrawal. This alternate backup protection is required (continued) when removing a CRD because this removal renders the withdrawn control rod incapable of being scrammed.
As described in LC0 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases. . LC0 As described in LC0 3.0.7, compliance with this Special Operations LC0 is optional. Operation in MODE 4 with the reactor mode switch in the refuel position can be performed in accordance with other LCOs (i.e., Special Operations LC0 3.10.2, " Reactor Mode Switch Interlock Testing") without k meeting this Special Operations LC0 or its ACTIONS. If a single control rod withdrawal is desired in MODE 4, controls consistent with those required during refueling must be O) ( implemented and this Special Operations LC0 applied.
" Withdrawal", in this application, includes the actual withdrawal of the control rod, as well as maintaining the control rod in a position other than the full-in position, and reinserting the control rod.
The refueling interlocks of LC0 3.9.2, " Refuel Position One-Rod-Out Interlock," required by this Special Operations LC0 will ensure that only one control rod can be withdrawn. At the time CRD removal begins, the disconnection of the position indication probe will cause LC0 3.9.4, " Control Rod Position Indication," and therefore, LC0 3.9.2 to fail to be met. Therefore, prior to commencing CRD removal, a control rod withdrawal block is required to be inserted to ensure that no additional control rods can be withdrawn and that compliance with this Special Operations LC0 is maintained. To back up the refueling interlocks (LC0 3.9.2) or the control rod withdrawal block, the ability to scram the withdrawn control rod in the event of an. inadvertent criticality is provided by the Special Operations LC0 ( (continued) v HATCH UNIT 1 B 3.10-17 REVISION D
Single Control Rod Withdrawal - Cold Shutdown l B 3.10.4 i i BASES h LC0 requirements in Item c.1. Alternatively, when the scram l (continued) function is not OPERABLE, or when the_CRD is to be removed, I a sufficient number of rods in the vicinity of the withdrawn I control rod are required to be inserted and made incapable I of withdrawal (Item c.2). This precludes the possibility of criticality upon withdrawal of this control rod. Also, once this alternate (Item c.2) is completed, the SDM requirement to account for both the withdrawn-untrippable control rod, and the highest worth control rod may be changed to allow the withdrawn-untrippable control rod to be the single highest worth control rod. APPLICABILITY Control rod withdrawals are adequately controlled in MODES 1, 2, and 5 by existing LCOs. In MODES 3 and 4, control rod withdrawal is only allowed if performed in accordance with Special Operations LCO 3.10.3, or this Special Operations LCO, and if limited to one control rod. This allowance is only provided with the reactor mode switch in the refuel position. During these conditions, the full insertion requirements for all other control rods, the one-rod-out interlock (LC0 3.9.2), control rod position indication (LC0 3.9.4), g and scram functions (LC0 3.3.1.1, " Reactor Protection System (RPS) Instrumentation," and LC0 3.9.5, " Control Rod OPERABILITY - Refueling"), or the added administrative controls in Item b.2 and Item c.2 of this Special Operations LCO, provide mitigation of potential reactivity excursions. ACTIONS A Note has been provided to modify the ACTIONS related to a single control rod withdrawal while in MODE 3. 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, 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 each requirement of the LC0 not met provide appropriate compensatory measures for separate requirements that are not (continued) HATCH UNIT 1 B 3.10-18 REVISION A i
i l UNIT 1 MARKUP OF CURRENT TECHNICAL l SPECIFICATIONS AND DISCUSSION OF CHANGES i i h i O i 2 i
?
i l O ,
e,,, ,,a
., , . ~ - - .. .-. .. - - - - -. . . -
r - x , N
. Insert 3.Or'(continued) l(~~V IM LCOL-3.O.4. When an LCOiis-not met, entry.into a MODE or other.
specified condition in.the Applicability shallinot. : sprs .be made except when the associated ACTIONS;to be .l WW. } entered permit continued operation in the MODE or. 1
-other specified condition in the Applicabilityffor
, an unlimited period ofLtime.' This' Specification ' shall not prevent changes.in' MODES,or other. . specified conditions-in the Applicability thatLare: 1 required;to comply with ACTIONS, or.that are part- : of a shutdown of the unit.- Exceptions to this' Specification are' stated in:the= individualiSpecifications. :These exceptions. allow: entry into MODES or'other specified~ conditions'in the Applicability when.the~ associated ACTIONS to j be entered allow unit operation in the MODE or 5
'other specified condition in the Applicability.
only for a limited period of-time. or th c ; LCO. -3.0.5 Equipmen removed from service or declared L, g inoperable to comply with ACTIONS may be returned ; to service nder administrative control solely to 1
. perform tes ng required to demonstrate-its A OPERABILITY, the OPERABILITY of other equipment,
-es=weedeinbes to be within ' limits. This is an g i exception to LCO-3.0.2=for the system: returned to- +
service under administrative control'to~ perform the. required testing. LCO- 3.0.6 When a supported system LCo is: not met solely; due R kM to:a-support. system LCO not being met,ithe: . R Conditions and Required Actions associated with j
.this supported' system'are not-required to be.
entered.. Only the support system LCO? ACTIONS"are. , required to be entered'.- This is an_' exception'to l LCO 3.0 . 2 for the supportedisystem'.: In_this H
. event, additional' evaluations and limitations;may" :)
be required in accordance:.with Specification. 'l 5.5.10, " Safety.' Function 1 Determination Program j (SFDP)." --If a loss of safety ~ function.is! l determined to exist byLthis program, the' . appropriate Conditions and Required Actions ofLthe LCO-in which the. loss of' safety function exists: are required.to be entered. i Hatch Unit'l Insert 3.0-1 .;
-i kh
~ - , , ,
1
,, DISCUSSION OF CHANGES x ) ITS: SECTION 3.1.4 - CONTROL R0D SCRAM TIMES TECHNICAL CHANGE - MORE RESTRICTIVE (continued)
M.3 An additional Frequency has been added (second Frequency to SR 3.1.4.1) to perform scram time tests on all control rods prior to exceeding 40% RTP after each reactor shutdown a 120 days. M.4 Two Surveillance Requirements have been added (SRs 3.1.4.3 and 3.1.4.4) requiring a scram time test after work on a control rod or CRD that could affect the scram time. SR 3.1.4.3 will require a scram time test, which may be done at any pressure, prior to declaring a control rod operable (and thus, enabling its withdrawal during a startup). SR 3.1.4.4 will require a scram time test after reactor pressure has reached 2 800 psig and prior to exceeding 40% RTP. To allow testing at less than normal operating pressures, a requirement for scram time limits at < 800 psig is included. These limits appear to k-bc less restrictive than the operating limits; however, due to lower reactor pressures not being available to assist the scram speed, the limits are reasonable for application as a test of operability at these conditions. Since these tests, and therefore any limits, are not applied in the existing Specification, this is an added restriction. (~~} V Furthermore, the existing scram time test requirement (performed at normal reactor operating pressure) is additionally required to be performed prior to exceeding 40% RTP. It is noted that if the control rod remains inoperable (which requires it to be inserted and disarmed) until normal operating pressures, a single scram time test'will satisfy both new Surveillance Requirements. TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 A " representative sample" of control rods is proposed to be tested each 120 days of power operation instead of the currently required "10% of the control rods" (current Surveillance 4.3.C.2.b). The proposed change adopts the BWR Standard Technical Specifications, NUREG 1433, position that these details be located within plant procedures and summarized in the Bases for the Surveillance. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process described in Chapter 5 of the Technical Specifications and changes to procedures will be controlled by the provisions of 10 CFR 50.59.
v HATCH UNIT 1 3 REVISION D
DISCUSSION OF CHANGES (X) v ITS: SECTION 3.3.1.2 - SRM INSTRUMENTATION ADMINISTRATIVE A.1 This description include all irradiated fuel assemblies since they must come from the spent fuel pool. Eliminating this discussion has no impact and is therefore considered administrative. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 The MODE 2 SRM requirements have been modified to require three SRMs instead of the current two SRMs. Additionally, the MODE 2 requirements are applied on Unit shutdown below Range 3 of the IRMs. These are - additional restrictions on plant operation and are consistent wilii the BWR Standard Technical Specifications, NUREG 1433. . M.2 Requirements have been added to ensure two SRMs are OPERABLE during MODE 3 and MODE 4. This ensures flux monitoring is available while shutdown. Appropriate ACTIONS (ACTION D) and Surveillance Requirements (SRs 3.3.1.2.3, 3.3.1.2.4, 3.3.1.2.6, and 3.3.1.2.7) have also been added. This is an additional restriction on plant operation and is consistent with the BWR Standard Technical Specifications, NUREG 1433. , N M3 Currently, no actions are provided, other than not to startup, refuel, or (d perform CORE ALTERATIONS, if the required SRMs are not OPERABLE. Proposed ACTIONS A, B, and C are provided to ensure proper actions are taken once a startup has commenced. These ACTIONS ensure the SRMs are restored to OPERABLE status, and if not, requires suspension of rod withdrawal and shut down of the unit. Proposed ACTION E is provided to ensure proper actions are taken during MODE 5 operations, including CORE ALTERATIONS. The proposed ACTIONS require suspension of CORE. ALTERATIONS, except for , rod insertion, and require control rods in core cells containing fuel l assemblies to be inserted. These are additional restrictions on plant ! operation and are consistent with the BWR Standard Technical i Specifications, NUREG 1433. M.4 Additional Surveillance Requirements have been added to ensure SRM OPERABILITY. Proposed SR 3.3.1.2.1 requires a CHANNEL CHECK to be performed every 12 hours during MODE 2. Proposed SR 3.3.1.2.2 requires the requirements of current Specification 3.10.C.1 (SRM location requirements) to be verified every 12 hours during CORE ALTERATIONS. Proposed SR 3.3.1.2.5 and SR ,.3.1.2.6 require a CHANNEL FUNCTIONAL TEST to be performed every 7 day. in MODE 5 and 31 days during MODE 2. : Proposed SR 3.3.1.2.7 requires a CHANNEL CALIBRATION to be performed every 18 months during MODE 2 and MODE 5. SR 3.3.1.2.4, SR 3.3.1.2.5, and A ! SR 3.3.1.2.6 have added requirements to verify and determine SRM signal- 3 L4Li to-noise ratio. These new Surveillances are additional restrictions on plant operation and are consistent with the BWR Standard Technical q Specifications, h0 REG 1433. 'L) HATCH UNIT 1 1 REVISION D
O g O O 2 3(' . O ).3.3 bhdQ A
= ~ +
3ddd'I e p p w N g e n alv -
$ INSTRUMENTATION WHICH PROVIDES SURVEILLANCE INFORMAT*0N Ref.
Required M-) j z Operable
] No. Instrument instrument ype and 1 (b) Channels Ranae Action Remarks
~ bnke Reactor vessel Wster Level 1g ecorder -150" to + 60"
}
y icator -150" to + 60" (c) (c) f (d) (d) 2 . cl Shroud Water level R order -317" to -17* tw @ (c) ((d) Q mgg 1,c., 2, I' Ind stor -317" to -17" (c) (d) l j 3' Reactor Pressure . 1 Reco er O to 1500 psig (c) il 2 indica r O to 1500 psig (c) () 44 Drywell Pressure 2 Recorde 10 to + 90 psig (c) (t i g
)D Jf Drywell Temperature Recorder to 500+F (c) (d Suppression er Air Temperature Recorder O t 500eF (d 7
N 9 -V Suppression Chember Water Temperature 2I Recorder O to 50cF (c) (d) Suppression Chamber Water Level 2 Indicator O to '-
- 3 %( b e (c) (d) 2 ( .I Recorder O to 3 * (c)(e) (d)
A SuporM Chamber PressureN Recorder -10 to 90 psig (d) h N_ Rod Posiheetjoformation SystemMREQ) W 28 Volt Indicating ights (d) 2 1 Hydrogen and Oxygen Analyrer Recorder O to 5%
^
% (d) go Post LOCA Radiation Monitoring System 1 Recorder (c) ' (d) 7
,C
- Irdicator 1 to 100 R, (c) (d) k$'
W . J ' J
~ e 1 pro.th.bos Me t sad e.
*$ g 13 a) Saf IRetief Valve Position Primary Q b
,M Indicat RV Indicating Light at 85 ps a (f) ! u* ' d Mec8ak 3.7. A .L . c, g b) Safety /Reli Va!ve Position Secondary
*j, 1 Recorder O to 600*F g (f) l&,,, di3 C indicator Q y JT3 %,9 , y
~
% s ro
~.
~g R-}
l DISCUSSION OF CHANGES - p) . (v ITS: SECTION 3.3.3.1 - POST-ACCIDENT MONITORING INSTRUMENTATION ADMINISTRATIVE (continued) A.5 Current Notes b, c and d to Table 4.2-11 have been deleted since these allowances are specified in proposed SR 3.0.1. A.6 The required number of Drywell High Range Radiation Function channels has been changed from two indicators and two recorders (an apparent total of : four channels) to two channels. The instrument design has an indicator, which is a remote indicating switch, that sends a signal to a recorder. The ITS is written such that the indicator is part of the associated recorder channel. That is, if the remote indicating switch is inoperable, such that the recorder does not receive a signal, the channel is , considered inoperable. This change is considered administrative, since each proposed channel consists of the indicator and its associated recorder. A.7 The current requirement to place the Unit in the cold shutdown condition (current Specification 3.7. A.8) if the H, and 0, analyzers are not restored to OPERABLE status has been deleted. The current Applicability for the analyzers is only " power operation" (e.g., > 1% rated thermal ~ power). Therefore, once power is reduced below 1%, the H, and 0, analyzers are no
/_ longer required, and the shutdown to cold shutdown does not have to be C} completed. Therefore, this deletion is considered administrative. (Note fi that the current Applicability is being revised to MODES I and 2; refer to La discussion M.I.)
RELOCATED SPECIFICATIONS R.1 The Suppression Air Temperature, Suppression Chamber Pressure, RFIS, Post LOCA Radiation Monitoring System, S/RV Position Indicators, Main Stack Post-Accident Effluent Monitor, and Reactor Building Vent Plenum Post-Accident Effluent Monitor are not credited as Category I or Type A variables. Further, the loss of these instruments is a non-significant risk contributor to core damage frequency and offsite release. Therefore, the requirements specified for these functions did not satisfy the NRC Policy Statement Technical Specification screening criteria as documented in the Application of Selection Criteria to the Hatch Unit 1 Technical ' Specifications and have been relocated to plant documents controlled in accordance with 10 CFR 50.59. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 Requirements for additional PAM Functions and channels are incorporated. These are included in accordance with NUREG 1433 guidelines to include all O Type A and Category 1 PAMs. Appropriate Action and Surveillance (/ Requirements are also added. HATCH UNIT 1 2 REVISION D ,
l DISCUSSION OF CHANGES
.i ITS: SECTION 3.3.3.1 . POST-ACCIDENT MONITORING INSTRUMENTATION TECHNICAL CHANGE - MORE RESTRICTIVE (continued)
M.2 The Applicability for the H, and 0, analyzers has been extended to encompass all of MODE 2, not just when > 1% rated thermal power (power operation). TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 Details of the system OPERABILITY requirements, description of the instruments, and methods to perform the Surveillances are relocated to the Bases, procedures, and the FSAR. 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 FSAR and procedures will be controlled by the provisions of 10 CFR 50.59.
LA.2 These details related to alternate monitoring methods are - more appropriately described (as necessary) in plant procedures (e.g., plant emergency procedures). Changes to the procedures will be controlled by the provisions of 10 CFR 50.59. LA.3 The use of alternate methods of monitoring is relocated to the Bases and procedures. The design features and system operation which dictate the methods are described in the FSAR. Additionally, 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. l
" Specific" L.1 The Frequencies for the CHANNEL CHECK and CHANNEL CALIBRATION (including recorders) are being changed to every 31 days and every 18 months, d
respectively. reliable, and they are providing indication only. These instruments (including recorders) are. highly I b. i No~ automatic actions A ; are performed by this instrumentation. The sensors and recorders are'also I la similar to others that are calibrated every 18 months. This Frequency is-also consistent with the BWR Standard Technical Specifications, NUREG 1433. HATCH UNIT 1 3 REVISION D
w DISCUSSION OF CHANGES (s ) ITS: SECTION 3.3.3.1 - POST-ACCIDENT MONITORING INSTRUMENTATION TECHNICAL CHANGE - LESS RESTRICTIVE (continued) L.2 The Required Action for one channel inoperable in one or more Functions for more than 30 days (current Table 3.2-11, Note c.1) is revised from requiring a shutdown to requiring a special report in accordance with the Administrative Control section of the Technical Specifications. The Required Action for one or more Drywell High Range Radiation channels , inoperable (current Table 3.2-11, Note g.1.a) is revised from requiring & channel to be restored within 7 days to requiring restoration within 30 days. The current requirement to submit a report if not restored is unchanged. Due to the passive function of this instrumentation and the operator's ability to respond to an accident utilizing alternate instraments and methods for monitoring, it is not appropriate to impose stringent shutdown requirements for out of service instrumentation. In some instances the existing A0T for these monitoring instruments is shorter than the A0T for the system which is needed to maintain the monitored parameter within limits. L.3 Post accident monitors (PAMs) are provided to assist in the diagnosis and preplanned actions required to mitigate design basis accidents which are assumed to occur in MODES 1 and 2. The probability of an event in MODE 3, p/ 4 or 5 that would require PAM instrumentation is sufficiently low that the q, PAMs are not required in these MODES. Therefore, the Action to be in MODE 4 if the post accident monitor is not restored to OPERABLE status within I D the appropriate time, has been deleted. L.4 Current Table 3.2-11, Note e.1 requires that if all suppression pool water level indication is lost, the indication must be restored within 6 hours or the reactor shall be in a Hot Shutdown condition (MODE 3) in the next 6 hours and in a cold shutdown in the following 18 hours. The requirement . related to indication of suppression pool level is covered by the suppression pool water level current Specification 3.7.A.I.a. Specification 3.7.A.1.a requires the water level to be within certain limits with the limits verified daily. Indication must be available to ; meet this verification. If all indications are inoperable such that suppression pool water level is not known, up to 24 hours (the maximum time before the next Surveillance could be due) may be allowed until Specification 3.7.A.I.a is declared not met. Once not met, Specification 3.7.A.8 requires the reactor to be in Hot Shutdown in 12 hours and in Cold Shutdown within the following 24 hours. Therefore, the deletion of n ms HATCH UNIT 1 4 REVISION D
- DISCUSSION OF CHANGES -( ITS: SECTION 3.3.3.1 - POST-ACCIDENT MONITORING INSTRUMENTATION TECHNICAL CHANGE - LESS RESTRICTIVE (continued)
L.4 (continued) current Table 3.2-11, Note e.1, effectively extends: 1) the time to restore indication by 18 hours; 2) the time to reach hot shutdown by an additional 6 hours; and 3) the time to reach cold shutdown by an additional 6 hours. Administratively, this change eliminates the situation of multiple Technical Specification ACTIONS for a single condition, with the corresponding potential for overlooking one or the other ACTIONS. f L.5 The current number of required channels for both the Reactor Vessel Water Level and Reactor Pressure is specified as 1 recorder and 2 indicators each. The proposed required number of channels for each Function is 2; consistent with the typical requirements for Post Accident Monitoring channels as recommended by Regulatory Guide 1.97. The Plant Hatch design is such that each reactor vessel water level and reactor pressure sensor is associated with one of two reactor vessel penetration " reference legs." Sensors from separate reference legs would satisfy the requirement for
/~
physical independence of the channels. The current requirement of three channels would necessarily involve one pair of channels that was not k (N ,/ physically independent. With the proposed revision, the required channels meet the minimum acceptable channels for compliance with Regulatory Guide l 1.97, and provide for channels that can meet the physical-independence j requirement. It is also noted that sever al additional channels are proposed to be added for appropriate compliance with Regulatory Guide 1.97 (see comment M.1). l l l k, , l a s HATCH UNIT 1 5 REVISION D
ll 1l1\ 1
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I l l DISCUSSION OF CHANGES j A)' ( ITS: SECTION 3.3.5.1 - ECCS INSTRUMENTATION - l ADMINISTRATIVE i (continued) A.8 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....") prov_ ides direction consistent with the intent of the existing Action for an inoperable ECCS instrumentation channel. Since this change only provides more explicit direction of the current interpretation of the existing specifications, this change is considered administrative. ' A.9 The ADS Instrumentation has been divided into two parts, Functions 4 and 5, with Function 4 being the ADS Trip System A and Function 5 being the ADS Trip System B. No technical changes are associated with this change. A.10 An equality sign has been added to the "less than" sign for the Allowable Values of the LCPI Pump Start Timers. The current Allowable Values in the Since this change setpoint calculations include the equality sign. results in an infinitesimally small and insignificant Allowable Value change, it is considered administrative in nature. p A.11 The Containment Spray System does not automatically actuate. 'This'LSFT Q requirement is in actuality related to the LPCI System, in that the containment spray and; cooling valves are required to close when a LOCA signal (Drywell Pressure - High or Reactor Vessel Water Level - Low, Low, Low, Level 1) is received. This requirement is embodied in the LSFT for l the LPCI System (which verifies all logic related to the LPCI subsystem functions properly). The Bases for the instrumentation specifically l identifies that containment spray and cooling valves receive a closed ; signal when a LOCA signal is received. Therefore, this change is l considered administrative. l i RELOCATED SPECIFICATIONS R.1 The HPCI turbine overspeed, exhaust pressure high, and pump suction pressure low functions, and the LPCI cross connect valve open annunciation , and valve selection timers are operational functions only and are not considered in any design basis accident or transient. The evaluation . summarized in the Hatch Unit 1 Application of Selection Criteria Report J determined the loss of these - Functions to be a non-significant risk j contributor to core damage frequency and offsite release. Therefore, the I requirements specified for this function did not satisfy the NRC Policy l Statement Technical Specification screening criteria as documented in the Application of Selection Criteria to the Hatch Unit .1 Technical Specifications and have been relocated to plant documents controlled in 1 accordance with 10 CFR 50.59. l {~ i HATCH UNIT 1 2 REVISION D l I
b DISCUSSION OF CHANGES
;f-s)
(, ITS: SECTION 3.3.5.1 - ECCS INSTRUMENTATION ~ i
. TECHNICAL CHANGES - MORE RESTRICTIVE -,
f M.1 .The allowance to place an inoperable channel in trip has been removed for some Functions. Placing a channel .in trip may not compensate for the > inoperability, or it may be a less safe action to take. Therefore, for these types of Functions, the channel must be restored; it is not allowed ; to be tripped. This applies to the following current Functions: .HPCI Functions 6 and 7; ADS Functions 3, 4, 5, and 6; LPCI Functions 3.a, 3.b, 6 and 7; and CS Functions 3 and 5. This is an additional restriction on > plant operation. 1 4 5 b HATCH UNIT 1 2A REVISION D'
q d
]' .
V' d '
. ~4
$ Notes for Table 4.2-1 (Cont'd) m-
- 2: .
t c
- 2: '
(IM) ' m
~4 No instrumentation is e ted from the instrument functio test definition. This instrument ' p funct I test will consist of inie ino a mir=datad electrical nianal inta ha measurement channels.)
- f. Siandard current so ce used which provides en inst nt channel alignment. Calibration using a ' '
3-radi 'on source shall be e once per operating cycle.
)-
rN SRL%L.I4 A* ' Logic system functional tests himulateNtomatic batiosi shall be performed once each operating cycle for the following: 7 to
- 1. Main Steam Une isolation Valves 8. Reactor Water Cleanup Isolation
- 2. Main Steam Line Drain Valves 9. Drywell isolation Valves '
- 3. Reactor Water Sample Velves 10. TIP Withdrawal e
4 RHR -Isolation Valve Control 11. Atmospheric Control Valves
- 5. Shutdown Cooling Valves 12. Sump Drain Valves
- 6. Deleted i3. Standby Gas Treatment SM 0,hu/ea sf C k g e4 W h ' 3 3 2.2 8
- 7. Drywell Equipment Sump Discharge to Redweste 14. Reactor Building Isoletion j, 4
# . .. ;; L~k.e. e -oretion of time deley sys and timers necessary c1 iogic system ru.m-- w N R P'N'*se) (M 3 for pr r functioning of the trip syst _
l
- 3 " A Eric '4 2 'es A'7 3.
o
- b P.
- (o ,
o Wh
- m. .
.D
_ _ _ . _ . - _ - _ . ~ -
DISCUSSION OF CHANGES () fm ITS: SECTION 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION ADMINISTRATIVE - A.1 Reformatting and renumbering requirements is in accordance with the BWR : Standard Technical Specifications, NUREG 1433. As a result, the Technical Specifications should be more readily readable, and therefore understandable by plant operators as well as other users. During this reformatting and renumbering process, no technical changes (either actual or interpretational) to the Technical Specifications were made unless they were identified and justified. During this process, the listing of the various tables has been deleted since it is found in the Table of Contents at the beginning of the document. Therefore, a new LC0 statement has been added, referencing the proper table. In addition, proposed Note 1 to the Surveillance Requirements has been added to identify the proper Surveillance Requirements for the PCIV Isolation Functions. I A.2 This action has been modified to place the Unit in MODE 2, instead of to close the MSIVs. This is essentially the same, since to close the MSIVs, the unit must be in MODE 2. Once in MODE 2, the Function is not required (as stated in the remarks section for the function), thus, the MSIVs are not required to be closed. Therefore, this change is considered administrative. A.3 An action to declare the affected system inoperable is an unnecessary v reminder that other Technical Specifications may be affected. This is essentially a " cross reference" between Technical Specifications that has been determined to be adequately provided through training. A.4 The format of the proposed Technical Specifications does not include a providing " cross-references". LC0 3.0.7 adequately prescribes the use of I E special operations LCOs without such references. Therefore, the existing references to Special Test Exceptions serve no functional purpose, and their removal is an administrative difference in presentation. A.5 The Frequency "once/ operating cycle" has been changed to "18 months". This change is administrative since 18 months is a normal operating cycle. A.6 Notes b.1 and c to Table 4.2-1 and Notes b, c, and d to Table 4.2-8 have , been deleted since these allowances are specified in proposed SR 3.0.1. 1 A.7 The HPCI and RCIC isolation valves have been added to the LSFT requirement. This change is considered administrative since they are ; currently required in Tables 4.2-2 and 4.2-3 (although the isolation instruments are not part of those Tables). 1 ? U HATCH UNIT 1 1 REVISION D
r '\ p DISCUSSION OF CHANGES l ITS: SECTION 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION l l TECHNICAL CHANGE - MORE RESTRICTIVE i (continued) M.3 A finite Completion Time of I hour has been provided to isolate the associated valves / systems / lines. Currently, no Completion Time is provided. This change is consistent with the BWR Standard Technical Specifications, NUREG 1433 and is considered more restrictive on plant operation. M.4 Isolation functions have been added (proposed Functions 1.f, 3.d, 3.e, 3.g, 4.d, 4.f, and 5.c). These Functions are added to ensure the safety analysis assumptions are met. Appropriate ACTIONS and Surveillance Requirements have also been added. This is an additional restriction on plant operation. M.5 The RHR Shutdown Cooling low water level isolation Function has been modified to require the isolation to be OPERABLE in MODES 4 and 5, since this is when it is needed to isolate if an inadvertent draindown event occurred. However, proposed footnote (d) to Table 3.3.6.1-1 only requires I one trip system to be OPERABLE (versus 2) if the RHR system is intact and integrity is maintained. This is consistent with the BWR Standard Technical Specifications, NUREG 1433, and is an additional restriction on
-[] plant operation.
%J M.6 The time provided to close the affected isolation valves has been decreased frcm 24 hours to I hour. In addition, the time to reach Hot Shutdown (MJDE 3) has been extended to 12 hours. However, since the overall time to reach MODE 3, assuming the valve associated with an inoperable instrument is not closed, has not increased Murrently it is 30 hours - 24 to close the valve and 6 to reach MODE 3, proposed is 13 hours
- 1 to close the valve and 12 to reach MODE 3) the change is considered more restrictive. These times are consistent with the time provided in the BWR Standard Technical Specifications, NUREG 1433 and is considered an additional restriction on plant operation.
M.7 An ACTION (ACTION G) is being added to provide the proper actions to take if the PCIV instrumentation is not restored within the 24 hour allowance of proposed ACTION B. The current ACTIONS for the Refueling Floor Monitor are only related to the secondary containment portion of the instrumentation function; no PCIV actions are provided. A portion of the Reactor Building Monitor ACTION requires closure of the PCIVs, which could result in a unit shutdown. The new Action requires a unit shutdown. Therefore, overall, this addition is more restrictive on plant operations. v HATCH UNIT 1 3 REVISION D
m ')ISCUSSION OF CHANGES ; (j ITS: SECTION 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION i TECHNICAL CHANGE - LESS RESTRICTIVE L.1 ) (continued) : simple set of actions that can be defined to cover all situations. The proposed Specifications have combined the ACTIONS for inoperable channels, independent of whether one or both trip systems are affected. This allows the conservative action of tripping the inoperable channels which is preferable to initiating a shutdown as is currently required in many cases. If all channels are not restored or tripped, then the ACTIONS referenced in the proposed Table are required, similar to the current TS. L.2 The time to close the MSIVs has been extended from 6 hours to 12 hours, and the time to reach MODE 4 (Cold Shutdown) has been extended from 24 to 36 hours. This provides the necessary time to close the MSIVs or shutdown the plant in a controlled and orderly manner that is within the capabilities of the unit, assuming the minimum required equipment is OPERABLE. This extra time reduces the potential for a unit upset that could challenge safety systems. This time is consistent with the BWR Standard Technical Specifications, NUREG 1433. L.3 The actions have been modified for when an RHR Shutdown Cooling (SDC) system reactor vessel low water level isolation channel is inoperable. (qj Currently, if the channel is not tripped within the appropriate time, the valves are required to be closed. This action however, will result in a loss of shutdown cooling, and could in fact, result in a more significant safety problem than if the valves were left open with inoperable channels. Therefore, the BWROG proposed new ACTIONS, and the NRC staff accepted these ACTIONS, as shown in the BWR Standard Technical Specifications, NUREG 1433. The new ACTIONS (proposed ACTION I) would require action to be immediately initiated to isolate the affected lines or to restore the channels . to OPERABLE status. The Bases describes circumstances under which each Required Action is to be taken. These new actions ensure that SDC is not interrupted when needed, yet also ensures action is continued to restore the channels if this is the case. L.4 The Required Action for when the RWCU low water level isolation Function is inoperable has been modified to allow the valves to be 1.solated in I hour, instead of requiring a unit shutdown. Isolation of the affected line returns the system to a status equivalent to the instrumentation " performing its function, thus continued operation should be allowed. L.5 The proposed Required Action if the Reactor Vessel Water Level-Low Low Low (Level 1) Function is inoperable is to allow isolation of the affected main steam line (currently a shutdown is required). Some conditions may affect the isolation logic for only one main steam line. In these cases, , it is not necessary to require a shutdown of the Unit; rather, isolation ! of the affected line returns the system to a status equivalent to the instrumentation performing its function, and continued operation is Q C allowed (although it may be at a reduced power level). The remainder of its isolation function is unaffected and still capable of performing its I function. HATCH UNIT 1 5 REVISION D
.oj DISCUSSION OF CHANGES ITS: SECTION 3.4.3 - SAFETY / RELIEF VALVES TECHNICAL CHANGE - LESS RESTRICTIVE (continued)
" Specific" L.1 An ACTION has been added (proposed ACTION A) allowing continued operation I for 14 days with one of the required S/RVs inoperable. This is justified i since the remaining S/RVs are capable of mitigating an event, assuming no 1 single failure. This change is consistent with Plant Hatch Unit 2, with i more recently licensed BWRs, and the BWR Standard Technical l Specifications, NUREG 1433. ;
l L.2 This requirement has been deleted. A failure of an S/RV is not ' significant enough to report, as shown by the lack of a specific reporting l requirement in 10 CFR 50.72 or 10 CFR 50.73 for this type of failure. l However, if the failure meets one of the reporting criteria deemed significant in 10CFR50.72 or 10CFR50.73 (e.g., the valve fails open and a shutdown is required), then 10CFR50.72 and 10CFR50.73 provide adequate ! reporting guidance. The two specific times in the current requirement ' coincide with the times in 10CFR50.72 and 10CFR50.73, respectively. L.3 The allowed lift setpoint tolerance has been increased from 1% to 3%. The i Q b vessel overpressure analysis uses this larger setpoint tolerance, as well as the transient analysis. In addition, when the setpoints are verified, j they are still required to be reset to 1% (proposed SR 3.4.3.1). Thus, I since the analyses still ensure that all limits are maintained even with the expanded tolerance, this change is considered acceptable. This change is also consistent with the BWR Standard Technical Specifications, NUREG , 1433. The following assement prov' as details of the analyses which were l performed to determine the accept .lity of tha change. j 1 DISCUSSION: The Edwin I. Hatch Nuclear Plant Unit 1 is designed with eleven SRVs and Edwin I. Hatch Nuclear Plant Unit 2 is also designed with eleven SRVs. Technical Specification 2.2A.1 (for Unit 1) and Technical Specification l 3.4.2.1 (for Unit 2) currently allow a setpoint error of 1% for each SRV. To reduce the number of forced outages and decrease maintenance and surveillance testing costs it is proposed to increase the SRV setpoint tolerance from 1% to 3%. Furthermore, when the setpoints are verified, they will still be required to be reset to i 1% (proposed SR , 3.4.3.1) to ensure the SRVs will not drift outside the proposed 3% l setpoint tolerance range. In addition, corresponding changes to the Bases I are proposed. v HATCH UNIT 1 3 REVISION D
l DISCUSSION OF CHANGES (m) ITS: SECTION 3.4.3 - SAFETY / RELIEF VALVES TECHNICAL CHANGE - MORE RESTRICTIVE (continued) EVALUATION: Georgia Power Company proposed a technical change to establish a SRV setpoint tolerance of 3%. To justify the change, licensing basis calculations were performed to show that with the proposed setpoint tolerance modifications, vessel overpressurization limits and Loss-of-Coolant Accident / Emergency Core Cooling System (LOCA/ECCS) performance requirements are satisfied. The calculations also show that the proposed change does not have a significant impact on thermal limits, Low-Low Set operation and containment performance. The following provide the details of the impact of the proposed change: Affects of Proposed Chance on Reactor Vessel and ECCS The proposed SRV setpoints (including the 3% tolerance) are below the reactor vessel design pressure of 1250 psig, satisfying the requirements of Article 9 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code - Section III, Nuclear Vessels. Evaluations'of 3 the overpressure transient (Main Steam Isolation Valve Closure with Flux [V Scram event) with SRV setpoints above the nominal setpoints plus the 3% tolerance and assuming one SRV inoperable demonstrate the capability of the remaining SRVs to maintain the reactor pressure vessel well below the ASME Code limit of 110% of the vessel design pressure (110% x 1250 psig - 1375 psig). The proposed setpoints (including the 3% tolerance) are low enough to ensure High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) rated flow is still achievable. The proposed setpoint tolerance change may result in a reduction in turbine overspeed margins during a HPCI/RCIC start-up transients. However, since HPCI/RCIC are not expected to actuate until reactor vessel pressure is within the normal HPCI/RCIC operating range, the reduction in the turbine overspeed margin does not have a significant impact on HPCI/RCIC performance. Also, since the proposed setpoints (including the 3% tolerance) are within the range of current setpoints, the overall likelihood of inadvertent valve opening (from downward setpoint drift) is not expected to change significantly. The impact of the SRV setpoint tolerance change on the ECCS/LOCA (SAFER /GESTR) analysis was reviewed. Large line breaks (including the design basis accident recirculation line break and the main steam line break), are not affected at all by the change since the vessel depressurizes so quickly that the SRVs never actuate. Small line breaks may experience SRV actuations following Group 1 isolation, but the impact
, of the higher SRV setpoints is a small increase in break flow and an insignificant change in peak clad temperature (PCT). The small line breaks are not limiting for Plant Hatch. Therefore, the proposed SRV A setpoint tolerance change does not have a significant impact on reactor
() vessel or ECCS operation. r l HATCH UNIT 1 3A REVISION D l 5
m DISCUSSION OF CHANGES ITS: SECTION 3.4.3 - SAFETY / RELIEF VALVES ~(u) TECHNICAL CHANGE - LESS RESTRICTIVE (continued) Affects of Prooosed Chanae on Thermal Limits The impact of the proposed SRV setpoint tolerance change was evaluated for the limiting thermal transient event. In this event, Load Rejection Without Bypass, peak vessel pressure occurs 1 to 2 seconds after the peak heat flux which determines the time of Minimum Critical Power Ratio (MCPR). iherefore, potential increases in SRV opening pressure have no impact on calculated fuel thermal limits. A decrease in the SRV opening would cause earlier SRV pressure actuation(for the same transient event.due toAnSRV ear downward setpoint drift) lier actuation redu rate of vessel pressurization and, therefore, the rate of void collapse. If SRV actuation occurs at or before the time of MCPR, the decreased rate of pressurization and void collapse will produce lower peak neutron and surface heat fluxes and, therefore, a smaller delta CPR. It should be noted that the reload transient analyses performed each cycle for Units 1 and 2 assumes a +/-3% SRV setpoint tolerance, and is therefore consistent with this proposed change. The results of the transient analyses are reported in the Core Operating Limits Report (COLR). ( Affects of Proposed Chanae on Low-low Set (LLS) c) The increased SRV opening pressure (due to the setpoint tolerance increase) will only affect the timing of the first SRV actuation. Once the logic is initiated, the opening and closing setpoints of preselected SRVs are automatically reset to lower values by the LLS logic. This logic is unaffected by the setpoint tolerance change since the logic acts on the relief mode of SRV actuation and not on the safety mode of operation. Affects of Proposed Chanae on Containment Structures The Plant Unique Analysis Reports (PUARs) for Hatch Units 1 and 2 were reviewed to determine the impacts of the proposed SRV setpoint tolerance change on the containment. SRV discharge loads have the potential to be affected by an increase in the SRV setpoint tolerance due to an increase in SRV flow rates. The evaluation performed for this change assessed the impact of the increased SRV setpoint tolerance on the actual load for the limiting load combination on a structure-by-structure basis as performed in the PUARs. The PUARs calculated the effects on the torus shell, torus support structures and torus attached piping assuming all SRVs actuate simultaneously. Each of the structures analyzed in the PUARs was reviewed to determine the impact of the SRV setpoint tolerance increase. Based on this review, it was determined that conservatism in the torus shell rm pressures used as input to the PVAR structural analyses would offset the ( increase in SRV loads with the increased SRV setpoint tolerance.
') Therefore, the resulting loads will not cause the stresses in these components to exceed allowable values.
HATCH UNIT 1 3B REVISION D
n DISCUSSION OF CHANGES o $ ITS: SECTION 3.4.3 - SAFETY / RELIEF VALVES V. TECHNICAL CHANGE - MORE RESTRICTIVE (continued) Thrust loads for SRV piping and T-quenchers were determined using the relief valve forced outage rate (RVFOR) computer. model. This computer model has been shown to overpredict water-clearing loads incident 'on submerged SRV piping and on T-quenchers and supports by 40% to 50%. However, the results of the Hatch Units I and 2 PUARS demonstrated adequate margins to the allowable stresses for these components to allow an increase in the SRV setpoint tolerances without exceeding the allowable stresses. Containment structures which could be affected by water jet or air bubble drag loads include such submerged structures as the vent header assembly, vent system supports, downcomer ties, vent line bellows, vent header deflectors and vent system penetrations. The results of the PUAR analyses showed that the limiting structures have large safety . mar ins. Considering these large safety margins, increasing the SRV set oint-tolerance will not result in submerged structure loads which excee the allowable loads. Based on these evaluations, the adequacy of the containment ' structures with the proposed SRV setpoint tolerance change has been ' demonstrated since allowable stresses will not be exceeded. Therefore, the proposed SRV setpoint tolerance change has no significant impact on containment - structures. A review of the , impact of the increased SRV setpoint tolerance on containment response was also performed. The most limiting drywell. pressure transient is the design basis LOCA, and the most limiting drywell temperature transient is the Main Steam ~Line Break event. Neither of these transients is affected by the increased SRV setpoint tolerance. For smaller steam line breaks, that require SRV actuations the resultant drywell temperatures are well below the limiting steam l'ine break. The peak drywell temperature occurs late in the event following many SRV actuations and is governed by the total energy released to the drywell. Since the SRVs will return to nominal 'setpoints following the first actuation, an increase in the SRV opening pressure will only affect the very beginning of the event-and will have a negligible impact on the total energy released to the drywell. Therefore, an increase in the SRV opening pressure (due to the proposed setpoint tolerance increase) will also have an insignificant impact of peak drywell temperature for the non-limiting drywell temperature events. Therefore, the proposed SRV setpoint tolerance change has no significant impact on the containment response. CONCLUSION: Based on the above evaluation, it is concluded that there is no significant safety impact on vessel overpressure margin, ECCS/LOCA
. performance, thermal limits, Low-Low Set operation, or containment structures due to operation with SRV setpoint tolerance' of 3%.
(- -L Therefore, these proposed. changes, including the corresponding changes to the Technical Specification Bases, are acceptable. HATCH UNIT 1 3C REVISION D
.6pJ,4r_n 4 0.) 7.f I
- LIMITING CONDITIONS FOR OPERATION SURVEILLARLL KLUULutRENT5 ]
* - 3.5.8.2. Doeration with inocerable 4.5.8.2. (Deleted)
Comoonents (Centinueo) S
- b. If one LPCI subsystem is inoper-able, the reactor may remain in
% 4
' rTC[/dAI b operation for a period not to i 1
exceed 7 da ovtoec t t - T ctive como nts of . l bS ' l l rema n P vstem. t CS 1 i t mv s>>vu o - A es (per 5 gene ifica 5 are o n 4.9.A. able
).
Ag lg
- c. When performing an inservice 3 hydrostatic or leakage test 3t g DJ3c g3, o p d,aIc, wita the reactor coolant )
temperature above or below k ITS*3*jo*j 212*F, comply with / e Specification 3.5.B.1.b. p3,q ggga N3d4JEk Tw 3,4,* r3 Oeruhuu, f Y Jr W . ] 3 + 1d .
't O..i V Amendment No. II. II, 747, HATCH - UNIT 1 3.5-4 1W,170 4dl[
- u. ... . -- .. . . . . . .. _ .. _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - _ _ _ _ - _ - - _ _ _ - _
i i i
, DISCUSSION OF CHANGES l ITS: SECTION 3.5.1 - ECCS - OPERATING l%w)t l l
TECHNICAL CHANGE - MORE RESTRICTIVE (continued) H.6 An explicit frequency (31 days) has been added to this Surveillance. Currently, no Frequency is listed, thus this change is more restrictive on plant operations. M.7 Explicit acceptance criteria have been added to describe this condition. The acceptance criteria 2 570 volts and s 606 volts is located in proposed SR 3.5.1.5. Since more explicit acceptance criteria is provided, this change is more restrictive on plant operations. TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 The details relating to system design and purpose have been relocated to A the Bases and/or plant controlled documents. The design features and I /DN system operation are also described in the FSAR. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process described in Chapter 5 of the Technical Specifications, and changes to b-plant controlled documents and the FSAR will be controlled by the (3 provisions of 10 CFR 50.59.
LA.2 This Surveillance Requirement has been relocated to plant procedures since the requirement is not included in the proposed Standard Technical Specifications, NUREG 1433. The system will continue to be required to perform its required safety function to be considered OPERABLE. Proposed SR 3.5.1.3 is added (refer to M.4) to address the important characteristic of whether there is sufficient air pressure available to permit . the actuation of the ADS valves should an accident occur. The surveillance being relocated will continue to be performed and will identify degradation of the ADS air system pressure retention capabilities. LA.3 The details relating to methods of performing surveillance test requirements have been 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 and changes to procedures will be controlled by the provisions of 10 CFR 50.59. LA.4 Any time the OPERABILITY of a system or component has been affected by repair, maintenance or replacement of a component, post maintenance testing is required to demonstrate OPERABILITY of the system or component. Explicit post maintenance Surveillance Requirements have, therefore, been deleted from the Specifications. Also, proposed SR 3.10.1 and SR 3.0.4 require Surveillances to be current prior to declaring components operable. O V HATCH UNIT 1 5 REVISION D
S PCRa% M t. 2-I I I __ tlMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0'JIREMENTS '
,~
4.7.,A.2.e. Tyne B Tests - Lesk Tests of Penh (V) e trations with Seals and Bel M l UCPo.R d L Cc M. I. "L (Continued) (Tables 3.7-2 at. 3.7-3) g,g pf,p,gg ggf,;4 jg ge L g ;,, J1) Primary containment components which seal or penetrate the H U+=p 4 D/* fcW % 2.-A % \ pressure containing boundary of the containment shall be f
/
I% 3.r..t.t'q
~M % c, q tested at a pressure not less b *" P Thes c on nt
- ' *f0ME /MT4,45 Sub. ' ,j ,
refueling shutdown or at
~
intervals not to exceed 2 o */*wdo4 g M y Jg EGFbed AtTbd C-(2) (a) The primary containneb,,-+,o,,r#m .; 3,4 j ,1 j airlock shall be tested at g g,g.) X-sontninte Is at r, by O p urizing t onpart- A.) ment ween the air a lock d The leakage et . shall not exceed 0.05 L.. APfl ic ;l;L &
' T/___.
~
b) If the prima containment airlock is open during eriods when prima contain-t integrity is no req d, the test requ ed by 4. .2.e.(2)(a)shall be perfo d at the end of such pertok. [ If the primary tainment) ('")' irlock is opened ring p iods when primary ontain-men niegrity is requ , it sha be tested withi daysof(Deinoopenedby 7 ressurtzing the gap Detween the door seals to 210 psig 4 g' for at least 15 minutes. The leakage for each set of door seals shall not exceed 0.01 Le d) If primary ntainment is required and e primary containment air k is being pened more frequ tly than ob e every 3 days, e test re ired by 4.7.A.2.e. 2)(c) shal be performed at 1 st once p 3 days during th period o frequent openings. ll Type B and Type C Le age Tests (i.e., Local L ak Rate Tests) that Y 11 '
'y(te.,testleakageissuc that an LER would be reg red) during an outa sha 1 de reported according 10 CFR 50.73 by one, 3 ay written report that s due within 30 days of he first leakage test fat re in the outage.
All et er leakage tesi failures iscovered during the outa will be reported in a re sion to the original rep t due within 30 days afte the end of the J foutage. ,m ( ) HATCH - UNIT 1 3.7-6 Amendment No. 9, 4 M , 449, 164 %) D
r3 DISCUSSION OF CHANGES (j ITS: SECTION 3.6.1.2 - PRIMARY CONTAINMENT AIR LOCK ADMINISTRATIVE I The existing Technical Specifi::ations contain the details for air lock A.1 leakage surveillances which are also found in 10 CFR 50 Appendix J. These regulations require licensee compliance, cannot be revised by the licensee, and are addressed by direct reference in the Technical Specifications. Therefore, these details of the regulations within the Technical Specifications are unnecessary. Furthermore, approved exemptions to the regulations, and exceptions presented within the regulations themselves, are also details which are adequately presented without repeating the details within the Technical Specifications. The only requirements that are necessary to be retained in the TS are the the door leakage rate (0.01L and test overall pressure leakage rate (0.05L'c)luded
/ time. These are in as SR 3.6.1.2.1.a and b. ,),
Therefore, retaining the requirement to meet the requirements of 10 CFR 50 Appendix J, as modified by approved exemptions, and eliminating the ; r Technical Specification details that are also found in Appendix J, is ! considered a presentation preference, which is administrative. I Clarifying Notes are proposed. The Notes, and a new Required Action C.1 (discussed in comment L.3), facilitate use and understanding of the intent of: V 1) (For SR 3.6.1.2.1 Note 1) the overall air lock acceptance criteria when one air lock door is inoperable. Since the inoperability is known to be only affecting one door, the barrel and the other OPERABLE door are providing a sufficient containment barrier. Even though the overall test could not be satisfied (SR 3.0.1 would normally require this to result in declaring the LCO not met - possibly requiring proposed Condition C to be entered), the note clarifies the intent that the previous test not be considered "not met." l
- 2) (For SR 3.6.1.2.1 Note 2) ensuring that the primary containment overall leakage is evaluated against the Appendix J acceptance criteria after testing the airlock for leakage. >
- 3) (For ACTIONS Note 2) considering the primary containment inoperable in the event air lock leakage results in Appendix J acceptance criteria being not met.
These clarifications are consistent with the intent and interpretation of the existing Technical Specifications, and are therefore considered administrative presentation preferences. A.2 This allowance provides an NRC approved interpretation for when a 10 CFR 50.73 written report is required following LLRT failures. This ' interpretation is discussed in the NRC SER for Unit 1 Amendment 149 and Unit 2 Amendment 86. This information is not directly related to a A Technical Specification requirement (i.e., TS does not require a written () report; 10 CFR 50.73 governs the report). This information is more HATCH UNIT 1 1 REVISION / D
N 5pec;1Uca 3.L.\ 3 _ l LIMITING CONDITIONS FOR OPERATION O SURVEILLANCE REOUIREMENTS 3.7.C.3. Violation of Secondarv Containment Intecrity 4.7.C.3. Surveillance After Intearity Violated 3 i
- a. Without Hatch-Unit 1 After a secondary containment viola-secondary containment tion is determined the standby gas integrity, restore treatment system will be operated Hatch-Unit I secondary innedtately after the affected zones 'i containment integrity are isolated from the remainder of within 4 hours, or' the secondary containment. The he hyg . perform the following (as ability to maintain the remainder applicable): of the secondary containment at 4be y' 1/4 inch of water vacuum pressure ,
fe m (1) Suspend irradiated fuel under calm (< 5 mph) wind conditions and/or fuel cask handling shall be confirmed. 3 +43, in the Hatch-Unit ! 5%g secondary containment.
%Nd
( w 4t 4 (2) Be in at least Hot Shutdown within the next 12 hours and
\ 5achod' meet the conditions of 3.7.C.I.a. within the next 24 hours.
- b. Without Hatch-Unit I secondary containment, refer to the follow-ing Hatch-Unit 2 Technical Specification, for LCOs to be followed for Hatch-Unit 2:
O (1) Section 3.6.5.1, O (2) Section 3.9.5.1. D. Primary Containment Isolation Valves D. Primary Containment isolation Valves -
- 1. Valves Reouired to be Ooerable .. Surveillance of Doerable Valves ;
NO3 S'g,3 al ing(feaRor pow r uu-Um d / rimary containment isoTation Surveillance of the primary con-tainment isolation valves shall be l sves and all reactor coolant performed as follows: system instrument line excess flow __ check valves shall be operable except a. At least Ance per operatin A'I ' as stated in Specification 3.7.D.2. Cycluthe operaDie isolation - valves that are power operated and automatically initiated ' h.S q jW ' M3 l.417 shall be tested automatic forrstmulated initiation and the I (,, 3g3.u.3.g closure times. l -(] HATCH - 1 3.7-13 Amendment No. 40, H, H. MO, H9, ISE u lof 13 -; L
i l Owp oa 3.i.i2 l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS k./'. 4.7.D.1. Surveillance of Operable Valves (Continued)
- b. At lame + GCEr~T6ng'-
cum 13.4he reactor cooia5T g 34 N 3 system instrument line excess A l /.El flow check valves shall be tested <dtr proper um .oon,,
. At least once pe uarter, I normally open p r-ope ed isolation va s (excep r the main ste line power- ated isolat 4I
, s) shall ully close ned, c nd a
- d. The isolation time of each a 5 N 3*g.),3.( main steam line isolation valve 1 shall be determined to be within its limit when tested, pursuant to Specification 4.6.K.
~_
Pro Poxd Au s yg g4dIWS At least once r week [ A,3 n steam line r-oper-
- ate lation valv shall Y#9*kE NOE 2.-lu Acpgs 4 ," be exere one at a ' b.I by partial e e and su Ls ent reopenino, yacogd 4 I-lo hQohs L,
- 2. Surveillance of Lines with an
/]y
(", 3.7.D.2. Doeration with Inocerable Valves Inonerable Valve W never an isolation valve m in w ie J.e-ots L. 4 In the event 6nv inninH nn valve . g*g inoperable the position of at Qpes.ified TAJable17-Ubecomes least one other isolation valve NM 1Mp1! Table, reactor power operation in each line having an inoperable It')Qred 4kg byC may continue provided at least one 6% isolation valve shall be verified Cliolation valvey n eacn inne naving dI.. / to be in its isolated position 4.y c,g l 1) an inoperable valve is in the mode a . t corresponding to the isolated con- a
*YPf*tasa %W T,' aab'Ly*ceh Mws h.\ 5 c.s L.
- 3. Shutdown Recuirements If Specification 3.7.D.1. and 3.7.D.2.
cannot be met,J an orderly shutdown Tnail be initiated and the reactor shall be placed in the Cold Shut- M' 7NW E b i down Condition within 24 hours. ; SE 5 M . t.7. 2, 3-L.I.L 3 3.t.i.r.y#g
%"f-a AGo@ 3.6.i.3./ gg.
p x & Prenod
?%*A he.W fm ,
( ) 3.7-14 Amendment No. 153
\_/- HATCH - UNIT 1 i
2A l3
-;.%. m,
; ~. . TABLE 3.7-
N. PRIMARY CONTAINMENT I LATIOll' VALVES WillCII
- r RECEIVE A PRIMARY CON INMENT SOLATION SIGNAL
[ $ A $ [n *l' b N> {hbD*= d Isolation- Number. F Power Maximum i ItparKT Action on - Ope ra tet VaIves Operating ] 06sition Inltlatle
-4 Croup.
a (b) Valve identlFicationj d1 Inside \0utside Time (seci l I fel Slonal J l I Hain steam-line II ,35T
-f5 0 cc (D21-F022 A,0,C,D, U21 028-A,D,C,D) 1 Main steam line tira 1 1 20' C SC l (D21-F016, B21-F0 ) '
Reactor water ample line 1 1 5 CC 1-(D31-F019, D -F020) 2 5 C SC 2 8 '8 Drywe l l irge inlet (ii8-F i ,.Tf8-F308)~ 2 C SC 28- Dr il main exhaust ( 8-F319, Tf8-F320) Drywell exhaust valve bypass to 2 5 C SC 2
- w standby gas treatment
* ( TriB-F3fi l, Tl8-F3fO) i i Drywell ni trogon make-up lino C SC b 1 5 Ch ,
(storma l ope ra tion) -
.(Ti8-F118A) l Suppression chamber purge inlet 2 5 C SC :
2 (T8 8-F309 Tf 8-F3284) 3 Suppression' chamber ma in exh st 2 5 SC 2 8 '8
'(Tia-F318, t i Tf8-F326) k a ,u.
g o 3, t+ N I
? ~ f s-h D, w
w . m-N v* Z . 2, = G ?* I
. . . _ .___ . . , _ - . . . . . . . . . . _ _ . . . . - . , . . . -.._ ..._ . , .. , _ . . . , . . _ , . . - _ . . , . . . . _ _ . . .__ ._.....--,_a.._____.______.._..__._.___.,___ _ _a
'iQ h tA 3. f .I. 3 .s LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS ! \
(/ 3.7.A.7. Primary Containment 4.7 A.7. Primary Containment Purce System Purae System
- a. ! When primary containment is a. In addition to the requirements required, all drywell and of Specification 4.7.0, each A.' suppression chamber 18 inch purge drywell and suppression chamber supply and exhaust isolation 18 inch purge supply and M hM'y* j A Lyalvet chall be coerable _ n in the fully closed position except w L* exhaust isolation valve shall be verified to be closed at g 3.(,g.),) when required for inerting, de- proposed least monthly, inerting, or pressure contro 44pg w b. Each refueling outage each EachdrywR1andsuppre ion ia ug g drywell and suppression chamber M4.l 3. f hamber 18 fRch purge su ly an g. 18 inch purge supply and g e aust isola n valve sh I hav p,,.y j exhaust isolation valve with a a1 tage rate a specified resilient material seat shall
' 4.7.A 7. .
be demonstrated operable by having its valve seat replaced k i witNn b 6 .b isolation dampers shall be . operable at all times when the/ p") c. At least once perW_ the Lk. U dampers will be Wisua i n vA--- bnit 1 primary teority containmW ILnquir nT the 18 Qnspected angAycledfo verifh bd *IE' b h isolatio to the tne Usmpera nave na damage ink dr 1 or suppression c ber are which renders them incapable of onan- __ perfoming their design functinn- - If these requirements cannot be met, close the drywell and suppression chamber 18 inch purge
,/ 3 Acyld supply and exhaust isolation (v) g valve (s) orjiithe_rwise itnlate th,er-- M S enetrat ns within 4 hours or yg i
u the requirements of Specification 3.7.A.8.
- 8. Shutdown Reauirements If Specification 3.7.A. cannot be met, an orderly shutdown shall be initiated and V
$0 g
the reactor shall be brought to Hot Shutdown within 12 hours and shall be in I the Cold Shutdown condition within the ' following 24 hours. l I r' (,) HATCH - UNIT 1 3.7-10a Amendment No. M, M, 99, MO,118 5 of13 . 2
Jfeci n dao $.t. 1 9 O g LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS A~' % Oi$c y,$n,3 of (e) At least once per 6 months it shall be verified that C(9 er-T'T) f 5 4 l.2, only one door in the air-can be opened at a
),; MM Nr -
t.a', a w - ff. Tvoe C Tests-tocal teak Tests of Containment Isolation Valves (Tables 3.7-1 and 3.7-4) Type C tests shall be perfomed f under the program established in Appendix J of 10 CFR Part 50. f g g 3,,,, of Containment isolation valves
>ft (except for main steam line iso-C(% grg lation valves) shall be tested d ,
at a pressure not less than P . M.\.1 P/e% Type C tests shall be performed at each major refueling shutdown
%g, y, g y or at intervals not to exceed IGchoi. 2 years.*
- g. Acceetance Criteria for Tvoe B and Tvoe C Tests The ccabined leakage rate of components subject to Type B and C tests shall be determined under the program established in Appendix fs 1 J of 10 CFR Part 50
( ) ( and shall not exceed 0.6 L..* _
- h. Main Steam Line Isolation Valves
{nP 3,g.3, g The main steam line isola- ik I Lg\. tion valves shall be tested at a pressure of 28 psic for leakage least, on .
,e no cycle a total Teak rate of 11.5 sef per hour for any one main steam line isolation valve is exceeded, repairs and retest shall be perfomed to correct this cc,ndition.
[ Mall Type 8 and Type C Leakage Tests (i.e., local Leak Rate Tests) that fai (i.e., test leakage is such that an LER culd be required) during an outage shall be reported according to 10 CFR 50.73 by one, 30-day written report that is due within 30 days of the first leakage test failure in the outage. All other leakage test failures discovered during the outage will be reported in a revision to the original report due within 30 days after the end of the octage. HATCH - UNIT I 3.7-6a Amendment No. 9, 4 M , 449, 449, (ql v 164 13413
n DISCUSSION OF CHANGES Q ITS: SECTION 3.6.1.3 - PRIMARY CONTAINMENT ISOLATION VALVES ADMINISTRATIVE A.1 The Frequency of "once per operating cycle" has been changed to "In accordance with the Inservice Testing Program" for proposed SR 3.6.1.3.5 (stroke time tests). Since the current IST program requires testing every 18 months (a normal operating cycle), this change is considered administrative. The Frequency of "once per operating cycle" has been changed to "18 months" for proposed SRs 3.6.1.3.5 (PCIV actuation test) I and 3.6.1.3.7 (EFCV tests). Since 18 months is a normal operating cycle, this change is considered administrative. A.2 This proposed Note (" 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 Note provides direction consistent with the intent of the existing Actions for inoperable isolation valves. A.3 The proposed ACTIONS include Notes 3 and 4. These Notes facilitate the use and understanding of the intent. Any system made inoperable by inoperable PCIVs is to have its ACTIONS also apply. Note 4 clarifies that these " systems" include the primary containment. With the proposed LC0 p 3.0.6, this intent would not necessarily apply. The clarification is (J consistent with the intent and interpretation of the existing Technical Specifications, and is therefore considered administrative. A.4 The current single Action for "any isolation valve" has been divided into three ACTIONS: proposed ACTION A for one valve inoperable in a penetration that has two valves; proposed ACTION B for two valves inoperable in a penetration that has two valves; and proposed ACTION C for one valve inoperable in a penetration that has only one valve. All technical changes are discussed elsewhere in this section. As such, this change is-considered administrative. A.5 The current Technical Specifications repeat most of the requirements, provisions and actions for purge valves and excess flow dampers .in a Specification separate from all other primary containment isolation valves. The propcsed Technical Specifications incorporate these requirements and associated restoration times into the primary containment isolation valve Specification. This is a presentation preference, except - as noted by other comments. A.6 The format of the proposed Technical Specifications does not include providing " cross-references". The existing reference to current Specification 4.7.A.2 serves no functional purpose since Specification 4.7.A.2 is applicable regardless of whether it is called out in this Specification. Therefore, its removal here is considered administrative. .p HATCH UNIT 1 1 REVISION D
. -~ m DISCUSSION OF CHANGES lv) ITS: SECTION 3.6.1.3 - PRIMARY CONTAINMENT ISOLATION VALVES ADMINISTRATIVE (continued) A.7 This Surveillance has been deleted since it is redundant to current Specification 4.7. A.2.f (proposed SR 3.6.1.1.1). Since the requirement is still maintained, this change is considered administrative in nature. A.8 The Frequency "at least once per operating cycle" has been replaced with "In accordance with 10 CFR 50, Appendix J, as modified by approved exemptions." The Appendix J requirements are essentially consistent with Technical Specification requirements. Therefore, this change is an administrative preference in presentation. A.9 Proposed LCO 3.6.1.3 applies to each PCIV, except reactor building-to-suppression chamber vacuum breakers. LC0 3.6.1.8 covers these vacuum k breakers and thus, they do not need to be considered in this LCO. Since the requirement is still maintair.ed, this change is considered administrative. A.10 The lim 5ts for main steam isolation valve stroke times are proposed to include " equal to" 3 seconds and 5 seconds. Currently the limits are stated as greater than 3 seconds and less than 5 seconds. This is an
/N inconsequential change that is considered administrative. Including the b equivalency is consistent with the safety analysis assumptions for MSIV D ,
closure times. The difference is realistically inconsequential since the j increase in the range of acceptable times is infinitesimal. This will ! result in cesistency with the Unit 2 requirements that are already stated l with the equivalency included. IECHNICAL CHANGE - MORE RESTRICTIVE M.1 The Applicability has been changed to include MODES 2 and 3 (in addition to the current MODE 1) as well as when associated instrumentation is required to be OPERABLE per LC0 3.3.6.1 (which adds a MODE 4 and 5 requirement to the RHR Shutdown Cooling System isolation valves). An appropriate ACTION is added (proposed ACTION F) for MODE 4 and 5 operation when the valves cannot be isolated (since the unit is already in MODE 4 or 5, the current actions provide no appropriate compensatory measures). M.2 In addition to checking the EFCVs for proper operation, proposed SR 3.6.1.3.8 checks to ensure that flow is restricted to within limits specified in test procedures. This is an additional restriction on plant operation. HATCH UNIT 1 2 REVISION D
~
" .i;
)
DISCUSSION OF CHANGES
;', -) ITS: SECTION 3.6.1.3 - PRIMARY CONTAINMENT ISOLATION VALVES (j
TECHNICAL CHANGE - MORE RESTRICTIVE (Continued) , M.3 The current Applicability of the isolation valves is "during reactor power operation", which is effectively MODE 1. Thus, if a valve is inoperable, once reactor power operation is exited, current TS do not. require the ! valves to be OPERABLE, thus a shutdown to cold shutdown is not: required. Power operation only must be exited, and up to 24 hours is allowed to do j this. With the change in Applicability described in comment M.1, the unit is now required to be placed in MODE 3 within 12 hours and MODE 4 within 36 hours. As such, this change is more restrictive on plant operation. . M.4 An ACTION has been added (ACTION D) providing actions to be taken for one , or more penetration flow paths with leakage not within limits. This is a - new requirement and as such, is an additional restriction on plant operation. t l i i HATCH UNIT 1 2A REVISION D w y -#- .
h DISCUSSION OF CHANGES l 7]' t ITS: SECTION 3.6.1.3 - PRIMARY CONTAINMENT ISOLATION VALVES > TECHNICAL CHANGE - MORE RESTRICTIVE : (continued) M.5 New Surveillance Requirements.have been added. SRs 3.6.1.3.2, 3.6.1.3.3 and 3.6.1.3.4 ensure PCIVs are in their proper position or state. SR 3.6.1.3.9 enrures the TIP squib valves will actuate if required. These- I S'_ SRs are additional restrictions on plant operation. M.6 This allowance, to only require the excess flow dampers OPERABLE if the 18 i inch purge valves are open, has been deleted. The dampers are now < required at all times in MODE 1, 2 and 3. This is an additional restriction on plant operation. M.7 The Frequency has been changed to 18 months to be consistent with the normal operating cycle length. This is an additional restriction on plant . operations. , M.8 An explicit listing of the types of isolation valves acceptable for use is added. Check valve isolation is not provided in proposed Action C. This is an additional restriction on plant operation. k: i TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 These valve exercising surveillances have been deleted since they are "r duplicative of inservice testing (IST) requirements.- If valves need to be cycled, the IST program covers this. The valves are still cycled periodically per Technical Specifications during proposed SRs 3.6.1.3.5, 3.6.1.3.6, and 3.5.1.3.7.
g, ; LA.2 The list of PCIVs has been relocated to the Technical Requirements Manual, ! consistent with the guidance provided in Generic Letter 91-08. Any change to the Technical Requirements Manual will be controlled by the provisions of 10 CFR 50.59. ; i LA.3 Details of visual inspections of valves and the purpose of the inspections l; have been relocated to plant procedures. This type of inspection is more appropriate for plant procedures. The valves are still required to be 1 cycled, which should ensure their operability. Any change to the
- procedures would be controlled by 10 CFR 50.59.
V q l HATCH UNIT 1 3 REVISION D l 1 l
(shg M [ CAT [o n I S * /* I StfRVEILLANCE RFnufREMENTS LIMITING CONDITIONS FOR OPERATION i 3.7.A.4. Pmssure Suncression Chamer to Drwell 4.7.A.4. Pressure suooression rh -har to
- Devwell Vamm Br==kars -
O Vacan Breakers a.fWhenprimarycontainmentisrequired, all pressure suppression chaser to
- a. The cham rbressure drywell vacuum suppression M M *
- i ghedJ'. - drywell vacun breakers shall be breakers shall 1 Lh.
operable and positioned in the fully 11 l c
- except that up a er s l
[ g ,p 7,6.l. g , noperante for opening pmvided A%.i that they are known to be in the closed position. ; f Ifeitherofthe sad position ] I.b Clos sitionisindicated} Act*
- indicating lights f a pmssum s sion chas er to 11 vacuan by main cent nt lights in the room which are br r is inoperable, tinued operated by separate .
rese operation is pers ible closed positi tches and : only 1 (1) the operabilit of circuits for sa acuum , the mdu t closed position aker. . If either i indicating ircuit is verified, ant position i sting .j and (2) a le test of the li is inoperable or s s i pressure supp sion chamber to that vacuum breaker is ! drywell vacuum aker system is stuck o , the affected ! [ C,) vacuum re r shall be satisfactorily pe ormed within 24 , rs. exercised w in two hours to -i' demonstrate ability of If e r of these re ts f the remaining ition- - M'1 cannot met, the react must icating circui and every ! be in the ld shutdown co itio 1 ays thereafter til , within 36 h s. dundant circut is f c The diff al pressure ichactuaEs c. Each pressure suppression C pmssure sten to dry- chamber to drywell vacuum i shall be 5psid t h*I breaker shall be tested M 7 d 14d 'i q y,4, f,g',3 wel acuus bre for or le f.is) proper pressureopening differen-each r2 fueling ,
- d. The tsui leakage between the dryell ' outage. ;
s and pmssure asipmssion chadier shall - be less than the equivalent leakage d. J ak test o pressure ! through a one-ind diameter orifice ; s sien ch to- ; at a differenttal pressure of one psi. drywe bre ) ;
-b stm nu be ed'a > i as saa e or macn refueling) c,) l outage andJ. . u ,. )
tion Q r t
'g 4.4 D.'S c.v55 <a n 'a[
ment of 3. . 4.b. I C L n g M 3T5 0 .6.f.) Prwry c A; rj p.3 ! k TMs reet*%. p)
$~R 3. L, l *V.
*0ne or more vacuum breakers may be open during surveillances or when performing their i ed function.-
t c o 3. 6.l.f O h 1ch - eh>1 1 2.1-. Amendment ho. 1.. lef 2_ j
= -
. 1
1 l DISCUSSION OF CHANGES l ITS: SECTION 3.6.1.8 - SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS j l TECHNICAL CHANGE - MORE RESTRICTIVE M.1 The LCO requirements have been modified to require ten of the twelve vacuum breakers instead~of the current nine. This change is consistent with the current accident analysis, which assumes nine vacuum breakers; thus, ten are required to meet the single failure criterion. - An Action has been provided limiting the time allowed when three vacuum breakers-are l inoperable to 72 hours (ACTION A)'. These changes are' additional. restrictions on plant operations. M.2 The time provided to "close" an "open" vacuum breaker has been decreased .c from 24 hours (as allowed in current TS 3.7.A.4 b(2)) to 2 hours. This- l ensures that bypass leakage between the drywell and suppression chamber- ! will be minimized. This is an additional restriction on plant operation.
~
M.3 A new Surveillance Requirement (proposed SR 3.6.1.8.1) has been added to ' verify the vacuum breakers are closed _ once every 14 days. This new SR ensures the " closed" requirement of the LC0 statement is being met. This , is an additional restriction on plant operation. , M.4 An additional Frequency is imposed: Within 12 hours after any S/RV l discharge to the suppression chamb'er, a vacuum breaker functional test is reauired. This is an additional rest-iction on plant operation. k. 1 IEQLMCAL CHANGE - LESS RESTRICTIVE ;
" Generic" -
I LA.1 The details comprising the opening setpoint of the vacuum breakers have. been moved to the Surveillances and Bases. . The opening setpoint is , explicitly required in proposed SR 3.6.1.8.3- and 'is not - needed to be : repeated in the LC0 statement. Changes to the Bases will be controlled by the provisions of.the proposed Bases Control Process described in Chapter ; 5 of the Technical Specifications. LA.2 Details _ of visual inspections of valves have been relocated to plant ~- ; procedures. This type of inspection is more appropriate' for plant l procedures. The valves are still required'to be cycled and their setpoint verified, which should ensure their operability. Any change to the i procedures will be controlled by 10 CFR 50.59.
~
j
)
i HATCH UNIT 1 1 REVISION D i
i l l DISCUSSION OF CHANGES ( ,) ITS: SECTION 3.6.3.2 - PRIMARY CONTAINMENT OXYGEN CONCENTRATION ADMINISTRATIVE A.1 This Applicability has been deleted since the startup test program, demonstration of plant electrical output, and the startup of February 22, 1983 have already been complet ed. As such, this deletion is administrative. A.2 This statement has been deleted since it is unnecessary. With the reactor in power operation, reactor coolant pressure will always be above 100 psig. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 The 24 hour time allowed to de-inert the drywell has been restricted to
" prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled I reactor shutdown." This provides more explicit requirements as to when the 24 hour time starts and represents additional restrictions on plant operation.
M.2 Current Specification 3.7.A.8 would allow up to 12 hours for the unit to be out of the Power Operation condition if oxygen is not within limits. (N Proposed ACTION B reduces that time to 8 hours, and explicitly states the V appropriate condition in which the unit must be placed. Therefore, the change represents an additional restriction on plant operation. TECHNICAL CHANGE - LESS RESTRICTIVE
" Specific" L.1 The 24 hour time allowance on startup has been changed to allow 24 hours after exceeding 15% RTP, instead of the current Run Mode (approximately 5%
RTP) requirement. This small difference provides some added time to inert the drywell, and provides consistency with the current Unit 2 requirement and the BWR Standard Technical Specifications, NUREG 1433. This minor change is justified, since the time allowed without an inerted drywell is only increased slightly, and the fact that at low power levels, hydrogen generation is very small compared to higher power levels. L.2 Currently, no time is provided to restore oxygen concentration to within the limit prior to requiring a plant shutdown. Proposed Required Action A.1 and associated Completion Time will allow 24 hours to restore oxygen to within the limit prior to requiring a plant shutdown. During this time, the CAD System is normally still 0PERABLE, thus a means to control hydrogen exists. This new ACTION would possibly prevent an unnecessary shutdown and the increased potential for transients associated with each a shutdown. V
, HATCH UNIT 1 1 REVISION D
DISCUSSION OF CHANGES () 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 because of the confusion associated with these definitions compared to its use in the respective LC0. 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 h applicable, not when it is not 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 91, 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 I secondary containment to maintain 1/4 inch vacuum, and SR 3.6.4.1.3 will ensure the vacuum is attained in 100 or 120 seconds, as applicable. SR 3.6.4.1.4 will ensure the SGT system maintains the vacuum for 1 hour. These new requirements are additional restrictions on plant operation.
,G, HATCH UNIT 1 1 REVISION D
kc'rik hw -3 L . 4. 2. LIMITING c0NDITIONS FOR OPERATION SURVElttANcf SFnUIREMElff5 C. Secondary Containment C. Secondary Cantainment
- 1. Normal Unit } Secondary 75urve111anceWhileIntaneity Containment
- Intaarity Naistained I
fNg.1 W CIlormal Unit I secondary con-tainment integrity shall be Normal Unit I secondary containment surveillance shall be performed as maintained during all medes indicated below: kI of Unit 1 elant_ operation g o m -.... saa ei _ following conditions are met:
- a. A normal Unit I secondary contain-(1) The reactor is sdicritical and ment capability test shall be Soecification 3.3.A. is set. conducted after isolating the ,
s normal Unit I secondary containment P#g &yih (2) The reactor meter tasperature and placing the standby gas treet- [vpplia is below 212'F and the reactor ment system filter trains in opera-bg coolant system is vented. tion. Such tests shall demonstrate the capability to maintain a mini-(3) No activity is being performed ones 1/4 inch of water vacuse moder editch can reduce the shutdown calmwind(<5 mph)conditionswith p.) sargin below that stated in each filter train flow rate not specification 3.3.A. more than 4000 cfa. propo)< ) (4) The(fuel Mor imultated fuelT b. Normal Unit I secondary containment , O pbfLJ AfPlu is not being sonr1 in the re- capability to maintain a minisme ! ( actor building. - lA2 1/4 inch of water vacuum under calm i wind (< 5 mph) conditions with each i rp) mi niildes betsman the normal l filter train flow rate not more laitt I smardary antaissent than 4000 cfm shall be demonstrated and thit 2 secmdary critain- at each refueling outage, prior to
/k(Mjpo4 Nap .
{- sent are closed ad saaled. (6) At least one der in endi ( refueling. gb'** " d p .pe M4 g/ access path behmen the normal
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ct Dy y p u'4',/ SEM,,1 @ thit I secondary contaivuont and thit 2 mcondary contairment l i,a 4.o 3d*# * (isclosed. j Seogq ' y b . Ife b . Inservim P0drostatic or leakap test of rector N m.ve )
% Lt.Flo l .
vessel is not in propuss anta reent 1 afstained during all or lhtti plant operations . Operst 1 Condition 4 as in the 2 Tedmical Speci M at- A
- Normal Unit I secondary containment includes the Unit I reactor building area below the refueling floor and the coenon Unit I and Unit 2 area above the refueling floor. For modified Unit I secondary containment conditions see
- Specification 3.7.C.2. L __ HATCH - UNIT 1 3.7-12 Amendment No. H, 44, M, M MB,160 loc L .
Ses&k s w s1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE pFfkIIREMENTS
>f& *i~ tens N eLt. 1;2+3}
3.7.C.3. Violation of Sacandarv .7.C.3. Surveillancehrter Intaarity violat2 contai m at Intmarity 7 hcN an Pr e. Without Hatch-Unit 1 After a secondary ontaissent ' viola-secondary conta t tion is determined standby gas integrity restore reatment system wil be operated Hatch-Un 1 seconda q tgnediately after the facted zones OL,i i containment int ity a isolated from the inder of (within 4 hour 1G or the ondary containment. The A6,03 0 8 perfom the following (as abili to maintain the ader applicable): of the ary containment i 1/4 inch water vacuou press (1) Suspend irradiated fuel < 5 mph) wind condit : WO and/or fuel cask handling in the Hatch-Unit I shall be con upe.J hree (undercalm d. secondary containment. Mou D . 2 2 (2) Se in at least Hot Shut'doun #"F'M within the next 12 hours and y'd Ws D . MScn meet the Conditions of 3.7.C.I.a.'within the next 24 hours. D. Without Hatch- it 1 secondary 7 Y - coatalnment, ref to the follow- r h .4.2.1,3 6 4.7.1,o j ing Hatch-Unit 2 ical ecification, for to be '4413 fd lowed for Hatch-Un 2: A. y V (1) tion 3.6.5.1. L (2) Secti 3.9.5 1 r p. Primary Containrent Isolation Valves D. Primary Containment Isolation Valves I
- 1. Valver Recuired to be Goerable 3. Surveillance of Goerable Valves During reactor power operation, Surveillance of the primary con-all primary containment isolation tainment isolation valves shall be valves and all reactor coolant perfomed as fo11aus:
system instrument line excess flow check valves shall be operable except a. At least once per operating as stated in Specification 3.7,0.2. cycle the operable isolation valves that are power operated and automatically initiated shall be tested for simulated automatic initiation and the. , j closure times. N See 1 % si.o g >
%es f-2.m u.o !
' % , 6 + % se.-h, .
O - . < . 7-1, . . . .. . 1,. m 1 l
l DISCUSSION OF CHANGES (J ,,'t ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENT ISOLATION VALVES I ADMINISTRATIVE A.1 The current definition of Secondary Containment Integrity requires all secondary containment isolation valves (SCIVs) to be OPERABLE or in their isolation position. Thus, the current secondary containment Specification ; encompasses the SCIV requirements. It is proposed to provide a separate Specification for SCIVs for clarity. Thus the new LC0 will require all , SCIVs to be OPERABLE, consistent with the current requirements. Therefore, this change is considered administrative. A.2 The applicability has been reworded consistent with the new definition of MODES and to have a positive statement as to when it .11 applicable, not when it is not 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 D.2) consistent with part 3. Since the proposed Applicability is effectively the same as the current Applicability, this change is considered administrative. A.3 The technical content of this Applicability is moved to Section 3.10 of the proposed Technical Specifications. Any technical changes to this requireinent are addressed in the Discussion of Changes associated with proposed Specification 3.10.1. v A.4 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 will be located in the Unit 2 Technical Specifications. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 An Applicability has been added. SCIVs are 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 D.3). This is an additional restriction on plant operation. M.2 Three Surveillance Requirements- have been added to ensure SCIV operability. SR 3.6.4.2.1 verifies that SCIVs are in the proper position every 31 days. SR 3.6.4.2.2 verifies that SCIVs isolate within the assumed times in accordance with the inservice testing program. SR 3.6.4.2.3 verifies that each SCIV actuates to its isolation positica on an , accident signal every 18 months. These are additional restrictions on 1 plant operation. M.3 An explicit ACTION listing acceptable methods, including specific types of A acceptable isolation devices, for restoring secondary centainment is Oh ' added. This is an additional restriction on plant operations. (q! HATCH UNIT 1 1 REVISION D j
l DISCUSSION OF CHANGES ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENT ISOLATION VALVES TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 Any time the operability of a system or component has been affected, testing is required to demonstrate operability of the system or component.
Explicit Surveillance Requirements to demonstrate operability of. a system or component being restored to OPERABLE status have therefore been deleted from the Specifications since they are governed by plant procedures. Any 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 shi aping cask). The current Plant Hatch heavy _ K-g) loads analysis covers all 'oads considered heavy, which includes ' com)onents much heavier than a spent fuel shipping cask. These loads will be landled 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 Changes)d in the same manner as other heavy loads which are not spent fuelThe performe ' assemblies.
" Specific" L.1 _ This Action has been changed to allow one valve in a penetration to be inoperable for up to 8 hours, instead of the current 4 hours. Pro)osed-ACTION A requires the penetration to- be isolated in 8 hours. Th s is .
justified since an OPERABLE valve in the penetration exists to isolate the penetration if needed,. thus the " leak tightness" of the secondary
. containment is maintained. The isolated penetration is required to be-verified every 31, further ensuring the continued " leak tightness" of the~
secondary containment. Proposed ACTION B will verify that if. both SCIVs ; in a penetration are inoperable, at-least one SCIV in a penetration 'it
- closed within 4 hours. This maintains ' consistency with the current ' .
requirements. Three Notes have been added. An allowance is proposed for intermittently opening closed secondary containment isolation valves under administrative control. The allowance is presented'in proposed ACTIONS-Note 1. Opening of secondary containment penetrations on an . intermittent basis -is required for performing Surveillances, repairs, routine evolutions, etc. vO l 1 HATCH UNIT'1 2 REVISION D
o DISCUSSION OF CHANGES Q ITS: SECTION 3.6.4.2 - SECONDARY CONTAINMENT ISOLATION VALVES 1 TECHNICAL CHANGE - LESS RESTRICTIVE L.1 (continued) 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 conjuncuan 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 l determined to be inoperable due to inoperable isolation valves, the affected systems Actions must be entered. With the proposed LC0 3.0.6, i 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 0 i l l l l 1 l l n U HATCH UNIT 1 2A REVISION D ______________________________j
- 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. A.3 This Surveillance has been deleted since there is no bypass valve in the system. The system has internal orifices for filter cooling. A.4 A new ACTION is proposed (ACTION D) which directs entry into LC0 3.0.3 if O two or more required standby gas treatment subsystems are inoperable in V Medes 1, 2, or 3. This avoids confusion as to the proper action if in Modes 1, 2, or 3 and simultaneously handling irradiated fuel, 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 C.2.2 and E.2) to suspend CORE ALTERATIONS, 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 B.1).to not only be in Cold Shutdown (MODE 4) within 36 hours, but to also be in Hot Shutdown (MODE 3) within 12 hours. This is an additional restriction on plant operation. ,m v k HATCH UNIT 1 1 REVISION D
-- u. . . . . , ~,. - .
- _ , . . . ~ . . . .- ,.
.!. Qec 19ebM 8* I'
?
4 LIMITING CIM)ITIONS FOR OPERATim SIRVEILLANCE REDLIIRDENTS 4.9.A.2. Standw AC power Suoolv (Diesel ' Generators IA. IB. and IC) j (Continued)
- 4. h erability (Continued) ,
.At least once per senths? ,
4 during shutdown, i diesel to an in ion- , 0 [,A. 'in with in cardienction thits
.. > manuf 's reccomenda ans ;
'for this lass of standby . ~
- ' service.* ;
- 4. ' A _
M3.g.).') . diesel generator capability to reject its largest single shutdown (emergency) load,: idille meintaining vol
'at 4160 t 420 volts. or .
1esel generator IA, is d be CS pimp 1A at a ra ow. For diesel-
. genera IB, this would be either the C or 2C Residual. ,
Heat Resoval ice Water !
) pinip at flow, or esel genera C, 3 this c 1 -
1d be E lat r=+w _fDuri the' - load rejection tes_ 11esei generator shall not ' k.9 exceed the nominal speed 1 iplus 795 of the difference a
. jbetween neednal speed and- i the overspeed trip setpoint, l
'i pr.155 above nominal s d Md g/cf#
Michswer is lessI* . j 4% _ .
- 5. At least once6sonths~
g 6Eb verify the ' diesel generator capability' , N}0 to reject a load of at : . least 277E w.without trip-WS pi J.p@gy . The generator voltage sha 1 not exceed 4e00 volts - g .O ' .during and following the; load rejection.*
- 6. At leasttGice per 18 months gryostE ggjeak Grurino shuth verify the' '
m.) g3,%.I'\g diesel tor operate o pin,ser%4cr for at east 24 hours. 4 ,g y
- During the first 2 hours- ,
of this test, the py\e.1
- 59-Q.\Il 18 For thewillIB nonths diesel satisfy generator, the requirements of Unit a single full load rejecton test every 1 Specification A .-
& @p 4.9.A.2.a.s and Unit 2 specification 4.e.l.l.2.d.4. A single partial - '
.l-ZEL:
g)4Ijload Unitrejection test 1 Specification every 18 4.9.A.2.a.4 and thit months will satisfy 2 Specification the rtquiriments of 4.8.1.1.2.d.3. :; y ~ ' - A O <1e diesel inspection will satisfy the requirements of Unit 1 Spec.ilcation 4.9.A.2.a.3 and Unit 2 Specification 4.8.1.1.2.d.l. : H4TCH - IMIT 1- 3.9-2a Aemndsent No. 447,178 'I
'3 M :
, ,, 'a ~, . . .- ~ , . .n -
- -.~ .- . - _ - . _ . _ . - - _ _ - _ - - . - - -
DISCUSSION OF CHANGES ITS: SECTION 3.8.1 - AC SOURCES - OPERATING 1% 'd TECHNICAL CHANGE - MORE RESTRICTIVE M.1 (continued) These ACTIONS are consistent with the requirements for a Unit I source, with the exception of the restoration time provided for a Unit 2 DG. The time provided is 7 days, which is consistent with the restoration time provided for in the LCOs for the individual components powered from Unit 2 sources. In addition, the SRs are also applicable to the Unit 2 A sources; thus, a proposed note applicable to all SRs and SR 3.8.1.19 has Ig been added to ensure Unit 2 sources are tested. Therefore, this change, is considered more restrictive on plant operations. M.2 A new Note has been added (proposed Note to SR 3.8.1.6) to restrict this Surveillance from being performed in MODES 1 and 2, since it could result in a grid perturbation and the potential for a Unit transient. However, , credit may be taken if unplanned events occur that satisfy this SR. M.3 Limitations on the operating power factor are added to the full load rejection test and to the 24-hour run Surveillance. These limitations ensure the DG is conservatively tested at as close to accident conditions r as reasonable, provided the power factor can be attained. A note is also added. Note 2 of SR 3.8.1.8 provides guidances for when the power factor k (]y cannot be attained. M.4 As with all other DG start requirements, proposed SR 3.8.1.10 is proposed to add the ar.ceptance criteria for voltage limits (upper and lower) and speed /frequet.y upper limit (lower limit included in the existing Surveillance). These acceptance criteria are consistent with all other DG start acceptance criteria. In addition, a time requirement has also been added, consistent with the accident analysis. Proposed SR 3.8.1.18 is proposed to add the voltage acceptance criteria. H.5 The DG fuel oil day tanks are proposed to have a Surveillance for the checking for and removal of accumulated water (proposed SR 3.8.1.4). This added restriction provides assurance that water will not degrade the performance of the diesel engine. M.6 A new, more restrictive requirement to be in MODE 3 (Hot Shutdown) within 12 hours has been added. This is consistent with the BWR Standard Technical Specifications, NUREG 1433. M.7 The restoration time for one offsite circuit or one unit specific DG (IA or IC) has been decreased to 72 hours, consistent with the BWR Standard Technical Specifications, NUREG 1433. The swing DG time remains at 7 days. -g ( HATCH UNIT 1 4 REVISION D
q DISCUSSION OF CHANGES 7, ITS: SECTION 3.8.1 - AC SOURCES - OPERATING i ) nj TECHNICAL CHANGE - LESS RESTRICTIVE (continued) LA.4 When the OPERABILITY of a system or component has been affected by repair, maintenance, or replacement of a component, post maintenance testing is required to demonstrate OPERABILITY of the system or component. 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 allowed as discussed in the Bases for SR 3.0.1. LA.5 The purpose of this SR is inherent ir, the manner in which the test is performed and is described in the Bases for SR 3.8.1.9 (load shedding) and SR 3.8.1.6 (auto bus transfer). Therefore, the description has been relocated to the Bases. Changes to the Bases will be controlled by the provision of the proposed Bases Control Process in Chapter 5 of the Technical Specifications. LA.6 The diesel generator accelerated test frequency requirements are relocated in their current licensing bases form to plant procedures, leaving the Technical Specifications periodic surveillance frequency as 31 days. A plant procedure implements the current Technical Specifications requirements for accelerated test frequency, as well as the requirements A and responsibilities for tracking emergency DG failures for the Cl determination and reporting of reaching trigger values specified in NUMARC 87-00. These requirements are more restrictive than those specified in NUREG 1433. In addition, Generic Letter 94-01, " Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators" allows Licensees to request removal from TS of provisions for accelerated testing and special reporting requirements for EDGs. Hatch proposes relocation only with no relaxation in ITS conversion. The allowances of GL 94-01 will be addressed separately, post ITS implementation.
" Specific" L.1 Note 2 to SR 3.8.1.2, Note 1 to SR 3.8.1.5 and the Note to SR 3.8.1.18 have been added to allow a prelube prior to starting the DG. - DG starts without prior engine prelube create unnecessary engine wear, thereby reducing overall reliability. The engine prelube does not result in an enhanced start performance which could mask the engine's ability to start in accident conditions without a prelube. In addition, Note 2 and SR 3.8.1.2 also allow a gradual DG warmup. This portion of the Note is allowed currently, because no startup time is specified in the current surveillance. /m HATCH UNIT 1 6 REVISION D
i i DISCUSSION OF CHANGES ITS: SECTION'3.8.1 - AC SOURCES - OPERATING TECHNICAL CHANGE - LESS RESTRICTIVE (continued) j
" Specific" l
L.2 The intent of a requirement for staggered testing is to increase reliability of.the component / system being tested. .A number,of studies-have been performed which have demonstrated. that. staggered testing has-negligible. impact on component reliability. .- These . . analytical . and i subjective analyses have determined that staggered testing 1) fis operationally difficult, 2) has . . negligible - impact- on component reliability, 3) is not as significant as initially thought,- 4)' has no impact on failure frequency, 5) introduces additional stress on components i such as DGs potentially causing increased component failure' rates and component wearout, 6) results in reduced redundancy during testing, and 7) ; increases -likelihood of human error by increasing testing intervals. ! Therefore, the majority of staggered testing requirements have been' deleted. j O , i 1] O 1 a LHATCH UNIT.I 6A REVISION D-
DISCUSSION OF CHANGES g3 ITS: SECTION 3.8.2 - AC SOURCES - SHUTDOWN V 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 is based on the current Unit I secondary containment Applicability, a little more i restrictive. However, based upon the proposed Secondary Containment Applicability, the requirements are 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. Currently, no actions are provided. The new ACTIONS (ACTIONS A and B) will either 1) require declaring the affected components inoperable and taking the ACTIONS of the applicable system LCO (Required Action A.1) or 2) will require suspending CORE ALTERATIONS, OPDRVs, and irradiated fuel movement in the secondary (] V containment, and initiating action to restore the inoperable source (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, Surveillances have been.added to ensure the OPERABILITY of the required AC sources. These new ACTIONS and Surveillances are additional restrictions on plant operation. 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 1-network and the onsite Unit 1 Class IE AC electrical power distribution subsystem (s) required by LCO 3.8.8. This means that all Unit 1 buses needing AC power must be capable of being powered from a qualified Unit 1 offsite source. In' addition, since certain Unit 2 equipment may be needed to meet Unit I accident analysis, this LCO also requires an OPERABLE Unit 2 DG (capable of powering the required equipment), and a Unit 2 offsite source (which is available to power to the required Unit 2 equipment). h These additional requirements adequately compensate for the reduction in the number of required DGs (from two to one). O j 1 I HATCH UNIT 1 1 REVISION D l
, , , n .u .n a an =. --~ = a >=+ - ~ ~ -
5pe;% 7.84
.;~ LIMITING CCM)ITIONS FOR OPERATICE Okeausm.a.&
, SURVFiltAKE Renu 1RDENTS I 3.9.A.3. 125/250 Volt DC Ememency Power 4.9.A.3. 125/250 Volt DC Enemancy Power System (Plant Batteries lA and System (Plant Batteries lA and Seed h S58b3 g g ' of C%e.3
% e r ns 3.3.9#. >
f80th IZ5/Z50 volt plant batteries . a. Idankly Surveillagt ' k %4_ (IA and 18) shall be operable and .g .Every week the specific gravity ~ Sech . shall have an operable battery 3,g~g and the voltage of tne pilot cell' h-charger and ventilation system andJoveren Dauery voltage shall a vailable for =
- y
%d
/EiFasasured and recorded. Each 125 volt battery shall have a =[,
pr9A m 3.g.(, wi thfr@. fainisam of 105 volts at the bat-tery teminals to be considemd i *
) 9 le. ,
pr g g k b. Monthly Surveillance ' . 1 m.t vv omnta measurements shall be ;
@ N 3 made of voltage of each cell to the nearest 0.1 volt and the spe- ;
- m. repwh k.M fie%k O ",g cific gravity of each cell. These j measurements shall be recorded. >
popo3e3 Liquid level shall be checked l g pr gasek Tohk 1E& Sd E4
- c. Ref'uefing Outape Surveillance . .
During each scheduled refueling.- ! outage, the batteries shall be " subjected to a rated load dis- ..
. charge test. The specific gravity ,.,
and voltage of each cell shall be ! detemined after the discharge
' and recorded. '
T E- ev 4100 voit uutes (it. ,. L . av 4150 Volt Eses (IE.
IF. and IG) IF. and IG). gw 0 %w, 'Theenemency4160voltbuses(IE. Theenergency4160voltbuses(IE,
- IF,andIG)shallbeenegizedand IF, and IG) shall be monitored to ;
c4Chu operable. the extent that they am shown to
. g h,, 3*3 he ready and capable of trans- j mitting the emergency load.
' 18.7 J' ja
]5. Ememency 600 Volt Buses (IC 5. Enemanev 600 Volt mm tic
.dh5e b. and.lD) and ID)
- Theemergency600voltbuses(IC The emagency 600 volt buses (IC' and 10) shall be energized and and ID) shall be sanitomd to the operable, extent that they are sham to be . j ready and capable of transmitting' j the emergency load._
f6. E--as 250 Volt DC to 600 Volt 6. L- v 250 volt ut to muu voit AC Inverters AC Inverters The emergency 250 volt DC to 600 a. The exegency 250 volt DC/600 . volt AC inverters shall be ener- volt AC inverters shall be seni- , gized and operable, tored to the extent that they are. i shown to be twady and capable of ~ ' transmitting the emergency load. ' b e 1 heub5fou e,Q @ ,3 08.1, Q5. y '
%%.<. +
10 -1 t HATCH - UNIT 1 3.9-3 Ausndment No. H , 48 .. i ok M '
, ,n -. - , - , -
DISCUSSION OF CHANGES ITS: SECTION 3.8.6 - BATTERY CELL PARAMETERS /_.\
,1 TECHNICAL CHANGE - MORE RESTRICTIVE M.1 (continued)
New Surveillance Requirements are being added, consistent with the BWR Standard Technical Specifications. A new Frequency is being added to proposed SR 3.8.6.2 to require all the cell parameters to be verified once within 24 hours after a battery overcharge > 150 V. In addition, the monthly check (of which no limit is provided) is now a 92 day check, consistent with IEEE-450. While this specific change is less restrictive, it is offset by the additional surveillances and requirements imposed. Proposed SR 3.8.6.3 requires a verification that electrolyte temperature is 2 65 F for each station service battery and 2 40*F for each DG battery every 92 days. This helps to ensure battery OPERABILITY. The Applicability of this new LC0 has been made "When associated DC electrical power subsystem is required to be OPERABLE." This covers the current MODES 1, 2, and 3, as well as new requirements for MODES 4 and 5 and fuel handling. _. p b o LJ HATCH UNIT 1 2 REVISION D
I l DISCUSSION OF CHANGES
- l. ITS: SECTION 3.8.6 - BATTERY CELL PARAMETERS O
A) o O HATCH UNIT I 3 REVISION D
i fg.,; w io.,.r,2- y 6.0 ADMINISTRATIVE CONTROLS l O< ' g,g ,l, ej d, The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the
. appropriate onsite manager; however, they shall have sufficient -
organizational freedom to ensure their independence from operating pressures. '? f,2 ,1 6.2.2 UNIT STAFF '
- a. ach on duty ift shall be e sed of at lea eminimu\mift
. M.! c composit n shown in table . 2. 2-1. '
g,g,g ,lo b. -At least one licensed Operator shall be in the control-room for each' reactor containing fuel. ;
- c. At least two licensed Operators shall be present in the control room
- f,2,1.b '
for each reactor in the process of start-up.- scheduled reactor.. i i shutdown and during recovery from reactor trips. ,
,f, 2,1,4 d. An individual qualified to implement radiation protection procedures shall be onsite when fuel is in either reactor.
- Ie. All CORE ALTERATI S shall be directly upervised by ett er a l g ,1, icensed Senior Rea tor Operator or Sent Reactor Operat Limited :
t Fuel Handling who s no other concurre responsibiliti during _thi operation. l T f. A.F e Team of at least fi members shall be ma tained onsite at 7 all t s. The Fire Team sh 1 not include the m imum shift crew I l h'* A necessa for safe shutdown on nits I and 2 or any ersonnel required or other essential fun ions during a fire rgency. f
)
- g. Administrative procedures shall be developed and implemented to I 2'1'4 limit the working hours of Unit staff who perfom safety-related functionsyne, sen1or rea or operators, re tor operators, faux ry opera rs, health p sicists, and key intenance-l g] mperso el.
Adequate shif t coverage shall be maintained without routine heavy ' use of overtime. The objective shall be to have operating. i personnel . work a nominal 40-hour week while the plant is : operating. However. in the event that unforeseen problems require i substantial' amounts of overtime to be used or during extended periods of shutdown for refueling, major maintenance, or major plant modifications, the following guidelines shall be followed on l a temporary basis: -
~t (1) An individual should not be pemitted to work more than !
16 hours straight, excluding shift turnover time. > q (2) An individual should not be permitted to work more than 'J 16 hours in any 24-hour period, nor more than 24 hours in any J HATCH - UNIT 1 6-2 Amendment No. I'#f, 55
- -- _. . . l-
, li c ens ed d m d n .' dp % med' N
a.-w.we s % ~ 4 ., yl g
%p rs wl < +c., ) , - .
./
.z d 7-
% , ./
a , . , - . ~ , - . . .
., - . .. - - . _ _ _ - - - . - - _ - - - ~ .
DISCUSSION OF CHANGES
.].-
ITS: SECTION 5.2 - ORGANIZATION I ADMINISTRATIVE o- A.I The definitions of the various . Operational Conditions are located in current Specification 1.0, Definitions, and in the- proposed Technical Specifications in ITS 1.1, " Definitions." Therefore, this table notation - , is unnecessary and has been deleted. l. A.2 The current Technical Specification provides examples of the Unit staff positions who perform safety-related functions and whose working hours are limited. Since these examples may not include all positions that could be limited and since these positions may change, the examples have been .
'j generalized. The modification of. these examples clarifies present requirements and thus is an administrative change, i
l o i 1 l L
'i O
HATCH UNIT 1 1 REVISION D
-- - . - - . - . _ .. . -_- ~ -- - -,c e c., C cde'or) . . h '.
ADMINISTRATIVE CONTROLS l 6.16 POST-ACCIDENT SAMPLING AND ANALYSIS A pr ran shall be established, implemented, and maintained to ensure the capab 11ty to obtain and analyze samples of reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere under accident conditions. S ee O h! .- ' The program shall include the following:
-f '
ed c % es 4a
...T s s 4 3 (1) Training of personnel, (2) Procedures for sampling and analysis, and ~ j (3) Provisions for maintenance of sampling and analysis equipment. 1
~
;- ~~~T.T7 OF5fTE DOSE CALCULAfitm ruuvun
(- 6.17.1 Licensee-initiated changes to the 00CM shall,
- a. Be documented and records of reviews performed shall be retained as required by Technical Specification 6.10.2.o. This documentation shall contain: ;
dee Ols us 3 ;.,
- f l
Ch 93 .fo, 1) gropriateanalysesoreval Sufficient information to tions su$ ortingthe justi he chahe change t$ (s),ether with the S : 5 3, / 2) A detemination that the change will maintain the level of f
. radioactive effluent control required by 10 CFR 20.1302, 40 CFR m 4 4.'s Se ef:.a Part 190, 10 CFR 50.36a, and Appendix 1 to 10 CFR Part 50 and !
not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. _
- b. Become effective after review and acceptance by the PRB and the f g
approval of the General Manager-Nuclear Plant.
- c. Be submitted to the Commission in the fom of a complete, legible ,
copy of the entire 00CM as a part of or concurrent with, the
.i Annual Radioactive Effluent Release keport for'the period of the i report in which any change to the 00CM was made. Each change shall !
be identified by markings in the margin of the affected pages, , clearly indicate indicating the date e.g., (themonth area/ year) of the thepage changethat was was changed, and shall implemented.
~ 6.18 RADIDACTIVE EFFLUENTS CONTROL PROGRAM 554 Ap ran shaii be estabiished, implemented, and maintained conforming with 0 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive i effluents as low as reasonab achievable. ' The rogram ( ) shall be l contained in the 00CM, (2) s 11 be implemented y operat ng procedures, and (3) shall include remedial actions to be taken whenever the program 1 limits are exceeded. The program shall include the following elements: _
3 f radioactive liquid and gaseous. {
/ ./ 5,5.4 9 1)LimitationsontheCQPERABILI monitori instrumente on neluding surveillance tests and setpoint '
detemina ion in accordance with the methodology in the 00CM,
' ' ~ 2) Limitatio Mn all ti on the concentrations of radioactive
- o. - 5.5.4 b material rereasec in uid effluents to UNRESTRICTED AREAS U confomi to 10 times the concentrations stated in 10 CFR Part 20, Appendix (to paragraphs 20.1001-20.2401), Table 2 Column 2,
- 3) Monitoring, sampling, and analysis of radioactive 11 id and aseous :J 5.5 d C affluents in accordance with 10 CFR 20.1302 and with he meth ology and parameters in the 00CM, j 6-23 Amendment No. 408, MG,190 HATCH - UNIT 1 4
i i i l
]
I0h 2,,,
l fgehe Ce .f $,9 ADMINISTRATIVE CONTROLS f} RADIOACTIVE EFFLUENTS CONTROL PROGRAM (Continued) G I* g* S y 4) Limitations on the annual and huarterly doses or dose comitment toa MEMBER i OF released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50, f* f* y' e 5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar the ODCM gear in accordance t least with the methodology and parameters in every 31 days, g*
- Y' y 6) Limitations on theQPERABILIT3and use of the liquid and gaseous effluent treatment systems w ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the i guidelines for the annual DOSE or dose commitment conforming to l Appendix I to 10 CFR Part 50,
{ f f,4, /9 7) Limitations on the dose rate resulting from radioactive material q released in aseous effluents - theSITEBOUgDARYasfollows:gromtnesitetoareasatang>beyond g, (
- t. For noble gases, less than or equal to a dose rate of 500 mrem / year to the total body and less than or equal to a dose rate of 3000 mrem / year to the skin, and a
- b. For Iodine-131 Iodine-133, tritium and all radionuclides in 1 particulate for,m with half-lives gre,ater than 8 days, less tha t {
nr coual to a dose rate of 1500 mrem / year to any organ,
' S' la [ 8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix 1 to 10 CFR Part 50,
_ f, f, Lf , d 9) Limitations on the annual and quarterly doses to a MEMBER OF THE (N PUBLIC from lodine-131 Iodine-133, tritium, and all radionuclides U) in particulate form wilh half-lives greater than 8 days in caseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix ! to 10 CFR Part 50, and f*f, Q.J 10) Limitations on the an THE PuBtiC eue to reinual dose ases oforradioactivity dose commitment and to toany MEMBER radiation OF from uranium fuel cycle sources conforming to 40 CFR Part 190. f6.19 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM A program shall be established, implemented, and maintained to monitor the radiation and radionuclides in the environs of the plant. The program shall provide: (1) representative measurements of radioactivity in the highest potential exposure pathways, and 2 verification of the. accuracy of the effluent monitoring program and m(od)eling of gTeQ**W environmental exposure pathways. The pro ram shall: (15 be contained in the ODCM c5 & g es and(3)Inc(2)dethefollowing:conformtotheguidanceofkppendixItoL0CFRPart50, lu Q C154.1% sampling analysis, and reporting of radiation and IW 4 1) Monitorinkdes radionucl in the, environment in accordance with the methodology and parameters in the ODCM, Sec k 2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and
- 3) that independent checks on the precision and accuracy of theParticipation in an Interlab i measurements of radioactive materials in environmental sample k matrices are performed as part of the quality assurance program for environmental monitoring.
n. ( ) v HATCH - UNIT 1 6-23a Amendment No. 190 hoO l
1 l m DISCUSSION OF CHANGES l ITS: SECTION 5.5.4 - RADI0 ACTIVE EFFLUENT CONTROLS. PROGRAM-
-l ADMINISTRATIVE A.1 Comment number not used.
4
. - J A.2 Consistent with NUREG 1433, the phrases, "at all times" and "from the site ..: -
1 to areas at and" have been deleted. The intent of the requirement ~ remains 2 the same and is thus considered administrative. q TECHNICAL CHANGE - LESS RESTRICTIVE i
" Specific" -]
L.1 The present TS uses the term " operability" when referring to radioactive- ; liquid and gaseous monitoring instrumentation and treatment systems. The l proposed TS uses the term " functional capability."_ - The proposed change is' 5j necessary because the Radioactive Effluent Controls Program 'is . located i outside the TS-in the ODCM. Use of the TS term ." operability" can be j confusing when used in programs which are not in the TS. The term i functional capability means that the component or system is capable'of- l 0, . performing its design function. Since it is not a TS defined term, the use of the " functional capability" is considered less restrictive than the i use of the TS term " operability." l l i _O ! A_/ l u HATCH UNIT 1 1 REVISION D. ,
Rct ICebe*m $$sh LIMITING COPOITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS
'll.6.K. STRUCTURAL INTEGRITY 4.6.K. STRUCTURAL INTEGRITY e
- 1. Normal Condition j",fj, Surveillance Reauir-te far4nhervve7 (mucecMosand testing of ASME Code .i The structural integrity of ASME u.>> 4. z, and 3 (equivalent) components
^
Code Class 1, 2, and 3 (equiva- 6shall be applicable as follows: lent) conpanents shall be maintained in accortlance with the 1. /Tn-krvice inspecQon of AS$ D Surveillance Requirements of Specification 4.6.K. W ' Code Elass 1, 2. m "I ' = 'ivalent U e A tt ans i serivce tes ing % a<f A5ME Code Class , 2, and 3
- 2. Off-Normal Conditions (&quivalent) pumps d valves s all -]
be rformed in acco ance with
- a. With the structural Sect on XI of the iler a >
integrity of any ASME Code Press re Vessel Code a applicab ' Class 1 conponent not Adde as required by 1 FR50, conforsing to the above Section 0.55a(g), except ere requirements, restors the specific elief has been structural integrity of granted b the Connission pu uant theaffectedcouponent(s) to 10CFR50, ection50.55a(g) , to within its limit or (6) (1). ; isolate the affected gg,g couponent(s)priorto - l increasing the Reactor 2. Pe Normance of 1 above i ervice Coolant System tenperature CiMpect1on; an]btes ing activ es
~
more than 50*F above the ,j snai in additio to other minian temperature required spect Surveillanc by 10T considerations.
@i nts.
- b. With the structural integrity 3. Nothing in the ASME Boiler and of any ASME Code Class 2 conponent(s) not conforsing Pressure Vessel Code st.all be
[.[ 3. d construed to supersede the . j /d un 1 to the above requirements, requirements of any Technical I O restore the structural-integrity of the affected couponent(s) to within its Specification. The Inservice Inspection ogram limit or isolate the affected for piping identified in ' component (s)priorto . neric Letter 88-01 shall ; increasing the Reactor Coolant ormed in accordance with he , System temperature above 212*F. sta f positions on schedule, f ,
, meth s and personnel, and saap
- c. With the structural integrity expans included in the generi of any ASME Code Class 3 letter, xcept where specific couponer.t(s) not conforming to written ief has been granted the above requirements, restore by the C ssio" the structural integrity of the 'i affectedcomponent(s)towithin its limit or isolate the .;
affected couponent(s) from service. A3 - 5 5.G.a A5M E ivr ve HSc ". p l G e q ve n c. .~ e.s 'b ( feeDi5am.ae.f j
- [ w. @ f f. 6. baN f ruis!.as pak r, tofwv a'"
SA 3 d ;
.Se< k .) }. 4. ^ TksY seYV5iAS ;
.j t
{' HATCH UNIT.1 3.6-10 Amendment No. M, 43, MS, 49,176 _
.[
SR 3feS N p* r.5.(,..c ita pww'dus to a5insernc ($ l sr. .p,, k,Lia , _-restQ a&Itlts ) A l 1&t .d
m DISCUSSION OF CHANGES
.%j1 ITS: SECTION 5.5.6 - INSERVICE TESTING PROGRAM ADMINISTRATIVE A.1 This requirement merely restates that all applicable requirements must be met. Repeating this overall requirement as a specific detail is redundant and unnecessary; therefore, this detail can be omitted without any technical change in the requirements.
A.2 A statement of applicability of SR 3.0.3 is needed to maintain allowances for surveillance frequency extensions contained in the proposed Technical Specifications since these SRs are not normally applied to frequencies identified in the Administrative Controls section of the. Technical Specifications. The addition of SR 3.0.3 is discussed in the proposed changes for ITS 3.0/4.0. The statement of applicability of SR 3.0.2 is needed to maintain current allowances in Definition 1.0.II for Surveillance Frequency since this allowance is not normally applied to frequencies identified in the Administrative Controls section. Since this change is a clarification needed to maintain provisions that would be allowed in the LC0 sections of the Technical Specifications, it is considered administrative in nature. A.3 The testing intervals for the Inservice Testing of pumps and valves are added to proposed Technical Specification 5.7.11. Since this information O is a part of the presently referenced ASME Section XI code for pump and D () valve testing, the addition is for clarification and considered administrative in nature. l HATCH UNIT 1 1 REVISION D l
^
, q l
ADMINISTRATIVE CONTROLS l [V\ f" 6, if MMTHLY OPERATING REPORT l 6.9.1.10 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly batitIto tne uirector, Orrice or nanagemen anc rro ram Analysis, U. 5. Nuclear Regulatory Commission, Washington, A'g D. C. 2 555, with a copy to the Recional Office of insoection and
;EnforcementIno later than the 15tn of each month following the calendar morith covered by the report.
g,$ CORE OPERATING LIMITS REPORT 4 f 6.,J ,4 6.9.1.11.a. Core operating limits shall be established and documented in ' the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following: , f.E.8e**b(1) Operation with a Limiting Control. Rod Pattern (for Rod Withdrawal Error, RWE) for Specification 3.3.F. . f,g,, J , a , 0 (2) .The Average Planar Linear Heat Generation Rate (APLHGR) for Specification 3.ll.A. T Linea Heat Genera onRate(LH%forSpecif on 3.1 % g g* g' g, p (4) The Minimum Critical Power Ratio (MCPR) for Specifications 3.3.F and 3.II.C and Surveillance Requirement 4.11.C. f",(n. f, /r b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the follawing documents f ,4. J , dr, h (1) NEDE-240ll-P-A, " General Electric Standard Application for Reactor gO Fuel," (applicable amendment specified in the CORE OPERATING LIMITS REPORT) f,4 , f. / ,2.) (2) " Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 157 to Facility Operating License DPR-57 " dated September 12, 1988. f,6, ,f, da c. The core, operatino limits chall be determined to that all anolicable ruei tHtrmal-meenanicts limits; 'corsnnermai-ny<traulic , O limits simin Nd.
, E .,
S limits, nuclear limits swch at shut h wr> margin. nnd - transient and accident scalysis limits) of the-sarety anaty>1s are met. + f,4, f, d d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto; shall be orovided upon issuance fnr each reload-cycle. to the NDrrDocument Control Desk with copies to the Regij {dministratorandResidentinspector. rTN SERT'M S M'3 A A) PTL/t L < v-HATCH - UNIT 1 6-15d Amendment No. 440, 449, MS, 190 hcN 10
- o. DISCUSSION OF CHANGES
() ITS: SECTION 5.6 - REPORTING REQUIREMENTS ADMINISTRATIVE (continued) i A.14 The reporting requirements for Inservice Inspection (ISI) evaluations are covered in programmatic requirements located outside the TS. The ISI requirements in CTS 4.6.K are being relocated to the ISI Program as described in the changes for ITS 5.5.6, " Inservice Testing Program." With the relocation of the ISI Program from the TS, the reporting requirements can also be relocated. A.15 The present Technical Specification requirement to provide a special report within 30 days when the reactor coolant radioactivity exceeds limits is not needed. Violations of Technical Specification limits are evaluated and reported in accordance with 10 CFR 50.73 as LERs, and do not need to be called out separately in the Technical Specifications as a Special Report. A.16 ITS 5.6, " Reporting Requirements," does not use the current Technical Specification subtitles of " Routine Reports," " Annual Reports," or .
"Special Reports." The ITS names each individual report rather than grouping reports under subtitles. This change does not change reporting requirements and only affects the format of the Technical Specifications.
Therefore, this change is considered to be administrative. I \ G A.17 Comment number not used. A.18 The general statement in current Technical Specification 6.9.2 to submit special reports within the time period specified for each report is not retained jn the ITS. Each special report contains requirements for submittal. This change merely deletes duplicate requirements in the Technical Specifications and is thus considered to be administrative in nature. 7 . w/ HATCH UNIT I 3 REVISION D
Syn.4;cd.on 5 W ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued) Records of radiation exposure for all individuals entering radiation
% control areas. -
- d. Records of gaseous and liquid radioactive material released to the environs.
- e. Records of transient or operational cycles for those unit compenents identified in Table 5.0.G-1.
- f. Records of reactor tests and experiments.
(Lovp 5IcM \ fo< CTU 1
- g. Records of training and qualification for current members of the unit staff.
- 6. A , # a/ N h. Records of in-service inspections performed pursuant to these gg, Technical Specifications.
- i. Records of Quality Assurance activities required by the QA Manual.
J. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
- k. Records of meetings of the PRB and the SRB.
- 1. Records for Environmental Qualification which are covered under the provisions of paragraph 6.15.
- m. Records of analyses required by the Radiological Environmental Monitoring Program.
- n. Records of the service lives of all safety-related hydraulic and mechanical snubbers including the date at which the service life commences and associated installation and maintenance records.
(,, o. Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.
$g f6.11 RAUlAllVNPROTECTIONPROGRAM}
of ch Io [y Procedures for personnel radiation protection shall be prepared consistent with the requirements of.10 CFR Part 20 and shall be approved, maintained and 95 b 'l l, L_a dhered to for all operations involving personnel radiation exposure. IN %s kb. ____ lL12 'HIGH RADIATION ARIA
- 6. M . I . In lieu of the " control device" or " alarm signal" required by paragrapt 20.401(a) of 10 CFR 20, eaM high radiation area in which the intensity of radLtion is greater than 106 mrem /hr but less than 1000 mrem /hr** shall be il.
barru aded and conspicuously posted as a high radiation area and entrance thered shall be controlled by requiring issuance of a Radiatfon Work Permit *. Any individual or group of indiwiduals ermitted h k ,* e, , , , A .1 r O - Incl., v.dva b gug6 -]* * ~?a&Geodeaeer Health Pysics P or per donel escortehby Health Physic p^-: - . , -
& person i
;n , . .; . . u . r.
-, m, . , , , . . __ - . -. 1 be exempt from the RWP--
issuance rehuirement during the perfownance of their assigned radidion j-protection daties, provided they comply with approved radiation protection b,
~t pr =4" C " f , y.hn . procedures foi entry into high radiation areas.
s
'J
-QN(L*G #) ** Measured at 30 cm from the radiation source or from any surface that t$e
.I radiation penetrates.
^
.( M~ HATCH - UNIT 1 6-20 Amendment No. 66, Gedee d6d." 10/2'.l", 444, 444, M 8, 190. ML
ADMINISTRATIVE CONTROLS .fF MC .M E 7 f so .nwr such areas shall be provided wth or accompan'ind by one or mor& c: [,} - the,following: L A radiation monitoring device which cantinuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continbously integrates the radiation dose rate in the area and alarms when a preset integrated j dose is Yaceived. Entry into such areas with this monitoring device may he made after the dose rate level in the area has been
[i
<. established imd personnel have been made knowledge 4ble of them.
I b, ,I c. An individual qua fledinradiationprotectionproce(ureswhois equipped with a ra tion dose rate monitoring device. This individual shall be re ponsible for providing positive clantrol over the activities within (the area and shall perform periodic radiatior.' surveillance at the frequency specified by the facility Heabh Physics supervision in the Radiation Work Permit. 6.12.2. The requirements of 6.12.1., above, shall also apply to ea radiatidq area in which the intensity ofNadiation is greater tha ci actuel c A* 1000 mrem /hr* but less than 500 rads in I hour.** In addition, lo ' Mn shall be provided to prevent unauthorized e into such areas and to keys shall be maintained under the administr e control of the Shifi
!,upervisor ont duty end/or the Laboratnev Forem ,on dirty.
6.13.
~
INTEGRITY OF SYSTEMS OUTSIDE EONTAINMENT The licensee shall implement a program to reduce leakage from systems outside k bicuni containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program ofch'YSfee shall include the following: (N ITM 5~. 5. 2>
- 1) Provisions establishing preventive maintenance and periodic visual
( ) Nk inspection requirements, and b 2) System leak test requirements, to the extent pemitted by system design and re!iological conditions, for each system at a frequency not to exceed re"ueling cycle intervals. The systems subject to this testing are (1) Residual Heat Removal, (2) Core Spray, (3) Reactor Water Cleanup, (4) HPCI, and (5) RCIC. J f 6.14. IODINEMONITORING} See asc>,s;w 1 g cL, u , Thelicenseeshallimplementaprokramwhichwillensurethecapabilityto 9 c.T % accurately determine the airborne odine concentration in vital areas *** under l ,
, accident conditions. This program shall include the following: 1 Sedd 1) Training of personnel,
- 2) Procedures for monitoring, and
- 3) Provisions for maintenance of sampling and analysis equipment.
~ *Nessurement made s at 30 centimeters frhm tne radiation source or from any sn. face that the hadiation penetrates.
**MeasNrement made at'J meter from the rad (ation source or frdg any surface
.that the r>M =tiv mawatrates, e
** Areas requiring personnel access for establishing hot shutdown condit I
i
,m .
) HATCH - UNIT 1 6-21 Amendment No. 65, M, MG,190 24 2-
n DISCUSSION OF CHANGES (m ,-) ITS: SECTION 5.7 - HIGH RADIATION AREA ADMINISTRATIVE A.1 Consistent with NUREG 1433, a phrase allowing individuals who are qualified in radiation protection procedures has been added. No change intent is made. Only qualified personnel will be permitted to perform the function. A.2 Consistent with NUREG 1433, the phrase "in accordance with approved emergency procedures" has been deleted. The intent of this footnote was to allow performance of assigned duties (e.g., surveys) to be performed without creating an undue administrative burden. A.3 Consistent with NUREG 1433 the words "or equal to" are added so that the intensity of radiation is covered. O 6 l l ,o ' HATCH UNIT 1 1 REVISION D
)
UNIT 1 NO SIGNIFICANT HAZARDS DETERMINATION O r b O O A
e NO SIGNIFICANT HAZARDS DETERMINATION , c 11S: SECTION 3.3.3.1 - POST-ACCIDENT MONITORING INSTRUMENTATION
. L.5 CHANGE 3 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 will revise the required number of channels for Reactor Vessel Water Level and Reactor Pressure Post Accident Monitoring (PAM). instruments. PAM instruments are not considered as an initiator for any accidents previously analyzed. Therefore, this change does not significantly increase the probability of a previously analyzed accident. : Also, this change does not further degrade the capability of the system to ; perform its required function under these circumstances, since there will be two physically independent channels required. Therefore, this change does not~ significantly increase the consequences of a previously analyzed accident.
- 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 modification to the plant. Therefore, it does not create the possibility of a new or different kind of accident from any ' accident previously evaluated.
- 3. Does this change involve a significant red'uction in a margin of safety? :
This change does not involve a significant reduction in a margin of safety since the monitors are not required to provide automatic response to any design basis accident. Sensors from separate reference legs would satisfy-the . requirement for physical independence of the channels. The current requirement of three channels , would necessarily involve 'one pair of channels that was not ' physically ' independent. This third channel- ' therefore would not necessarily satisfy the recommendations of Regulatory Guide 1.97. With the proposed revision, the required channels meet the minimum acceptable channels for compliance with Regulatory Guide 1.97, and provide for channels that can meet the physical-independence requirement. i O
- HATCH UNIT 1 5 REVISION D u
r P NO SIGNIFICANT HAZARDS DETERMINATION . ( ITS: SECTION 3.8.2 - AC SOURCES-SHUTDOWN L.1 CHANGE In accordance with the criteria set forth in 10 CFR 50.92, Georgia Power Company l 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 die:;e1 generators (DGs) are used to support mitigation of an accident; however, they are not considered the initiator of any previously analyzed accident. This change deletes the requirement for one of . the . two I > currently required DGs; that is, one DG will still be required. The DG requirement is being replaced with a requirement for. a qualified circuit between the offsite transmission network and the onsite Unit 1 Class IE AC l electrical power distribution subsystem (s) required by proposed LCO 3.8.8. This ensures that all of the Unit 1 buses needing AC power can be powered I from a Unit I source. Not only is this a more strict requirement, but the offsite circuits historically have a lower failure rate ~ than the DGs (which are very reliable). In addition, certain Unit 2 sources are also required to be '0PERABLE ~ to power Unit 2 loads required by Unit 1. . Therefore, due to these additional requirements which replace the one DG f] being deleted from this LCO, these changes do not involve a significant V increase in the consequences of any 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 mde of plant operation and does not involve physical modification to the plant. Therefore, it does not create the possibility of a new or diffarent kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
This change does not involve a significant reduction in a margin to safety since all AC buses needing power will continue to . be powered. In addition, in many cases, more AC Sources will be required than currently. Therefore, any reduction -in a margin of safety will be offset by the benefit gained from requiring additional, less failure prone offsite sources. O l HATCH UNIT 1 1 REVISION D
w
; c r :- .
- ]
i
-i NO-SIGNIFICANT' HAZARDS DETERMINATION 'i O 11s: s.e.e - 8at'<av ce u eaaaatTras -
L.1' CHANGE '
- Comment' number,not used, l 1 h
-t
-l i
l
.- j O .
s
?
r a 8 a
'l 1!
i
")q L
- HATCH UNIT .l -1 REVISION ; D'.'
- I
. - - , - . - , . . . . . . . - . . . . , . . - - . _ - . . ~ . ,
.. s, _ ,- , - - ,
- k. -.-
i O UNIT 2 IM. PROVED TECHNICAL SPECIFICATIONS l l 1 l l l l l O I l O ! l i l i
LCO Applicability l 3.0 < 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LC0 3.0.7. LC0 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LC0 3.0.5 and LC0 3.0.6. If the LC0 is met or is no longer applicable prior to expiration of the specified Completion Time (s), completion of the Required Action (s) is not required, unless otherwise stated. LC0 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if airected by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LC0 is not applicable. Action shall be initiated within I hour to place the unit, as applicable, in: (3 a. MODE 2 within 7 hours; V b. MODE 3 within 13 hours; and
- c. MODE 4 within 37 hours.
Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LC0 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, and 3. LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required (q,,) (continued) HATCH UNIT 2 3.0-1 REVISION A ! i
LCO Applicability 3.0 3.0 LC0 APPLICABILITY LCO 3.0.4 to comply with ACTIONS or that are part of a shutdown of the (continued) unit. Exceptions to this Specification are stated in the individual Specifications. These exceptions allow entry into MODES or other specified conditior,s in the Applicability when the associated ACTIONS to be entered allow unit operation in the MODE or other specified condition in the Applicability only for a limited period of time. LC0 3.0.5 Equipment removed from service or declare;l inoperable to comply with ACTIONS may be returned to ser vice under administrative control solely to perform sesting required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LC0 3.0.2 for the system k returned to service under administrative control to perform the required testing. LC0 3.0.6 When a supported system LCO is not met solely due to a support system LC0 not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LC0
. ACTIONS are required to be entered. This is an exception to LC0 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.5.10, " Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. (continued) HATCH UNIT 2 3.0-2 REVISION D
SDM 3.1.1 ( ') 3.1 REACTIVITY CONTROL SYSTEMS ! 'wd l 3.1.1 SHUTDOWN MARGIN (SDM) LCO 3.1.1 SDM shall be: ;
- a. 2 0.38% &/k, with the highest worth control rod analytically determined; or
- b. 2 0.28% M /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 rm 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. D. SDM not within limits D.1 Initiate action to Immediately in MODE 4. fully insert all insertable control rods. AND
- (continued)
RJ HATCH UNIT 2 3.1-1 REVISION A P
SDM 3.1.1 l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) D.2 Initiate action to I hour restore Unit 2 secondary containment to OPERABLE status. AND D.3 Initiate action to I hour restore one Unit 2 - standby gas treatment (SGT) subsystem to OPERABLE status. AND D.4 Initiate action to I hour restore isolation capability in each ld required Unit 2 secondary 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 D
SDM. 3.1.1 V ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. (continued) E.3 Initiate action to I hour restore Unit I secondary containment to OPERABLE status. AND E.4 Initiate action to I hour restore two SGT subsystems to OPERABLE status. AND E.5 Initiate action to I hour restore isolation capability in each d-required Unit I secondary containment penetration flow path () v not isolated. i n U HATCH UNIT 2 3.1-3 REVISION D
1
)
SDM l 3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY I 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 j i
control Rod Scram Times 3.1.4
.( SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 (continued) Prior to exceeding 40% RTP after each reactor shutdown 2 120 days SR 3.1.4.2 Verify, for a representative sample, each 120 days tested control rod scram time is within the cumulative limits of Table 3.1.4-1 with reactor steam operation in dome pressure 2 800 psig. MODE 1 SR 3.1.4.3 Verify each affected control rod scram time Prior to is .within the limits of Table 3.1.4-1 v:ith declaring any reactor steam dome pressure. control rod OPERABLE after 7h t work on control 'd rod or CRD System that could affect scram time SR 3.1.4.4 Verify each affected control _ rod scram time Prior to is within the limits of Table 3.1.4-l' with exceeding reactor steam dome pressure 2 800 psig. 40% RTP after work on control rod or CRD System that' could affect scram time x_)
HATCH UNIT 2 3.1-13 REVISION A
'l
1 Control Rod Scram Times j 3.1.4 I Table 3.1.4-1 (page 1 of 1) Control Rod Scram Times h j __________________________-----------NOTES------------------------------------
- 1. OPERABLE control rods with scram times not within the limits of this Table I are considered " slow." l
- 2. Enter applicable Conditions and Required Actions of LCO 3.1.3, " Control Rod OPERABILITY," for control rods with scram times > 7 seconds to notch position 06. These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered
_____________________________________" slow." SCRAM TIMES WHEN REACTOR STEAM DOME PRESSURE 2 800 psig (a)(b) NOTCH POSITION (seconds) 46 0.44 A 36 1.08 26 1.83 06 3.35 (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero. (b) When reactor steam dome pressure < 800 psig, established scram time limits apply. d O HATCH UNIT 2 3.1-14 REVISION D 4 \'
SRM Instrumentation ~ 3.3.1.2
-s (s) s_
SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.2.4 ------------------NOTES------------------
- 1. Not required to be met with less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies in the associated core quadrant.
- 2. Not required to be met during spiral unloading.
Verify count rate is 2: 3.0 cps with a 12 hours during signal to noise ratio 2: 2:1. CORE ALTERATIONS ILN.Q 24 hours A
-- J SR 3.3.1.2.5 Perform CHANNEL FUNCTIONAL TEST and 7 days determination of signal to noise ratio.
SR 3.3.1.2.6 ------------------NOTE------------------- Not required to be performed until 12 hours after IRMs on Range 2 or below. Perform CHANNEL FUNCTIONAL TEST and 31 days /hi determination of signal to noise ratio. SR 3.3.1.2.7 ------------------NOTES------------------
- 1. Neutron detectors are excluded.
- 2. Not reqrired to be performed until 12 hours after IRMs on Range 2 or below.
Perform CHANNEL CALIBRATION. 18 months
, ~\
x HATCH UNIT 2 3.3-13 REVISION D
i SRM Instrumentation , 3.3.1.2 l l Table 3.3.1.2 1 (page 1 of 1) Source Range Monitor Instrumentation APPLICABLE MODES OR OTHER REQUIRED SURVEILLANCE FUNCTION SPECIFIED CONDITIONS CHANNELS REQUIREMENTS
- 1. Source Range Monitor 2(*) 3 SR 3.3.1.2.1 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1.2.7 3,4 2 SR 3.3.1.2.3 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1.2.7 5 2(b)(c) SR 3.3.1.2.1 SR 3.3.1.2.2 SR 3.3.1.2.4 SR 3.3.1.2.5 SR 3.3.1.2.7 (a) With IRMs on Range 2 or below.
(b) Only one SRM channel is required to be OPERABLE during spiral offload or reload when the fueled re?'m includes only that SRM detector. (c) Special movable detectors may be used in place of SRMs if connected to normal ,$RM circuits. I l i l 1 1 l HATCH UNIT 2 3.3-14 REVISION A l
l Control Rod Block Instrumentation 3.3.2.1
<-s
/ \ ACTIONS (continued)
\ j CONDITION REQUIRED ACTION COMPLETION TIME E. One or more Reactor E.1 Suspend control rod Immediately Mode Switch -- Shutdown withdrawal .
Position channels inoperable. AND E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or . more fuel assemblies. 1 SURVEILLANCE REQUIREMENTS
-------------------------------------NOTES------------------------------------
- 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod
. -~g Block Function.
l l () 2. When an RBM channel is placed in an inoperable status solely for 3 ' performance of required Surveillances, entry into associated Conditions l and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability. j SURVEILLANCE FREQUENCY SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. 92 days i (continued) ( HATCH UNIT 2 3.3-17 REVISION A
Control Rod Block Instrumentation 3.3.2.1 l l SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.2.1.2 ------------------NOTE------------------- Not required to be performed until I hour after any control rod is withdrawn at
< 10% RTP in MODE 2.
[ Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.2.1.3 ------------------NOTE------------------- Not required to be performed until I hour after THERMAL POWER is < 10% RTP in MODE 1. Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.2.1.4 ------------------NOTE------------------- Neutron detectors are excluded. h Verify the RBM: 18 months
- a. Low Power Range - Upscale Function is not bypassed when THERMAL POWER is 2 29% and < 64% RTP.
- b. Intermediate Power Range - Upscale Function is not bypassed when THERMAL POWER is 2 64% and < 84% RTP.
- c. High Power Range - Upscale Function is not bypassed when THERMAL POWER is 2 84% RTP.
(continued) O HATCH UNIT 2 3.3-18 REVISION D
Remote Shutdown System 3.3.3.2 (O,
, SURVEILLANCE REQUIREMENTS
-------------------------------------NOTE-------------------------------------
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. SURVEILLANCE FREQUENCY SR 3.3.3.2.1 Perform CHANNEL CHECK for each required 31 days A instrumentation channel that is normally S energized. SR 3.3.3.2.2 Verify each required control circuit and 18 months I /bi transfer switch is capable of performing the intended function. () SR 3.3.3.2.3 Perform CHANNEL CALIBRATION for each required instrumentation channel. 18 months I[ki l k- J HATCH UNIT 2 3.3-27 REVISION D
i l EOC-RPT Instrumentation 3.3.4.I j l l 3.3 INSTRUMENTATION llk 3.3.4.I End of Cycle Recirculation Pump Trip (E0C-RPT) Instrumentation LC0 3.3.4.1 a. Two channels per trip system for each E0C-RPT instrumentation Function listed below shall be OPERABLE:
- 1. Turbine Stop Valve (TSV) -- Closure; and
- 2. Turbine Control Valve (TCV) Fast Closure, Trip Oil Pressure -- Low.
OB
- b. LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," limits for inoperable E0C-RPT as specified in the COLR are made applicable.
APPLICABILITY: THERMAL POWER 2: 30% RTP. ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each channel. g CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.I Restore channel to 72 hours inoperable. OPERABLE status. DE A.2 --------NOTE--------- Not applicable if inoperable channel is the result of an inoperable breaker. Place channel in 72 hours trip. (continued) O HATCH UNIT 2 3.3-28 REVISION A i
Primary Containment Isolation Instrumentation 3.3.6.1 b d SURVEILLANCE REQUIREMENTS
-------------------------------------NOTES------------------------------------
- 1. Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary 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. IE SURVEILLANCE FREQUENCY SR 3.3.6.1.1 Perform CHANNEL CHECK. 12 hours SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.6.1.3 Perform CHANNEL CALIBRATION. 92 days i
SR 3.3.6.1.4 Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.6.1.5 Perform CHANNEL CALIBRATION. 18 months SR 3.3.6.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months (continued) l l f% N HATCH UNIT 2 3.3-53 REVISION D
Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.6.1.7 ------------------NOTE------------------- Radiation detectors may be excluded. Verify the ISOLATION SYSTEM RESPONSE TIME 18 months on a. is within limits. STAGGERED TEST BASIS O t HATCH UNIT 2 3.3-54 REVISION A l l l
i i Secondary Containment Isolation Instrumentation 3.3.6.2
- l / 3.3 INSTRUMENTATION
. \ ,. 3.3.6.2 Secondary Containment Isolation Instrumentation LCO 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
,,\
(b AN.Q 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 I hour associated Completion associated zone (s). Time of Condition A or B not met. QB (continued)
!"~'\
HATCH UNIT 2 3.3-59 REVISION A ;
Secondary Containment Isolation Instrumentation ' 3.3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.1.2 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. OE 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 ineperable status solely for performance of required Surveillances, entry 'nto associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains isolation capioility. 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
ECCS - Shutdown 3.5.2 l\ ') 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.2 ECCS - Shutdown LC0 3.5.2 Two low pressure ECCS injection / spray subsystems shall be OPERABLE. 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 /'N 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 OPDRVs. subsystems inoperable. AND C.2 Restore one ECCS 4 hours injection / spray subsystem to OPERABLE status. (continued) i f HATCH UNIT 2 3.5-7 REVISION A- !
)
1 ECCS - Shutdown 3.5.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action C.2 D.1 Initiate action to Immediately and associated restore Unit 2 Completion Time not secondary containment met in MODE 4. to OPERABLE status. AND D.2 Initiate action to Immediately restore one Unit 2 standby go, treatment subsystem to OPERABLE status. AND D.3 Initiate action to Immediately restore isolation capability in each k required Unit 2 secondary containment penetration flow path not isolated. E. Required Action C.2 E.1 Initiate action to Immediately and associated restore Unit I and Completion Time not Unit 2 secondary met in MODE 5. containment to OPERABLE status. NQ E.2 Initiate action to Immediately restore three standby gas treatment subsystems to OPERABLE status. AND 1 (continued) i eI HATCH UNIT 2 3.5-8 REVISION D
ECCS - Shutdown 3.5.2 (- ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. (continued) E.3 Initiate action to Immediately. restore isolation capability ir, each lk ' required Unit I and Unit 2 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 p suppression pool water level is 2 146 inches. , i 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 l
- b. -----------------NOTE-----------------
Only one required CS subsystem may take credit for this option during OPDRVs. Condensate storage tank water level is 2 12 ft. , (continued) l %) HATCH UNIT 2 3.5-9 REVISION D
i 1 ECCS - Shutdown 3.5.2 SURVEILLANCE REQUIREMENTS (continued) I SURVEILLANCE FREQUENCY ! l 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 considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable. 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 NO. CORRESPONDING OF TO A REACTOR SYSTEM FLOW RATE PVMPS PRESSURE OF CS 2 4250 gpm 1 2 113 psig LPCI 2 7700 gpm 1 2 20 psig 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. O HATCH UNIT 2 3.5-10 REVISION A
I l Primary Containment Air Lock j 3.6.1.2 !
) SUR/EILLANCE REQUIREMENTS
{ SURVEILLANCE FREQUENCY l SR 3.6.1.2.1 ------------------NOTES----------------- .
- 1. An inoperable air lock door does not I
{ invalidate the previous successful 1 performance of the overall air lock leakage test.
- 2. Results shall be evaluated against acceptance criteria of SR 3.6.1.1.1 in accordance with 10 CFR 50, g j Appendix J, as modified by approved exemptions.
( ---_-______----_____----_____---_----____ Perform required primary containment air -----NOTE------ lock leakage rate testing in accordance SR 3.0.2 is not with 10 CFR 50, Appendix J, as modified applicable by approved exemptions. --------------- The acceptance criteria for air lock In accordance
~ testing are: with 10 CFR 50,
( Appendix J, as
\ a. Overall air lock leakage rate is modified by s 0.05 L when tested at 2 P., approved l
exemptions l ! b. For each door, leakage rate is j s 0.01 L. when the gap between the -l door seals is pressurized to i 2 10 psig for at least 15 minutes. i l SR 3.6.1.2.2 ------------------NOTE------------------- Only required to be performed upon entry or exit through the primary containment i air lock when the primary containment is de-inerted. Verify only one door in the primary 184 days I containment air lock can be opened at a ' time. I g - l HATCH UNIT 2 3.6-7 REVISION D
1 PCIVs 3.6.1.3-3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs) LC0 3.6.I.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE. [- APPLICABILITY: MODES 1, 2, and 3, When associated instrumentation is required to be OPERABLE per LC0 3.3.6.1, " Primary Containment Isolation Instrumentation." ACTIONS
NOTES------------------------------------
- 1. Penetration flow paths except for 18 inch purge valve 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 PCIVs.
- 4. Enter applicable Conditions and Required Actions of LC0 3.6.1.1, " Primary Containment," when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria.
CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE--------- A.1 Isolate the affected 4 hours except Only applicable to penetration flow path for main steam penetration flow paths by use of at least line with two PCIVs. one closed and de-activated automatic N EN2 One or more valve, closed manual valve, blind flange, 8 hours for main h penetration fiow paths u check valve with steam line with ona PCIV flow through the inoperable except due valve secured. I to leakage not within l limit. AND ' (continued) O HATCH UNIT 2 3.6-8 REVISION D
PCIVs 3.6.1.3 ( 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 the affected Once per 31 days penetration flow path for isolation is isolated. devices outside primary containment . bHQ Frior to entering MODE 2 or 3 from MODE 4 if primary p) containment was v de-inerted while in MODE 4, if not performed within the previous > 92 days, for isolation devices inside primary containment (continued) ' n HATCH UNIT 2 3.6-9 REVISION A
PCIVs 3.6.1.3 . l ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME j B. ---------NOTE--------- B.1 Isolate the affected I hour l Only applicable to penetration flow path l penetration flow paths by use of at least with two PCIVs. one closed and de- 3 activated automatic ' valve, closed manual One or more valve, or blind i penetration flow paths flange. I with two PCIVs inoperable except due ; to leakage not within limit. C. ---------NOTE---- ---- C.1 Isolate the affected 4 hours except ; Only applicable to penetration flow path for excess flow ; penetration flow paths by use of at least check valve ; with only one PCIV. one closed and de- (EFCV) line 1 activated automatic 1 valve, closed manual One or more valve, or blind A_tiQ ! penetration flow paths flange. with one PCIV 12 hours for inoperable except due EFCV line A to leakeage not within
- limits. J AND i
C.2 --------NOTE-------- Valves and blind flanges in high radiation areas may be verified by use of administrative means. Verify the affected Once per 31 days penetration flow path is isolated. (continued) O HATCH UNIT 2 3.6-10 REVISION D l
PCIVs 3.6.1.3 ,a (v) SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.1.3.3 ------------------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 PCIVs that are open under administrative controls.
Verify each primary containment manual Prior to isolation valve and blind flange that is entering located inside primary containment and is MODE 2 or 3 required to be closed during accident from MODE 4 if conditions is closed. primary containment was de-inerted while in MODE 4, if not
~
performed / 'T within the (_) previous 92 days SR 3.6.1.3.4 Verify continuity of the traversing incore 31 days probe (TIP) shear isolation valve explosive charge. i SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated and each automatic PCIV, except with the jh for MSIVs, is within limits. Inservice Testing Program i i (continued) ) RATCH UNIT 2 3.6-13 REVISION D
~
PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.6.1.3.6 Verify the isolation time of each MSIV is In accordance 2 3 seconds and s 5 seconds, with the Inservice d Testing Program SR 3.6.1.3.7 Verify each automatic PCIV, excluding EFCVs, 18 months Id actuates to the isolation position on an - actual or simulated isolation signal. SR 3.6.1.3.8 Verify each reactor instrumentation line 18 months lA EFCV actuates to restrict flow to within limits. SR 3.6.1.3.9 Remove and test the explosive squib from 18 months on a l each shear isolation valve of the TIP STAGGERED TEST System. BASIS SR 3.6.1.3.10 Verify the combined leakage rate for all ------NOTE----- SR 3.0.2 is not 1A secondary containment bypass leakage paths is s 0.009 L, when pressurized to applicable. a p,, _ _ _ _ - - ________ In accordance with 10 CFR 50, Appendix J, as modified by approved exemptions (continued) O HATCH UNIT 2 3.6-14 REVISION D
r PCIVs 3.6.1.3 a i c Iv). ' ' SURVEILLANCE REQUIREMENTS (continued) - SURVEILLANCE . FREQUENCY SR 3.6.1.3.11 Verify leakage rate through each MSIV is s 100 scfh, and a combined maximum
------NOTE-----
SR 3.0.2 is not Ik. pathway leakage s 250 scfh for all four applicable. main steam lines, when tested at --------------- . 2: 28.8 psig. In accordance However, the leakage rate acceptance with 10 CFR 50,. criteria for the first test following Appendix J, as discovery of leakage through an MSIV not modified by meeting the 100 scfh limit, shall be approved s 11.5 scfh for that MSIV. exemptions SR 3.6.1.3.12 Replace the valve seat of each 18 inch purge valve having a resilient material 18 months Ik
~
seat. O SR 3.6.1.3.13 Cycle each 18 inch excess flow isolation damper to the fully closed and fully open 18 months Ik - position. 7 i H
.. I L
L i i HATCH UNIT 2 3.6-15 REVISION D:- I J
Drywell Pr:ssure 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell Pressure LCO 3.6.1.4 Drywell pressure shall be s 1.75 psig. l APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell pressure not A.1 Restore drywell I hour within limit. pressure to within i limit. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l SR 3.6.1.4.1 Verify drywell pressure is within limit. 12 hours O HATCH UNIT 2 3.6-16 REVISION [
~ Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.8-IT 3.6 CONTAINMENT SYSTEMS
.v '
3.6.1.6 Suppression Chamber-to-Drywell Vacuum Breakers LC0 3.6.1.8 Ten suppression chamber-to-drywell vacuum breakers shall be OPERABLE for opening. 8hD Twelve suppression chamber-to-drywell vacuum breakers shall be closed, except when performing their intended function. APPLICABILITY: MODES 1, 2, and 3. ACTIONS , CONDITION REQUIRED ACTION COMPLETION TIME r A. One. required A.1 Restore one vacuum 72 hours suppression chamber- breaker to OPERABLE to-drywell vacuum status. O breaker inoperable for opening.
-B. One suppression B.1 Close the open vacuum 2 hours chamber-to-drywell breaker.
vacuum breaker not' < closed. C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. AND 4 C.2 Be in MODE 4. 36 hours i, r HATCH UNIT 2 3.6-23 REVISION A 4 l J
Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.8.1 ------------------NOTE-------------------- Not required to be met for vacuum breakers that are open during Surveillances. Verify each vacuum breaker is closed. 14 days SR 3.6.1.8.2 Perform a functional test of each required 31 days vacuum breaker. MQ Within 12 hours after any discharge of A steam to the [Q suppression chamber from the S/RVs SR 3.6.1.8.3 Verify the opening setpoint of each 18 months required vacuum breaker is s 0.5 psid. O HATCH UNIT 2 3.6-24 REVISION D
Secondary Containment-0perating 3.6.4.1 ( ,
) SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify all Unit I and Unit 2 secondary 31 days A containment equipment hatches are closed and sealed.
A\ SR 3.6.4.1.2 Verify each Unit 1 and Unit 2 secondary 31 days containment access 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 Verify each Unit 2 standby gas treatment 18 months on a (SGT) subsystem will draw down the Unit 2 STAGGERED TEST secondary containment to 2 0.25 inch of BASIS vacuum water gauge in s 120 seconds. (GD SR 3.6.4.1.4 Verify two SGT subsystems will draw down 18 months on a the Unit I secondary containment to STAGGERED TEST a 0.25 inch of vacuum water gauge in BASIS
$ 120 seconds.
SR 3.6.4.1.5 Verify each Unit 2 SGT subsystem can 18 months on a maintain 2 0.25 inch of vacuum water STAGGERED TEST gauge in the Unit 2 secondary containment BASIS for 1 hour at. a flow rate s 4000 cfm. SR 3.6.4.1.6 Verify two SGT subsystems can maintain 18 months on a 2 0.25 inch of vacuum water gauge in the STAGGERED TEST Unit I secondary containment at a flow BASIS rate s 4000 cfm for each subsystem. (_) HATCH UNIT 2 3.6-39 REVISION D
Secondary Containment-0PDRVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS 3.6.4.2 Secondary Containment-0PDRVs h LCO 3.6.4.2 The Unit 2 secondary containment shall be OPERABLE. APPLICABILITY: During operations with a potential for draining the reactor vessel (0PDRVs). ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Unit 2 secondary A.I Initiate action to Immediately containment suspend OPDRVs. inoperable. O O' HATCH UNIT 2 3.6-40 REVISION A 4
Secondary Containment-OPDRVs 3.6.4.2' I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l SR 3.6.4.2.1 Verify all Unit 2 secondary containment 31 days A
- equipment hatches are closed and sealed. I LEG SR 3.6.4.2.2 Verify each Unit 2 secondary containment 31 days access door is. closed, except when the access opening is being used for entry and exit, then at least one door shall be closed.
SR 3.6.4.2.3 Verify each Unit 2 standby gas treatment 18 months on a (SGT) subsystem will draw down the STAGGERED TEST l Unit 2 secondary containment to BASIS j 2 0.25 inch of vacuum water gauge , in s 120 seconds. J o l I O SR 3.6.4.2.4 Verify each Unit 2 SGT subsystem can 18 months on a maintain 2 0.25 inch of vacuum water STAGGERED TEST gauge in the Unit 2 secondary containment BASIS 1 for I hour at a flow rate s 4000 cfm. l I HATCH UNIT 2 3.6-41' REVISION D u--- -- ----- ----.---- ------_ --_-- --- --- -- -----------------------_ ---__-------_ ---- _----- _ -- ---- _ ------- ---- _
Secondary Containment-Refueling l 3.6.4.3 l 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Secondary Containment - Refueling LC0 3.6.4.3 The Unit I secondary containment shall be OPERABLE. APPLICABILITY: During movement of irradiated fuel assemblies in the Unit I secondary containment, During CORE ALTERATIONS. q ACTIONS
\
CONDITION REQUIRED ACTION COMPLETION TIME i A. Unit I secondary A.1 --------NOTE-------- j containment LC0 3.0.3 is not i inoperable. applicable. l Suspend movement of Immediately irradiated fuel assemblies in the Unit I secondary containment. bid l A.2 Suspend CORE Immediately ALTERATIONS. > 4 O HATCH UNIT 2 3.6-42 REVISION A 1 l C__________________--__.---__ - . . - - _ . _ _ _ _ _ _ . - - - - - - - - - - - - - - - - - - - - - _ - - - _ - - - - - - - - - - - - - - - - - - - - _ - - - - - _ - - - - - -
I Secondary Containment-Refueling 3.6.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 1 SR 3.6.4.3.1 Verify all Unit I secondary containment 31 days equipment hatches are closed and sealed.
~
I SR 3.6.4.3.2 Verify each Unit I secondary containment 31 days access door is closed, except when the access opening is being used for entry-and exit, then at least one door shall be closed. SR 3.6.4.3.3 Verify two standby gas treatment 18 months on a (SGT) subsystems will-draw down the STAGGERED TEST-Unit I secondary containment to BASIS 2 0.25 inch of vicuum water gauge in s 100 seconds. O SR 3.6.4.3.4 Verify two SGT subsystems can maintain 18 months on a a 0.25 inch of vacuum water gauge in the STAGGERED TEST Unit 1 secondary containment for I hour- BASIS at a flow rate s 4000 cfm for each
-subsystem.
l l O HATCH UNIT 2 3.6-43 REVISION D i
P SCIVs-0perating 3.6.4.4 3.6 CONTAINMENT SYSTEMS 3.6.4.4 Secondary Containment Isolation Valves (SCIVs) - Operating LC0 3.6.4.4 Each Unit I and Unit 2 SCIV shall be OPERABLE. APPLICABILITY: MODES I, 2, and 3. 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 i. I 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 k, i i valve, or blind l flange.
.l ANQ (continued)
O HATCH UNIT 2 3.6-44 REVISION D
SCIVs--Operating 3.6.4.4 ( )j 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 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 h ( valve, or blind
\ flange.
C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. AND. C.2 Be in MODE 4. 36 hours . fh V 1ATCH UNIT 2. 3.6-45 REVISION D
SCIVs--Operating 3.6.4.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.4.1 ------------------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 Unit I and Unit 2 secondary 31 days containment isolation manual valve and blind flange that is required to be closed during accident conditions is closed. SR 3.6.4.4.2 Verify the isolation time of each 92 days required power operated and each automatic Unit I and Unit 2 SCIV is within limits. SR 3.6.4.4.3 Verify each automatic Unit I and Unit 2 18 months SCIV actuates to the isolation position on an actual or simulated actuation signal. Ol. HATCH UNIT 2 3.6-46 REVISION A i 1
r' l SCIVs--0PDRVs i 3.6.4.5 ! l f% i 3.6 CONTAINMENT SYSTEMS (J 3.6.4.5 Secondary Containment Isolation Valves (SCIVs) -- OPDRVs LC0 3.6.4.5 Each Unit 2 SCIV shall be OPERABLE. 1 APPLICABILITY: During operations with a potential for draining the reactor-vessel (0PDRVs). ACTIONS
-------------------------------------NOTES------------------------------------
I. 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.
(- (m-) CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Unit 2 A.I 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 /hs valve, closed manual valve, or blind flange. AND (continued)
\~ J HATCH UNIT 2- 3.6-47 REVISION D i
SCIVs-0PDRVs 3.6.4.5 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 the affected Once per 31 days penetration flow path is isolated. B. One or more Unit 2 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 k valve, or blind fl ange. h C. Required Action and C.1 Initiate action to Immediately associated Completion suspend OPDRVs. Time not met. O HATCH UNIT 2 3.6-48 REVISION D
SCIVs-0PDRVs 3.6.4.5 iv SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.5.1 ------------------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 Unit 2 secondary containment 31 days isolation manual valve and blind flange that is required to be closed during accident conditions is closed. SR 3.6.4.5.2 Verify the isolation time of each 92 days required power operated and each f- automatic Unit 2 SCIV is within limits. SR 3.6.4.5.3 Verify each automatic Unit 2 SCIV 18 months actuates to the isolation position on an actual or simulated actuation signal. v HATCH UNIT 2 3.6-49 REVISION A \;
SCIVs-Refueling 3.6.4.6 3.6 CONTAINMENT SYSTEMS 3.6.4.6 Secondary Containment Isolation Valves (SCIVs) - Refueling LC0 3.6.4.6 Each Unit 1 SCIV shall be OPERABLE. APPLICABILITY: During movement of irradiated fuel assemblies in the Unit I secondary containment, During CORE ALTERATIONS. ACTIONS
-------------------------------------NOTES------------------------------------
- 1. Penetration flow paths may be unisolated intermittently under administrative controls.
l
- 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 e A. One or more Unit 1 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 B valve, closed manual l valve, or blind flange. MQ (continued) HATCH UNIT 2 3.6-50 REVISION D
SCIVs--Refueling ! E3.6.4.6 (")\ (, 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 the affected Once per 31 days penetration flow path is isolated. B. One or more Unit 1 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 /h
,-. valve, closed manual valve, or blind
(\-} fl ange. C. Required Action and C.1 --------NOTE-------- associated Completion LC0 3.0.3 is not Time not met. applicable. Suspend movement of Immediately irradiated fuel assemblies in the Unit I secondary containment. AND C.2 Suspend CORE- 'Immediately ALTERATIONS. /; L] HATCH UNIT 2 3.6-51 REVISION D
SCIVs-Refueling 3.6.4.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.6.1 ------------------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 Unit I secondary containment 31 days isolation manual valve and blind flange that is required to be closed during accident conditions is closed. SR 3.6.4.6.2 Verify the isolation time of each 92 days required power operated and each automatic Unit 1 SCIV is within limits. SR 3.6.4.6.3 Verify each automatic Unit 1 SCIV 18 months actuates to the isolation position on an actual or simulated actuation signal.
\
O HATCH UNIT 2 3.6-52 REVISION A
AC Sources - Operating 3'8.1' _ SURVEILLANCE REQUIREMENTS _________________________------------NOTE------------------------------------- SR 3.8.1.1 through SR 3.8.1.18 are applicable only to the Unit 2 AC sources. 4 SR 3.8.1.19 is applicable only to the Unit 1 AC sources. M SURVEILLANCE FREQUENCY ! SR 3.8.1.1 Verify correct breaker alignment and 7 days indicated power availability for each required offsite circuit. SR 3.8.1.2 -------------------NOTES-------------------
- 1. Performance of SR 3.8.1.5 satisfies this SR.
- 2. All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.
- 3. A modified DG start involving idling d and gradual acceleration to synchronous speed may be used for this SR as recommended by the manufacturer.
When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.5.a must be met. ,
- 4. For the swing DG, a single test will satisfy this Surveillance for both units, using the starting circuitry of.
Unit 2 and synchronized to 4160 V bus 2F for one periodic test, and the starting circuitry of Unit 1 and synchronized to 4160 V bus'1F during the next periodic test. ,
- 5. DG loadings may include gradual loading as recommended by the manufacturer. ,
- 6. Starting transients above the upper voltage limit do not invalidate this test.
(continued) HATCH UNIT 2 3.8-7 REVISION D L
AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) _ SURVEILLANCE FREQUENCY SR 3.8.1.2 NOTES (continued)
- 7. Momentary transients outside the load range do not invalidate this test.
- 8. This Surveillance shall be conducted on only one DG at a time.
Verify each DG: 31 days
- a. Starts from standby conditions and achieves steady state voltage 2 3740 V and s 4243 V and frequency a 58.8 Hz and s 61.2 Hz; and
- b. Operates for 2 60 minutes at a load 2 1710 kW and s 2000 kW.
SR 3.8.1.3 Verify each day tank contains 1900 gallons 31 days of fuel oil. SR 3.8.1.4 Check for and remove accumulated water from 184 days each day tank. , (continued) 1 O' A HATCH UNIT 2 3.8-8 REVISION
AC Sources - Operating 3.8.1 (,,) - SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.5 -------------------NOTES-------------------
- 1. All DG starts may be preceded by an engine prelube period.
- 2. DG loadings may include gradual loading as recommended by the manufacturer.
- 3. Momentary load transients outside the load range do not invalidate this test.
- 4. This Surveillance shall be conducted l on only one DG at a time.
- 5. For the swing DG, a single test will l '
satisfy this Surveillance for both units, using the starting circuitry of , Unit 2 and synchronized to 4160 V bus 2F for one periodic test and the O k,/ starting circuitry of Unit I and synchronized to 4160 V bus IF during the next periodic test. Verify each DG:
- a. Starts from standby conditions and achieves, in s 12 seconds, voltage 2 3740 V and frequency 2 58.8 Hz and 184 days after steady state conditions are reached, maintains voltage a 3740 V and s 4243 V and frequency 2 58.8 Hz and s 61.2 Hz; and
- b. Operates for a 60 minutes at a load 2 2764 kW and s 2825 kW for DG 2A, a 2360 kW and s 2425 kW for DG 1B, and a 2742 kW and s 2825 kW for DG 2C.
(continued) O V HATCH UNIT 2 3.8-9 REVISION [
\
AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.6 ------------------NOTE--------------------- This Surveillance shall not be performed in MODE 1 or 2. However, credit may be taken for unplanned events that satisfy this SR. Verify automatic and manual transfer of 18 months unit power supply from the normal offsite circuit to the alternate offsite circuit. SR 3.8.1.7 ------------------NOTES--------------------
- 1. This Surveillance shall not be performed in MODE 1 or 2, except for the swing DG. For the swing DG, this Surveillance shall not be performed in
!!0DE 1 or 2 using the Unit 2 controls.
Credit may be taken for unplanned events that satisfy this SR.
- 2. For the swing DG, a single test at the specified Frequency will satisfy this Surveillance for both units.
Verify each DG rejects a load greater than 18 months or equal to its associated single largest I b post-accident load, and
- a. Following load rejection, the frequency is s 65.5 Hz; and
- b. Within 3 seconds following load rejection, the voltage is 2: 3740 V and s 4580 V.
(continued) O HATCH UNIT 2 3.8-10 REVISION D
AC Sources - Operating 3.8.1 m I) v SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY l l l SR 3.8.1.8 ------------------NOTES-------------------- l l
- 1. This Surveillance shall not be performed in MODE 1 or 2, except for l the swing DG. For the swing DG, this l Surveillance shall not be performed in MODE 1 or 2 using the Unit 2 controls.
Credit may be taken for unplanned events that satisfy this SR.
- 2. If grid conditions do not permit, the I powei factor itait is not r6 quired to -
be met. Under this condition, the power factor shall be maintained as close to the limit as practicable.
- 3. For the swing DG, a single test at the l D specified frequency will satisfy this b ___________________________________________
Surveillance for both units. Verify each DG operating at a power factor 18 months s 0.88 does not trip and voltage is maintained s 4800 V during and following a load rejection of 2: 2775 kW. (continued) q Q.) HATCH UNIT 2 3.8-11 REVISION D _ _ _ _ _ . - . _ _ _ _ - _ _ _ _ _ _ _ _ - - 1
AC Sources - Operating ! 3.8.1 J SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.9 -------------------NOTES-------------------
- 1. All DG starts may be preceded by en engine prelube period.
- 2. This Surveillance shall not be performed in MODE 1, 2, or 3.
However, credit may be taken for unplanned events that satisfy this SR. Verify on an actual or simulated loss of 18 months offsite power signal:
- a. De-energization of amergency buses;
- b. Load shedding from emergency buses; and
- c. DG auto-starts from standby condition and:
- 1. energizes permanently connected loads in s 12 seconds,
- 2. energizes auto-connected shutdown loads through automatic load sequence timing devices,
- 3. maintains steady state voltage 2 3740 V and s 4243 V,
- 4. maintains steady state frequency 2 58.8 Hz and s 61.2 Hz, and
- 5. supplies permanently connected and auto-connected shutdown loads for 2 5 minutes.
(continued) O HATCH UNIT 2 3.8-12 REVISION A I
l AC Sources - Operating 3.8.1 t ( SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.17 -------------------NOTES-------------------
- 1. All DG starts may be preceded by an engine prelube period.
- 2. This Surveillance shall not be performed in MODE 1, 2, or 3.
However, credit may be taken for unplanned events that Latisfy this SR. Verify, on an actual or simulated loss of 18 months offsite power signal in conjunction with an actual or simulated ECCS initiation signal:
- a. De-energization of emergency buses;
- b. Load shedding from emergency buses; and
- c. DG auto-starts from standby condition and:
pN R
- 1. energizes permanently connected loads in s 12 seconds,
- 2. energizes auto-connected emergency loads through automatic 1c,ad sequence timing devices,
- 3. achieves steady state voltage 2 3740 V and s 4243 V,
^
- 4. achieves steady state frequency 2 58.8 Hz and s 61.2 Hz, and
- 5. supplies permanently conr.ccted and auto-connected emergency loads for a 5 minutes.
1 (continued)
./
HATCH UNIT 2 3.8-17 REVISION A , I l l
AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.8.1.18 -------------------NOTE-------------------- All DG starts may be preceded by an engine prelube period. Verify, when started simultaneously from 10 years standby condition, the Unit 2 DGs achieve, in s 12 secords, voltage 2 3740 V and frequency 2 58.8 Hz. SR 3.8.1.19 For required Unit 1 AC Sources, the SRs of In accordance Unit 1 Technical Specifications are with applicable applicable, except SR 3.8.1.6, SR 3.8.1.10, SR 3.8.1.11, SR 3.8.1.15, SR 3.8.1.17, and SRs g SR 3.8.1.18. O O HATCH UNIT 2 3.8-18 REVISION D
AC Sources - Operating 3.8.1 O l t O Y
\
.O , c
-HATCH UNIT 2 3.8-19 REVISION [
I
AC Sources - Shutdown 3.8.2 3.8 ELECTRICAL POWER SYSTEMS 3.8.2 AC Sources - Shutdown h 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 2 Class IE AC electrical I k ,
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 Unit 2 Class IE AC electrical power distribution subsystem (s) required by LC0 3.8.8;
- c. One qualified circuit connected between the offsite transmission network and the onsite Unit 1 Class IE AC electrical power distribution subsystem (s) needed to support the Unit 1 equipment required to be OPERABLE by LC0 3.6.4.9, " Standby Gas Treatment (SGT)
System-Refueling," LC0 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 I equipment required to be OPERABLE by LC0 91 1
3.6.4.9, LC0 3.7.4, and LC0 3.7.5. APPLICABILITY: MODES 4 and 5, During movement of irradiated fuel assemblies in the Unit I secondary containment. I 4 i e l HATCH UNIT 2 3.8-20 REVISION D t =
AC Sources - Shutdown 3.8.2 m ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required ------------NOTE------------- lA offsite circuit (s) Enter applicable Condition 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. lJ 0.8 A.2.1 Suspend CORE Immediately ALTERATIONS. (l L/ AND A.2.2 Suspend movement of Immediately irradiated fuel assemblies in the Unit I secondary containment. AND A.2.3 Initiate action to Immediately suspend operations with a potential for draining the rr ator vessel (0PDRVs). AND A.2.4 Initiate' action to Immediately restore required offsite power . circuit (s) to OPERABLE status. f (continued) r )
\d
' HATCH UNIT 2 3.8- M R) REVISION D
AC Sources - Shutdown 3.8.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. One or more required B.1 Suspend CORE Immediately I DG(s) inoperable. ALTERATIONS. AND B.2 Suspend movement of Immediately irradiated fuel assemblies in Unit I secondary containment. > AND i B.3 Initiate action to Immediately suspend OPDRVs. AND B.4 Initiate action to Immediately restore required 1 DG(s) to OPERABLE status. 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 i 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 LCO 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) g HATCH UNIT 2 3.8-22 REVISION D
.i
'l Battery Cell Parameters )
3.8.6 Table 3.8.5-1 (page 1 of 2)
"- Battery Cell Parameter Requirements- 1 i
CATEGORY A: CATEGORY 8: CATEGORY C: ) LIMITS FOR EACH LIMITS FOR EACH LIMITS .. 1 DESIGNATED PILOT 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) , s 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 BlQ average of all. s connected cells Average of all connected cells AND.
> 1.205 ,
Average of'all connected cells , a 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 not required, however, when on float charge battery charging is < 1 amp 1 i 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. i maximum of 7 days. . bihen charging current is used to satisfy specific ; A gravity requirements, specific gravity of each connected cell shall be i V measured prior to expiration of the 7 day allowance. HATCH UNIT 2 3.8-39 REVISION D , 1
Distribution Systems - Operating 3.8.7 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Distribution Systems - Operating LCO 3.8.7 The following AC and DC electrical power distribution subsystems shall be OPERABLE:
- a. Unit 2 AC and DC electrical power distribution !
subsystems comprised of: I
- 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 LC0 3.6.4.7, " Standby Gas Treatment (SGT)
System-Operating," 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."
W 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. l subsystems inoperable. (continued) O HATCH UNIT 2 3.8-40 A REVISION [
I RHR --High Water Level 3.9.7
,~
-( ) ,
. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
)
l B. (continued) 8.3 Initiate action to Immediately I restore two standby gas treatment subsystems to OPERABLE status. AND B.4 Initiate action to Immediately restore isolation capability in each required Unit I 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 7-,;s i operation. by an alternate reactor coolant A- method. circulation AND One per 12 hours thereafter N A..D C.2 Monitor reactor Once per hour coolant temperature.
\ _)
HATCH UNIT 2 3.9-11 REVISION D
RHR -High Water Level 3.9.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify one RHR shutdown cooling subsystem 12 hours is operating. O O HATCH UNIT 2 3.9-12 REVISION A
i i RHR - Low Water Level 3.9.8 3.9 REFUELING OPERATIONS ( _. / 3.9.8 Residual Heat Removal (RHR) - Low Water Level I i LCO 3.9.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and or,e 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 l
/ A. One or two required A.1 Verify an alternate 1 hour
- k. 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 l
i i B. Required Action and B.1 Initiate action to Immediately associated Completion restore Unit 1 Time of Condition A secondary containment i not met. to OPERABLE status. AND B.2 Initiate action to Immediately restore two standby gas treatment subsystems to OPERABLE status. SHD (continued) v HATCH UNIT 2 3.9-13 REVISION A
RHR - Low Water Level 3.9.8 ; ACTIONS , CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.3 Initiate action t'o Immediately restore isolation capability in each l[ required Unit I 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. SURVEILLANCE REQUIREMENTS l l SURVEILLANCE FREQUENCY l SR 3.9.8.1 Verify one RHR shutdown cooling subsystem 12 hours
- is operating.
l ! O HATCH UNIT 2 3.9-14 REVISION D
Inservice Leak and Hydrostatic Testing Operation 3.10.1 ( ) 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 considered not to be in MODE 3; and the requirements of A LCO 3.4.8, " Residual Heat Removal (RHR) Shutdown Cooling I dh System - Cold Shutdown," may be suspended, to allow performance of an-inservice leak or hydrostatic test provided the following MODE 3 LCOs are met:
- a. LC0 3.3.6.2, " Secondary Containment Isolation Instrumentation," Functions 1, 3, and 4 of Table 3.3.6.2-1;
- b. LC0 3.6.4.1, " Secondary Containment - Operating";
- c. LC0 3.6.4.4, " Secondary Containment Isolation Valves I (SCIVs) -
Operating"; and l
- d. LCO 3.6.4.7, " Standby Gas Treatment (SGT) System -
Operating." fT O APPLICABILITY: MODE 4 with average reactor coolant temperature > 212 F. HATCH UNIT 2 3.10-1 REVISION D
Inservice Leak and Hydrostatic Testing Operation 3.10.1 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.I --------NOTE--------- above requirements not Required Actions to met. be in MODE 4 include reducing average reactor coolant temperature to s 212 F. Enter the applicable Immediately Condition of the affected LCO. 08 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. i \ O i HATCH UNIT 2 3.10-2 REVISION A l I 2
Organization 5.2 W Q 5.2 ~ Organization 5.2.2 Unit' Staff
- a. (continued) the required PMs shall be assigned to each reactor containing fuel.
- b. At least one licensed Reactor Operator (RO) shall be present in the control room for each unit that contains fuel in the reactor. In addition, while the unit is in MODE 1, 2, or 3,:
at least one licensed Senior Reactor Operator (SRO) shall be present in the control room,
- c. The minimum shift crew composition shall be in accordance with 10 CFR 50.54(m)(2)(i). Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(1) and 5.2.2.a for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on duty shift-crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
A d. An. individual qualified to implement radiation protection procedures shall be on site when fuel is in the reactor. '() The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is take.n to fill the required position,
- e. Administrative procedures sha'll-be developed.and implemented to limit the working hours of unit' staff who perform safety related functions (e.g., licensed and non-licensed .A-operations ~ personnel, health physics technicians,~ key .R-maintenance personnel, etc.).
Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a nominal 40 hour week while the unit is operating. However, in the event that unforeseen l problems require substantial amounts of overtime to be used,- or during extended periods of shutdown for refueling, major: maintenance, or major plant modification, on a temporary basis the following' guidelines shall be.followed:
- 1. An individual should not be permitted to work 'more than 16 hours straight, excluding shift turnover time;-
-(continued)
QJ HATCH UNIT 2 5.0-3 REVISION D -
Organization 5.2 5.2 Organization h 5.2.2 Unit Staff (continued)
- e. (continued)
- 2. An individual should not be permitted to work more than 16 hours in any 24 hour period, nor more than 24 hours in any 48 hour period, nor more than 72 hours in any 7 day period, all excluding shift turnover time;
- 3. A break of at least 8 hours should be allowed between work periods, including shift turnover time;
- 4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
Any deviation from the above guidelines shall be authorized by the AGM-P0, Assistant General Manager-Plant Support (AGM-PS), or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the AGM-P0, AGM-PS, or designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.
- f. The Operations Manager shall hold an active or inactive SR0 license,
- g. The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor (SS) in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. In addition, the STA shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.
O HATCH UNIT 2 5.0-4 REVISION A l
Programs and Manuals 5.5 4 m c.5 Programs and Manuals (continued) (v) 5.5.4 Radioactive Effluent Controls Proaram This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
- a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation, including surveillance tests and setpoint determination, in accordance with the methodology in the ODCM;
- b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, A conforming to 10 times the concentrations stated in 10 CFR 20, Appendix B (to paragraphs 20.1001-20.2401),
I [A Table 2, Column 2;
- c. Monitoring, sampling, and analysis of radioactive liquid and f]
V gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;
- d. Limitations on the annual ano quarterly doses or dose commitment'to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I;
- e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year, in accordance with the methodology and parameters in the ODCM, at least every 31 days;
- f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2Y, of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I; g) (continued) j V I HATCH UNIT 2 5.0-9 REVISION D
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Procram (continued) 9 Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary as follows:
- 1) For noble gases, less than or equal to a dose rate of 500 inrem/ year to the total body and less than or equal to a dose rate of 3000 mrem / year to the skin, and ; -
- 2) For Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days, less tLn or equal to a dose rate of 1500 mrem / year to any organ;
- h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;
- i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and J. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.
5.5.5 Comoonent Cyclic or Transient limit This program provides controls to track FSAR Section 5.2, cyclic and transient occurrences, to ensure that reactor coolant pressure boundary components are maintained within the design limits. 5.5.6 Inservice Testina Procram This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports,
- a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda are as follows:
k I (continued) HATCH UNIT 2 5.0-10 REVISION D
Programs and Manuals - 5.5 5.5 Programs and Manuals 5.5.6 Inservice Testina Proaram (continued) t ASME Boiler and Pressure Vessel Code and Applicable Required Frequencies for Performing Inservice Addenda Terminology for Inservice Testina Activities Testina Activities . Weekly At least once.per 7 days ! Monthly At least'once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days. Yearly or annually At least once per 366 days
- b. The provisions of SR 3.0.2 are applicable to the frequencies l' /ks for performing inservice testing activities;
- c. The provisions of SR 3.0.3 are applicable to inservice 'l/hii testing activities; and ,
- d. Nothing in the ASME Boiler and Pressure Vessel ' Code 'shall be 1[h construed to supersede the requirements of any Technical ,
Specification. O , L F f ,+, .1 S' ~'- (continued) HATCH UNIT 2 .5.0-10A. REVISION D j l j
i Reporting Requirements 5.6 ; (( ); 5.6fReporting Requirements (continued) i 5.6.5- CORE OPERATING LIMITS REPORT (COLR)
- a. Core operating limits shall be established prior to each reload cycle, or )rior to any remaining portion of a reload
. cycle, and shall se documented in the COLR for the i following:
s- ;
~
- 1) Control Rod Block Instrumentation - Rod Block Monitor for Specification 3.3.2.1.
- 2) The Average Planar Linear Heat Generation Rate for SpecificTtion 3.2.1.
- 3) The Miniuum Critical Power Ratio for Specifications .
3.2.2 and 3.3.2.1.
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following :
documents:
- 1) NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel," (applicable amendment specified in
'p \j the COLR).
- 2) " Safety Evaluation by the Office of Nuclear Reactor :
Regulation Supporting Amendment Nos.151 and 89 to Facility Operating Licenses DPR-57 and NPF-5," dated , January 22, 1988. c.. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, i core thermal hydraulic limits, Energency Core Cooling A i Systems l(ECCS) limits, nuclear 1,mits such as SDM, transient M j analysis limits and accident analysis limits) of the safety analysis are met.
- d. The COLR, including any mid-cycle revisions or supplements, '
shall be provided upon issuance for each reload cycle to the I NRC. l kM 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS ! REPORT (PTLR) l l
- a. RCS pressure and temperature limits for heatup, cooldown, l low temperature operation, criticality, pd hydrostatic l testing as well as heatup and cooldown ra tes shall be- )
i
% (continued)
Lf \ HATCH ONIT 2 5.0-19 REVISION D
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
- a. (continued) established and documented in the PTLR for LC0 3.4.9, "RCS Pressure and Temperature (P/T) Limits."
- b. The analytical methods used to determine the RCS pressure and temperature limits shall be determined in accordance with Regulatory Guide 1.99,
- c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluency period and for any revision or supplement thereto.
5.6.7 Post Accident Monitorino (PAM) Instrumentation Report When a report is required by LC0 3.3.3.1, " Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. HATCM UNIT 2 5.0-20 REVISwN[
High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS O).- (_ 5.7 High Radiation Area 5.7.1 Pursuant to 10 CFR 20, paragraph 20.1601, in lieu of the requirements of 10 CFR 20.1601a, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is
> 100 mrem /hr but < 1000 mrem /hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, shall be barricaded and conspicuously posted as a high radiation area. Entrance thereto shall be controlled by requiring issuance of.a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Health Physics Technicians) or personnel continuously escorted by such . individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates
< 1000 mrem /hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.
Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the i following:
- a. A radiation monitoring device that continuously indicates l the radiation dose rate in the area.
O- b. A radiation monitoring device that continuously integrates l' the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.
- c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the '
activities within the area and shall perform periodic radiation surveillance at the frequency specified by the-facility Health Physics supervision in the RWP. (continued)' U l HATCH UNIT 2 5.0-21 REVISION D j
1 High Radiation Area l 5.7 j 5.7 High Radiation Area (continued) 1 5.7.2 In addition to the requirements of Specification 5.7.1, areas with radiation levels 2: 1000 mrem /hr, measured at 30 cm from the radiation source or from any surface the radiation penetrates, but less than 500 Rads in I hour measured at 1 meter from the radiation source or from any surface that the radiation penetrates, shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervision on duty or Health Physics supervision. O O HATCH UNIT 2 5.0-22 REVISION D
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i RCS Pressure SL ; B 2.1.2 B 2.0 SAFETY LIMITS (SLs) ; B 2.1.2 Reactor Coolant System (RCS) Pressure SL. BASES 1 BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor ' coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam l dome pressure ensures continued RCS integrity. Per i 10 CFR 50, Appendix A, GDC 14, " Reactor Coolant Pressure Boundary," and GDC 15, " Reactor Coolant System Design" < (Ref.1), the reactor coolant pressure boundary (RCPB) shall ' be designed with sufficient margin to ensure.that the design conditions are not exceeded during normal operation and anticipated operational occurrences (A00s). i During normal operation and A00s, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are O hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core may be done under , LCO 3.10.1, " Inservice Leak and Hydrostatic Testing : Operation." Following inception'of unit operation, RCS ' components shall be pressure tested in accordance with the ; requirements of ASME Code, Section XI (Ref. 3). Overpressurization of the RCS could result in a breach of ' the RCPB, reducing the number of protective barriers designed to prevent radioactive releases from exceeding the ! limits specified in 10 CFR 100, " Reactor Site Criteria" (Ref. 4). If this occurred in conjunction with a~ fuel cladding failure, fission products could enter the containment atmosphere. Ik (continued) HATCH UNIT 2 B 2.0-7 REVISION D
RCS Pressure SL B 2.1.2 BASES (continued) h APPLICABLE The RCS safety / relief valves and the Reactor Protection SAFETY ANALYSES System Reactor Vessel Steam Dome Pressure - High Function have settings established to ensure that the RCS pressure SL will not be exceeded. The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME, Boiler and Pressure Vessel Code,1968 Edition, including Addenda through the Summer of 1970 (Ref. 5), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig. The SL of 1325 psig, as measured in the reactor steam dome, is equivalent to 1375 psig at the lowest elevation of the RCS. The RCS is designed to Section III of the ASME, Boiler and Pressure Vessel Code,1980 Edition, including addenda through Winter 1981 (Ref. 6), for the reactor recirculation piping, which permits a maximum pressure transient of 110% of design pressures of 1250 psig for suction piping and 1450 psig for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes. O SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure , vessel under the ASME Code, Section III, is 110% of design ' pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 110% of design pressures 1 of 1250 psig for suction piping and 1450 psig for discharge piping. The most limiting of these two allowances is the , 110% of the reactor vessel and recirculation suction piping j design pressures; therefore, the SL on maximum allowable RCS : pressure is established at 1325 psig as measured at the I reactor steam dome. l APPLICABILITY SL 2.1.2 applies in all MODES. (continued) HATCH UNIT 2 B 2.0-8 REVISION A
V-
.l 1
i LCO Applicability l B 3.0 < l f BASES ; LCO 3.0.3 assemblies in the spent fuel storage pool." Therefore, this '
' (continued) LC0 can be. applicable in any or all MODES. If the LCO and the Required Actions of LCO 3.7.8 are not met while in i MODE 1, 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of LCO 3.7.8 of " Suspend movement of irradiated fuel .
assemblies in the spent fuel storage pool" is the ! appropriate Required Action to complete in lieu of the : actions of LCO 3.0.3. These exceptions are addressed in the : individual Specifications. , LCO 3.0.4 LC0 3.0.4 establishes limitations on changes in MODES or ; other specified conditions in the Applicability when an LC0 : is not met. It precludes placing the unit in a MODE or other specified condition stated in that LC0's Applicability (e.g., Applicability desired to be entered) when the following exist:
- a. Plant conditions are such that the requirements of an l LCO would not be met in the Applicability desired to be entered; and
- b. Continued noncompliance with these LCO requirements, if that Applicability were entered, would result in the unit being required to exit the Applicability ;
desired to be entered to comply with the Required ! Actions. i 4 Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a
.. MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change. Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions. '
The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good ; practice of restoring systems or components to OPERABLE 1 status before unit startup. The provisions of LC0 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability (continued) j j HATCH UNIT 2 B 3.0-5 REVISION A i i I
LCO Applicability B 3.0 BASES h LC0 3.0.4 that are required to comply with ACTIONS. In addition, the (continued) provisions of LC0 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. Exceptions to LC0 3.0.4 are stated in the individual Specifications. Exceptions may apply to all the ACTIONS or to a specific Required Action of a Specification. Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing MODES or other specified conditions while in an ACTIONS Condition, either in compliance with LC0 3.0.4 or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO. LC0 3.0.5 LLO 3.0.5 establishes the allowance for restoring equipment O to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with the applicable Required Action (s)) to allow the performance of SRs to demonstrate:
- a. The OPERABILITY of the equipment being returned to service; or i
- b. The OPERABILITY of other equipment.
The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the allowed SRs. This Specification does not provide time to perform any other preventive or corrective maintenance. (continued) HATCH UNIT 2 B 3.0-6 REVISION D C
SDM B 3.1.1 {gj BASES (continued)
' ACTIONS M 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, and the low probability of an event occurring during this interval.
M 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. r~T Cl L.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 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) Unit 2 secondary containment is OPERABLE; 2) at least one Unit 2 Standby Gas Treatment (SGT) subsystem is OPERABLE; A
- 3) and secondary containment isolation capability is available (i.e., at least one secondary containment isolation valve and associated instrumentation Il (continued)
V HATCH UNIT 2 o B 3.1-3 REVISION D
SDM-B 3.1.1 BASES ACTIONS D.I. D.2. D.3. and D.4 (continued) are OPERABLE, or other acceptable administrative controls to assure isolation capability) in each associated Unit 2 g secondary 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. 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 W 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 I I reactivity of the core and therefore do not have to be inserted. Action must also be initiated within I hour to provide means for control of potential radioactive releases. This includes ensuring: 1) Unit I secondary containment is iA (continued) HATCH UNIT 2 B 3.1-4 REVISION D
SDM B 3.1.1 pg (J BASES OPERABLE; 2) at least tv:o SGT subsystems are OPERABLE (any combination of Unit 1 and Unit 2 subsystems); 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 I l-1
)
O (continued) HATCH UNIT 2 B 3.1.4A REVISION D
SDM B 3.1.1 O v BASES ACTIONS E.1. E.2. E.3. E.4. and E.5 (continued) , Unit I secondary containment penetration flow path not isolated that is assumed to be isolated to mitigate ld radioactivity releases. This may be. performed as an s 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 Id required component is ino)erable, then it must be restored to OPERABLE status. In t11s 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 (Oj testing before or during the first startup after fuel movement or shuffling within the reactor pressure vessel, ce 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 control 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 (B0C) test must also account for changes in core reactivity during the cycle. Therefore, to obtain the SDM, the initial value "R", which is the difference must betweenbe changed by the the calculated value, value o f minimum SDM during the operating cycle and the calculated BOC 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. l i (continued) t
]
HATCH UNIT 2 B 3.1-5 REVISION D
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 LCO 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 the 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 additiona.' 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-24011-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
i Control Rod Scram Times B 3.1.4
,o (j BASES APPLICABLE The scram function of the CRD System protects the MCPR SAFETY ANALYSES Safety Limit (SL) (see Bases for SL 2.1.1, " Reactor Core (continued) SLs" and LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)")
and the 1% cladding plastic strain fuel design limit (see Bases for LC0 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded. Above 800 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL, during the analyzed limiting power transient. Below 800 psig, the scram function is assumed to perform during the control rod drop accident (Ref. 5) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LC0 3.1.6, " Rod Pattern Control"). For the reactor vessel overpressure protection analysis, the scram function, along with the safety / relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits. Control rod scram times satisfy Criterion 3 of the NRC A Policy Statement (Ref. 8). I p_\ .V() LC0 The scram times specified in Table 3.1.4-1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met (Ref. 6). To account for single failures and " slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis. The scram times have a margin that allows up to approximately 7% of the control rods (e.g.,137 x 7
== 10) to have scram times exceeding the specified limits (i.e., " slow" control rods) assuming a single stuck control rod (as allowed by LC0 3.1.3, " Control Rod OPERABILITY") and an additional control rod failing to scram per the single failure criterion. The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes (" pickup") when the index tube passes a specific location and then opens
(" dropout") as the index tube travels upward. Verification (continued) B 3.1-23 HATCH UNIT 2 REVISION D l 1 i
Control Rod Scram Times B 3.1.4 BASES h LCO of the specified scram times in Table 3.1.4-1 is (continued) accomplished through measurement of the " dropout" times. To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed
" slow" control rods may occupy adjacent locations.
Table 3.1.4-1 is modified by two Notes, which state that control rods with scram times not within the limits of the Table are considered " slow" and that control rods with scram times > 7 seconds are considered inoperable as required by SR 3.1.3.4. This LC0 applies only to OPERABLE control rods since inoperable control rods will be inserted and disarmed (LC0 3.1.3). Slow scramming control rods may be conservatively declared inoperable and not accounted for as " slow" control rods. APPLICABILITY In MODES 1 and 2, a scram is assumed to function during transients and accidents analyzed for these plant conditions. These events are assumed to occur during startup and power operation; therefore, the scram function of the control rods is required during these MODES. In MODES 3 and 4, with the mode switch in shutdown control rod block prevents withdrawal of control rods. This provides adequate requirements for control rod scram capability during these conditions. Scram requirements in MODE 5 are contained in LC0 3.9.5, " Control Rod OPERABILITY - Refuel ing. " ACTIONS Ad When the requirements of this LC0 are not met, the rate of negative reactivity insertion during a scram may not be within the assumptions of the safety analysis. Therefore, 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 MODE 3 within 12 hours. 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. (continued) HATCH UNIT 2 B 3.1-24 REVISION A
Control Rod Scram Times l B 3.1.4
,-.x .
BASES (continued) SURVEILLANCE The four SRs of this LC0 are modified by a Note stating that REQUIREMENTS during a single contral rod scram time Surveillance, the CRD pumps shall be isolated from the associated scram accumulator. With the CRD pump isolated, (i.e., charging valve closed) the influence of the CRD pump head does not affect the single control rod scram times. During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times. SR 3.1.4.1 The scram reactivity used in DBA and transient analyses is based on an assumed control rod scram time. Measurement of the scram times with reactor steam dome pressure 2 800 psig demonstrates acceptable scram times for the transients analyzed in References 3 and 4. Maximum scram insertion times occur at a reactor steam dome pressure of approximately 800 psig because of the competing effects of reactor steam dome pressure and stored accumulator energy. Therefore, demonstration of adequate L scram times at reactor steam dome pressure 2 800 psig ensures that the measured scram times will be within the specified limits at higher pressures. Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and l to allow a range of pressures over which scram time testing ! can be performed. To ensure that scram time testing is l performed within a reasonable time following fuel movement within the reactor pressure vessel or after a shutdown 2120 days or longer, control rods are required to be tested before. exceeding 40% RTP. In the event fuel movement is limited to selected core cells, it is the intent of this SR that only those CRDs associated with the core cells affected by the fuel movements are required to be scram time tested. This Frequency is acceptable considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by work on , l control rods or the CR0 System, n (continued) {} HATCH UNIT 2 B 3.1-25 REVISION A _ - - __ - _____-___-_ _ ___-_____ _________-________________f
l Control Rod Scram Times B 3.1.4 BASES h SURVEILLANCE SR 3.1.4.2 REQUIREMENTS (continued) Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods. The sample remains representative if no more than 20% of the control rods in the sample tested are determined to be " slow". With more than 20% of the sample declared to be " slow" per the criteria in Table 3.1.4-1, additional control rods are tested until this 20% criterion (i.e., 20% of the entire sample size) is satisfied, or until the total number of " slow" control rods (throughout the core, from all Surveillances) exceeds the LCO limit. For planned testing, the control rods selected for the sample should be different for each test. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample. The 120 day Frequency is based on operating experience that has shown control rod scram times do not significantly change over an operating cycle. This Frequency is also reasonable based on the additional Surveillances done on the CRDs at more frequent intervals in accordance with LC0 3.1.3 and LC0 3.1.5, " Control Rod Scram Accumulators." SR 3.1.4.3 O When work that could affect the scram insertion time is aerformed on a control rod or the CRD System, testing must
>e done to demonstrate that each affected control rod retains adequate scram performance over the range of-applicable reactor pressures from zero to the maximum permissible pressure. The scram testing must be performed once before declaring the control rod OPERABLE. The required scram time testing must demonstrate the affected control rod is still within acceptable limits. The limits for reactor pressures < 800 psig, required by footnote (b), 4 are included in the Technical Requirements Manual Ref. 7 IP1 and are established based on a high probability of(meeting) the acceptance criteria at reactor pressures a 800 psig.
The limits for reactor pressures 2 800 psig are found in Table 3.1.4-1. If testing demonstrates the affected control rod does not meet the.se limits, but is within the 7 second limit of Table 3.1.4-1, Note 2, the control rod can be declared OPERABLE and " slow." l l Specific examples of work that could affect the scram times I are (but are not limited to) the following: removal of any l CRD for maintenance or modification; replacement of a (continued) HATCH UNIT 2 B 3.1-26 REVISION D
Control Rod Scram Times i B 3.1.4 s ) BASES SURVEILLANCE SR 3.1.4.3 (continued) REQUIREMENTS control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator, isolation valve or check valve in the piping required for scram. The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY. SR 3.1.4.4 When work that could affect the scram insertion time is performed on a control rod or CRD System, testing must be done to demonstrate each affected control rod is still within the limits of Table 3.1.4-1 with the reactor steam dome pressure 2 800 psig. Where work has been performed at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 can be satisfied with one test. However, for a (] control rod affected by work performed while shutdown, a V zero pressure test and a high pressure test may be required. This testing ensures that, prior to withdrawing the control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions. Alternatively, a control rod scram test during hydrostatic pressure testing could also satisfy both criteria. I The Frequency of once prior to exceeding 40% RTP is acceptable because of the capability to test the control rod i over a range of operating conditions and the more frequent , surveillances on other aspects of control rod OPERABILITY. l This test is also used to demonstrate control rod OPERABILITY when a 40% RTP after work that could affect the scram insertion time is performed on the CRD System. REFERENCES 1. 10 CFR 50, Appendix A, GDC 10. l
- 2. FSAR, Section 4.2.3.2. l
- 3. FSAR, Supplement 5A.4.3.
q (continued) l () HATCH UNIT 2 B 3.1-27 REVISION A
Control Rod Scram Times B 3.1.4 BASES h REFERENCES 4. FSAR, Section 15.1. (continued)
- 5. NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel," (revision specified in the COLR).
- 6. Letter from R. F. Janecek (BWROG) to R. W. Starostecki (NRC), "BWR Owners' Group Revised Reactivity Control Systems Technical Specifications," BWROG-8754, September 17, 1987.
- 7. Technical R>quirements Manual. Ib
- 8. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
I k O 1 O HATCH UNIT 2 B 3.1-28 REVISION D
RPS Instrumentation B 3.3.1.1
,,m
'( ) BASES SURVEILLANCE time required to perform channel Surveillance. That REQUIREMENTS analysis demonstrated that the 6 hour testing allowance does (continued) not significantly reduce the probability that the RPS will trip when necessary. SR 3.3.1.1.1 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 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. (i 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 upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.1.1.2 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Frequency of once per 7' days is based on minor changes in LPRM sensitivity, which could affect the APRM reading between performances of SR 3.3.1.1.8. ( (continued) HATCH UNIT 2 B 3.3-25 REVISION A l
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3. 3.1 M (continued) REQUIREMENTS A A restriction to satisfying this SR when < 25% RTP is I la provided that requires the SR to be met only at 2 25% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when < 25% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At 2 25% RTP, the Surveillance is required to have been satisfactorily performed within the last 7 days, in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after reaching or exceeding 25% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. SR 3.3.1.1.3 The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function uses the recirculation loop . drive flows to vary the trip setpoint. This SR ensures that I the total loop drive flow signals from the flow units used to vary the setpoint is appropriately compared to an injection test flow signal to verify the flow signal trip setpoint and, therefore, the APRM Function accurately reflects the required setpoint as a function of flow. If the flow unit signal is not within the appropriate limit, i one required APRM that receives an input from the inoperable i flow unit must be declared inoperable. The Frequency of 7 days is based on engineering judgment, operating experience, and the reliability of this instrumentation. 1 SR 3.3.1.1.4 ) 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. (continued) , HATCH UNIT 2 B 3.3-26 REVISION D '
i 1 SRM Instrumentation B 3.3.1.2 m () BASES l SURVEILLANCE SR 3.3.1.2.2 (continued) REQUIREMENTS where CORE ALTERATIONS are being performed, and the other OPERABLE SRM must be in an adjacent quadrant containing fuel. Note 1 states that the SR is required to be met only during CORE ALTERATIONS. It is not required to be met at other times in MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a review of plant logs to ensure that SRMs required to be OPERABLE for given CORE ALTERATIONS are, in fact, OPERABLE. In the event that only one SRM is required to be OPERABLE (when the fuoled region encompasses only one SRM), per Table 3.3.1.2-1, footnote (b), only the a. portion of this SR is required. Note 2 clarifies that more than one of the three requirements can be met by the same OPERABLE SRM. The 12 hour Frequency is based upon operating experience and supplements operational controls over refueling activities I that include steps to ensure that the SRMs required by the LCO are in the proper quadrant. SR 3.3.1.2.4 o b This Surveillance consists of a verification of the SRM instrument readout to ensure that the SRM reading is greater than a specified minimum count rate, which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. With few fuel assemblies loaded, the SRMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient " source" material, in the form of irradiated fuel assemblies, to establish the minimum count rate. To accomplish this, the SR is modified by a Note (Note 1) that states that the count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblies are in the associated core quadrant. With four or less fuel assemblies loaded around each SRM and no other fuel assemblies in the associated core quadrant, even with a control rod withdrawn, I the configuration will not be critical. In addition, Note 2 l states that this requirement does not have to be met during i spiral unloading. If the core is being unloaded in this l manner, the various core configurations encountered will not i be critical. ( (continued) HATCH UNIT 2 B 3.3-39 REVISION A L - _ _ _ _ _ _ _ __
SRM Instrumentation B 3.3.1.2 BASES h SURVEILLANCE SR 3.3.1.2.4 (continued) REQUIREMENTS The Frequency is based upon channel redundancy and other information available in the control room, and ensures that the required channels are frequently monitored while core reactivity changes are occurring. When no reactivity changes are in progress, the Frequency is relaxed from 12 hours to 24 hours. SR 3 3.1 4.5 and SR 3.3.1.2.6 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated channel will function properly. SR 3.3.1.2.5 is required in MODE 5, and the 7 day frequency ensures that the I channels are OPERABLE while core reactivity changes could be in progress. This Frequency is reasonable, based on operating experience and on other Surveillances (such as a CHANNEL CHECK), that ensure proper functioning between CHANNEL FUNCTIONAL TESTS. SR 3.3.1.2.6 is required in MODE 2 with IRMs on Range 2 or below, and in MODES 3 and 4. Since core reactivity changes do not normally take place in MODES 3 and 4, and core reactivity changes are due only to control rod movement in MODE 2, the Frequency has been extended from 7 days to i 31 days. The 31 day Frequency is based on operating I experience and on other Surveillances (such as CHANNEL CHECK) that ensure proper functioning between CHANNEL FUNCTIONAL TESTS. Verification of the signal to noise ratio also ensures that the detectors are inserted to an acceptable operating level. In a fully withdrawn condition, the detectors are sufficiently removed from the fueled region of the core to essentially eliminate neutrons from reaching the detector. d Any count rate obtained while the detectors are fully withdrawn is assumed to be " noise" only. The Note to the SR 3.3.1.2.6 allows the Surveillance to be delayed until entry into the specified condition of the Ik Applicability (THERMAL POWER decreased to IRM Range 2 or bel ow) . The SR must be performed within 12 hours after IRMs are on Range 2 or below. The allowance to enter the Applicability with the 31 day frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. (continued) HATCH UNIT 2 B 3.3-40 REVISION D l l l
Control Rod Block Instrumentation B 3.3.2.1 l g () BASES j BACKGROUND The purpose of the RWM is to control rod patterns during l (continued) startup and shutdown, such that only specified control rod i sequences and relative positions are allowed over the operating range from all control rods-inserted to 10% RTP. The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses feedwater flow and steam flow signals to determine when the reactor power is above the preset power level at which the . RWM is automatically bypassed (Ref. 2). The RWM is a single channel system that provides input into both RMCS rod block circuits. With the reactor modo switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This < Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is recuired to be in the (. O) shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a , control rod block to all control rods. APPLICABLE 1. Rod Block Monitor SAFETY ANALYSES, LCO, and The RBM is designed to prevent violation of the MCPR APPLICABILITY SL and the cladding 1% plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event. The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 3. A statistical analysis of RWE events was performed to determine the RBM response for both channels for each event. From these responses, the fuel thermal performance as a function of RBM Allowable Value was determined. The Allowable Values are chosen as a function of power level. Based on the specified Allowable Values, operating limits are established. ,/3 Q (continued) HATCH UNIT 2 B 3.3-43 REVISION A
Control Rod Block Instrumentation B 3.3.2.1 BASES h APPLICABLE 1. Rod Block Monitor (continued) SAFETY ANALYSES, LCO, and The RBM Function satisfies Criterion 3 of the NRC Policy APPLICABILITY Statement (Ref. 10). I b Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Values, to ensure that no single instrument failure can preclude a rod block from this Function. The setpoints are calibrated consistent with applicable setpoint methodology (nominal trip setpoint). Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Vilues between successive 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 power), 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 W from the limiting values of the process 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. The RBM is assumed to mitigate the consequences of an RWE event when operating a 29% RTP. Below this power level, the consequences of an RWE event will not exceed the MCPR SL and, therefore, the RBM is not required to be OPERABLE (Ref. 3). When operating < 90% RTP, analyses (Ref. 3) have shown that with an initial MCPR a 1.70, no RWE event will result in exceeding the MCPR SL. Also, the analyses demonstrate that when operating at a 90% RTP with MCPR a 1.40, no RWE event will result in exceeding the MCPR (continued) HATCH UNIT 2 B 3.3-44 REVISION D
Control Rod Block Instrumentation B 3.3.2.1 m Q BASES APPLICABLE 1. Rod Block Monitor (continued) SAFETY ANALYSES, LCO, and SL (Ref. 3). Therefore, under these conditions, the RBM is . APPLICABILITY also not required to be OPERABLE.
- 2. Rod Worth Minimizer The RWM enforces the banked position withdrawal sequence (BPWS) to ensure that the initial conditions of the CRDA analysis are not violated. The analytical methods and assumptions used in evaluating the CRDA are summarized in References 4, 5, 6, and 7. In addition, the Reference 6 analysis (Generic BPWS analysis) may be modified by plant specific evaluations. The BPWS requires that control rods be moved in groups, with all control rods assigned.to a l specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, " Rod Pattern Control ."
The RWM Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10). Ib ' Since the RWM is a system designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref. 7). ' Special circumstances provided for in the Required Action of LC0 3.1.3, " Control Rod OPERABILITY," and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. The tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh RWM may be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed. Compliance with the BPWS, and therefore OPERABILITY of the RWM, is required in MODES I and 2 when THERMAL POWER is
< 10% RTP. When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal /gm fuel damage limit during a CRDA (Refs. 5 and 7). In MODES 3 and 4, all control rods.are required to be inserted into the core; therefore, a CRDA cannot occur. In MODE 5, since only a O
v (continued) HATCH UNIT 2 B 3.3-45 REVISION D
Control Rod Block Instrumentation B 3.3.2.1 BASES h' APPLICABLE 2. Rod Worth Minimizer (continued) SAFETY ANALYSES, single control rod can be withdrawn from a core cell LCO, and l APPLICABILITY containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor ' will be subcritical. l
- 3. Reactor Mode Switch - Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch - Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.
The Reactor Mode Switch - Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement (Ref.10). Ib Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when required. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on reactor mode switch position. During shutdown conditions (MODE 3, 4, or 5), no positive reactivity insertion events are analyzed because assumptions are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, the control rod withdrawal block is required to be OPERABLE. During MODE 5 with the reactor mode switch in the refueling position, the refuel position one-rod-out interlock (LC0 3.9.2, " Refuel Position One-Rod-Out Interlock") provides the required control rod withdrawal blocks. ACTIONS A_d With one RBM channel inoperable, the remaining OPERABLE i channel is adequate to perform the control rod block function; however, overall reliability is reduced because a (continued) l HATCH UNIT 2 B 3.3-46 REVISION D
Control Rod Block Instrumentation B 3.3.2.1 Q) BASES ACTIONS E.1 and E.2 (continued) required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted. SURVEll. LANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Control Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1. The Surveillances are modified by a second Note to indicate that when an RBM 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 control rod block 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 4 This Note is based on the reliability analysis (Ref. 8[. /] assumption of the average time required to perform channel v Surveillance. That analysis demonstrated that the 6 hour D testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary. SR 3.3.2.1.1 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended , function. It includes the Reactor Manual Control System ' input. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 92 days is based on reliability analyses (Ref. 8). i
) (continued)
HATCH UNIT 2 B 3.3-49 REVISION D
Control Rod Block In:trumentation B 3.3.2.1 BASES (continued) h i SR 3.3.2.1.2 and SR 3.3.2.1.3 l A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs. This test is performed as soon as possible after the applicable conditions are entered. As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until I hour O (continued) , HATCH UNIT 2 B 3.3-49A REVISION D
Control Rod Block Instrumentation B 3.3.2.1 BASES l
. SURVEILLANCE SR 3.3.1.2 and SR 3.3.2.1.3 (continued) i after any control rod is withdrawn at < 10% RTP in MODE 2, and SR 3.3.2.1.3 is not required to be performed until.
1 A 1 1 hour after THERMAL POWER is < 10% RTP in MODE 1. This allows entry into MODE 2 (and if entered during a shutdown, concurrent power reduction to < 10% RTP) for SR 3.3.2.1.2 g and THERMAL POWER reduction to < 10% RTP in MODE 1 for SR 3.3.2.1.3 to perform the required Surveillances if the 92 day Frequency is not met per SR 3.0.2. The I hour allowance is based on operating experience and in consideration of 4 providing a reasonable time in which to complete the SRs. l The 92 day Frequencies are based on reliability analysis (Ref. 8). SR 3.3.2.1.4-The RBM setpoints are automatically varied as a function of power. Three Allowable Values are specified in i Table 3.3.2.1-1, each within a specific power range. The , power at which the control rod block Allowable Values !A.d automatically change.are based on the APRM signal's input to each RBM channel. Below the minimum power setpoint, the RBM { l
- is automatically bypassed. These power Allowable Values must be verified periodically to be less than or equal to the .specified values. If any power range setpoint is nonconservative, then the affected RBM channel is considered inoperable. Alternatively, the power range channel can be placed in the conservative condition (i.e., enabling the proper RBM setpoint). If placed in this condition, the SR is met and the RBM channel is not considered inoperable. As noted,- neutron detectors are excluded from the Surveillance because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal.
Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8. The 18 month Frequency is based on the actual trip setpoint methodology utilized for these channels. SR 3.3.2.1.5 The RWM is automatically bypassed when power is above-a i specified value. The power level is determined from V (continued) HATCH UNIT 2 B 3.3-50 REVISION D
1 Control Rod Block Instrumentation B 3.3.2.1 BASES h feedwater flow and steam flow signals. The automatic bypass setpoint must be verified periodically to be 2: 10% RTP. If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power O (continued) HATCH UNIT 2 B 3.3-50A REVISION D
Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES 6. NED0-21231, " Banked. Position Withdrawal Sequence," (continued) January 1977.
- 7. NRC SER, " Acceptance of Referencing of 1.icensing
' Topical Report NEDE-24011-P-A," " General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.
- 8. NEDC-30851-P-A, " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"
October 1988.
- 9. GENE-770-06-1, " Bases for Changes To Surveillance Test Intervals And Allowed Out-Of-Service Times For Selected Instrumentation Technical Specifications,"
g February 1991.
- 10. NRC No. 93-102, " Final Policy Statement on Technical lA Specification Improvements," July 23, 1993.
O
' HATCH UNIT 2 B 3.3-53 REVISION D
Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 B 3.3 INSTRUMENTATION h B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation BASES BACKGROUND The feedwater and main turbine high water level trip instrumentation is designed to detect a potential failure of the Feedwater Level Control System that causes excessive feedwater flow. With excessive feedwater flow, the water level in the reactor vessel rises toward the high water level setpoint, causing the trip of the two feedwater pump turbines and the main turbine. Reactor Vessel Water Level-High signals are provided by level sensors 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 in the reactor vessel (variable leg). Three channels of Reactor Vessel Water Level-High instrumentation are provided as input to a two-out-of-three initiation logic that trips the two feedwater pump turbines and the main turbine. The channels include electronic equipment (e.g., trip relays) that compare measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a main feedwater and turbine trip signal to the trip logic. A trip of the feedwater pump turbines limits further increase in reactor vessel water level by limiting further addition of feedwater to the reactor vessel. A trip of the main turbine and closure of the stop valves protects the , turbine from damage due to water entering the turbine. APPLICABLE The feedwater and main turbine high water level trip SAFETY ANALYSES instrumentation is assumed to be capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref. 1). The high level trip indirectly initiates a reacter scram from the main turbine trip (above 30% RTP) and trips the feedwater pumps, thereby terminating the event. The reactor scram mitigates the reduction in MCPR. (continued) HATCH UNIT 2 B 3.3-54 REVISION A
i Remote Shutdown System B 3.3.3.2 b BASES ACTIONS M (continued) The Required Action is to restore the Function to OPERABLE status within 30 days. The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room. M If the Required Action and associated Completion Time of Condition A are not 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. The allowed Completion Time is reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when an instrument channel is placed in an inoperable status (V_') solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours. 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. The Note is based upon. a NRC Safety Evaluation Report (Reference 1) which concluded that the 6 hour testing allowance does not significantly reduce the probability of monitoring required parameters, when necessary. SR 3.3.3.2.1 Performance of the CHANNEL CHECK once every 31 days 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 g channels. It is based on the assumption that instrument channels monitoring the same parameter should read - approximately the same value. Significant deviations ( ( (continued) HATCH UNIT 2 B 3.3-75 REVISION D
Remote Shutdown System B 3.3.3.2 BASES 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 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 sensor or the signal processing equipment has drifted outside its limit. As specified in the Surveillance, a CHANNEL CHECK is only required for those channels that are i normally energized. The Frequer.cy is based upon plant operating experience that demonstrates channel failure is rare. SR 3.3.3.2.2 1d SR 3.3.3.2.2 verifies each required Remote Shutdown System I transfer switch and control circuit performs the intended function. This verification is performed from the remote shutdown panel and locally, as appropriate. Operation of equipment from the remote shutdown panel is not necessary. The Surveillance can be satisfied by performance of a continuity check, or in the case of the DG controls, the I routine Surveillances of LC0 3.8.1 (since local control is utilized during the performance of some of the Surveillances i I (continued) HATCH UNIT 2 B 3.3-75A REVISION D
Remote Shutdown System B 3.3.3.2
'O y/ BASES SURVEILLANCE SR 3.3.3.2.2 (continued) l REQUIREMENTS of LC0 3.8.1). This will ensure that if the control room becomes inaccessible, the plant can be placed and maintained in MODE 3 from the remote shutdown panel and the local control stations. 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 demonstrates t that Remote Shutdown System controls usually pass the Surveillance when performed at the 18 month Frequency.
SR 3.3.3.2.3 Ih CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies the channel responds to measured parameter values with the necessary range and accuracy. The 18 month Frequency is based upon operating experience p and consistency with the typical industry refueling cycle. d I REFERENCES 1. 10 CFR 50, Appendix A, GDC 19. l
- 2. Technical Requirements Manual.
- 3. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
i I o HATCH UNIT 2 B 3.3-76 REVISION D I 1 a _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ - - . _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ _ ._-_______________-._____________________________--__A
;)
Primary Containment Isolation Instrumentation- l B 3.3.6.1 l 1 BASES l ACTIONS f_d (continued) For the RWCU Area and Area Ventilation Differential Temperature-High Functions, the affected penetration flow i path (s) may be considered isolated by isolating only that portion of the system in the associated room monitored by the inoperable channel. That is,.if the RWCU pump room A - area channel is inoperable, the pump room A area can be isolated while allowing continued RWCU operation utilizing , the B RWCU pump. Alternately, if it is not desired to isolate the affected penetration flow path (s) (e.g., as in the case where isolating the penetration flow path (s) could result in a reactor scram), Condition G must be entered and its' Required Actions taken. The I hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for personnel , to isolate the affected penetration flow path (s). G.1 and G.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, or.any Required Action of Condition F is not met and the' associated Completion Time has expired, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant 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, plant conditions from full power conditions in an orderly manner and without > challenging plant systems. , i (continued) HATCH UNIT 2 B 3.3-169 REVISION A ,
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES h ACTIONS H.1 and H.2 (continued) If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the SLC System is declared inoperable or the RWCU System is isolated. Since this Function is required to ensure that the SLC System performs its intended function, sufficient remedial measures are provided by declaring the SLC System inoperable or isolating the RWCU System. The 1 hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for personnel to isolate the RWCU System. I.1 and I.2 If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated penetration flow path should be closed. However, if the shutdown cooling function is needed to provide core cooling, these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the channel to OPERABLE status or to isolate the RHR Shutdown Cooling System (i.e., provide alternate decay heat removal capabilities so the penetration flow path can be isolated). Actions must continue until the channel is restored to OPERABLE status or the RHR Shutdown Cooling System is isolated. SURVEILLANCE As noted at the beginning of the SRs, the SRs for each REQUIREMENTS Primary Containment Isolation instrumentation Function are found in the SRs column of Table 3.3.6.1-1. 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 naquired Actions may be delayed for up to A 6 hours provided the associated Function maintains isolation 1la 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. 4 and 5) assumption of the (continued) h HATCH UNIT 2 B 3.3-170 REVISION D
p Secondary Containment Isolation Instrumentation B 3.3.6.2 m (,) BASES ACTIONS B.1 (continued) SGT System (s). 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 the associated penetration flow path (s) and one SGT subsystem in each associated SGT System 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 I hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channel s. C.l.l. C.l.2. C.2.1. and C.2.2 If any Required Action and associated Completion Time of Condition A or B are not met, the ability to isolate the
/ secondary containment (s) and start the associated SGT (7j' System (s) cannot be ensured. Therefore, further actions must be performed to ensure the ability to maintain the secondary containment function. Isolating the associated zone (s) (closing the ventilation supply and 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.I.2 and C.2.2) is also acceptable since the Required Actions of the respective LCOs (LC0 3.6.4.4, LC0 3.6.4.5, LCO 3.6.4.6, LC0 3.6.4.7, LC0 3.6.4.8 and LC0 3.6.4.9) provide appropriate actions for the inoperable components. Since each trip system can affect two SGT subsystems (one Unit I and one Unit 2), 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 A.ction C.2.1) while the other SGT subsystem can be declared inoperable (Required Action C.2.2). ( (continued) HATCH UNIT 2 B 3.3-181 REVISION A
Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES ACTIONS C.1.1. C.1.2. 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. 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, d 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 the 6 hour testing allowance does not significantly reduce 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 is normally a ccmparison 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 D
S/RVs B 3.4.3 l B 3.4 REACTOR COOLANT SYSTEM (RCS) [] B 3.4.3 Safety / Relief Valves (S/RVs) BASES BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of S/RVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB). The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell . The S/RVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the spring loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve. This rm satisfies the Code requirement. ( Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool. The S/RVs that provide the relief mode are the low-low set (LLS) valves and the Automatic Depressurization System (ADS) valves. The LLS requirements are specified in LC0 3.6.1.6,
" Low-Low Set (LLS) Valves," and the ADS requirements are specified in LC0 3.5.1, "ECCS - Operating."
APPLICABLE The overpressure protection system must accommodate the most SAFETY ANALYSES severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 1). For the purpose of the analyses,11 S/RVs are assumed to operate in the safety mode. The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure well below the ASME Code limit of 110% of vessel
) (continued)
HATCH UNIT 2 B 3.4-13 REVISION A
l l S/RVs B 3.4.3 l BASES APPLICABLE design pressure (110% x 1250 psig = 1375 psig). Sensitivity 1 SAFETY ANALYSES analyses have demonstrated that 8 or 9 S/RVs operating in l (continued) the pressure relief mode will maintain the reactor vessel ' below 1375 psig. This LC0 helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event. From an overpressure standpoint, the design basis events are A-bounded by the MSIV closure with flux scram event described 2p_\ above. Reference 2 discusses additional events that are expected to actuate the S/RVs. S/RVs satisfy Criterion 3 of the NRC Policy Statement (Ref. i 4). 1 i 1 LC0 The safety function of eleven S/RVs are required to be j OPERABLE to satisfy the assumptions of the safety analysis (Refs. I and 2), although margins to the ASME Vessel Overpressure Limit are substantial. The requirements of this LC0 are applicable only to the capability of the S/RVs to mechanically open to relieve excess pressure when the ! lift setpoint is exceeded (safety function). The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME l' Code specifications require the lowest safety valve setpoint l to be at or below vessel design pressure (1250 psig) and the ' highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the FSAR are based on these setpoints, but also include the ; additional uncertainties of 3% of the nominal setpoint i drift to provide an added degree of conservatism. Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more l severe reactor response to a transient than predicted, ; possibly resulting in the ASME Code limit on reactor ' pressure being exceeded. l l \ } (continued) O I i HATCH UNIT 2 B 3.4-14 REVISION D 1___--__--.- _ - _ . - - . - - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
S/RVs B 3.4.3 BASES (continued) O(/ - APPLICABILITY In MODES 1, 2, and 3, all S/RVs must be OPERABLE, since considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time + that the Residual Heat Removal (RHR) System is capable of dissipating the core heat. In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents. In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The S/RV function is not needed during these conditions. r ACTIONS Ad With the safety function of one S/RV inoperable, the remaining OPERABLE S/RVs are capable of providing the , ( necessary overpressure protection. However, the overall i reliability of the pressure relief system i.e reduced because additional failures in the remaining OPERABLE S/RVs could result in failure to adequately relieve pressure during a ' limiting event. For this reason, continued operation is ! permitted for a limited time only. The 14 day Completion Time to restore the' inoperable S/RV to OPERABLE status is based on the relief capability of the remaining S/RVs, the low probability of an event requiring
- S/RV actuation, and a reasonable time to complete the :
Required Action. i B.1 and B.2 With more than one S/RV inoperable, a transient may result ; in the violation of the ASME Code limit on reactor pressure. ; If the safety function of the inoperable S/RV'cannot be ' restored to OPERABLE status within the associated Completion l Time of Required Action A.1, or if the safety function of ; two or more S/RVs is inoperable, the plant must be brought to a MODE in which the LCO does not apply. To rchieve this l status,-the plant must be brought to MODE 3 within 12 hours [ (continued) HATCH UNIT 2 B 3.4-15 REVISION A ! i
7 S/RVs B 3.4.3 BASES h ACTIONS L 1 and B.2 (continued) and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillance requires that the S/RVs will open at the pressures assumed in the safety analysis of Reference 1. The demonstration of the S/RV safety lift settings must be performed during shutdown, since this is a bench test, to be done in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is - 3% for OPERABILITY; however the valves are reset to i 1% during the Surveillance to allow for drift. Performance of this SR in accordance with the Inservice Testing Program requires an 18 month Frequency. The 18 month Frequency was selected because this Survoillance must be performed during shutdown conditions and is based on W the time between refuelings. SR 3 . 4 . 3._J A manual actuation of each S/RV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine centrol valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the S/RVs divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at A whichthistestistobeb.erformedis920psig(the pressure recommended by t e valve manufacturer). Adequate f[h steam flow is represented by at least 1.25 turbine bypass valves open, or total steam flow 2: IE6 lb/hr. Plant (continued) HATCH UNIT 2 B 3.4-16 REVISION D
-l r
RCS Operaticnal LEAKAGE . B 3.4.4'
~
-t BASES.'(continued)
. ACTIONS Ad With RCS unidentified or total LEAKAGE greater. than =the j limits, actions must be taken to reduce the leak. Because- .
the LEAKAGE limits are conservatively below the LEAKAGE that '! would constitute a critical crack' size, 4 hours is allowed 1 to reduce the. LEAKAGE rates before the reactor must be shut-
-) J down. .If an unidentified LEAKAGE has been-identified and- ,l quantified, it may be reclassified and considered as .
7 identified LEAKAGE;. however, the total LEAKAGE would remain ! unchanged. The total LEAKAGE must be averaged over the previous 24 hours for comparison to the limit.
]
B.d .f An unidentified LEAKAGE increase of > 2 gpm within a'24 hour-period is an indication of a potential flaw in the' RCPB and; .. must be quickly evaluated. .Although the increase does not ! necessarily violate the absolute unidentified LEAKAGE;11mit,- certain susceptible components must be determined not to be l the source of the LEAKAGE increase within the required Completion Time. - The 4 hour Completion Time is reasonable to properly reduce ; the LEAKAGE increase before the reactor must be shut down ; without unduly jeopardizing plant safety.
]
i C.1 and C.2 If any Required Action and associated Completion Time of- l Condition A or B is not met or if pressure boundary-LEAKAGE : exists, the plant must be brought to a MODE in which the LCO ; does not apply. To-achieve this status,-the plant'must.be l brought to MODE 3 within 12 hours'and to MODE'4 within- 1 36 hours. The allowed Completion Times are reasonable, . j based on operating experience,:to reach the ~ required; plant - conditions = from. full' power conditions in an' orderly manner ,
.and without challenging plant safety systems.
]
- / j
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J (continued)' ; HATCH UNIT 2 B 3.4 REVISION A l
RCS Operational LEAKAGE B 3.4.4 BASES (continued) h SURVEILLANCE SR 3.4.4.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to provide alarms when LEAKAGE is indicated and to quantify the various types of LEAKAGE. Leakage detection instrumentation is discussed in more detail in the Bases for LC0 3.4.5, "RCS Leakage Detection Instrumentation." Sump level and flow rate are typically monitored to determine actual LEAKAGE rates; however, any method may be used to quantify LEAKAGE within the guidelines of Reference 7. In conjunction with alarms and other administrative controls, a 12 hour Frequency for this Surveillance is appropriate for identifying LEAKAGE and for tracking required trends (Ref. 8). The identified portion of the total LEAKAGE is usually determined by the drywell equipment drain sump monitoring system which collects expected leakage, not indicative of a degraded RCS boundary. The system equipment and operation A is identical to that of the drywell floor drain monitoring Ah system described in the Bases for LC0 3.4.5, "RCS Leakage Detection Instrumentation." If a contributor to the unidentified LEAKAGE has been identified and quantified, it may be reclassified and considered as identified LEAKAGE. REFERENCES 1. 10 CFR 50.2.
- 2. 10 CFR 50.55a(c).
- 3. 10 CFR 50, Appendix A, GDC 55.
- 4. GEAP-5620, " Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall flaws," April 1968.
- 5. NUREG-75/067, " Investigation and Evaluation of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactors," October 1975.
- 6. FSAR, Section 5.2.7.5.2.
- 7. Regulatory Guide 1.45, May 1973.
- 8. Generic Letter 88-01, Supplement 1, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping,"
February 1992.
- 9. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
O HATCH UNIT 2 B 3.4-22 REVISION D
]
RCS Leakage Detection Instrumentation B 3.4.5
%)
LCO systems) provide early alarms to the operators so closer (continued) examination of other detection systems will be made to determine the extent of any corrective action that may be required. With the leakage detection systems inoperable, monitoring for LEAKAGE in the RCPB is degraded. APPLICABILITY In MODES 1, 2, and 3, leakage detection systems are required to be OPERABLE to support LC0 3.4.4. This Applicability is consistent with that for LC0 3.4.4. ACTIONS Ad With the drywell floor drain sump monitoring system inoperable, no other form of sampling can provide the equivalent information to quantify leakage. However, the primary containment atmospheric activity monitor will provide indication of changes in leakage. With the drywell floor drain sump monitoring system
,_) inoperable, but with RCS unidentified and total LEAKAGE (V being determined every 12 hours (SR 3.4.4.1), operation may continue for 30 days. The 30 day Completion Time of Required Action A.1 is acceptable, based on operating experience, considering the multiple forms of leakage detection that are still available. Required Action A.1 is modified by a Note that stater, that the provisions of LC0 3.0.4 are not applicable. As a result, a MODE change is allowed when the drywell floor drain sump monitoring system is inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage.
Acceptable methods for quantifying both identified and unidentified LEAKAGE include but are not limited to the following:
- 1) With a drifting sump monitoring system integrator, the g sump can be manually pumped down with integrator readings taken before and after pumpdown. The difference in readings determines total gallons pumped. Using time elapsed since last pumpdown, sump inleakage rate can be calculated; and
,m
'i
( m) (continued) HATCH UNIT 2 B 3.4-25 REVISION KD
RCS Leakage Detection Instrum:ntation B 3.4.5 BASES h Ad 2) With an inoperable sump monitoring system integrator, (continued) the sump can be manually pumped down and the time for pumpdown recorded. Utilizing pump flow rate, total A gallons pumped is determined. Using time elaased M since last pumpdown, sump inleakage rate can 3e calculated. B.1 and B.2 With both gaseous and particulate primary containment atmospheric monitoring channels inoperable (i.e., the required containment atmospheric monitoring system), grab samples of the primary containment atmosphere must be taken and analyzed to provide periodic leakage information. Provided a sample is obtained and analyzed once every 12 hours, the plant may be operated for up to 30 days to allow restoration of at least one of the required monitors. O l 1 i I (continued) HATCH UNIT 2 B 3.4-25A REVISION D
RCS Leakage Detection Instrumentation . B 3.4.5 ' vO BASES - ACTIONS B.1 and 8.2 (continued) The 12 hour interval provides periodic information that is adequate to detect LEAKAGE. The 30 day Completion Time for ' restoration recognizes that at least one other form of leakage detection is available. 3 The Required Actions are modified by a Note that states that the provisions of LC0 3.0.4 are not applicable. As a result, a MODE change is allowed when both the gaseous and , particulate primary containment atmospheric monitoring channels are inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage. C.) 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 MODE 4 within 36 hours. The allowed Completion '
/ Times are reasonable, based on operating experience, to C] perform the actions in an orderly manner and without challenging plant syr.tems.
Ed With all required monitors inoperable, no required automatic means of monitoring LEAKAGE are available, and-immediate plant shutdown in accordance with LC0 3.0.3 is required. SURVEILLANCE The Surve111ances are modified by a Note to indicate that REQUIREMENTS 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 other required instrumentation (either -; the drywell floor drain sump monitoring system or the ['. . . primary containment atmospheric monitoring channel, as ' applicable) ic OPERABLE. 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. C-s (continued) HATCH UNIT 2 B 3.4-26 REVISION A
)
RHR Shutdown Cooling System - Cold Shutdown B 3.4.8 ('J ) BASES APPLICABILITY In MODES 1 and 2, and in MODE 3 with reactor steam dome (continued) pressure greater than or equal to the RHR low pressure permissive pressure, this LCO is not applicable. Operation of'the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at reactor pressures greater than or equal to the RHR low pressure permissive pressure is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LC0 3.5.1, "ECCS - Operating") do not allow placing the RHR shutdown cooling subsystem into operation. I The requirements for decay heat removal in MODE 3 below the RHR low pressure permissive pressure and in MODE 5 are discussed in LC0 3.4.7, " Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown"; LC0 3.9.7,
" Residual Heat Removal (RHR) -- High Water Level"; and LC0 3.9.8, " Residual Heat Removal (RHR) - Low Water Level."
p C/ ACTIONS A Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems. Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent divisions, 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 Required Actions I 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 shutdown cooling subsystems provide appropriate compensatory measures for separate inoperable shutdown cooling > subsystems. As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem. i 1 (\ () (continued) HATCH UNIT 2 B 3.4-41 REVISION D e
RHR Shutdown Cooling System - Cold Shutdown B 3.4.8 BASES ACTIONS Ad (continued) With one of the two required RHR shutdown cooling subsystems inoperable, except as permitted by LC0 Note 2, 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 RHR 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 I 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 these alternate method (s) must be reconfirmed every 24 hours thereafter. This will provide assurance of continued heat removal capability. The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Decay , heat removal by ambient losses can be considered as, or contributing to, the alternate method capability. Alternate methods that can be used include (but are not limited to) the Condensate / Main Steam Systems (feed and bleed) and the Reactor Water Cleanup System. B.1 and B.2 With no RHR shutdown cooling subsystem and no recirculation pump in operation, except as permitted by LC0 Note 1, and until RHR or recirculation pump operation is re-established, an alternate method of reactor coolant circulation must be placed into service. This will provide the necessary circulation for monitoring coolant temperature. The I hour Completion Time is based on the coolant circulation function and is modified such that the I hour is applicable separately for each occurrence involving a loss of coolant circulation. Furthermore, verification of the functioning of the alternate . method must be reconfirmed every 12 hours thereafter. This will provide assurance of continued temperature monitoring capability. (continued) HATCH UNIT 2 B 3.4-42 REVISION A
RCS P/T Limits B 3.4.9 l,_T BASES J BACKGROUND as necessary, based on the evaluation findings and the (continued) recommendations of Reference 5. The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the , span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions. The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls. The criticality limits include the Reference 1 requirement that they be at least 40 F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic (T) C. testing. The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. ASME Code, Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits. APPLICABLE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws l to propagate and cause nonductile failure of the RCPB, a ; condition that is unanalyzed. The PTLR references the I methodology for determining the P/T limits. Since the l {q; (continued) HATCH UNIT 2 8 3.4-45 REVISION D
RCS P/T Limits B 3.4.9 BASES h APPLICABLE P/T limits are not derived from any DBA, there are no SAFETY ANALYSES acceptance limits related to the P/T limits. Rather, the (continued) P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition. RCS P/T limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 8). Id LC0 The elements of this LC0 are:
- a. RCS pressure, temperature, and heatup or cooldown rate are within the limits specified in the PTLR during RCS heatup, cooldown, and inservice leak and hydrostatic testing;
- b. The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is within the limit of the PTLR during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow;
- c. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel meets the limit of the PTLR during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow;
- d. RCS pressure and temperature are within the criticality limits specified in the PTLR, prior to achieving criticality; and
- e. The reactor vessel flange and the head flange temperatures are within the limits of the PTLR when tensioning the reactor vessel head bolting studs.
These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure. The rate of change of temperature limits controls the thermal gradient through the vessel wall and is used as input for calculating the heatup, cooldown, and inservice (continued) HATCH UNIT 2 B 3.4-46 REVISION D
RCS P/T Limits. B 3.4.9 BASES ACTIONS. C.1 and C.2 (continued) Dperation outside the P/T limits in other than MODES 1. 2, and 3 (including defueled conditions) must be corrected so ' that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be-L initiated without delay and continued until the limits are ! restored. Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation.must verify that the RCP8 : integrity. is acceptable and must be completed before , approaching criticality or heating up to > 212*F. Severai 4 methods may be used, including comparison with_ pre-analyzed- 1 transients, new analyses, or inspection of the_ components. J' ASME Code, Section XI, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is' restricted .to evaluation of the beltline.
, ~ Condition C is modified by a Note requiring Required Action C.2 be completed whenever the Condition is entered. The Note emphasizes the need to perform'the evaluation of.the O effects of the excursion outside the allowable limits.
Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB-integrity. SURVEILLANCE SR 3.4.9.1 REQUIREMENTS - Verification that operation is within' PTLR limits is required'every 30 minutes when RCS pressure and temperature .j conditions ~are undergoing planned changes. This Frequency :1 is considered reasonable in view of the control-room indication available to monitor RCS status. Also, since temperature rate of change limits are. specified in hourly increments,' 30 minutes permits a reasonable time for - assessment and correction of minor deviations. Surveillance for heatup, cooldown, or inservice leakage and'. hydrostatic testing may be discontinued when the criteria given:in the relevant plant procedure for ending.the
; activity are satisfied.
(continued) HATCH UNIT 2 B 3.4-49 REVISION A
.4
____________i.___________.__m.__ _ _ _ m J
RCS P/T Limits B 3.4.9 BASES h, SURVEILLANCE SR 3.4.9.1 (continued) REQUIREMENTS This SR has been modified with a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and RCS inservice leakage and hydrostatic testing. SR 3.4.9.2 A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical. Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal. SR 3.4.9.3 and SR 3.4.9.4 Differential temperatures within the applicable PTLR limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 7) are satisfied. Ik Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start. An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.4 is to compare the temperatures of the operating recirculation loop and the idle loop. (continued) h HATCH UNIT 2 B 3.4-50 REVISION D
e ; i RCS P/T Limits B 3.4.9 t BASES j' SURVEILLANCE SR 3.4.9.3 and SR 3.4.9.4 (pontinued) REQUIREMENTS i
'SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that i requires the Surveillance to be performed only in MODES 1, !
'2, 3, and 4. In MODE 5, the overall stress on limiting -
components is lower. Therefore, AT limits are not required. , i SR 3.4.9.5. SR 3.4.9.6. and SR 3.4.9.7 i i Limits on the reactor vessel flange and head flange ! temperatures are generally bounded by the other P/T limits l during system heatup and cooldown. However, operations " approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal- to certain specified values , l require assurance that these temperatures meet the LC0 limits. j r The flange temperatures must be verified to be above the ; limits 30 minutes before and while tensioning the. vessel ; head bolting studs to ensure that once the head is tensioned ; the limits are satisfied. When in MODE 4 with RCS : q temperature s 80*F, 30 minute' checks of the flange temperatures are required because '.,f the reduced margin to Q the limits. When in MODE 4 with RCS temperature s 100'F, r monitoring of the flange temperature is required every ! 12 hours to ensure the temperature is within the limits
- specified in the PTLR. !
The 30 minute Frequency reflects the urgency of maintaining. ! the temperatures within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour L Frequency is reasonable based on the rate of temperature change possible at these temperatures. i SR 3.4.9.5 is modified by a Note that requires the Surveillance to be performed only when tensioning the , reactor vessel head bolting studs. SR 3.4.9.6 is modified , by a Note that requires the Surveillance to be initiated 30 minutes after RCS temperature s 80'F in Mode 4. SR . 3.4.9.7 is modified by a Note that requires the Surveillance to be initiated 12 hours after RCS temperature $ 100*F in , Mode 4. The Notes contained in these SRs are necessary to , (Continued) HATCH UNIT 2 B 3.4-51 REVISION A 7
RCS P/T Limits B 3.4.9 BASES h SURVEILLANCE SR 3.4.9.5. SR 3.4.9.6. and SR 3.4.9.7 (continued) REQUIREMENTS specify when the reactor vessel flange and head flange temperatures are required to be verified to be within the limits specified in the PTLR. REFERENCES 1. 10 CFR 50, Appendix G.
- 2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.
- 3. ASTM E 185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," July 1982.
- 4. 10 CFR 50, Appendix H.
- 5. Regulatory Guide 1.99, Revision 2, May 1988.
- 6. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E. g
- 7. FSAR, Section 15.1.26. IA
- 8. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
iM 1 O HATCH UNIT 2 B 3.4-52 REVISION D
ECCS - Operating B 3.5.1 (q g BASES 1 SURVEILLANCE SR 3.5.1.7. SR 3,5.1.8. and SR 3.5.1.9 (continued) ) 4 REQUIREMENTS The flow tests for the HPCI System are performed at two I different pressure ranges such that system capability to prn rated flow is tested at both the higher and lower op w ing ranges of the system. The pump flow rates are ve,1fied against a system head corresponding to the RPV pressure. The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel , discharge, the piping friction losses, and RPV pressure. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass 1 valves to continue to control reactor pressure when the HPCI l System diverts steam flow. The reactor steam pressure must j be 2 920 psig to perform SR 3.5.1.8 and 2150 psig to j perform SR 3.5.1.9. Adequate steam flow for SR 3.5.1.8 is ; represented by at least two turbine bypass valves open, or i a 200 MWE from the main turbine-generator; and for SR 3.5.1.9 adequate steam flow is represented by at least 1.25 turbine bypass valves open, or total steam flow 2 IE6 lb/hr. Therefore, sufficient time is allowed after adequate pressure and flow are achieved to perform these tests. Reactor startup is allowed prior to performing the
.O V
low pressure Surveillance test because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance test is short. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure test has been satisfactorily , completed and there is no indication or reason to believe l that HPCI is inoperable. Therefore, SR 3.5.1.8 and ! SR 3.5.1.9 are modified by Notes that state the l Surveillances are not required to be performed until 12 hours after the reactor steam pressure and flow are adequate to perform the test. Therefore, implementation of these Notes requires these tests to be performed during reactor startup within 12 hours after adequate steam pressure 9,)J flow are achieved. The Frequency for SR 3.5.1.7 and SR 3.5.1.8 is consistent ' l with the Inservice Testing Program pump testing requirements. The 18 month Frequency for SR 3.5.1.9 is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startu) from a plant outage. Operating experience has shown that t1ese components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. (continued) HATCH UNIT 2 B 3.5-13 REVISION D
ECCS - Operating B 3.5.1 BASES h SURVEILLANCE SR 3.5.1.10 REQUIREMENTS (continued) The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI, CS, and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions. This SR also ensures that the HPCI System will automatically restart on an RPV low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the section is automatically transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LC0 3.3.5.1 overlaps this Surveillance to provide complete testing of the assumed safety function. The 18 month Frequency is based on the need to perform the 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 that these components usually O W pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency w.s concluded to be acceptable from a reliability standpoint. This SR is modified by a Note that excludes vessel injection / spray during the Surveillance. Since all active , components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance. SR 3.5.1.11 The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e., solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper i actuation of all the required components. SR 3.5.1.12 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LC0 3.3.5.1 (continued) HATCH UNIT 2 B 3.5-14 REVISION A
l l ECCS - Operating l B 3.5.1 '
)
(y Q BASES SURVEILLANCE SR 3.5.1.11 (continued) REQUIREMENTS overlap this Surveillance to provide complete testing of the assumed safety function. The 18 month Frequency is based on the need to perform the 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 that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability - standpoint. This SR is modified by a Note that excludes valve actuation. This prevents an RPV pressure blowdown. SR 3.5.1.12 A manual actuation of each ADS valve is performed to verify that the valve and solenoid are functioning properly and r'N that no blockage exists in the S/RV discharge lines. This () is demonstrated by the response of the turbine control or bypass valve or by a change in the measured steam flow or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the ADS valves divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this SR. Adequate pressure at which this SR is to be performed is a 920 psig (the pressure recommended by the valve manufacturer). Adequate steam flow is represented by at least 1.25 turbine bypass valves o)en, g or total steam flow a IE6 lb/hr. Reactor startup is a' lowed prior to performing this SR because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test. The 12 hours allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions and provides adequate time to complete the Surveillance. SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LC0 3.3.5.1 O (U (continued) I HATCH UNIT 2 B 3.5-15 REVISION D
ECCS - Operating B 3.5.1 BASES SURVEILLANCE 3R 3.5.1.12 (continued) REQUIREMENTS overlap this Surveillance to provide complete testing of the assumed safety function. The Frequency of 18 months is based on the need to perform the Surveillance under the conditions that apply just prior to or during startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. FSAR, Section 6.3.2.2.3.
- 2. FSAR, Section 6.3.2.2.4.
- 3. FSAR, Section 6.3.2.2.1.
- 4. FSAR, Section 6.3.2.2.2.
- 5. FSAR, Section 15.1.39.
- 6. FSAR, Section 15.1.40.
- 7. FSAR, Section 15.1.33.
- 8. 10 CFR 50, Appendix K.
- 9. FSAR, Section 6.3.3.
- 10. NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Analysis," December 1986.
- 11. 10 CFR 50.46.
- 12. Memorandum from R.L. Baer (NRC) to V. Stello, Jr.
(NRC), " Recommended Interim Revisions to LCOs for ECCS i Components," December 1, 1975. '
- 13. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
O' HATCH UNIT 2 8 3.5-16 REVISION A
ECCS - Shutdown B 3.5.2 {gt 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 (3 potential for fission product release. Actions must U continue until OPDRVs are suspended. C.1. C.2. 0,1. 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 OPDRVs are Espended. 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 1er Trassure 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 h ACTIONS C.l. C.2. D.l. D.2. D.3. E.1. E.2. and E.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. If the unit is in MODE 4, this includes ensuring: 1) Unit 2 secondary containment is OPERABLE; 2) one Unit 2 standby gas treatment (SGT) subsystem is OPERABLE; 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 Unit 2 secondary containment penetration flow path not isolated that is assumed to be isolated to mitigate radioactivity releases. If the unit is in MODE 5, this includes ensuring: 1) both Unit I and Unit 2 secondary containments are OPERABLE; 2) three SGT subsystems (any combination of Unit I and Unit 2 subsystems) are OPERABLE;
- 3) and secondary containment isolation capability is available (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other ac'ceptable administrative controls to assure isolation capability) in each Unit I and Unit 2 secondary containment W
flowpath not isolated that is assumed to be isolated to mitigate radioactivity releases. The Unit I requirements are not required when the unit is in MODE 4 since the vessel head is not detensioned and the reactor is depressurized, thus the potential for a fission product release in the Unit I secondary containment is negligible. OPERABILITY may be verified by an administrative check, or by examining logs or A other information, to determine whether the components are out of service for maintenance or other reasons. It is not lA 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. (continued) HATCH UNIT 2 B 3.5-20 REVISION D I l I
ECCS - Shutdown B 3.5.2
,e ~y .
() BASES 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 (~~
's l
l l l ([] (continued) HATCH UNIT 2 B 3.5-20A REVISION D
- _ _ _ _ _ _ _ - - _ - _ _ _ _ - _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ -_-______________-__----_______---_A
I RCIC System B 3.5.3
,n.
BASES (] SURVEILLANCE SR 3.5.3.2 (continued) REQUIREMENTS The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve position i would affect only the RCIC System. This Frequency has been shown to be acceptable through operating experience. SR 3.5.3.3 and SR 3.5.3.4 The RCIC pump flow rates ensure that the system can maintain reactor coolant inventory during pressurized conditions with the RPV isolated. The required flow rate (400 gpm) is the pump design flow rate. Analysis has demonstrated that RCIC can fulfill its design function at a system flow rate of 360 gpm (Reference 4). The pump flow rates are verified against a system head equivalent to the RPV pressure. The total system pum) outlet pressure is adequate to overcome the elevation lead pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV e pressure. The flow tests for the RCIC System are performed (' at two different pressure ranges such that system capability to provide rated flow is tested both at the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Reactor steam pressure must be 2 920 psig to perform SR 3.5.3.3 and 2 150 psig to perform SR 3.5.3.4. Adequate steam flow is represented by at least one turbine bypass valve open, or for SR 3.5.3.3 2 200 MWE'from the main turbine-generator and for SR 3.5.3.4 total steam flow 2 IE6 lb/hr. Therefore, sufficient time is allowed after adequate pressure and flow are achieved to perform these SRs. Reactor startup is allowed prior to performing the low pressure Surveillance because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance is short. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low )ressure Surveillance has been satisfactorily completed and t1ere is no indication or reason to believe that RCIC is inoperable. Therefore, these SRs are modified by Notes that state the Surveillances are not required to be performed until 12 hours after the reactor steam pressure and flow are adequate to perform the test. Therefore, implementation of these Notes require {n} (continued) HATCH UNIT 2 B 3.5-27 REVISION D
RCIC System B 3 5.3 BASES SURVEILLANCE SR 3.5.3.3 and SR 3.5.3.4 (continued) REQUIREMENTS these tests to be performed during reactor startup within 12 hours after the reactor steam pressure and flow are adequate to perform the test. A 92 day Frequency for SR 3.5.3.3 is consistent with the Inservice Testing Program requirements. The 18 month Frequency for SR 3.5.3.4 is based on the need to perform the Surveillance under conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. SR 3.5.3.5 The RCIC System is required to actuate automatically in order to verify its design function satisfactorily. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of the RCIC System will cause the system to operate as designed, including actuation of the system throughout its emergency operating sequence; that is, g automatic pump startup and actuation of all automatic valves to their required positions. This test also ensures the RCIC System will automatically restart on an RPV low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LC0 3.3.5.2 overlaps this Surveillance to provide complete testing of the assumed safety function. The 18 month Frequency is based on the need to perform the 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 that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. (continued) HATCH UNIT 2 B 3.5-28 REVISION A
Primary Containment Air Lock B 3.6.1.2 O % ,J BASES ACTIONS 3.1. B.2. and B.3 (continued) typically restricted. Therefore, the probability of misalignment of the door, once it has been verified to be in the proper position, is small. C.I. C.2. and C.3 If the air lock is inoperable for reasons other than those described in Condition A or 8, Required Action C.1 requires action to be immediately initiated to evaluate containment overall leakage rates using current air lock leakage test results. An evaluation is acceptable since it is overly conservative to immediately declare the primary containment inoperable if both doors in the air lock have failed a seal test or if the overall air lock leakage is not within limits. In many instances (e.g., only one seal per door has failed), primary containment remains OPERABLE, yet only 1 hour (according to LC0 3.6.1.1) would be provided to restore the air lock door to OPERABLE status prior to requiring a plant shutdown. In addition, even with both doors failing the seal test, the overall containment leakage Q,A rate can still be within limits. Required Action C.2 requires that one door in the primary containment air lock must be verified closed. This action must be completed within the 1 hour Completion Time. This specified time period is consistent with the ACTIONS of LCO 3.6.1.1, which require that primary containment be restored to OPERABLE status within I hour. Additionally, the air lock must be restored to OPERABLE status within 24 hours. The 24 hour Completion Time is reasonable for restoring an inoperable air lock to OPERABLE status considering that at least one door is maintained closed in the air lock. D.1 and D.2 If the inoperable primary containment air lock cannot be restored to OPERABLE status within the associated 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 3 (continued) [O HATCH UNIT 2- B 3.6-11 REVISION A
i Primary Containment Air Lock B 3.6.1.2 BASES h ACTIONS D_d_and D.2 (continued) 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. SURVEILLANCE SR 3.6.1.2.1 REQUIREMENTS Maintaining primary containment air locks OPERABLE requires compliance with the leakage rate test requirements of 10 CFR 50, Appendix J (Ref. 3), as modified by approved exemptions. This SR reflects the leakage rate testing requirements with respect to air lock leakage (Type B leakage tests). The acceptance criteria were established as a small fraction of the total allowable containment leakage. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall primary containment leakage rate. The Frequency is required by 10 CFR 50, Appendix J (Ref. 3), as modified by approved exemptions. Thus, SR 3.0.2 (which allows Frequency extensions) does not apply. The SR has been modified by two Notes. Note I states that an inoperable air lock door does not invalidate the previous Id successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR, requiring the results to be evaluated against the acceptance criteria of SR 3.6.1.1.1. This ensures that air lock leakage is g properly accounted for in determining the overall primary containment leakage rate. { i l SR 3.6.1.2.2 > The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed l to withstand the maximum expected post accident primary (continued) HATCH UNIT 2 B 3.6-12 REVISION D
Primary Containment Air Lock B 3.6.1.2 ? BASES SURVEILLANCE SR 3.6.1.2.2 (continued) REQUIREMENTS containment pressure, closure of either door will support primary containment OPERABILITY.- Thus, the interlock feature supports primary containment OPERABILITY while the air lock.is being used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not O v i l l l l (Continued) HATCH UNIT 2 B 3.6-12A REVISION D' l v - , ,
PCIVs B 3.6.1.3 I n . {) BASES ! BACKGROUND The primary containment purge supply lines are 18 inches and I l (continued) 20 inches in diameter; exhaust lines are 18 inches in l diameter. The 18 inch primary containment purge valves are ) normally maintained closed in MODES 1, 2, and 3 to ensure I I the primary containment boundary is maintained. However, the 18 inch valves are qualified for use and may be opened , I when used for inerting, de-inerting, pressure control, ALARA { or air quality considerations for personnel entry, or l Surveillances that require the valves to be open. These j valves are qualified to be open because two additional I redundant excess flow isolation dampers are provided on the i vent line upstream of the Standby Gas Treatment (SGT) System filter trains. These isolation dampers, together with the PCIVs, will prevent high pressure from reaching the SGT System filter trains in the unlikely event of a loss of coolant accident (LOCA) during venting. Closure of the excess flow isolation dampers will not prevent the SGT System from performing its design function (that is, to maintain a negative pressure in the secondary containment). To ensure that a vent path is available, a 2 inch bypass line is provided around the dampers. The isolation valves on the 18 inch exhaust lines have 2 inch bypass lines around O them for use during normal reactor operation or when the 18 O inch valves cannot be opened. APPLICABLE The PCIVs LC0 was derived from the assumptions related to SAFETY ANALYSES minimizing the loss of reactor coolant inventory, and establishing the primary containment boundary during major accidents. As part of the primary containment boundary, PCIV OPERABILITY supports leak tightness of primary containment. Therefore, the safety analysis of any event requiring isolation of primary containment is applicable to this LCO. The DBAs that result in a release of radioactive material for which the consequences are mitigated by PCIVs are a LOCA and a main steam line break (MSLB). In the analysis for each of these accidents, it is assumed that PCIVs are either closed or close within the required isolation times following event initiation. This ensures that potential paths to the environment through PCIVs (including primary containment purge valves) are minimized. Of the events h (v (continued) HATCH UNIT 2 B 3.6-15 REVISION D
i PCIVs B 3 6.1.3 BASES APPLICABLE analyzed in Reference 1, the MSLB is the most limiting event SAFETY ANALYSIS due to radiological consequences. The closure time of the (continued) main steam isolation valves (MSIVs) is a significant variable from a radiological standpoint. The MSIVs are a required to close withi. 3 to 5 seconds since the 5 second im ' closure time is assumed in the analysis. The safety analyses assume that the purge valves were closed at event initiation. Likewise, it is assumed that the primary containment is isolated such that O (continued) HATCH UNIT 2 B 3.6-15A REVISION D
i PCIVs B 3.6.1.3- l .g y BASES ; APPLICABLE release of fission products to the environment is SAFETY ANALYSES controlled. i (continued) i The single failure criterion required to be imposed in the > conduct of unit safety analyses was considered in the l original design of the primary containment purge valves. Two valves in series on each purge line provicie assurance that both the supply and exhaust lines could be isolated [ ' even if a. single failure occurred. > PCIVs satisfy Criterion 3 of the NRC Policy Statement ! (Ref. 6). ; LCO PCIVs form a part of the primary containment boundary. The PCIV safety function in related to minimizing the loss of reactor coolant inventory and establishing the primary ; containment boundary dJring a DBA. The power operated and the automatic isolation valves are required to have isolation times within limits and the automatic isolation valves actuate on an automatic isolation signal. While the reactor building-to-suppression chamber vacuum breakers isolate 3rimary containment penetrations, , they are excluded from t11s Specification. Controls on i their isolation . function are adequately addressed in LCO j 3.6.1.7, " Reactor Building-to-Suppression Chamber Vacuum Breakers." The valves covered by this LCO are listed with their associated stroke times in Reference 2. i The normally closed PCIVs are considered OPERABLE when manual valves are closed, or open in accordance with appropriate administrative controls, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact. These passive isolation valves and devices are those listed in Reference 2. Secondary containment bypass valves and MSIVs must meet additional leakage rate requirements. Other PCIV leakage rates are addressed by LCO 3.6.1.1, " Primary Containment," as Type B or C testing. This LC0 provides assurance that the PCIVs will perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the primary containment boundary during accidents. (continued) HATCH UNIT 2 B 3.6-16 REVISION D
PCIVs B 3.6.1.3 v
) BASES (continued)
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, most PCIVs are not required to be OPERABLE and the primary containment purge valves are not required to be sealed closed in MODES 4 and 5. Certain valves, however, are required to be OPERABLE to prevent inadvertent reactor vessel draindown. These valves are those whose associated instrumentation is required to be OPERABLE per LC0 3.3.6.1, " Primary Containment Isolation Instrumentation." (This does not include the valves that isolate the associated instrumentation.) ACTIONS The ACTIONS are modified by a Note allowing penetration flow - path (s) except for 18 inch purge valve flow path (s) to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the (Q penetration can be rapidly isolated when a need for primary t/ containment isolation is indicated. Due to the size of the primary containment purge supply and exhaust line penetrations and the fact that those penetrations exhaust directly from the containment atmosphere to the environment (via the SGT Sy',tems), the penetration flow path containing these valves is not allowed to be opened under administrative controls. A second Note has been added to provide clarification that, for the purpose of this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable PCIV. Complying with the Required Actions may allow for continued operation, and subsequent inoperable PCIVs are governed by subsequent Condition entry and application of associated Required Actions. The ACTIONS are modified by Notes 3 and 4. Note 3 ensures that appropriate remedial actions are taken, if necessary, if the affected system (s) are rendered inoperable by an inoperable PCIV (e.g., an Emergency Core Cooling System subsystem is inoperable due to a failed open test return () LJ (continued) HATCH UNIT 2 B 3.6-17 REVISION A
PCIVs B 3.6.1.3 BASES h ACTIONS valve). Note 4 ensures appropriate remedial actions are (continued) taken when the primary containment leakage limits are exceeded. Pursuant to LC0 3.0.6, these actions are not required even when the associated LC0 is not met. Therefore, Notes 3 and 4 are added to require the proper actions be taken. lg A.1 and A.2 With one or more penetration flow paths with one PCIV inoperable except for inope ability due to leakage not within a limit specified in an SR to this LCO, the affected penetration flow paths must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, a blind flange, and a check valve with flow through the valve secured. For a penetration isolated in accordance with Required Action A.1, the device used to isolate the penetration should be the closest available valve to the primary containment. The Required Action must be completed within the 4 hour Completion Time (8 hours for main steam lines). The Completion Time of 4 hours is reasonable considering the time required to isolate the penetration and the relative importance of supporting primary containment OPERABILITY during MODES 1, 2, and 3. For main steam lines, an 8 hour Completion Time is allowed. The Completion Time of 8 hours for the main steam lines allows a period of time to restore the MSIVs to OPERABLE status given the fact that MSIV closure will result in isolation of the main steam line(s) and a potential for plant shutdown. For affected penetrations that have been isolated in accordance with Required Action A.1, the affected penetration flow path must be verified to be isolated on a periodic basis. This is necessary to ensure that primary containment penetrations required to be isolated following an accident, and no longer capable of being automatically isolated, will be in the isolation position should an event occur. This Required Action does not require any testing or device manipulation. Rather, it involves verification that those devices outside containment and capable of potentially being mispositioned are in the correct position. The (continued) HATCH UNIT 2 B 3.6-18 REVISION D
]
PCIVs B 3.6.1.3 BASES v ACTIONS A.1 and A.2 (continued) Completion Time of "Once per 31 days for isolation devices outside primary containment" is appropriate because the devices are operated under administrative controls and the probability of their misalignment is low. For the devices inside primary containment, the time period specified " Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of the inaccessibility of the , devices and other administrative controls ensuring that device misalignment is an unlikely possibility. Condition A is modified by a Note indicating that this Condition is only applicable to those penetration flow paths with two PCIVs. For penetration flow paths with one PCIV, Condition C provides the appropriate Required Actions. Required Action A.2 is modified by a Note that applies to isolation devices located in high radiation areas, and allows them to be verified by use of administrative means. rQ Allowing verification by administrative means is considered C' 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. EL1 With one or more penetration flow paths with two PCIVs inoperable except due to leakage not within limits, either > the inoperable PCIVs must be restored to OPERABLE status or the affected penetration flow path must be isolated within I hour. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration. The 1 hour Completion Time is consistent with the ACTIONS of LC0 3.6.1.1. (g) (continued) HATCH UNIT 2 B 3.6-19 REVISION A
PCIVs j 8 3.6.1.3 BASES h ACTIONS fl.1 (continued) Condition B is modified by a Note indicating this Condition is only applicable to penetration flow paths with two PCIVs. For penetration flow paths with one PCIV, Condition C provides the appropriate Required Actions. C.1 and C.2 With one or more penetration flow paths with one PCIV inoperable, except due to leakage not within limits, the iJ inoperable valve must be restored to OPERABLE status or-the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration. Required Action C.1 must be completed within 4 hours for lines other than excess flow check valve (EFCV) lines and 12 hours for EFCV lines. The Completion Time of 4 hours is reasonable considering the relative stability of the closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of supporting primary containment OPERABILITY during MODES 1, 2, and 3. The Completion Time of 12 hours is reasonable considering the instrument to act as a penetration isolation boundary and the small pipe diameter of the affected penetrations. In the event the affected penetration flow path is isolated in accordance with Required Action C.1, the affected penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that primary containment penetrations required to be isolated following an accident are isolated. The Completion Time of once per 31 days for verifying each affected penetration is isolated is appropriate because the valves are operated under administrative controls and the probability of their misalignment is low. Condition C is modified by a Note indicating that this Condition is only applicable to penetration flow paths with only one PCIV. For penetration flow paths with two PCIVs, Conditions A and B provide the appropriate Required Actions. 1 (continued) i HATCH UNIT 2 B 3.6-20 REVISION D 1 l
r PCIVs B 3.6.1.3 (v ) BASES ACTIONS C.1 and C.2 (continued) Required Action C.2 is modified by a Note that applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is low. D_d With the secondary containment bypass leakage rate or MSIV leakage rate not within limit, the assumptions of the safety analysis may not be met. Therefore, the leakage must be restored to within limit within 4 hours. Restoration can be accomplished by isolating the penetration that caused the limit to be exceeded by use of one closed and de-activated automatic valve, closed manual valve, or blind flange. When a penetration is isolated, the leakage rate for the isolated penetration is assumed to be the actual pathway leakage (7 ( ,/ through the isolation device. If two isolation devices are used to isolate the penetration, the leakage rate is assumed to be the lesser actual pathway leakage of the two devices. The 4 hour Completion Time is reasonable considering the time required to restore the leakage by isolating the penetration and the relative importance to the overall containment function. E.1 and E.2 If any Required Action and associated Completion Time cannot be met 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 conditions in an orderly manner and without challenging plant systems. j l Q (continued) HATCH UNIT 2 B 3.6-21 REVISION A . l l 1
PCIVs B 3.6.1.3 BASES h ACTIONS F.1 and F.2 (continued) If any Required Action and associated Completion Time cannot be met, the unit must be placed in a condition in which the LCO does not apply. 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 subsequent potential for fission product release. Actions must continue until 0PDRVs are suspended and the valve (s) are restored to OPERABLE status. If suspending an OPDRV would result in closing the residual heat removal (RHR) shutdown cooling isolation valves, an alternative Required Action is provided to immediately initiate action to restore the valve (s) to OPERABLE status. This allows RHR shutdown cooling to remain in service while actions are being taken to restore the valve. SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS This SR ensures that the 18 inch primary containment purge valves are closed as required or, if open, are open for an allowable reason. If a purge valve is open in violation of this SR, the valve is considered inoperable (Condition A applies). The SR is modified by a Note stating that the SR is not required to be met when the 18 inch purge valves are open for the stated reasons. The Note states that these valves may be opened for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open. The 18 inch purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The 31 day Frequency is consistent with other PCIV requirements discussed in SR 3.6.1.3.2. I SR 36.1.31 This SR verifies that each primary containment isolation manual valve and blind flange that is located outside primary containment and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits. (continued) h HATCH UNIT 2 B 3.6-22 REVISION D
=
P PCIVs. B 3.6.1.3 l
' BASES SURVEILLANCE SR 3.6.1.3.2 (continued) '
REQUIREMENTS This SR does not require any testing or valve manipulation. Rather, it involves verification that those isolation i devices outside primary containment, and capable of being mispositioned, are in the correct position. Since : verification of valve position for isolation devices outside ' primary containment is relatively easy, the 31 day Frequency , was chosen to provide added assurance that the isolation . devices are in the correct positions. i Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas 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 l isolation devices, once they have been verified to be in the " proper position, is low. A second Note has been included te , clarify that PCIVs that are open under administrative controls are not required to meet the SR during the time ; that the PCIVs are open. SR 3.6.1.3.3 This SR verifies that each primary containment manual ! isolation valve and blind flange that is located inside primary containment and is required to be closed during i accident conditions is closed. The SR helps-to ensure that- ! post accident leakage of radioactive fluids or gases outside j the primary containment boundary is within design limits. i For these isolation devices inside primary containment, the i Frequency defined as " Prior to entering MODE 2 or 3 from ! MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 22 days" is appropriate since these isolation devices are operated under administrative controls and the probability of their misalignment is low. I Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas to , be verified by use of administrative controls. Allowing i verification by administrative controls is considered ) 1 l (continued) HATCH UNIT 2 B 3.6-23 REVISION A_ j
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.3 (continued) REQUIREMENTS < acceptable since the primary containment is inerted and access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA and personnel safety reasons. Therefore, the probability of misalignment of these isolation devices, once they have been verified to be in their proper position, is low. A second Note has been included to clarify that PCIVs that are open under administrative controls are not required to meet the SR during the time that the PCIVs are open. t SR 3.6.1.3.4 : The traversing incore probe (TIP) shear isolation valves are actuated by explosive charges. Surveillance of explosive charge continuity provides assurance that TIP valves will actuate when required. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The 31 day Frequency is based on operating experience that has demonstrated the reliability of the explosive charge continuity. SR 3.6.1.3.5 Verifying the isolation time of each power operated and each automatic PCIV is within limits is required to demonstrate OPERABILITY. MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.6. A The isolation time test ensures that each valve will isolate ; in a time period less than or equal to that listed in the i FSAR and that no degradation affecting valve closure since the performance of the last surveillance has occurred. j (EFCVs are not required to be tested because they have no ' specified time limit). The Frequency of this SR is in 3 accordance with the requirements of the Inservice Testing Program. SR 3.6.1.3.6 Verifying that the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY. The isolation time test ensures that the MSIV will isolate k (continued) h HATCH UNIT 2 B 3.6-24 REVISION D
PCIVs B 3.6.1.3 C) BASES SURVEILLANCE SR 3.6.1.3.6 (continued) REQUIREMENTS in a time period that does not exceed the times assumed in the DBA analyses. This ensures that the calculated radiological consequences of these events remain within 10 g CFR 100 limits. The frequency of this SR is in accordance with the requirements of the Inservice Testing Program. SR 3.6.1.3.7 lk Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position bl v I s []J t (continued) HATCH UNIT 2 B 3.6-24A REVISION D
PCIVs B 3.6.1.3 im (v) BASES SURVEILLANCE SR 3.6.1.3.7 (continued) e 1 REQUIREMENTS on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.1.6 overlaps this SR to provide complete testing of the safety function. The 18 month Frequency was developed considering it is prudent that this Surveillance be performed only during a unit outage since isolation of penetrations would eliminate cooling water flow and disrupt the normal operation of many critical components. Operating experience has shown that these ! components usually pass this Surveillance when performed at I the 18 month Frequency. Therefore, the Frequency was I concluded to be acceptable from a reliability standpoint. SR 3.6.1.3.8 I This SR requires a demonstration that each reactor instrumentation line excess flow check valve (EFCV) is OPERABLE by verifying that the valve reduces flow to within limits on an actual or simulated instrument line break condition. This SR provides assurance that the
/ instrumentation line EFCVs will perform as designed. The i C] 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant l
outage and the potential for an unplanned transient if the {' Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass this Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. SR 3.6.1.3.9 l The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Frequency of 18 months on ' STAGGERED TEST BASIS is considered adequate given the
, ad- strative controls on replacement charges and the frequent checks of circuit continuity (SR 3.6.1.3.4).
O) b (continued) HATCH UNIT 2 B 3.6-25 REVISION D
PCIVs B 3.6.1.3 BASES h SURVEILLANCE SR 3.6.1.3.10 l b REQUIREMENTS (continued) This SR ensures that the leakage rate of secondary containment bypass leakage paths is less than the specified leakage rate. This provides assurance that the assumptions in the radiological evaluations that form the basis of the FSAR (Ref. 3) are met. The secondary containment bypass leakage paths are: 1) main steam condensate drain, penetration 8; 2) reactor water cleanup, penetration 14; 3) equipment drain sunp discharge, penetration 18; 4) floor drain sump discharge, penetration 19; and 5) chemical drain sump discharge, penetration 55. The leakage rate of each bypass leakage path is assumed to be the maximum pathway leakage (leakage through the worse of the two isolation valves) unless the penetration is isolated by use of one closed and de-activated automatic valve, closed manual valve, or blind flange. In this case, the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two valves. This method of quantifying maximum pathway leakage is only to be used for this SR (i.e., Appendix J maximum pathway- & leakage limits are to be quantified in accordance with W Appendix J). The Frequency is required by 10 CFR 50, Appendix J, as modified by approved exemptions (and therefore, the Frequency extensions of SR 3.0,2 may not be applied), since the testing is an Appendix J, Type C test. This SR simply impo. s additional acceptance criteria. SR 3.6.1.3.11 Ik The analyses in References 1 and 4 are based on leakage that is less than the specified leakage rate. Leakage through 1 each MSIV must be s 100 scfh, and a combined maximum pathway leakage s 250 scfh for all four main steam lines, when tested at 2: 28.8 psig. In addition, if any MSIV exceeds the 100 scfh limit, the as left leakage shall be s 11.5 scfh for that MSIV. The MSIV leakage rate must be verified to be in accordance with the leakage test requirements of 10 CFR 50, Appendix J (Ref. 5), as modified by approved exemptions. This ensures that MSIV leakage is properly accounted for in determining the overall primary containment leakage rate. (continued) HATCH UNIT 2 B 3.6-26 REVISION D
l PCIVs B 3.6.1.3 l { BASES ; SURVEILLANCE SR 3.6.1.3.11 (continued) Ik REQUIREMENTS ! The Frequency is required by 10 CFR 50, Appendix J, as ! modified by approved exemptions; thus, SR 3.0.2 (which allows Frequency extensions) does not apply. SR 3.6.1.3.12 lb The valve seats of each 18 inch purge valve (supply and exhaust) having resilient material seats must be replaced every 18 months. This will allow the opportunity for repair , before gross leakage failure develops. The 18 month Frequency is based on engineering judgment and operational ! experience which shows that gross leakage normally does not i occur when the valve seats are replaced on an 18 month l Frequency. i SR ~3.6.1.3.13 l l The Surveillance Requirement provides assurance that the i excess flow isolation dampers can close following an isolation signal. The 18 month Frequency is based on vendor recommendations and engineering judgment. Operating experience has shown that these dampers 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, Chapter 15. , t
- 2. Technical Requirements Manual.
- 3. FSAR, Section 15.1.39.
- 4. FSAR, Section 6.2. *
- 5. 10 CFR 50, Appendix J.
- 6. NRC No. 93-102, " Final Policy Statement on Technical ,
Specification Improvements," July 23, 1993. i O ! i HATCH UNIT 2 B 3.6-27 REVISION D ( I
Drywell Pressure B 3.6.1.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.4 Drywell Pressure BASES BACKGROUND The drywell pressure is limited during normal operations to preserve the initial conditions assumed in the accident analysis for a Design Basis Accident (DBA) or loss of coolant accident (LOCA). APPLICABLE Primary containment performance is evaluated for the entire SAFETY ANALYSES spectrum of break sizes for postulated LOCAs (Ref.1). Among the inputs to the DBA is the initial primary containment internal pressure (Ref.1). Analyses assume an initial drywell pressure of 1.75 psig. This limitation l ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that , the peak LOCA drywell internal pressure does not exceed the ' maximum allowable of 62 psig. The maximum calculated drywell pressure occurs during the reactor blowdown phase of the DBA, which assumes an W instantaneous recirculation line break. The calculated peak drywell pressure for this limiting event is 48.7 psig l (Ref. 1). Drywell pressure satisfies Criterion 2 of the NRC Policy Statement (Ref. 2). LC0 In the event of a DBA, with an initial drywell pressure s 1.75 psig, the resultant peak drywell accident pressure l will be maintained below the drywell design pressure. APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 (continued) HATCH UNIT 2 B 3.6-28 A REVISION /
j LLS Valves , B 3.6.1.6 1
. BASES APPLICABLE assumption that simultaneous S/RV openings occur only on the I SAFETY ANALYSES initial actuation for DBAs. Even though four LLS S/RVs are (continued) specified, all four LLS S/RVs do not operate in any DBA analysis.
LLS valves satisfy Criterion 3 of the NRC Policy Statement (Ref. 3). l l LC0 Four LLS valves are required to be OPERABLE to satisfy the I assumptions of the safety analyses (Ref. 1). The requirements of this LC0 are applicable to the mechanical and electrical / pneumatic capability of the LLS valves to function for controlling the opening and closing of the S/RVs. APPLICABILITY In MODES 1, 2, and 3, an event could cause pressurization of the reactor and opening of S/RVs. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. O Therefore, maintaining the LLS valves OPERABLE is not required in MODE 4 or 5. ACTIONS- &J With one LLS valve inoperable, the remaining OPERABLE LLS valves are adequate to perform the designed function. However, the overall reliability is reduced. The 14 day Completion Time takes into account the redundant capability-afforded by the remaining LLS valves and the low probability of an event in which the remaining LLS valve capability would be inadequate. B.1 and B.2 If two or more LLS valves are inoperable or if the inoyerable LLS valve cannot be restored to OPERABLE status wit 11n the required Completion Time, .the plant must be brought to a MODE in which the LCO does not apply. To (continued) HATCH UNIT 2 B 3.6-35 REVISION A
LLS Valves B 3.6.1.6 BASES h, ACTIONS B.1 and 8.2 (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. SURVEILLANCE SR 3.6.1.6.1 REQUIREMENTS A manual actuation of each LLS valve is performed to verify that the valve and solenoids are functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control or bypass valve, by a change in the measured steam flow, or by any other method that is suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Adequate pressure at which this test is to be performed is a 920 psig ; (the pressure recommended by the valve manufacturer). Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the LLS valves divert steam flow upon opening. Adequate steam flow is represented by at least 1.25 turbine bypass valves open, or total steam flow 2 IE6 lb/hr. The o 18 month Frequency was based on the S/RV tests required by the ASME Boiler and Pressure Vessel Code, Section XI (Ref. 2). Operating experience has shown that 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. ! Lince steam pressure is required to perform the i Surveillance, however, and steam may not be available during i a unit outage, the Surveillance may be performed during the startup following a unit outage. Unit startup is allowed ; prior to performing the test because valve OPERABILITY and l the setpoints for overpressure protection are verified by ASME Section XI testing prior to valve installation. After adequate reactor steam pressure and flow are reached, 12 hours is allowed to prepare for and perform the test. Adequate pressure at which this test is to be performed is , consistent with the pressure recommended by the valve ' manufacturer. 1 (continued) HATCH UNIT 2 B 3.6-36 REVISION D
Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.8 ; R
~h. l (Q - BASES (continued)
ACTIONS M With one of the required vacuum breakers inoperable for . opening (e.g., the vacuum breaker is not open and may be ! stuck closed or not within its opening setpoint limit, so that it would not function as designed during an event that depressurized the drywell), the remaining nine OPERABLE. vacuum breakers are capable of providing the vacuum relief function. However, overall system reliability is reduced because a single failure in one of the remaining vacuum breakers could result in an excessive suppression chamber-to-drywell differential pressure during a DBA. Therefore, with one of the 10 required vacuum breakers inoperable, 72 hours is allowed to restore at least one of the inoperable vacuum breakers to OPERABLE status so that plant conditions are consistent with those assumed for the design I basis analysis. The 72 hour Completion Time-is considered acceptable due to the low probability of an event in which the remaining vacuum breaker capability would not be adequate.
.A M V
An open vacuum breaker allows communication between the drywell and suppression chamber airspace, and, as a result, there is the potential for suppression chamber overpressurization due to this bypass leakage if a LOCA were to occur. Therefore, the open vacuum breaker must be closed. A short time is allowed to close the vacuum breaker due to the low probability of an event that would pressurize primary containment. If vacuum breaker position indication is not reliable, an alternate method of verifying.that the vacuum breakers are closed is to verify that a differential pressure of > 0.5 psid between the drywell and suppression chamber is maintained for 1-hour without makeup. The required 2 hour Completion Time is considered adequate to perform this test. C.1 and C.2 If any Required Action and associated Completion Time 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
.(continued)
HATCH UNIT 2 B 3.6-47 REVISION A
Suppression Chamber-to-Drywell Vacuum Breakers ' B 3.6.1.8 BASES ACTIONS C.1 and C.2 (continued) 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. SURVEILLANCE SR 3.6.1.8.1 REQUIREMENTS Each vacuum breaker is verified closed to ensure that this potential large bypass leakage path is not present. This Surveillance is performed by observing the vacuum breaker position indication or by verifying that a differential pressure of 0.5 psid between the drywell and suppression chamber is maintained for 1 hour without makeup. The 14 day Frequency is based on engineering judgment, is considered adequate in view of other indications of vacuum breaker status available to operations personnel, and has been shown to be acceptable through operating experience. A Note is added to this SR which allows suppression chamber-to-drywell vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers. SR 3.6.1.8.2 Each required vacuum breaker must be cycled to ensure that it opens adequately to perform its design function and returns to the fully closed position. This ensures that the safety analysis assumptions are valid. The 31 day Frequency of this SR was developed, based on Inservice Testing Program requirements to perform valve testing at least once every 92 days. A 31 day Frequency was chosen to provide additional assurance that the vacuum breakers are OPEPABLE, since they are located in a harsh environment (the suppression chamber airspace). In addition, this functional test is required within 12 hours after a discharge of steam D to the suppression chamber from the safety / relief valves. 1 (Continued) HATCH UNIT 2 B 3.6-48 REVISION D l
Secondary Containment-Operating B 3.6.4.1
) BASES (continued)
APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to Unit I and Unit 2 secondary containments. Therefore, Unit I secondary containment and Unit 2 secondary containment opt ABILITY is required during the same operating conditions that require Unit 2 primary containment OPERABILITY. Secondary Containment requirements for MODES 4 and 5 are covered by LCOs 3.6.4.2 and 3.6.4.3, " Secondary Containment- OPDRVS" and "-Refueling," respectively. ACTIONS L1 If one or both units secondary containments are inoperable, it must be 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 maintaining secondary containment during MODES 1, 2, and 3. , This time period also ensures that the probability of an accident (requiring Unit I and Unit 2 secondary containments 3 OPERABILITY) occurring during periods where secondary (V containment is inoperable is minimal. B.1 and B.Z If both secondary containments cannot be restored to OPERABLE status within the required Completion Time, the plant r,;ust 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. i N (continued) l HATCH UNIT 2 B 3.6-83 REVISION A
Secondary Containment-0perating B 3.6.4.1 BASES (continued) h SURVEILLANCE SR 3.6.4.1.1 and SR 3.6.4.1.2 REQUIREMENTS Verifying that Unit 1 and Unit 2 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 containments will not occur. SR 3.6.4.1.1 also requires A equipment hatches to be sealed. In this application, the /B term " sealed" has no connotation of leak tightness. Maintaining Unit I and Unit 2 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 modified Unit I secondary containment configuration is used, these SRs also include verifying the hatches and doors separating the common refueling floor zone from the Unit I reactor building. 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. SR 3.6.4.1.4. SR 3.6.4.1.5. and 3.6.4.1.6 The Unit 2 SGT System exhausts the Unit 2 secondary containment atmosphere and the Unit 1 and Unit 2 SGT Systems exhaust the Unit I secondary containment atmosphere to the environment through appropriate treatment equipment. To ensure that all fission products are treated, SR 3.6.4.1.3 and SR 3.6.4.1.4 verify that the appropriate SGT System will rapidly establish and maintain a pressure in the secondary containments that is less than the lowest postulated l pressure external to the secondary containment boundary. i This is confirmed by demonstrating that one Unit 2 SGT subsystem will draw down the Unit 2 secondary containment and two SGT subsystems (one Unit I and one Unit 2) will draw down the Unit 1 secondary containment to 2: 0.25 inch of vacuum water gauge in s 120 seconds. This cannot be accomplished if the secondary containment boundary is not intact. SR 3.6.4.1.5 and SR 3.6.4.1.6 demonstrate that one Unit 2 SGT subsystem and two SGT subsystems (one Unit 1 and l l (continued) i HATCH UNIT 2 B 3.6-84 REVISION D i
Secondary Containment-Operating-B 3.6.4.1
. , i
(>%,
) BASES (continued) one Unit 2) can maintain Unit 2'and Unit I secondary containments, respectively, 2 0.25 inch of vacuum water
_ gauge for I hour at a flow rate s ,4000 cfm for. Unit 2 - secondary containment and. s 8000 cfm (s .4000 cfm for each of
- two subsystems) for Unit 1 secondary containment. ,The Y
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HATCH UNIT 2 B 3.6-84A REVISION D +
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i Secondary Containment-0PDRVs B 3.6.4.2 (j. BASES (continued) APPLICABILITY 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 Unit 2 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), since this condition could lead to an inadvertent vessel draindown event. Secondary containment requirements for MODES 1, 2 and 3, and during other conditions for which significant releases of radioactive material can be postulated, are covered by LCOs 3.6.4.1 and 3.6.4.3, " Secondary Containment-0perating" and
-Refueling," respectively.
ACTIONS .A_d OPDRVs can be postulated to cause fission product release to the Unit 2 secondary containment. In such cases, the Unit 2 secondary containment is the only barrier to release of fission products to the environment. Therefore, if the Unit 2 secondary containment is inoperable 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. SURVEILLANCE SR 3.6.4.2.1 and SR 3.6.4.2.2 REQUIREMENTS Verifying that Unit 2 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 3.6.4.2.1 also requires equipment hatches to be sealed. In this application, the term " sealed" has no connotation of leak tightness. hD Maintaining Unit 2 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). The 31 day Frequency for these SRs has been shown to be adequate, based on operating experience, and is considered adequate in view of
) (continued)
HATCH UNIT 2 B 3.6-87 REVISIONgg
Secondary Containment-0PDRVs B 3.6.4.2 BASES SURVEILLANCE SR 3.6.4.2.1 and SR 3.6.4.2.2 (continued) REQUIREMENTS the other indications of door and hatch status that are available to the operator. SR 3.6.4.2.3 and SR 3.6.4.2.4 The Unit 2 SGT System exhausts the Unit 2 secondary l containment atmosphere to the environment through ' appropriate treatment equipment. To ensure that all fission products are treated, SR 3.6.4.2.3 verifies that the Unit 2 SGT System will rapidly establish and maintain a pressure in the Unit 2 secondary containment that is less than the l lowest postulated pressure external to the secondary containment boundary. This is confirmed by demonstrating that one Unit 2 SGT subsystem will draw down the Unit 2 secondary containment to 2 0.25 inch of vacuum water gauge in s 120 seconds. This cannot be accomplished if the Unit 2 i secondary containment boundary is not intact. SR 3.6.4.2.4 i demonstrates that one Unit 2 SGT subsystem can maintain 2 0.25 inch of vacuum water gauge for 1 hour at a flow rate ' s 4000 cfm. The 1 hour test period allows secondary containment to be in thermal equilibrium at steady state conditions. Therefore, these two tests are used to ensure Unit 2 secondary containment boundary integrity. Since these SRs are secondary containment tests, they need not be performed with each Unit 2 SGT subsystem. The Unit 2 SGT subsystems are tested on a STAGGERED TEST BASIS, however, to ensure that in addition to the requirements of LC0 3.6.4.8, either Unit 2 SGT subsystem will perform this test. 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. 1 REFERENCES 1. FSAR, Section 15.1.39.
- 2. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
O HATCH UNIT 2 8 3.6-88 REVISION A
Secondary Containment-Refueling B 3.6.4.3 l BASES ( ) SURVEILLANCE SR 3.6.4.3.1 and SR 3.6.4.3.2 (continued) REQUIREMENTS containment will not occur. SR 3.6.4.3.1 also requires equipment hatches to be sealed. In this application, the term " sealed" has no connotation of leak tightness. Maintaining Unit I 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 modified Unit 1 secondary containment configuration is used, these SRs also include verifying that hatches and doors separating the common refueling floor zone from the Unit I reactor building. 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 3.6.4.3.3 and SR 3.6.4.3.4 The Unit 1 and Unit 2 SGT Systems exhaust the Unit I p/ t secondary containment atmosphere to the environment through appropriate treatment equipment. To ensure that all fission products are treated, SR 3.6.4.3.3 verifies that the Unit I and Unit 2 SGT Systems will rapidly establish and maintain a pressure in the Unit I secondary containment that is less than the lowest postulated pressure external to the secondary containment boundary. This is confirmed by demonstrating that two SGT subsystems (one Unit I and one Unit 2) will draw down the Unit I secondary containment to a 0.25 inch of vacuum water gauge in s 100 seconds. This cannot be accomplished if the secondary containment boundary is not intact. SR 3.6.4.3.4 demonstrates that two SGT subsystems can maintain 2 0.25 inch of vacuum water gauge : for I hour at a flow rate s 4000 cfm for each subsystem. The 1 hour test period allows secondary containment to be in thermal equilibrium at steady state conditions. Therefore, these two tests are used to ensure Unit I secondary
- containment boundary integrity. Since these SRs are secondary containment tests, they need not be performed with 1 each SGT subsystem. The SGT subsystems are tested on a !
STAGGERED TEST BASIS, however, to ensure that in addition to l 1 l j q (continued) l 1 HATCH UNIT 2 B 3.6-91 REVISION D j i
l u Secondary Containment-Refueling I B 3.6.4.3 BASES h the requirements of LC0 3.6.4.9, each SGT subsystem combination will perform this test. 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 (continued) HATCH UNIT 2 8 3.6-91A REVISION D
Secondary Containment-Refueling B 3.6.4.3 BASES (continued) REFERENCES 1. FSAR, Section 15.1.41.
- 2. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
{N \ HATCH UNIT 2 B 3.6-92 REVISION A
SGT System-Operating B 3.6.4.7 B 3.6 CONTAINMENT SYSTEMS (} B 3.6.4.7 Standby Gas Treatment (SGT) System-Operating 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 for subsystem OPERABILITY). 7.... ( 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 (not required for subsystem OPERABILITY);
- 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. A centrifugal fan.
The sizing of the SGT Systems equipment and compunents is based on the results of an infiltration analysis, as well as an exfiltration analysis of the Unit 1 and Unit 2 secondary containments. The internal pressure of the SGT Systems (continued) HATCH UNIT 2 B 3.6-111 REVISION A
SGT System-Operating B 3.6.4.7 BASES h 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 prov;Jed to remove entrained water in the air, while the electric heater (no credit is taken for heater OPERABILITY) rcduces 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 I and Unit 2 SGT Systems SAFETY ANALYSES during MODES 1, 2, and 3 is to mitigate the consequences of a loss of coolant accident (Refs. 2, 3, and 4). For this event, the SGT Systems are shown to be automatically initiated to reduce, via filtration and adsorption, the radioactive material released to the environment. One SGT subsystem is required to draw-down the Unit 2 secondary containment and two SGT subsystems are_ required to draw-down the Unit I secondary containment. The need for Unit I secondary containment during a Unit 2 LOCA arises because of potential leakage past the Unit 2 drywell head onto the k refueling floor (i.e., into the Unit I secondary containment). The SGT System satisfies Criterion 3 of the NRC Policy Statement (Ref. 5). (continued) HATCH UNIT 2 B 3.6-112 REVISION D
i 1 SGT System 10perating i B 3.6.4.7 1 l
/N
-( j BASES i
.l.
LC0~ Followingla LOCA, a minimum of three SGT subsystems are : l required to maintain'the Unit I and Unit'2 secondary . j containments at a negative pressure with respect to the environment and to process gaseous releases. Meeting-the- ! LCO requirements for four OPERABLE subsystems ensures i operation of at least.three SGT subsystems in the. event:of ~a , single active failure. ,
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NATCH'. UNIT 2 B 3.6-112A REVISION'D-l u I l
-..a_.. . . _ . . . - . _ . . .. - -_
. . . _ - - ~ . . _ _ _ _ _ _-- __. -
AC Sources - Operating B 3.8.1 BASES ACTIONS L1 (continued) With two or more Unit 2 and swing DGs inoperable, with an assumed loss of offsite electrical power, insufficient standby AC sources are available to power the minimum required ESF functions. Since the offsite electrical power system is the only source of AC power for the majority of ESF equipment at this level of degradation, the risk associated with continued operation for a very short time could be less than that associated with an immediate controlled shutdown. (The immediate shutdown could cause grid instability, which could result in a total loss of AC power.) Since any inadvertent unit generator trip could also result in a total loss of-offsite AC power the time allowed for continued operation is severely restricted. The intent here is to avoid the risk associated with an immediate controlled shutdown and to minimize the risk associated with this. level of degradation. According to Regulatory Guide 1.93 (Ref. 6), with two or more DGs inoperable, operation may continue for a period that should not exceed 2 hours. (Regulatory Guide 1.93 . /3 assumed the unit has two DGs. Thus,.a loss of both DGs V results in a total loss of onsite power. Therefore, a loss of more than two DGs, in the Plant Hatch design, results in degradation no worse than that assumed in Regulatory Guide 1.93. In addition, the loss of a required Unit 1 DG concurrent with the loss of a Unit 2 or swing DG, is analogous to the loss of a single DG in the Regulatory Guide i 1.93 assumptions, thus, entry into this condition is not ' required in this case). G.1 and G.2 1 If the inoperable AC electrical power sources cannst be restored to OPERABLE status within the associated Completion - Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be ! brought to at least MODE 3 within 12 hours and to MODE ~4 i within 36 hours. -The allowed Completion Times are 1 reasonable, based on operating experience, to reach the ' required plant conditions from full power conditions in an orderly manner and without challenging plant systems. )
'l l
'l (continued)
HATCH UNIT 2 B 3.8-17 REVISION A l l l
=l
AC Sources - Operating l B 3.8.1 ! BASES i ACTIONS lid (continued) Condition H corresponds to a level of degradation in which all redundancy in the AC electrical power supplies has been lost. At this severely degraded level, any further losses in the AC electrical power system will cause a loss of-function. Therefore, no additional time is justified for continued operation. The unit is required by LC0 3.0.3 to commence a controlled shutdown. SURVEILLANCE The AC sources are designed to permit inspection and REQUIREMENTS testing of all important areas and features, especially those that have a standby function, in accordance with 10 CFR 50, GDC 18 (Ref. 8). Periodic component tests are supplemented by extensive functional tests during refueling outages under simulated accident conditions. The SRs for demonstrating the OPERABILITY of the DGs are generally consisttnt with the recommendations of Regulatory Guide 1.9 (Ref. 3), Regulatory Guide 1.108 (Ref. 9), and Regulatory Guide 1.137 (Ref. 10), although Plant Hatch Unit 2 is not committed to Regulatory Guides 1.108 or 1.137. Specific commitments relative to DG testing is described in FSAR Section 8.3 (Ref. 2). Where the SRs discussed herein specify voltage and frequency tolerances, the following summary is applicable. The allowable values for achieving steady state voltage are specified within a range of minus 10 percent (3740V) and plus 2 percent (4243V) of 4160 V. The Allowable Value of 3740 V is consistent with Regulatory Guide 1.9 for demonstrating that the diesel generator is capable of attaining the required voltage. A more limiting value of 4243 V is specified as the allowable value for overvoltage due to overvoltage limits on the 600 V buses. The plus 2 percent value maintains the required overvoltage limits. The specified minimum and maximum frequencies of the DG are 58.8 Hz and 61.2 Hz, respectively. These values are equal to 2% of the 60 Hz nominal frequency and are derived from the recommendations found in Regulatory Guide 1.9 (Ref. 3). The SRs are modified by a Note to indicate that SR 3.8.1.1 A through SR 3.8.1.18 apply only to the Unit 2 AC sources, and El that SR 3.8.1.19 applies only to the Unit 1 AC sources. (continued) HATCH UNIT 2 B 3.8-18 REVISION D l l I
1 AC Sources - Operating-B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2 (continued) I REQUIREMENTS .
)
Note 6 modifies the Surveillance by stating that. starting ' transients above the upper voltage limit do not invalidate this test. Notes 7 modifies this Surveillance by stating that momentary ; load transients because of changing bus loads do not , invalidate this, test. ! Note 8 indicates that this Surveillance is required to be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations. The normal 31 day Frequency for SR 3.8.1.2 is consistent l l with Regulatory Guide 1.108 (Ref 9). This Frequency provides adequate assurance of DG OPERABILITY, while minimizing degradation resulting from testing. . SR 3.8.1.3 1
, This SR provides verification that the level of fuel oil in the day tank is at or above the level at which fuel oil is automatically added. The level is expressed as an ,
equivalent. volume in gallons, and is selected to ensure r adequate fuel oil for a minimum of I hour of DG operation at full load plus 10%. The actual amount required to meet the SR (900 gallons) will provide approximately 3.5 hours of DG operation at full load. The 31 day Frequency is adequate to ensure that a sufficient supply of fuel oil is available, since low level alarms are provided and operators would be aware of any large uses of. ; fuel oil during this period. , l SR 3.8.1.4
- Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the-fuel oil day tanks once every 184 days eliminates the necessary environment for bacterial survival.
i (continued)-
- HATCH UNIT 2 B 3.8-21 REVISION /
? -c-er r - -
AC Sources - Operating B 3.8.1 BASES h SURVEILLANCE SR 3.8.1.4 (continued) REQUIREMENTS This is a means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment O (continued) HATCH UNIT 2 B 3.8-21A REVISION B
AC Sources - Operating B 3.8.1 [tj BASES SURVEILLANCE SR 3.8.1.4 (continued) REQUIREMENTS in the fuel oil during DG operation. Water in the day tank may come from condensation, rain water, contaminated fuel oil, and breakdown of the fuel oil by bacteria. Checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequency is based on engineering judgment and has shown to be acceptable through operating experience. This SR is for preventive maintenance. The presence of water does not necessarily represeat a failure of this SR provided that accumulated water is removed during performance of this Surveillance. SR 3.8.1.5 This SR helps to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and maintain the unit in a safe shutdown condition. This Surveillance verifies that the DGs are capable of a " fast cold" start, synchronizing, and accepting a load more r~ closely simulating accident loads. A minimum run time of ()T 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source. SR 3.8.1.5 requires that, at a 184 day Frequency, the DG starts from standby conditions and achieves required voltage and frequency within 12 seconds. The 12 second start requirement supports the assumptions in the design-basis LOCA analysis of FSAR, Chapter 6 (Ref. 4). For the purposes of this testing, the DGs are started from standby conditions. Standby conditions for a DG mean that the diesel engine coolant and oil are being continuously circulated and temperature is.being maintained consistent with manufacturer recommendations. Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0. The 0.8 value is the design rating of the machine, while 1.0 is an operational limitation. ()
%j (continued)
HATCH UNIT 2 B 3.8-22 REVISION A l l
AC Sources - Operating B 3.8.1 (g) BASES SURVEILLANCE SR 3.8.1.7 (continued) REQUIREMENTS the overspeed trip. The largest single load for each DG is a residual heat removal service water pump at rated flow (1225 bhp). This Surveillance may be accomplished by: I a) tripping the DG output breaker with the DG carrying greater than or equal to its associated single largest post-accident load while paralleled to offsite power or while lA solely supplying the bus, or b) tripping its associated A single largest post-accident load with the DG solely C supplying the bus. Although Plant Hatch Unit 2 is not committed to IEEE-387-1984, (Ref. 11), this SR is consistent with the IEEE-387-1984 requirement that states the load rejection test is acceptable if the increase in diesel speed does not exceed 75% of the difference between synchronous speed and the overspeed trip setpoint, or 15% above synchronous speed, whichever is lower. For all DGs, this represents 65.5 Hz, equivalent to 75% of the difference between nominal speed and the overspeed trip setpoint. The voltage and frequency specified are consistent with the nominal range for the DG. SR 3.8.1.7.a corresponds to the ( maximum frequency excursion, while SR 3.8.1.7.b is the b] voltage to which the DG must recover following load rejection. The 18 month Frequency is consistent with the recommendation of Regulatory Guide 1.108 (Ref. 9). This SR is modified by two Notes. The reason for Note 1 is that, during operation with the reactor critical, performance of this SR could cause perturbati'ns to the
. electrical distribution systems that could ch 11enge continued steady state operation and, as a result, plant ;
safety systems. Credit may be taken for unplanned events that satisfy this SR. In order to ensure that the DG is tested under load . conditions thht are as close to design basis conditions as possible, testing is performed with only the DG providing power to the associated 4160 V ESF bus. The DG is not synchronized with offsite power. ] l To minimize testing of the swing DG, Note 2 allows a single i test (instead of two tests, one for each unit) to satisfy l the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing A g (continued) HATCH UNIT 2 B 3.8-25 REVISION D
AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.7 (continued) REQUIREMENTS the test on either unit (no unit specific DG components are being tested). If the swing DG fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit. O I I (continued) HATCH UNIT 2 REVISION A B3.8-2/8A
AC Sources - Operating B 3.8.1
) BASES SURVEILLANCE SR 3.8.1.8 REQUIREMENTS (continued) This Surveillance demonstrates the DG capability to reject a full load without overspeed tripping or exceeding the predetermined voltage limits. The DG full load rejection may occur because of a system fault or inadvertent breaker tripping. This Surveillance ensures proper engine generator load response under the simulated test conditions. This test simulates the loss of the total connected load that the DG experiences following a full load rejection and verifies ,
that the DG does not trip upon loss of the load. These acceptance criteria provide DG damage protection. While the DG is not expected to experience this transient during an event, and continues to be available, this response ensures that the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be corrected or isolated. In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, testing must be performed using a power factor s 0.88. This power factor is chosen to be representative of O V the actual design basis inductive loading that the DG would experience. The 18 month Frequency is consistent with the recommendation of Regulatory Guide 1.108 (Ref. 9) and is intended to be consistent with expected fuel cycle lengths. This SR is modified by thr2e Notes. The reason for Note 1 Ib is that during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that would challenge continued steady state operation and, as a result, plant safety systems. Credit may be taken for unplanned events x that satisfy this SR. Note 2 is provided in recognition I dh that if the offsite electrical power distribution system is lightly loaded (i.e., system voltage is high), it may not be possible to raise voltage without creating an overvoltr.ge condition on the ESF bus. Therefore, to ensure the bus voltage, supplied ESF loads, and DG are not placed in an unsafe condition during this test, the power factor limit does not have to be met if grid voltage or ESF bus loading does not permit the power factor limit to be met when the DG is tied to the () / (continued) HATCH UNIT 2 B 3.E-26 REVISION D
AC Sources - Operating B 3.8.1 (gj BASES SURVEILLANCE SR 3.8.1.8 (continued) REQUIREMENTS grid. When this occurs, the power factor should be maintained as close to the limit as practicable. To A minimize testing of the swing DG, Note 3 allows a single I LD_\ test (instead of two tests, one for each unit) to satisfy the requirements for both units. This is allowed since the main purpose of the Surveillance can be met by performing the test on either unit (no unit specific DG components are being tested). If the swing DG fails one of these Surveillances, the DG should be considered inoperable on both units, unless the cause of the failure can be directly related to only one unit. SR 3.8.1.9 This Surveillance demonstrates the as designed operation of ; the standby power sources during loss of the offsite source
~
and is consistent with Regulatory Guide 1.108 (Ref. 9), paragraph 2.a.(1). This test verifies all actions encountered from the loss of offsite power, including shedding of the nonessential loads and energization of the V] / emergency buses and respective loads from the DG. It further demonstrates the capability of the DG to automatically achieve the required voltage and frequency within the specified time. The DG auto-start time of 12 seconds is derived from requirements of the accident analysis for responding to a design basis large break LOCA. The Surveillance should be ; continued for a minimum of 5 minutes in order to demonstrate that all starting transients have decayed and stability has been achieved. The requirement to verify the connection and power supply of permanent and auto-connected loads is intended to satisfactorily show the relationship of these loads to the DG loading logic. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation. For instance, Emergency Core Cooling Systems (ECCS) injection valves are not desired to be stroked open, or systems are not capable of being operated at full flow, or RHR systems performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation. In lieu of p (continued) HATCH UNIT 2 B 3.8-27 REVISION D 9
1 AC Sources - Operating f B 3.3.1 BASES h SURVEILLANCE SR 3.8.1.9 (continued) REQUIREMENTS actual demonstration of the connection and loading of these loads, testing that adequately shows the capability of the ' DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified. For the purpose of this testing, the DGs shall be started from standby conditions, that is, with the engine coolant and oil being continuously circulated and j temperature maintained consistent with manufacturer ' recommendations. The Frequency of 18 months is consistent with the recommendations of Regulatory Guide 1.108 (Ref. 9), paragraph 2.a.(1), takes into consideration plant conditions { required to perform the Surveillance, and is intended to be i consistent with expected fuel cycle lengths. ! This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. The reason for Note 2 is that performing the Surveillance would remove a required offsite circuit from service, perturb the i electrical distribution system, and challenge safety ' systems. Credit may be taken for unplanned events that l satisfy this SR. This Surveillance tests the applicable l logic associated with the Unit 2 swing bus. The comparable - test specified in the Unit 1 Technical Specifications tests the applicable logic associated with the Unit I swing bus. Consequently, a test must be performed within the specified Frequency for each unit. The Note specifying the l restriction for not performing the test while the unit is in MODE 1, 2, or 3 does not have applicability to Unit 1. As the Surveillance represents separate tests, the Unit 2 Surveillance should not be performed with Unit 2 in MODE 1, i 2, or 3 and the Unit 1 test should not be performed with ' Unit 1 in MODE 1, 2, or 3. l SR 3.8.1.10 This Surveillance demonstrates that the DG automatically starts and achieves the required voltage and frequency within the specified time (12 seconds) from the design basis t (continued)
- HATCH UNIT 2 B 3.8-28 REVISION A I
L AC Sources - Operating B 3.8.1 BASES SURVElLLANCE SR 3.8.1.17 -(continued) REQUIREMENTS electrical distribution system, and challenge safety i systems. Credit may be taken for unplanned events that satisfy this SR. This Surveillance tests the applicable logic associated with the Unit 2 swing bus. The comparable test specified in_ the Unit 1 Technical Specifications tests the applicable logic associated with the Unit I swing bus. Consequently, a test must be performed within the specified Frequency for each unit. The Note specifying the restriction for not performing the test while the unit is in MODE 1, 2, or 3 does not have applicability to Unit 1. As-the Surveillance represents separate tests, the Unit 2 Surveillance should not be performed with Unit 2 in MODE 1, . 2, or 3 and the Unit I test should not be performed with ! Unit 1 in MODE 1, 2, or 3. j l SR 3.8.1.18 This Surveillance demonstrates that the DG starting independence has not been compromised. Also, this Surveillance demonstrates that each engine can achieve-b'~h proper speed within the specified time wher the DGs are i started simultaneously. For the purpose of this testing, .{ the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations. It is permissible to place all three DGs l in test simultaneously, for the performance of this l Surveillance. The 10 year Frequency is consistent with the recommendations of Regulatory Guide 1.108 (Ref. 9). This SR is modified by_ a Note. The reason for the Note is to minimize wear on the DG during testing. SR 3.8.1.19
;, With the exception of this Surveillance, all other "- Surveillances of this Specification (SR 3.8.1.1 through SR 3.8.1.18) are applied only to the Unit 2 DG and offsite-circuits, and swing DG. This Surveillance is provided to direct that the appropriate Surveillances for the required Unit 1 DG and offsite circuit are governed by the Unit 1 (continued)
HATCH UNIT 2 B 3.8-37 REVISION:A
AC Sources - Operating B 3.8.1 BASES h $URVEILLANCE SR 3.8.1.19 REQUIREMENTS Technical Specifications. Performance of the applicable Unit 1 Surveillances will satisfy both any Unit I requirements, as well as satisfying this Unit 2 Surveillance requirement. Several exceptions are noted to the Unit 1 SRs: SR 3.8.1.6 is excepted since only one Unit I circuit i8 is required by the Unit 2 Specification (therefore, there is not necessarily a second circuit to transfer to); A SRs 3.8.1.10,11,15 and 17 are excepted since they relate to the DG response to a Unit 1 ECCS initiation signal, which
]p is not a necessary function for support of the Unit 2 requirement for an OPERABLE Unit 1 DG; and SR 3.8.1.18 is excepted since there is only one Unit 1 DG required by the 4 Unit 2 Specification (therefore, there are not necessarily 14 multiple DGs for simultaneous start).
The Frequency required by the applicable Unit 1 SR also governs performance of that SR for both Units. 9 (continued) HATCH UNIT 2 B 3.8-38 REVISION D
AC Sources - Shutdown B 3.8.2 d,~ BASES APPLICABLE During MODES 1, 2, and 3, various deviations from the SAFETY ANALYSES analysis assumptions and design requirements are allowed (continued) within the ACTIONS. This allowance is in recognition that certain testing and maintenance activities must be conducted, provided an acceptable level of risk is not 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 administrative 1y 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.
- 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
(]c systems.
- d. Maintaining, to the extent practical, the ability to perfor:n required functions (even if not meeting MODES 1, 2, and 3 OPERABILITY requirements) with systems assumed to function during an event.
In the event of an accident during shutdewn, this LC0 ensures the capability of supporting systems necessary for avoiding immediate difficulty, assuming either a loss of all offsite power or a loss of all onsite (diesel generator (DG)) power. 1 The AC sources satisfy Criterion 3 of the NRC Policy Statement (Ref. 1). LCO One Unit 2 offsite circuit capable of supplying the onsite Class IE power distribution subsystem (s) of LC0 3.8.8, Ik .
" Distribution Systems - Shutdown," ensures that all required j Unit 2 loads are powered from offsite power. An OPERABLE l Unit 2 DG, C,, (continued) l HATCH UNIT 2 B 3.8-41 REVISION D l
? AC Sources - Shutdown B 3.8.2 BASES LCO associated with a Distribution System Engineered Safety (continued) Feature (EST) 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 2 are powered from Unit I sources (e.g., Standby Gas Treatment (SGT) System). Therefore, one qualified circuit between the offsite transmission network and the onsite Unit 1 Class IE Distribution System, and one Unit 1 DG capable of supplying power to one of the required Unit I subsystems of each of the required components, must also be OPERABLE. Together, OPERABILITY of the required 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 consist of incoming breaker and disconnect to the 10 or ID and the 2C or 2D startup auxiliary transformers (SATs), associated IC or 10 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 l 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 capabilities are required to be met from a variety of i 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. k (continued) i HATCH UNIT 2 B 3.8-42 REVISION D
AC Sources - Shutdown B 3.8.2 - m
) BASES LC0 Proper sequencing of loads, including tripping of (continued) nonessential loads, is a required function for DG OPERABILITY.
T' N O I ( (continued) HATCH UNIT 2 B 3.8-42A REVISION D
AC Sources - Shutdown B 3.8.2 1 ( ) BASES LCO 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. I APPLICABILITY The AC sources are required to be OPERABLE in MODES 4 and 5 and during movement of irradiated fuel assemblies in the i Unit I secondary containment to provide assurance that: f
- a. Systems providing adequate coolant inventory makeup are available for the irradiated fuel assemblies in I
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 to mitigate the effects of events that can lead to core damage during shutdown are available; and O 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 MODES 1, 2, and 3 are covered in LC0 3.8.1. ACTIONS Ad An offsite circuit is considered inoperable if it is not available to one required ESF 4160 V bus. If two or more g 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 draining the reactor vessel. By the allowance of the option to declare required features inoperable with no offsite power available, appropriate restrictions can be implemented in accordance with the j d affected required feature (s) LCOs' ACTIONS.
.g (continued)
(
)
HATCH UNIT 2 B 3.8-43 REVISION D
- - - - - - - - - - - - - - - - - - - - - - - - - - .- _ __ 3
AC Sources - Shutdown B 3.8.2 BASES h ACTIONS A.2.1. A.2.2. A.2.3. A.2.4. B.1. B.2. B.3. and B.4 Ib (continued) With one or more offsite circuits not available to all required 4160 V ESF buses, the option still exists to IA 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 Id sources is not available. It is, therefore, required to suspend CORE ALTERATIONS, movement of irradiated fuel assemblies in the Unit I secondary containment, and 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 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 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 LC0 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 (continued) HATCH UNIT 2 B 3.B-44 REVISION D
Diesel Fuel Oil and Transfer, Lube Oil, and Starting Air B 3.8.3 v l ( BASES LC0 addressed in LC0 3.8.1, "AC Sources - Operating," and (continued) LC0 3.8.2, "AC Sources - Shutdown." The starting air system is required to have a minimum > capacity for five successive DG start attempts without recharging the air start receivers. Only one air start receiver per DG is required, since each air start receiver has the required capacity. APPLICABILITY The AC sources (LC0 3.8.1 and LC0 3.8.2) are required to - ensure the availability of the required power to shut down the reactor and maintain it in a safe shutdown condition after an A00 or a postulated DBA. Because stored diesel fuel oil and transfer, lube oil, and starting air subsystem support LC0 3.8.1 and LC0 3.8.2, stored diesel fuel oil and transfer, lube oil, and starting air are required to be within limits when the associated DG is required to be OPERABLE. O V ACTIONS The ACTIONS Table is modified by a Note indicating that l[ separate Condition entry is allowed for each DG. .This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable DG subsystem. Complying with the Required Actions for one inoperable DG subsystem may allow for continued operation, 1A and subsequent inoperable DG subsystem (s) are governed by separate Condition entry and application of associated Required Actions. Ad With one or more required DGs with one fuel oil transfer pump inoperable, the inoperable pump must be restored to OPERABLE status within 30 days. With the unit in this condition, the remaining OPERABLE fuel transfer pump is . adequate to perform the fuel transfer function. However, ] the overall reliability is reduced because a single failure ; in the OPERABLE pump could result in loss of the associated 1
- (v (continued) i HATCH UNIT 2 B 3.8-49 REVISION D c
Diesel Fuel Oil and Transfer, Lube Oil, and Starting Air B 3.8.3 BASES h ACTIONS M (continued) DG and loss of the fuel oil in the respective tank. The 30 day Completion Time is based on the remaining fuel oil transfer capability, and the low probability of the need for the DG concurrent with a worst case single failure. M In this condition, the 7 day fuel oil supply for a required DG is not available. However, the Condition is restricted to fuel oil level reductions that maintain at least a 6 day supply. These circumstances may be caused by events such as:
- a. Full load operation required for an inadvertent start while at minimum required level; or
- b. Feed and bleed operations that may be necessitated by increasing particulate levels or any number of other oil quality degradations.
This restriction allows sufficient time for obtaining the requisite replacement volume and performing the analyses required prior to addition of the fuel oil to the tank. A period.of 48 hours is considered sufficient to complete restoration of the required level prior to declaring the DG inoparable. This period is acceptable based on the rema ning capacity (> 6 days), the fact that procedures will be initiated to obtain replenishment, and the low probability of an event during this brief period. C.1 With a required DG lube oil inventory < 400 gal, sufficient lube oil to support 7 days of continuous DG operation at full load conditions may not be available. However, the Condition is restricted to lube oil volume reductions that maintain at least a 6 day supply. This restriction allows sufficient time for obtaining the requisite replacement volume. A period of 48 hours is considered sufficient to ' complete restoration of the required volume prior to (continued) HATCH UNIT 2 B 3.8-50 REVISION A
DC Sources - Shutdown B 3.8.5 (m (v) 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 I 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). APPLICABILITY The DC electrical power sources required to be OPERABLE in [~] v MODES 4 and 5 and during movement of irradiated fuel assemblies in the Unit I secondary containment provide assurance that:
- a. Required features to provide adequate coolant inventory makeu) are available for the irradiated fuel assemblies in tie 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.
bv (continued) HATCH UNIT 2 B 3.8-69 REVISIONf[ p
DC Sources - Shutdown B 3.8.5 BASES l APPLICABILITY The DC electrical power requirements for MODES 1, 2, and 3 l (continued) 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 distribution subsystem is requird according to LC0 3.8.8, the DC subsystems remaining OPEPABLE with one or more DC power sources inoperable may be capattle 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 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 actions is made (i.e., to suspend CORE ALTERATIONS, moverr.ent of irradiated fuel assemblies in the Unit I secondary 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 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) HATCH UNIT 2 B 3.8-70 REVISION A
DC Sources--Shutdown B 3.8.5 (m) v BASES SURVEILLANCE SR 3.8.5.1 (continued) REQUIREMENTS the corresponding Bases for LC0 3.8.4 for a discussion of each SR. This SR is modified by a Note. The reason for the Note is ! to preclude requiring the OPERABLE DC sources from being discharged below their capability to provide the required power supply or otherwise rendered inoperable during the l performance of SRs. It is the intent that these SRs must i still be capable of being met, but actual performance is not i required. ! SR 3.8.5.2 This Surveillance is provided to direct that the appropriate Surveillances for the required Unit 1 DC sources are governed by the Unit 1 Technical Specifications. l Performance of the applicable Unit 1 Surveillances will satisfy both any Unit I requirements, as well as satisfying this Unit 2 Surveillance Requirement. The Frequency i - required by the applicable Unit 1 SR also governs (, performance of that SR for both Units. REFERENCES 1. FSAR, Chapter 6.
- 2. FSAR, Chapter 15.
- 3. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
f\ HATCH UNIT 2 B 3.8-71 REVISION A
i Battery Cell Parameters B 3.8.6 8 3.8 ELECTRICAL POWER SYSTEMS B 3.8.6 Battery Cell Parameters BASES ! BACKGROUND This LC0 delineates the limits on electrolyte temperature, A level, float voltage, and specific gravity for the DC l SS electrical power subsystems batteries. A discussion of these batteries and their OPERABILITY requirements is provided in the Bases for LC0 3.8.4, "DC Sources - Operating," and LC0 3.8.5, "DC Sources - Shutdown." APPLICABLE The initial conditions of Design Basis Accident (DBA) and SAFETY ANALYSES transient analyses in FSAR, Chapter 6 (Ref.1) and Chapter 15 (Ref. 2), assume Engineered Safety Feature systems are OPERABLE. The DC electrical power subsystems provide 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 is based upon meeting the design basis of the unit. This includes ' maintaining at least one division of DC sources OPERABLE during accident conditions, in the event of:
- a. An assumed loss of all offsite AC or all onsite AC power; and
- b. A postulated worst case single failure.
Since battery cell parameters sup) ort the operation of the DC electrical power subsystems, taey satisfy Criterion 3 of the NRC Policy Statement (Ref. 4). I - LC0 Battery cell parameters must remain within acceptable limits to ensure availability of the required DC power to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence or a postulated DBA. Cell parameter limits are established to allow continued DC electrical system function even with Category A and B limits not met. l (continued) HATCH UNIT 2 B 3.8-72 REVISION D l
!l Battery Cell Parameters- !
B 3.8.6
- BASES (continued)
APPLICABILITY The battery cell parameters are required solely for the support of the associated DC electrical power subsystem. Therefore, these cell parameters a're only required when the DC power source is required to be OPERABLE. Refer to the ! Applicability discussions in Bases for LC0 3.8.4 and LC0 3.8.5. ACTIONS A Note has been added providing that, for this LCO, separate Condition entry is allowed for each battery. This is acceptable, since the Required Actions for each Condition , provide appropriate compensatory actions for each inoperable . battery. Complying with the Required Actions for battery cell parameters allows for restoration and continued . operation, and subsequent out of limit battery cell ! parameters may be governed by separate Condition entry and application of associated Required Actions. 1 A.1. A.2. and A.3 f With parameters of one or more cells in one or more ( batteries not within limits (i.e., Category A limits not met : or. Category B limits not met, or Category A and B limits not met) but within the Category C limits specified in , Table 3.8.6-1, the battery is degraded but'there is still sufficient capacity to perform the intended function. Therefore, the affected battery is not required to be considered inoperable solely as a result of Category A or B . limits not met, and continued operation is permitted for a l limited period. The pilot cell electrolyte level and float voltage are required to be verified to meet the Category C limits within I hour (Required Action A.1). This check provides a quick indication of the status of the remainder of the battery cells. One hour provides time to inspect the electrolyte level and to confirm the float voltage of the pilot cells. , One hour is considered a reasonable amount of time to perform the required verification. < Verification that the Category C limits are met (Required ; Action A.2) provides assurance that during the time needed to restore the parameters to the Category A and B limits, (continued) HATCH UNIT 2 B 3.8-73 REVISION A
Battery Cell Parameters B 3.8.6 BASES h ACTIONS A.l. A,2. and A.3 (continued) the battery is still capable of performing its intended function. A period of 24 hours is allowed to complete the initial verification because specific gravity measurements I must be obtained for each connected cell. Taking into consideration both the time required to aerform the required verification and the assurance that the sattery cell parameters are not severely degraded, this time is considered reasonable. The verification is repeated at 7-day intervals until the parameters are restored to Category A and B limits. This periodic verification is consistent with the normal Frequency of pilot cell surveillances. Continued operation is only permitted for 31 days before battery cell parameters must be restored to within Category A and B limits. Taking into consideration that, while battery capacity is degraded, sufficient capacity exists to perform the intended function and to allow time to fully restore the battery cell parameters to normal limits, this time is acceptable for o)eration prior to declaring the associated DC battery inoperaale. M When any battery parameter is outside the Category C limit for any connected cell, sufficient capacity to supply the maximum expected load requirement is not ensured and the corresponding DC electrical power subsystem must be declared inoperable. Additionally, other potentially extreme conditions, such as not completing the Required Actions of Condition A within the required Completion Time or average electrolyte temperature of representative cells falling below the appropriate limit (65 F for station service and 40 F for DG batteries), also are cause for immediately declaring the associated DC electrical power subsystem inoperable. SURVEILLANCE SR 3.8.6.1 REQUIREMENTS i This SR verifies that Category A battery cell parameters are ! consistent with IEEE-450 (Ref. 3), which recommends regular 1 battery inspections (at least one per month) including (continued) HATCH UNIT 2 B 3.8-74 REVISION D
Battery Cell ~ Parameters B 3.8.6
' BASES i SURVEILLANCE SR 3.8.6.1 (continued) l REQUIREMENTS A voltage, specific gravity, and electrolyte level of pilot I fp_\
cells. SR 3.8.6.2 The 92 day inspection of specific gravity, cell voltage, and' l level is consistent with IEEE-450 (Ref. 3). In addition,- within 24 hours of a battery overcharge > 150 V, the battery must be demonstrated to meet Category B limits. This inspection is also consistent with IEEE-450 (Ref. 3), which recommends special inspections following a severe overcharge, to ensure that no significant degradation of the battery occurs as a consequence of such overcharge. SR 3.8.6,3 7 This Surveillance verification that the average temperature of representative cells is within limits is consistent with ; a recommendation of IEEE-450 (Ref. 3) that states that the temperature of electrolyte in representative cells should be determined on a quarterly basis. Lower than normal temperatures act to inhibit or reduce , battery capacity. This SR ensures that the operating : temperatures remain within an acceptable operating range. This limit is based on IEEE-450 or the manufacturer's recommendations when provided. Table 3.8.6-1 This table delineates the limits on electrolyte level, float A voltage, and s)ecific gravity for three different I& categories. T1e meaning of each category is discussed below. Category A defines the normal' parameter limit for each designated pilot _ cell in each battery. The cells selected as pilot cells are those whose temperature, voltage, and electrolyte specific gravity approximate the state of charge of the entire battery. d A G (continued) , HATCH UNIT 2 B 3.8-75 REVISION D 5
Battery Cell Parameters B 3.8.6 BASES ?JRVEILLANCE Table 3.8.6-1 (continued) REQUIREMENTS The Category A limits specified for electrolyte level are based on manufacturer's recommendations and are consistent with the guidance in IEEE-450 (Ref. 3), with the extra
% inch allowance above the high water level indication for operating margin to account for temperature and charge effects. In addition to this allowance, footnote a to Table 3.8.6-1 permits the electrolyte level to be above the specified maximum level during equalizing charge, provided it is not overflowing. These limits ensure that the plates suffer no physical damage, and that adequate electron transfer capability is maintained in the event of transient conditions. IEEE-450 (Ref. 3) reenmmends that electrolyte level readings should be made only after the battery has been at float charge for at least 72 hours.
The Category A limit specified for float voltage is a 2.13 V per cell. This value is based on the recommendation of IEEE-450 (Ref. 3), which states that prolonged operation of cells below 2.13 V can reduce the life expectancy of cells. The Category A limit specified for specific gravity for each pilot cell is a 1.200 (0.015 below the manufacturer's fully charged nominal specific gravity) or a battery charging current that had stabilized at a low value. This value is characteristic of a charged cell with adequate capacity. According to IEEE-450 (Ref. 3), the specific gravity readings are based on a temperature of 77 F (25*F). The specific gravity readings are corrected for actual electrolyte temperature and level. For each 3*F (1.67*C) above 77 F (25 C), 1 point (0.001) is added to the reading; 1 point is subtracted for each 3 F below 77 F. The specific gravity of the electrolyte in a cell increases with a loss of water due to electrolysis or evaporation. Level correction will be in accordance with manufacturer's recommendations. Category B defines the normal parameter limits for each connected cell. The term " connected cell" excludes any battery cell that may be jumpered out. The Category B limits specified for electrolyte level and float voltage are the same as those specified for Category A d and have been discussed above. The Category B limit (continued) h HATCH UNIT 2 B 3.8-76 REVISION D
i Battery Cell Parameters- , B 3.8.6 ' 1 l i BASES !
- x. ,
f specified.for specific gravity.for'each connected cell is 2: 1.195 (0.020 below the manufacturer's fully charged,- nominal specific gravity)- with the average of-all connected i cells 1.205 (0.010 below the manufacturer's. fully charged,- ' nominal specific gravity). These values are based on manufacturer's recommendations. The minimum' specific. ' gravity value required for each cell ensures that the effects of a highly charged or newly' installed cell do'not mask overall degradation of the battery. -i Category C defines the limits for each connected cell. l' These values, although reduced, provide assurance that sufficient capacity exists to perform the intended function . and maintain a margin of safety.- When any battery parameter " is outside the Category C limit, the assurance of sufficient . capacity described above no longei exists, and the battery must be declared inoperable. The Category C limits specified for electrolyte level (above the top of the plates and not overflowing) ensure that the . plates suffer no physical damage and maintain adequate 3 n V l I I l q f / (continued) HATCH' UNIT 2 B 3.8-76A REVISION D ) I
Battery Cell Parameters B 3.8.6 I, ,I BASES v SURVEILLANCE Table 3.8.6-1 (continued) REQUIREMENTS electron transfer capability. The Category C limit for voltage is based on IEEE-450 (Ref. 3), which states that a cell voltage of 2.07 V or below, under float conditions and not caused by elevated temperature of the cell, indicates internal cell problems and may require cell replacement. The Category C Allowable Value of average specific gravity 2: 1.195, is based on manufacturer's recommendations (0.020 below the manufacturer's recommended fully charged, nominal specific gravity). In addition to that limit, it is required that the specific gravity for each connected cell must be no less than 0.020 below the average of all connected cells. This limit ensures that the effect of a highly charged or new cell does not mask overall degradation of the battery. The footnotes to Table 3.8.6-1 that apply to specific gravity are applicable to Category A, B, and C specific gravity. Footnote b of Table 3.8.6-1 requires the above (') v' mentioned correction for electrolyte level and temperature, with the exception that level correction is not required A when battery charging current, while on float charge, is Su
< 1 amp for station service batteries and < 0.5 amp for DG batteries. This current provides, in general, an indication of overall battery condition.
Because of specific gravity gradients that are produced during the recharging process, delays of several days may occur while waiting for the specific gravity to stabilize. ; A stabilized charger current is an acceptable alternative to ; specific gravity measurement for determining the state of ; charge of the designated pilot cell. This phenomenon is discussed in IEEE-450 (Ref. 3). Footnote c to Table 3.8.6-1 allows the float charge current to be used as an alternate i to specific gravity for up to 7 days following a battery , recharge. 4 REFERENCES 1. FSAR, Chapter 6.
- 2. FSAR, Chapter 15.
- 3. IEEE Standard 450 - 1987.
,n (continued) v)
HATCH UNIT 2 B 3.8-77 REVISION D
l Battery Cell Parameters B 3.8.6 , 1 BASES (continued) h I
- 4. NRC No. 93-102, " Final Policy Statement on Technical I !D\
Specification Improvements," July 23, 1993. j O l l l O HATCH UNIT 2 B 3.8-77A REVISION D i
Distribution Systems - Operating B 3.8.7 O V B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.7 Distribution Systems - Operating BASES BACKGROUND The onsite Class IE AC and DC electrical power distribution system is divided into redundant and independent AC and DC electrical power distribution subsystems. The primary AC distribution system consists of three 4.16 kV Engineered Safety Feature (ESF) buses each having an offsite source of power as well as a dedicated onsite diesel generator (DG) source. Each 4.16 kV ESF bus is normally connected to a normal source startup auxiliary transformer (SAT) (2D). During a loss of the normal offsite power source to the 4.16 kV ESF buses, the alternate supply breaker from SAT 2C attempts to close. If all offsite suurces are unavailable, the onsite emergency DGs supply power to the 4.16 kV ESF buses. The secondary plant distribution system includes 600 VAC emergency buses 2C and 2D and associated load centers, and transformers. (3 V There are two independent 125/250 VDC station service electrical power distribution subsystems and three independent 125 VDC DG electrical power distribution subsystems that support the necessary powe. for ESF functions. A description of the Unit 1 AC and DC electrical power distribution system is provided in the Bases for Unit 1 LCO 3.8.7, " Distribution System-Operating." g, The list of required Unit 2 distribution buses is presented in LC0 3.8.7. l APPLICABLE The initial conditions of Design Basis Accident (DBA) and SAFETY ANALYSES transient analyses in the FSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume ESF systems are OPERABLE. The j AC and DC electrical power distribution systems are designed , to provide sufficient capacity, capability, redundancy, and I reliability to ensure the availability of necessary power to i ( (continued) l HATCH UNIT 2 B 3.8-J4' 78 REVISION
Refuel Position One-Rod-Out Interlock B 3.9.2 g] 1 BASES
' ACTIONS A.1 and A.2 (continued) containing no fuel assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted.
SURVEILLANCE SR 3.9.2.1 REQUIREMENTS Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel. During control rod withdrawsl in MODE 5, improper positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks. Therefore, this Surveillance imposes an additional level of assurance that the refueling position one-rod-out interlock will be OPERABLE when required. By " locking" the reactor mode switch in the proper position (i.e., removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel), an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation. The Frequency of 12 hours is sufficient in view of other administrative controls utilized during refueling operations to ensure safe operation. SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST.on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual-signal indicative of a required condition is injected into the logic. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping,-or total channel steps so that the entire channel is tested. The 7 day Frequency is considered adequate because of demonstrated circuit reliability, procedural controls on control rod withdrawals, and visual and audible indications available in the control room to alert the operator. to control rods not fully inserted. To perform the required testing, the applicable condition must be entered (i.e.,- a control rod must be withdrawn from its full-in position). Therefore, . SR 3.9.2.2 has been modified by a Note that states the 1 (continued) HATCH UNIT 2 B 3.9-7 REVISION D
Refuel Position One-Rod-Out Interlock B 3.9.2 BASES h SURVEILLANCE SR 3.9.2.2 (contir9ed) REQUIREMENTS CHANNEL FUNCTIONAL TEST is not required to be performeo until I hour after any contr ol rod is withdrawn. REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.
- 2. FSAR, Section 7.6.1.
- 3. FSAR, Section 15.1.13.
- 4. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
O O HATCH UNIT 2 B 3.9-8 REVISION A
l RHR - High Water Level B 3.9.7
,j ) BASES LC0 An OPERABLE RHR shutdown cooling subsystem consists of an I (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. m
/
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_d 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 j O b/ (continued) HATCH UNIT 2 B 3.9-23 REVISION A
RHR - High Water Level B 3.9.7 BASES h ACTIONS A.J (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.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 g 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) Unit I secondary containment is OPERABLE;
- 2) two standby gas treatment subsystems (any combination of Unit I and Unit 2 subsystems) are OPERABLE; and 3) secondary containment isolation capability is available (i.e., one secondary containment isolation valve and associated instrumentation are OPERABLE or other acceptable A administrative controls to assure isolation capability) in /E each associated Unit I 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 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 ! (continued) HATCH UNIT 2 B 3.9-24 REVISION D i l
RHR - Low Water Level B 3.9.8 I BASES (l) LCO 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. /'% U 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 I/8 inches above the RPV flange are given in LC0 3.9.7, " Residual Heat Removal (RHR) - High Water Level." ACTIONS Ad 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 i reliability is reduced. Therefore an alternate method of ' decay heat removal must be provided. With both required RHR i i (O (continued) %) HATCH UNIT 2 B 3.9-27 REVISION A
RHR - Low Water Level B 3.9.8 BASES h ACTIONS L1 (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 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) Unit I secondary containment is OPERABLE; 2) two standby gas treatment subsystems (any combination of Unit I and Unit 2 subsystems) are OPERABLE; and 3) secondary containment isolation capability is A available (i.e., one secondary containment isolation valve [Q J and associated instrumentation are OPERABLE or other i acceptable administrative controls to assure isolation capability) in each associated Unit I 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 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 (continued) HATCH UNIT 2 B 3.9-28 REVISION D
Inservice Leak and Hydrostatic Testing Operation B 3.10.1 O U B 3.10 SPECIAL OPERATIONS ; B 3.10.1 Inservice Leak and Hydrostatic Testing Operation BASES BACKGROUND The purpose of this Special Operations LCO is to allow certain reactor coolant pressure tests to be performed in MODE 4 when the metallurgical characteristics of the reactor pressure vessel (RPV) require the pressure testing at temperatures > 212 F (normally corresponding to MODE 3). ] same as 1 t System inservicehyd.ostatic testing leakage tests) and system pressure tests leakage require (d by l Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref.1) are performed prior to the reactor going critical after a refueling outage. Inservice system leakage tests are performed at the end of each refueling outage with the system set for normal power operation. Some parts of the Class 1 boundary are not pressurized during these system tests. System hydrostatic tests are required once per ~; interval and include all the Class 1 boundary unless the ! test is broken into smaller portions. Recirculation aump ! O operation and a water solid RPV (except for an air bu)ble i I ( for pressure control) are used to achieve the necessary j temperatures and pressures required for these tests. The j minimum temperatures (at the required pressures) allowed for ! these tests are determined from the RPV pressure and I temperature (P/T) limits required by LC0 3.4.9, " Reactor IA i Coolant System (RCS) Pressure and Temperature (P/T) Limits." 2 These limits are conservatively based on the fracture ) toughness of the reactor vessel, taking into account J anticipated vessel neutron fluence. The hydrostatic test requires increasing pressure to approximately 1106 psig. The system leakage test requires increasing pressure to.
'd l4
{ approximately 1005 psig. l l With increased reactor vessel fluence over time, the minimum ) allowable vessel temperature increases at a given pressure. l Periodic updates to the RCS P/T limit curves are performed j as necessary, based upon the results of analyses of irradiated surveillance specimens removed from the vessel. 1 APPLICABLE Allowing the reactor to be considered in MODE 4 during SAFETY ANALYSES hydrostatic or leak testing, when the reactor coolant temperature is > 212*F, effectively provides an exception to MODE 3 requirements, including 0PERABILITY of primary
) (continued) l HATCH UNIT 2 B 3.10-1 REVISION D
Inservice Leak and Hydrostatic Testing Operation B 3.10.1 BASES APPLICABLE containment and the full complement of redundant Emergency SAFETY ANALYSES Core Cooling Systems. Since the hydrostatic or leak tests (continued) are performed nearly water solid (except for an air bubble for pressure control), at low decay heat values, and near MODE 4 conditions, the stored energy in the reactor core will be very low. Under these conditions, the potential for failed fuel and a subsequent increase in coolant activity above the LC0 3.4.6, "RCS Specific Activity," limits are minimized. In addition, the secondary containment will be OPERABLE, in accordance with this Special Operations LCO, and will be capable of handling any airborne radioactivity or steam leaks that could occur during the performance of hydrostatic or leak testing. The required pressure testing conditions provide adequate assurance that the consequences of a steam leak will be conservatively bounded by the consequences of the postulated main steam line break outside of primary containment described in Reference 2. Therefore, these requirements wil3 conservatively limit radiation releases to the environment. In the event of a large primary system leak, the reactor vessel would rapidly depressurize, allowing the low pressure core cooling systems to operate. The capability of the low pressure coolant injection and core spray subsystems, as required in MODE 4 by LC0 3.5.2, "ECCS - Shutdown," would be more than adequate to keep the core flooded under this low decay heat load condition. Small system leaks would be detected by leakage inspections before significant inventory loss occurred. For the purposes of this test, the protection provided by normally required MODE 4 applicable LCOs, in addition to the secondary containment requirements required to be met by this Special Operations LCO, will ensure acceptable consequences during normal hydrostatic test conditions and < during postulated accifient conditions. As described in LC0 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases. (continued) HATCH UNIT 2 B 3.10-2 REVISION A
=
l Inservice Leak and Hydrostatic Testing Operation B 3.10.1 7 Q BASES (continued) LCO As described in LC0 3.0.7, compliance with this Special Operations LC0 is optional. Operation at reactor coolant temperatures > 212 F can be in accordance with Table 1.1-1 for MODE 3 operation without meeting this Special Operations LCO or its ACTIONS. This option may be required due to P/T limits, however, which require testing at temperatures
> 212 F, while the ASME system hydrostatic test itself requires the safety / relief valves to be gagged, preventing their OPERABILITY.
If it is desired to perform these tests while complying with this Special Operations LCO, then the MODE 4 applicable LCOs and specified MODE 3 LCOs must be met. This Special Operations LC0 allows changing Table 1.1-1 temperature limits for MODE 4 to "NA" and suspending the requirements of A LC0 3.4.8, " Residual Heat Removal (RHR) Shutdown Cooling I OD , System - Cold Shutdown." The additional requirements for secondary containment LCOs to be met will provide sufficient protection for operations at reactor coolant temperatures
> 212 F for the purpose of performing either an inservice leak or hydrostatic test.
D This LC0 allows primary containment to be open for frequent (d unobstructed access to perform inspections, and for outage activities on various systems to continue consistent with the MODE 4 applicable requirements that are in effect immediately prior to and immediately after this operation. ' APPLICABILITY The MODE 4 requirements may only be modified for the performance of inservice leak or hydrostatic tests so that these operations can be considered as in MODE 4, even though the reactor coolant temperature is > 212 F. The additional requirement for secondary containment OPERABILITY according to the imposed MODE 3 requirements provides conservatism in the response of the unit to any event that may occur. Operations in all other MODES are unaffected by this LCO.
~"
(continued) HATCH UNIT 2 B 3.10-3 REVISION D
inservice Leak and Hydrostatic Testing Operation B 3.10.1 BASES (continued) ACTIONS A Note has been provided to modify the ACTIONS related to inservice leak and hydrostatic testing operation. 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, 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 each requirement of the LCO not met provide appropriate compensatory measures for separate requirements that are not met. As such, a Note has been provided that allows separate Condition entry for each requirement of the LCO. A,.1 If an LC0 specified in LC0 3.10.1 is not met, the ACTIONS applicable to the stated requirements are entered immediately and complied with. Required Action A.1 has been modified by a Note that clarifies the intent nf another LCO's Required Action to be in MODE 4 includes reducing the average reactor coolant temperature to s 212 F. A.2.1 and A.2.2 Required Action A.2.1 and Required Action A.2.2 are alternate Required Actions that can be taken instead of Required Action A.1 to restore compliance with the normal MODE 4 requirements, and thereby exit this Special Operation LCO's Applicability. Activities that could further increase reactor coolant temperature or pressure are suspended immediately, in accordance with Required Action A.2.1, and the reactor coolant temperature is reduced to establish normal MODE 4 requirements. The allowed Completion Time of 24 hours for Required Action A.2.2 is based on engineering judgment and provides sufficient time to reduce the average reactor coolant temperature from the highest expected value to s 212 F with normal cooldown procedures. The Completion Time is also consistent with the time provided in LC0 3.0.3 to reach MODE 4 from MODE 3. (continued) HATCH UNIT 2 B 3.10-4 REVISION A
Single Control Rod Withdrawal - Cold Shutdown B 3.10.4 BASES APPLICABLE on the withdrawn control rod, are inserted and incapable SAFETY ANALYSES of withdrawal. This alternate backup protection is required-(continued) when removing a CRD because this removal renders the withdrawn control rod incapable of being scrammed. As described in LCO 3.0.7, compliance with Special . Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs.- A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases. LCO As described in LC0 3.0.7, compliance with this Special ! Operations LCO is optional. Operation in MODE 4 with the I reactor mode switch in the refuel position can be performed A in accordance with other LCOs (i.e., Special Operations 1 2 LC0 3.10.2, " Reactor Mode Switch Interlock Testing") without meeting this Special Operations LCO or its' ACTIONS. If a single control rod withdrawal is desired in MODE 4, controls' l consistent with those required during refueling must be implemented and this Special Operations LC0 applied.
" Withdrawal", in this' application, includes the actual withdrawal' of the control rod, as well as maintaining the control rod in a position other than the full-in position, and reinserting the control rod.
The refueling interlocks of LC0 3.9.2, " Refuel Position One-Rod-Out Interlock," required by this Special Operations LC0 will ensure that only one control rod.can be withdrawn. At the time CRD removal begins, the. disconnection of the position indication probe will cause LCO 3.9.4, " Control Rod Position Indication," and therefore, LC0 3.9.2 to fail to be met. Therefore, prior to commencing CRD removal, a control-rod withdrawal block is required'to be inserted to ensure-that no additional. control rods can be withdrawn and that . compliance with this Special Operations LCO is maintained. To back up the refueling interlocks (LC0 3.9.2) or the control rod withdrawal block, the ability to scram the -i withdrawn control rod in the event of an inadvertent criticality is provided by the-Special Operations LCO
-i (continued)
HATCH UNIT 2 B 3.10-17 REVISION D ! i
Single Control Rod Withdrawal - Cold Shutdown B 3.10.4 BASES LC0 requirements in Item c.1. Alternatively, when the scram (continued) function is not OPERABLE, or when the CRD is to be removed, a sufficient number of rods in the vicinity of tha withdrawn control rod are required to be inserted and made incapable of withdrawal (Item c.2). This precludes the possibility of criticality upon withdrawal of this control rod. Also, once this alternate (c.2) is completed, the SDM requirement to account for both the withdrawn-untrippable control rod, and the highest worth control rod may be changed to allow the withdrawn-untrippable control rod to be the single highest worth control rod. APPLICABILITY Control rod withdrawals are adequately controlled in MODES 1, 2, and 5 by existing LCOs. In MODES 3 and 4, control rod withdrawal is only allowed if performed in accordance with Special Operations LC0 3.10.3, or this Special Operations LCO, and if limited to one control rod. This allowance is only provided with the reactor mode switch in the refuel position. During these conditions, the full insertion requirements for all other control rods, the one-rod-out interlock (LC0 3.9.2), control rod position indication (LC0 3.9.4), and scram functions (LC0 3.3.1.1, " Reactor Protection System (RPS) Instrumentation," and LC0 3.9.5, " Control Rod OPERABILITY - Refueling"), or the added administrative controls in Item b.2 and Item c.2 of this Special Operations LCO, provide mitigation of potential reactivity excursions. ACTIONS A Note has been provided to modify the ACTIONS related to a single control rod withdrawal while in MODE 3. 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, 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 each requirement of the LC0 not met provide appropriate compensatory measures for separate requirements that are not (continued) HATCH UNIT 2 B 3.10-18 REVISION A
0- '-----
*, - n-B- + - - ,c,,4> a wr ,+lasmM - i4 -4 4 e- 6- - - -
~ UNIT 2 MARKUP OF CURRENT TECHNICAL ~ i SPECIFICATIONS AND DISCUSSION OF CHANGES
, i t
i
'f I
i O i I O ; l ) ,
,s Insert 3.0D i
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)
LCO' 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS;to be entered permit continued operation in the MODE o'. L[z other specified condition in the Applicability for an unlimited period of time. This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS, or that are part of a shutdown of the unit. ( f Exceptions to this Specification are stated in the individual Specifications. These exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered allow unit operation in the MODE or other specified condition in the Applicability only for a limited period of time. Insert 3.0E Or LCO 3.0.5 Equipment removed from service or declared
,_ inoperabl to comply with ACTIONS may be returned L.3 to service under administrative control solely to
/T perform tes ing required to demonstrate its
'ss) OPERABILITY, the OPERABILITY of other equipment,. I
<n variables to be within limits. This is an D I
exception to LCO 3.0.2 for the system returned to service under administrative control to perform the required testing. I i i l J i I 4 Hatch Unit 2 Insert 3/4 0-l( I ) sos 8
DISCUSSION OF CHANGES ITS: SECTION 3.1.4 - CONTROL R0D SCRAM TIMES ADMINISTRATIVE A.1 The Surveillance Frequency has been modified to require testing after fuel movement within the reactor pressure vessel. This is ?quivalent to CORE ALTERATIONS excluding normal control rod movement (which should not affect. scram time) and excluding control rod removal / replacement, which is covered in SRs 3.1.4.3 and 3.1.4.4. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 The pressure at which the control rods must be tested has been changed to be 2 800 psig. This pressure corresponds to the limiting pressure for CRD scram testing for the Hatch Unit 2 system. " Limiting" refers to - the maximum scram times experienced at or below this pressure because of the competing effects of the reactor vessel pressure and the accumulator pressure scram forces. The scram time requirements are related to - transients analyzed at rated reactor pressure (assumed _ to be > 950 psig); however, if the scram times are demonstrated at pressures above 800 psig, the measured times are conservative with respect to the conditions assumed in the design basis transient and accident analyses. In addition, a Note , has been added to the SURVEILLANCE REQUIREMENTS' Table requiring that, (7 during a single control rod scram time Surveillance, the CRD pumps be V isolated from the associated accumulator. This ensures that accumulator ' pressure alone is scramming the rod, not the CRD pump pressure (which can , improve the scram times). :; M.2 In the Surveillance Requirement "for specifically affected" CRDs, deleting , the flexibility to delay post maintenance testing until reactor pressure-is a 950 psig is proposed, to ensure adequate testing is performed prior 1 to declaring the control rod operable, and entering MODE 2. In support of
. the proposed restriction, an additional surveillance is proposed. (SR '
3.1.4.3). This new surveillance will require a scram time test, which may be done at any reactor pressure, prior to declaring the control rod , operable (and thus, enabling its withdrawal during a startup). f To allow testing at less than normal operating pressures, a requirement' for scram time limits at < 800 psig is included. These limits appear less i restrictive than the operating limits; however, due to reactor pressure i not being available to assist the scram speed, the limits are reasonable , for application as a test of operability at these conditions. Since this ! test, and therefore any limits, are not applied in the existing j Specification, this is an added restriction. Furthermore, the existing l scram time test requirement (performed at normal operating reactor l pressure) is additionally required to be performed prior to exceeding 40% l RTP. -It is noted that if the control rod-~ remains inoperable (which requires it to be inserted and disarmed) until normal operating pressures, p a single scram time test will satisfy both Surveillance Requirements. d HATCH UNIT 2 1 REVISION D
m I3 g) (j
*/ s Ki,leb:3.1.) - )
% TABLE 3.3.1-1
--e .
O z REACTOR PROTECTION SYSTEM INSTRUMENTATION APPUCABLE MINtMUM NUMBER h OPERATIONAL OPERABLE CHANNELS y FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTE# ACTION N 1. Intermediate Range Monitors: y3 ,g dfiSE1-K60M. B. C. DR F. G, H) - A_- M. C& } Db .
- e. Neutron Flux - High pse I I
__ , 3 GfH 1
- b. Inoperative gdch f) 3 (,-/,4 lM 4w m m a
'3 *3 .3
- 2. Average Power Renee Moniter:
Q2C5J-K605 A32 C O,4 t_ .4
- e. Neutron Flux - Upecolo,15% 2 2 [y1
- b. Flow Referenced Simuleted Thermal Power - Upecele 1 2 F' 3
- c. Fixed Neutron Flux -
g Upecole,118% 1 L, 4 2' F3
% C 4:' Inoperative 1. 2 2 (r 4 i g e. Downocale 1 2 r3 Ed N
C CPEM Ms 4 *
- 3. Reector Vessel Steam Dome Pressure -
High M1-N671lL4. B.WD 1, 2 2 (f- 5 g 4. Reector Vessel Water Level - LA.3 2 ys m Low (Level al@550%e. C. O 1. 2
- o. Q
=
- 5. Mein Steam Une tooletion Velve -
Cio.or.& Q W S -6 t= 3 n c. i
. -- e . 'r -
- l. p~
t g x Dry.eil Pr ore- High 1 2 (y- s c? g ceu-Nsun. q e g L
$8 '
[ ' v .E . w~ l l- _. __ _ . _ _ . _ _ . _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ = _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _
.1
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TABLE AAM ICoritinuse - hg g n REACTOR PROTECTION SYSTEM fMSTRUMENTATION SURVEILLANCE REQUIREMENTS % f, ').J.J )
. 5 6 3.3.l.l.4. D LI 53 f '5.3.l. a. s o 4 8
Se 7.3.t 1.) CNNM.s *3I1'3 OFEftATIONAL (CONDmONS IN WHICH Q FlJNCTIONAL UNfT CHANNEL
~ CHECK FUNCTIONAL TEST CHANNEL CALAIRATIOld SURVEILLANCE IEQUIRED
' 7X Tintdne Stop Valve . Cloewe NA O -1 (3s we i
. $ Hl. Turtdne Deewe.Cortrol Trip OilVelve Presome Feet- pgF Low NA Gi l3 ,R ,g gg 1 jo -M. Remotor Mode Switch in Shutdown .NA RM1. NA 1.2,3,4,5 m -L 0 W. ManuelSerern NA WI NA 1.2,3,4.E
- e. Neutron datastore may be enoluded from CHANNEL CALIBRATION $( ). 3.I 1.80, O d ~ Se L11Lli e dI .-
. _. 24 _
em us. If M_ L-O wisNn 4_ ' 7 deve kO
- o. The APftM. IRM and SRM ehennale sheE he compared for overlap dwing each overtup,if not SA. 3 3-I 14 y 3. 2.t L 7 HL. _ within the provisue 7 days.
W d. from OOODmON 1 to CONDITION 2 perform the required survelsence within 12 heure .54 3 3.l.t.4,wk $ p 3,3,g.3.go g g y yy ,g,,,, y W e. . This esswesen ehd eeneiet of the 9 -- of the APftM ehennel to eenform to the power weluse k esicalmeed by a host belonce dudng CONDITION 1 when THERMAL POWER 2,25% of RATED THERMAL POWER. SE 3.3. i.r. t Adjust the APRM ohannelIf the ebeelute efferense 12% .
- f. TNe eastwegion shall sonalet of the edluotment of the APRM Sow Merenood elmulated thermal SR 7 3. t. I. 3 % .54 3. 3.i.r. :y power ehennel to eenform to e enebrated flow eierd.
- g. e sheE be salawesed at least enee per wer home IEFPHI using the M J.J. l . . B
[Q , _ A, and notegegn er enttches for metrumome mH-NUl3AkC. D. f g. 2 ,a*. a #
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e . c: m-m 2 _C_m._ .__.: hh .-w__:..i. _ m-__h a-_ _m_m._-_ 2-. m _ _m_::.i._ - m=z. _22_____ __u________:_____2 -eme.a. _ . - _ .um
INSTRUMENTATION 6g ec; g3 g 3 . 3. ) 7-
'N SOURCE RANGE MONITORS
[O LIMITING CONDITION FOR OPERATION 3.3.6.5 ange monitors shall be OPERABLE. 3 I' APPLICABILITY: CONDITIONS , 3 and 4. M* M' ACTION: nop3.J ATua 8 Ay a. In CONDITION
- with one of the above required source range monito g inoperable, restore 3 source rang 2 monitors to OPERABLE status within4hoursogbeinatleastHOTSHUTDOWNwithinthenext ou g mtA C -
- b. In CONDITION 3 or 4, with 6o or mor f the above required source range monitors ino erable,cretirn all control rods to be fully inserted kW D in the core and the reactor mode switch in the Shutdown position within one hour. ,, . t.
pow L c. One instrument channel may be inoperable for up to 6 hours to perform 4W g%"), required surveillances prior to entering other applicable ACTIONS. t p-) SURVEILLANCE RE0VIREMENTS Joeud _ wa i io s%%. , hire ~.As W.D 4.3.6.5 Each of the above required source range monTtors shall De demon-strated OPERABLE by:
- a. Performance of a:
- 1. CHANNEL CHECK at least once per: 73/. g a SR.3.LI.L t (a) 12 hours in CONDITION * , and so.3.t.2 3(b) 24 hours in CONDITION 3 or 4.
SfD .3.12.7 2. CHANNEL CALIBRATI0f6 at least once per 18 months %uroaea ns
- b. Performance of a CHANNEL FUNCTIONAL TEST: Q.r., z . 7 _
2
- 1. Within 24 hours prior to ving the reactor mode s tch from t hutdown position if not formed within the pre ' s7 L. 3 days, St.3332A 2. At least once per 31 days. upted IMe 4o S R f. 3.12 (All 5/4 R^mkh h f3 *With IRMs on range 2 or below. Td/e 734 2 -), M a_
**May exclude neutron detectors. di( 4 5/2 3 3.r. 2 7 HATCH-UNIT 2 3/4 3-56 Amendment No. 125 f.? 4
INSTRUMENTATIO!f L
%h9 33I2 f-~3 SVRVEILLANCE RE0VIREMENTS CONTINVED
- b. Performance of a CHANNEL FUNCTIONAL TEST:
bn 24 hours hier to the star _t nf CORNTERATIONS, and) t Q g33.\.25{2. At least once per 7 days. (5/u Anno W e-
/nrmificok).
O3,g,q c . Verify that the channel count rate is at least 3 cps t least once per 12 hours during CORE ALTERATIONS, and at least once per 24 hours, except: ex4e 2 H 1. The 3 cps is not required during core alterations involving sp_3 314 *l on_1y fuel unloadingJrovided the SRMs were confirmed to reaa gtleast3cpsinitiallyandwerecheckedforneutronresponse 4
- 2. The 3 cps is not required initially on a full core reload.
Prior to the reload, up to four fuel assemblies will be loaded N6 O into core positions next to each of the 4 SRMs to obtain the SR- _ required count rate. (These as!.emolles may be any which have been 3.Yq shown to meet the criteria given in Section 5.6.1 of these (Technical ( Specifications for storage in the spent fuel pool. ('~' . Verifying that the RPS itry " shorting links" have n removed that the RPS circuitry a non-coincidence trip m ithin 8 ho prior to starting CORE A IONS or shutdown margi demonstr ons. U4 fropeh SR 3 3.f.2 7 M.3
,rm .
N) HATCH - UNIT 2 3/4 9-4 Amendment No. 46, 39, 89 L4 eC H
l i CC l "k QAl e * *
-c INSTRUMENTATION REMOTE SHUTDOWN MONITORING INSTRUMENTATION l LIMITING CONDITION FOR OPERATION LLo 3 ' 3 h 2-3.3.6.3 The remote shutdown monitoring instrumentation channels . 4.4 . g eMie 3.AL3-yhall be OPERABLE jwiv%reaaouRaisplayea sternal i.o)
@e sontdroonr f-APPLICABILITY: CONDITIONS 1, @
ACTION: 7 h6 Ag a. ith one or more of the above required remote shutdown monitoring instrumentation channels inoperable, either restore the inoperable hannel(s) to OPERABLE status within 30 days or be in at least HOT-k d j g $ SHUTDOWN within the next 12 hours g in COLO M TDOWN withifhthe L-(.coTtowing zmun. ~ go Te W b. One instrument channel may be inoperable for up to 6 hours to ggiue perform required surveillances prior to entering other applicable GM g ed$ ACTIONS. p g jgic. The provisions of Specification 3.0.4 are not applicable. l625) ' w hc # SURVEILLANCE RE0VIREMENTS A f.$.3.T* \ 9(333- 4.3.6.3. Each of the above required remote shutdown monitoring instrumen-tation channels shall be demonstrated OPERABLE by performance of the GIANNEL CHECK ahD CHANNEL CALIBRATION operations at the frequencies (snag s in Table 43.6.3-Td --LA.g
@ p sea .s 9 3 3 3.2 h g i
\
V , HATCH - UNIT 2 3/4 3-50 Amendment No. 125 i IcS 3
- i9
-g -- I ABLE 5.3-1
- -4 y . /
$ - REMO7% SHUTDOWN MO ITORING INSTRUMENTATION SURV LANCE REQUIREMENTS i -/ @.s. ,.,.
g-E QHANNE CHANNR Q NNCTIONAL UNIT THECK 3 E3. 3.3,,q , f CAllBRATION
" f
'1. Reector Vessel Pressure - M [
- 2. Reactor Vessel Water le
- 3. Suppression Chambe eter Level -l M R
- 4. . Suppression C e Water Temperature M R
- 5. Drywell Pres re . M Q
- 6. Drywell sture M R W 7. R w M -
s w 8. R Flow M y N E:. Q.. r-
+
k L s w t h w
.b f
L I- . .
gm DISCUSSION OF CHANGES ( /
) ITS: SECTION 3.3.3.2 - REMOTE SHUTDOWN SYSTEM ADMINISTRATIVE A.1 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 A ("one or more required Functions") provides direction consistent with the intent of the existing Action for an inoperable remote shutdown instrumentation channel. Since this change only provides more explicit direction of the current interpretation of the existing specifications, this change is considered administrative.
A.2 Some instrumentation channels are deenergized during normal operation. No specific acceptance criteria would apply to the CHANNEL CHECK (since the instruments would not be indicating). Therefore, this Surveillance Requirement is modified to exclude the CHANNEL CHECK requirement on these deenergized channels. This change is considered administrative (since the channels are normally deenergized and any CHANNEL CHECK requirement would i be essentially equivalent to no requirement). TECHNICAL CHANGE - MORE RESTRICTIVE ,q (.,/ M.1 A new Surveillance has been added (SR 3.3.3.2.2) to verify each required l control circuit and transfer switch is capable of performing the intended function once per 18 months. This is an additional restriction on plant operation. TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 Details relating to system design and operation (e.g., specific instrument listings) are unnecessary in the LC0 and have been relocated to the Bases and procedures. 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 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. j l
HATCH UNIT 2 1 REVISIONy'D
pec &b 3. 3 4. 2_ INSTRUMENTATION J 3/4.3.9 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (]
\
ATWS RECIRCULATION PVMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION g . M 'L-3.3.9.1 The anticipated transient without scram recirculation pump trip (ATWS-RPT) s OPERABLF ' n~ useir ~ ystem L instrumentation ip setpoint set channels shown in Table 3.3.9.1-1 shall be j [ rip Setpoin olumn of Table 3.3.9_. _AconsistentwitQaiuessnowninIQ ' l APPLICABILITY: OPERATIONAL CONDITION 1. l
~
ACTION: cT)om MoY l a. With an aTWS recirculation pump trip system instrumentation channel I l Mmom A, trip setpoint less conservative than the value shown in the Allowable G,.4 C Values column of Table 3.3.9.1-2, declare the channel inoperable l until the channel is restored to OPERABL E status Alt l in% setpoint aajuRed consistentMth tie Trio Setpoin he alu channel) fe A41 l <M cM A L . b With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both y,i tri systems, lac he inoperable channel in the tripped condition , wit in gr5 6.) (yea mk % we l c. With the num er of OPERABLE channels two less than required by the 'G Minimum OPERABLE Channels per Trip System requirement for one trip Q system and,
- 1. If the inoperable channels consist of one reactor vessel water B.1 l$
RgWweb b9 level channel and one reactor vessel pressure channelinoperable .g channels in the tri
- 2. If the inoperable channels include two reactor vessel water level O.gual M W channels or two reactor vessel pressure channels,(aeclare the (trip s_vstem inoperabiG With one trip Ywithinherable, sysle stem R / p ,6. y d. to OPERABLE status - or restore be in at least theSTARTUP inoperable w trip khin 9,1 4 the next 6 hours. M*e wt D e. With to OPERABLE skatus both tri N'swithino $erable, restore at least one trihourorbeinatleastSTARTUPbibstem in the d"f # CA p-(next6 hours, g.g g--
Rg? f. One instrument channel may be inoperable for up to 6 hours to perfop b.i g A.% required surveillances prior to entering other applicable ACTIONS. M > gff L.3 ) G V HATCH - UNIT 2 3/4 3-66 Amendment No. 69, 14, 125 Inf 1 I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ J
^
l I (- y DISCUSSION OF CHANGES ITS: SECTION 3.3.5.1 - ECCS INSTRUMENTATION ADMINISTRATIVE (continued) A.8 This' requirement has been moved to proposed LC0 3.3.6.3, LLS j Instrumentation; a new Specification which lists all the LLS Instrumentation requirements. The discussion of any changes is provided with the new Specification LCO 3.3.6.3. JECHNICAL CHANGE - MORE RESTRICTIVE . M.1 The allowance to place an inoperable channel in trip has been removed for some Functions. Placing a channel in trip may not compensate for the inoperaulity, or it may be a less safe action to take. Therefore, for these types of Functions, the channel must be restored; it is not allowed
- to be tripped. This applies to the following current Functions
- 1.c, 2.d, 2.e, 2.f, 3.f, 4.c, 4.d, 4.f, and 4.g. This is an additional restriction-k:
on plant operation. M.2 Additional Functions are included to provide requirements for the ECCS i pumps' minimum flow instrumentation. The logic of this instrumentation is important to the proper functioning of the ECCS in response to a design -
,._T basis accident. Appropriate ACTIONS and Surveillance Requirements have (V also been added.
M.3 A Note (proposed Note b) has been added to ensure the DG and the PSW turbine building isolation valves are also covered by the associated i instruments. Thus, when a channel is not restored, the af%cted DG or PSW ' valve will be declared inoperable in addition to the affected ECCS subsystem. , M.4 The Allowable Value for the CST Level-Low Function has been modified to reference water level to a different reference point. In addition, due to recent analysis, the Allowable Value is also being increased to ensure ; 10,000 gallons of useable water exists at the swap-over setpoint, instead , of the current 10,000 gallons of total water. The Allowable Value for the
. Suppression Pool Water Level-High Function has been decreased to 154 ,
inches to correspond with the Unit 1 Value. These changes are additional restrictions on plant operation. M.5 An upper limit to the CS and RHR Discharge Pressure-High Allowable Values-for ADS has been provided. This will ensure that the setpoint is below the shutoff head of the low pressure ECCS pump. M.6- The ADS timer setpoints have been decreased to the proper Allowable Value. l The current TS value is the Analytical Limit. This is an additional -: restriction on_ plant operation. ; q D 1 HATCH UNIT 2 2 REVISION Q
- DISCUSSION OF CHANGES (v) ITS: SECTION 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION t
l ADMINISTRATIVE 1 l A.1 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 i inoperable isolation instrumentation channel. Since this change only l provides more explicit direction of the current interpretation of the I existing specifications, this change is considered administrative. l A.2 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 to place a channel in trip within a certain time is already covered by another, more explicit ACTION, this change is considered administrative. A.3 The Specification 3.0.4 exception has been deleted since proposed LC0 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 LC0 3.0.3 states it is only applicable in MODES 1, 2, and 3 (and this statement is applicable in MODE 5). (] v A.4 This section has been divided into two sections, Main Steam Line Isolation (Function 1), and Primary Containment Isolation (Function 2). The appropriate individual Functions have been placed with the proper isolation. A.5 This line item has been deleted since it is redundant to the current trip Function 2.c (proposed LCO 3.3.6.2, Function 1) and current trip Function 3.e (proposed LC0 3.3.6.1, Function 5.d). No new requirements are added by this specific line item; therefore, the deletion is considered administrative. A.6 This requirement (Secondary Containment Isolation) has been moved to proposed LC0 3.3.6.2, Secondary Containment Isolation Instrumentation. Any technical changes are discussed in the Discussion of Changes section for the new LCO. A.7 The SLC centrol switch inputs to the RWCU isolation logic (one channel). Thus, the number of channels has been changed to 1, to correspond to this input channel. This is the current logic, (and the current interpretation would require the Function to be declared inoperable if the input was inoperable) thus, this change is considered administrative. Ol HATCH UNIT 2 1 REVISION y _ - - - - - - - - - - - - - a I
o DISCUSSION OF CHANGES I l ITS: SECTION 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION V ADMINISTRATIVE (continued) I A.8 An Required Action has been added (proposed Required Action H.1) which allows the SLC System to be declared inoperable if RWCU System isolation is not desired. Since this is what would be required if the RWCU System could not be isolated (i.e., the Function's purpose is to ensure the SLC System functions properly and the injected boron is not removed from the reactor coolant system), the change is considered administrative. A.9 An action to " declare the affected system inoperable" is an unnecessary reminder that other Technical Specifications may be affected. This is essentially a " cross reference" between Technical Specifications that has been determined to be adequately provided through training. A.10 These valves have been deactivated and locked in the closed position in accordance with current LC0 3.6.3 (proposed LC0 3.6.1.3), Primary Containment Isolation Valves. This action is redundant to one that is always required and met. Therefore, this action has been deleted and is considered administrative. A.11 The maximum response time for these Functions corresponds with the diesel generator start delay time. While the time specified in this (~)h (. Specification is one second longer than the Technical Specification start times for the DGs, the additional one second is in error and the response times should have been the same as the diesel generator start times. This is consistent with the
- footnote. 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 recommends 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 since they are redundant to the diesel generator tests in proposed LC0 ;
3.8.1. l In addition, the HPCI and RCIC Steam Line High Flow-High Functions have a minimum time specified. This time appears consistent with the minimum closure time of a MSIV, to ensure these valves do not close faster than 4 MSIVs. However, the HPCI and RCIC valves are motor-operated and cannot physically close in that short of time. Thus, their deletion is also considered administrative. A.12 The CHANNEL FUNCTIONAL TEST (CFT) has been deleted since it is redundant to the LOGIC SYSTEM FUNCTIONAL TEST (LSFT). The SLC System Initiation channels have no adjustable setpoints, but are based on switch manipulation. Therefore, the LSFT, which test all contacts, will provide proper testing of the channels tested by a CFT. Therefore, this deletion p is considered administrative. V HATCH UNIT 2 2 REVISION / D ,
1 .n DISCUSSION OF CHANGES -( ) ITS: SECTION 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION ACTUATION INSTRUMENTATION TECHNICAL CHANGE - LESS RESTRICTIVE (continued) LC.1 The logic power monitors do not necessarily relate directly to the respective system operability. In general the BWR Standard Technical Specifications, NUREG 1433, does not specify indication-only equipment to be operable to support operability of a system or component. Control of the availability of, and necessary compensatory activities if not available, for indications, monitoring instruments, and alarms are addressed by plant operational procedures and policies. Therefore, this instrumentation, along with the supporting surveillances and actions are removed from the Technical Specifications.
" Specific" L.1 The current ACTIONS differentiate between whether channels are inoperable in one or both trip systems. With channels out in both trip systems,'the current ACTIONS do not allow all inoperable channels to be placed in the tripoed condition even if this would not cause an isolation. Because of the wied logic in isolation actuation systems there is no relatively simple set of actions that can be defined to cover all situations. The proposed Specifications have combined the ACTIONS for inoperable channels,
_h independent of whether one or both trip systems are affected. This allows (V the conservative action of tripping the inoperable channels which is preferable to initiating a shutdown as is currently required in many cases. If all channels are not restored or tripped, then the ACTIONS referenced in the proposed Table are required, similar to the current TS. L.2 The Required Action if the Required Action and associated Completion Time of Conditions A or B are not met for the Reactor Vessel Water Level-Low, A Low, Low (Level 1) is proposed to allow isolation of the affected main steam line (currently a shutdown is required). Some conditions may affect g the isolation logic for only one main steam line. In these cases, it is not necessary to require a shutdown of the unit; rather, isolation of the affected line returns the system to a status where it can perform the remainder of its isolation function, and continued operation is allowed (although it may be at a reduced power level). g- [ N A HATCH UNIT 2 6 REVISION
I l O DISCUSSION OF CHANGES l Q 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 procedures. The design features and system I operation are also described in the FSAR. 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 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. p Changes to the MPL numbers will be controlled by the provisions of 10 CFR ! 50.59. LA.3 Details of the methods for performing Surveillances are relocated to the l 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 (d) controlled by the provisions of 10 CFR 50.59. l 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 Required Action B.2) has also been added, ; since this is the effect of two inoperable channels. . l
" 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 Unit !
1 Secondary Containment, and operations with a potential for draining the i reactor vessel, are maintained. l l l HATCH UNIT 2 3 REVISION Q
1 l
,3 DISCUSSION OF CHANGES )
(w) ITS: SECTION 3.4.3 - SAFETY / RELIEF VALVES TECHNICAL CHANGE - LESS RESTRICTIVE (continued) I i LA.2 The details relating to system design and purpose have been relocated to the Bases. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process described in Chapter 5 of the Technical Specifications. l l
" Specific" l L.1 The allowed lift setpoint to.lerance has been increased from 1% to 3%. The '
vessel overpressure analysis uses this larger setpoint tolerance, as well as the transient analysis. In addition, when the setpoints are verified, they are still required to be reset to 1% (proposed SR 3.4.3.1). Thus, since the analyses still ensure that all limits are maintained even with the expanded tolerance, this change is considered acceptable. This change is also consistent with the BWR Standard Technical Specifications, NUREG 1433. The following assessment provides details of the analyses which were performed to determine the acceptability of the change. DISCUSSION: (] The Edwin I. Hatch Nuclear Plant Unit 1 is designed with' eleven SRVs and I Edwin I. Hatch Nuclear Plant Unit 2 is also designed with eleven SRVs. Technical Specification 2.2A.1 (for Unit 1) and Technical Specification 3.4.2.1 (for Unit 2) currently allow a setpoint error of 1% for each ; SRV. To reduce the number of forced outages and decrease maintenance and - surveillance testing costs it is proposed to increase the SRV setpoint tolerance from 1% to 3%. Furthermore, when the setpoints are verified, they will still be required to be reset to 1% (proposed SR 3.4.3.1) to ensure the SRVs will not drift outside the proposed 3% setpoint tolerance range. In addition, corresponding changes to the Bases are proposed. EVALUATION: Georgia Power Company proposed a technical change to establish a SRV setpoint tolerance of i 3%. To hstify the change, licensing basis calculations were performed to sFow that with the proposed setpoint i tolerance modifications, vessel everpressurization limits and Loss-of- l Coolant Accident / Emergency Core Cooling System (LOCA/ECCS) performance requirements are satisfied. The calculations also show that the proposed I change does not have a significant impact on thermal limits, Low-Low Set
,q operation and containment performance.
'b The following provide the details of the impact of the proposed change: HATCH UNIT 2 2 ( REVISION)D .
n DISCUSSION OF CHANGES Iv) ITS: SECTION 3.4.3 - SAFETY / RELIEF VALVES TECHNICAL CHANGE - LESS RESTRICTIVE L.1 (continued) I Affects of Proposed Chance on Reactor Vessel and ECCS The proposed SRV setpoints (including the 3% tolerance) are below the reactor vessel design pressure of 1250 psig, satisfying the requirements , of Article 9 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code - Section III, Nuclear Vessels. Evaluations of the overpressure transient (Main Steam Isolation Valve Closure with Flux Scram event) with SRV setpoints above the nominal setpoints plus the 3% tolerance and assuming one SRV inoperable demonstrate the capability of , the remaining SRVs to maintain the reactor pressure vessel well below the , ASME Code limit of 110% of the vessel design pressure (110% x 1250 psig = j 1375 psig). The proposed setpoints (including the 3% tolerance) are low . enough to ensure High Pressure Coolant Injection (HPCI) and. Reactor Core Isolation Cooling (RCIC) rated flow is still achievable. The proposed setpoint tolerance change may result in a reduction in turbine overspeed margins during a HPCI/RCIC start-up transients. However, since HPCI/RCIC are not expected to actuate until reactor vessel pressure is within the normal HPCI/RCIC operating range, the reduction in the turbine overspeed (~') margin does not have a significant impact on HPCI/RCIC performance. Also, l V since the proposed setpoints (including the 3% tolerance) are within the range of current setpoints, the overall likelihood of inadvertent valve opening (from downward setpoint drift) is not expected to change significantly. The impact of the SRV setpoint tolerance change on the i ECCS/LOCA (SAFER /GESTR) analysis was reviewed. Large line breaks (including the design basis accident recirculation line break and the main steam line break), are not affected at all by the change since the vessel depressurizes so quickly that the SRVs never actuate. Small line breaks may experience SRV actuations following Group 1 isolation, but the impact of the higher SRV setpoints is a small increase in break flow and an insignificant change in peak clad temperature (PCT). The small line breaks are not limiting for Plant Hatch. Therefore, the proposed SRV ; setpoint tolerance d ange does not have a significant impact on reactor i vessel or ECCS operation, j i Affects of Proposed Chanae on Thermal limits f The impact of the proposed SRV setpoint tolerance change was evaluated for the limiting thermal transient event. In this event, Load Rejection Without Bypass, peak vessel pressure occurs 1 to 2 seconds after the peak heat flux, which determines the time of Minimum Critical Power Ratio (MCPR). Therefore, potential increases in SRV opening have no impact on calculated fuel thermal limits. A decrease in the SRV opening pressure v HATCH UNIT 2 [.M REVISION
-m DISCUSSION OF CHANGES {} ITS: SECTION 3.4.3 - SAFETY / RELIEF VALVES TECHNICAL CHANGE - LESS RESTRICTIVE L.1 (continued) due to SRV downward setpoint drift) would cause earlier SRV actuation for the same transient event. An earlier actuation reduces the rate of vessel , pressurization and, therefore, the rate of void collapse. If SRV ! actuation occurs at or before the time of MCPR, the decreased rate of l pressurization and void collapse will produce lower peak neutron and I surface heat fluxes and, therefore, a smaller delta CPR. It should be noted that the reload transient analyses performed each cycle for Units 1 and 2 assumes a +/-3% SRV setpoint tolerance, and is therefore consistent with this proposed change. The results of the transient analyses are i reported in the Core Operating Limits Report (COLR). ! i Affects of Proposed Chanae on Low-low Set (LLS) ! The increased SRV opening pressure (due to the setpoint tolerance I increase) will only affect the timing of the first SRV actuation. Once I the logic is initiated, the opening and closing setpoints of preselected l SRVs are automatically reset to lower values by the LLS logic. This logic is unaffected by the setpoint tolerance change since the logic acts on the /] (, relief mode of SRV actuation and not on the safety mode of operation. l l Affects of Proposed Chanae on Containment Structures i The Plant Unique Analysis Reports (PUARs) for Hatch Units 1 and 2 were reviewed to determine the impacts of the proposed SRV setpoint tolerance change on the containment. SRV~ discharge loads have the potential to be affected by an increase in the SRV setpoint tolerance due to an increase in SRV flow rates. The evaluation performed for this change assessed the impact of the increased SRV setpoint tolerance on the actual load for the i limiting load combination on a structure-by-structure basis as performed I in the PUARs. The PUARs calculated the effects on the torus shell, torus support structures and torus attached piping assuming all SRVs actuate simultaneously. Each of the structures analyzed in the PUARs was reviewed to determine the impact of the SRV setpoint tolerance increase. Based on this review, it was determined that conservatism in the torus shell pressures used as input to the PUAR structural analyses would offset theincrease in SRV loads with the increased SRV setpoint tolerance. Therefore, the resulting loads will not cause the stresses in these components to exceed allowable values. v HATCH UNIT 2 [M REVISION /[
n DISCUSSION OF CHANGES V) t ITS: SECTION 3.4.3 - SAFETY / RELIEF VALVES TECHNICAL CHANGE - LESS RESTRICTIVE j L.1 (continued) Thrust loads for SRV piping and T-quenchers were determined using the relief valve forced outage rate (RVFOR) computer model. This computer model has been shown to overpredict water-clearing loads incident 'on submerged SRV piping and on T-quenchers and supports by 40% to 50%. However, the results of the Hatch Units 1 and 2 PVARS demonstrated adequate margins to the allowable stresses for these components to allow an increase in the SRV setpoint tolerances without exceeding the allowable stresses. Containment structures which could be affected by water jet or air bubble drag loads include such submerged structures as the vent header assembly, vent system supports, downcomer ties, vent line bellows, vent header. , deflectors and vent system penetrations. The results of the PVAR analyses showed that the limiting structures have large safety margins. Considering these large safety margins, increasing the SRV setpoint tolerance will not result in submerged structure loads which exceed the allowable loads. /m Based on these evaluations, the adequacy of the containment structures d with the proposed SRV setpoint tolerance change has been demonstrated since allowable stresses will not be exceeded. Therefore, the proposed SRV setpoint tolerance change has no significant impact on containment structures. A review of the impact of the increased SRV setpoint tolerance on containment response was also performed. The most limiting drywell pressure transient is the design basis LOCA, and the most limiting drywell ! temperature transient is the Main Steam Line Break event. Neither of these tr,ansients is affected by the increased SRV setpoint tolerance. For smaller steam line breaks, that require SRV actuations, the resultant drywell temperatures are well below the limiting steam line break. The l peak drywell temperature occurs late in the event following many SRV actuations and is governed by the total energy released to the drywell. Since the SRVs will return to nominal setpoints following the first , actuation, an increase in the SRV opening pressure will only affect the i very beginning of the event and will have a negligible impact on the total energy released to the drywell. Therefore, an increase in the SRV opening l pressure (due to the proposed setpoint tolerance increase) will also have ' an insignificant impact of peak drywell temperature for the non-limiting drywell temperature events. Therefore, the proposed SRV setpoint tolerance change has no significant impact on the containment response, i G I) HATCH UNIT 2 [cSC, REVISION /
I o I
.e)
Q DISCUSSION OF CHANCES . ITS: SECTION 3.4.3 - SAFETY / RELIEF VALVES i TECHNICAL CHANGE - LESS RESTRICTIVE ; s L.1 (continued) . CONCLUSION-Based on' the above evaluation, it is concluded that there is no i : significant safety impact on vessel overpressure margin, ECCS/LOCA- .___.- performance, thermal limits, Low-Low Set operation, or containment > structures due to operation with SRV setpoint tolerance of
- 3%.-
Therefore, these proposed changes, including the corresponding changes to i the Technical Specifications Bases, are acceptable. L.2 .This requirement has been deleted. A failure of an S/RV is not significant enough to report, as shown by the lack of a specific reporting requirement in 10 CFR 50.72 or 10 CFR 50.73 for this failure. However, if .: the failure meets one of the reporting criteria deemed significant in 1 10 CFR 50.72 or 10 CFR 50.73 (e.g., the valve fails open and a shutdown is i required), then 10 CFR 50.72 and 10 CFR 50.73 provide adequate reporting guidance. The two specific times in the current requirement coincide with - k the times in 10 CFR 50.72 and 10 CFR 50.73, respectively. ; O
~
i 4 ; 1 A U D 1 HATCH UNIT 2 [c2 D REviSiONp'
ELECTRICAL POWER SYSTEMS 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION Cf%%Q 3 01 3.8.2.1 The following A.C. distribution system buses, inverters and-motor generator (MG) sets shall be OPERABLE with breakers open between redundant buses:
- a. 4160 volt Essential Buses 2E, 2F, and 2G,
- b. 600 volt Essential Buses 2C and 20,
- c. 120/208 volt Essential Cabinets 2A and 28,
- d. 120/208 volt Instrument Buses 2A and 28, and
- e. A.C. inverters 2R44-5002 and 2R44-5003. -Lco 351 LICABILITY: CONDITIONS 1,2and)
ACTION:
- a. (With one of the inverters in 3.8.2.1.e inoperable, restore the inverter .['
EQod (to an OPERABLE status within a period not'to exceed seven (7) consecutive days or be in at least HOT SHUTOOWN within the next 12 ed (hours and be in COLD SHUTOOWN within the following 24 hours. l-
- b. With one of the above required A.C. distribution sy-tem buses
.seeAy,g inoperable, restore the inoperable bus to OPERABLE status within 8 g
.f q g hours or be in at least HOT SHUTDOWN within the next 12 hours and in m g ,7 COLD SHUTDOWN within the following 24 hours.
DW4de c. With two or more of the above required A.C. distribution system buses Metysh g or inverters inoperable, restore at least all except one of the l Qo seh.g,, inoperable buses.and inverters to OPERABLE status within 2 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN Q thin the following 24 hours. 3URVEILLANCE REQUIREMENTS C.4.8.2.1 TheaboverequiredA.C.distributionsystembusesand.inverterssha$ be determined OPERABLE:
"{a . At least once per 7 days by verifying correct breaker alignment and' pg[,-
l indicated power availability, and ,
. b. 'At least once per 31 days by determining that the 250 volt DC/600 volt-
~ M161.T AC inverters 2R44-5002 and 2R4_4-5003 are OPERABLE by verifying inverter output voltage ofQlts $while supplying their respective
- 5 .
buses. HATCH - UNIT 2 3/4 8-10 Amendment No. 23, 36 104 jo
i 4
;fm DISCUSSION OF CHANGES ITS:'SECTION 3.5.1 - ECCS - OPERATING TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 The details relating to system design and purpose have been relocated to -
the Bases. The design features and system operation are described in the FSAR. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process described in Chapter 5 of the Technical Specifications and changes to the FSAR will be controlled by the provisions of 10 CFR 50.59. LA.2 The details relating to methods of performing surveillance test i requirements have been 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 and changes to procedures will be controlled by the provisions'of 10 CFR 50.59. LA.3 This surveillance requirement has been relocated to plant procedures since the requirement is not included in the proposed Standard Technical ' Specifications, NUREG 1433. The system will continue to be required to
~T- perform its required safety function to be considered OPERABLE. Proposed '
(V SR 3.5.1.3 is added (refer to M.2) to address the important characteristic of whether there is sufficient air pressure available to permit the actuation of the ADS valves should an accident occur.- The surveillance being relocated will continue to be performed and will identify degradation of the ADS air system pressure retention capabilities. LA.4 The CTS contain two Surveillances regarding power availability of the LPCI inverter (4.8.2.1.a and 4.B.2.1.b). The-latter Surveillance (4.8.2.1.b) is retained as ITS SR 3.5.1.S and required a 31-day verification of- . specific output voltage to -the. associated LPCI bus. The first Surveillance (4.8.2.1.a) is a more frequent check, but only of " breaker alignment and indicated power availability." This more frequent ;
. verification is not retained in ITS, however, it will remain procedurally .p required as part - of routine operator checks on plant ; status. This
- relocation from the Technical Specifications still provides adequate assurance of the necessary power to the LPCI bus due ~ to the continuous ,
alarm capability on loss of bus power, as well as the routine (at least-shiftly) operator verification of safety system status. Any changes to the procedures that require formal documentation of " breaker alignment and ; indicated power availability" will be controlled by the provisions. of 10 CFR 50.59. HATCH UNIT 2 3 REVISION D
-~, DISCUSSION OF CHANGES ITS: SECTION 3.5.1 - ECCS - OPERATING (s.s)
TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" (continued)
LC.1 The Core Spray header delta-P instrumentation does not necessarily relate directly to CS system operability. The BWR Standard Technical Specifications, NOREG 1433, does not specify indication-only equipment to be operable to support operability of a system or component. Control of the availability of, and necessary compensatory activities if not available, for indication instruments, monitoring instruments, and alarms are addressed by plant operational procedures and policies. Therefore, this instrumentation, along with the supporting Surveillances and Actions are removed from the Technical Specifications. i f G
<-'3 \.j -
HATCH UNIT 2 h( 3 A REVISION D
e-l i CONTAINMENT SYSTEMS AQ 3.r .l. L i
'n ) PRIMARY CONTAINMENT AIR LOCK v' LIMITING CONDITION FOR OPERATION
# g.i.> l 3.6.1.3 The primary containment airlock shall be OPERABLE with:
(d . Both rs closed exc normal tansit entry an when the airl kisbeinguseh mit through th containment, th at I I least oneNirlock door s 11 be closed, an J L. . i t;,R h.bh b. An overall airlock leakage rate of s 0.05 La at Pa. 57.5 psig .2 APPLICABILITY: CONDITIONS 1,2kand3. ) , g p note. Q ACTION-WuPed %1 +9 % M WS 2[
- 4. Acw4 % a. ith one primary containment airlock door inoperable.CeaintaTh at
- A bW wigadthe least OPERABLE airlock door closedmmarr-the tacceralt yane +npumutunwithin 24 hours or lock the OPERABLE N al.rlock' door closeducper uen may snengonunue un @ eriormsacs Q, of Tned ert reoutred.overa irlock leakah test / proviced that the OPERABLE airlock door is verified to be locked closed at least once g per he-grovfMons or imiiir.sh 1 0 5 erwui appi m p
/ b. With the primary containment airlock incoerable, except as a i
M"g h ip result of an inoperable airlock door (4 HTal]F at least one airloc d/ C. door closed; restore the inoperable irlock to OPERABLE status within 24 hours. gw6gp wx (D w c. Otherwise, be in at least HOT SHUT 00WN within the next 12 hours and I Q1 'b in COLD SHUT 00WN within the following 24 hours. SURVEILLANCE REQUIREMENTS
'p6>fcxb MleM1 4.6.1.3 The primary containment airlock shall be demonstrated OPERABLE: l g g. % .t M 7 g '1, 4,.k l l a. The primary containment airlock shall be tested at 6-month i intervals at Pa by pressurizing the compartment between the ;
, two airlock doors. The leakage shall not exceed 0.05 La-
{ A.1 b. If the primary containment airlock is opened during periods when primary containment integrity is not required, the test required by 4.6.1.3.a. shall be performed at the end of such f periods.
- c. If the primary containment airlock is opened during periods when !
primary containment integrity is required, it shall be tested ; within 3 days of being opened by pressurizing the gap between the i doors seals to a 10 psig for at least 15 minutes. The leakage for each set of doors seals shall not exceed 0.01 La- i
- d. If primary containment is required and the primary containment j j~
airlock is being opened more frequently than once every 3 days, , the test required by 4.6.1.3.c. shall be performed at least once C per 3 days during the period of frecuent openings. >, L'g Q3 4112, e. At least once per 6 months by verifying that only one door in l the airlock can be opened at a time. [d ; yr% Q)
$R h.ps , s m Q g 5pec gl Test Ex Q tion 3h 0.1._ k HATCH - UNIT 2 3/4 6-6 Amendment No. 101
!ad
r l
<m DISCUSSION OF CHANGES
{
) ITS: SECTION 3.6.1.2 - PRIMARY CONTAINMENT AIR LOCK ADMINISTRATIVE ,
A.1 The format of the proposed Technical Specifications does not include cross references. The existing reference to "See Special Test Exception 3.10.1" serves no purpose, and therefore its removal is an administrative difference in presentation. A.2 The existing Technical Specifications contain the details for air lock leakage surveillances which are also found in 10 CFR 50 Appendix J. These regulations require licensee compliance, cannot be revised by the licensee, and are addressed by direct reference in the Technical Specifications. Therefore, these details of the regulations within the Technical Specifications are unnecessary. Furthermore, approved exemptions to the regulations, and exceptions presented within the regulations themselves, are also details which are adequately presented without repeating the details within the Technical Specifications. The only requirement is that the overall leakage rate (0.05L,) and the door leakage rate (0.0ll ) and test pressure / time be in Technical Specifications. These are included as SR 3.6.1.2.1.a and b. Therefore, retaining the requirement to meet the requirements of 10 CFR 50 Appendix J, as modified by approved exemptions, anu eliminating the Technical Specification details that are also found in Appendix J, is (^)N
- n. considered a presentation preference, which is administrative.
Clarifying Notes are proposed. These Notes facilitate use and understanding of the intent of:
- 1) (For SR 3.6.1.2.1 Note 1) the overall air lock acceptance criteria when one air lock door is inoperable. Since the inoperability is known to be only affecting one door, the barrel and the other operable door are providing a sufficient containment barrier. Even though the overall test could not be satisfied (SR 3.0.1 would normally require this to result in declaring the LC0 not met - possibly requiring ,
proposed Condition C (current ACTION C) to be entered), the note clarifies the intent that the previous test not be t.onsidered "not met."
- 2) (For ACTIONS Note 2) considering the primary containment inoperable in the event air lock leakage results in Appendix J acceptance criteria being not met. '
- 3) (For Required Action C.1) ensuring that the primary containment overall leakage is evaluated against the Appendix J acceptance criteria if an airlock is inoperable.
- 4) (For SR 3.6.1.2.1 Note 2) ensuring that the primary containment r overall leakage is evaluated against the Appendix J acceptance (3) cirteria after testing the air lock for leakage.
HATCH UNIT 2 1 REVISIONf( D
CONTAINMENT SYSTEMS C,. 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES 5 pec A 6 3 . G l. 3 LIMITING CONDITION FOR OPERAT R FA . 9 ) uo g ;,.s. 3 -
~
3.6.3 The@rimarycontainmentisolationvaltes)andthereactorinstru-mentation line excess flow check valves 4pect4.ied ihQable 3%.3-y shall be OPERABLE (wxp isohttion timst as snown in iaoiewo.a-1.L i 1 APPLICABJLITY: CONDITIONS 1, 2 and 3.. --{[iote>ed 16 Neli M [' ACTION: D e M ei u .w [ - b el % L .1 y % g,,, l a. With one or more of the primary containment isolation valves gcp g' 'pecXfied in TAN.6.3-1 inoperable, operation may continue A AA (ar.a t ne wo v i s i ons o i 5 pew fication M.4 arewt. appiinpliii WD @ro(loed thaAat leasT. yne i>vi giva voiv is u.ainu ina OPERABLE in ea% affecteh penetratton that open, f i eith . o valve [ restore Q OPERABLE Q u )
- 2. Each affected penetration is isolated within 4 hours by !
, - use of at least one deactivated automatic valve secured (mj in the isolation position, or
<gq gn 3 ,, _,,g
- 3. Each affected penetration is isolated within 4 hours by 'o r. d 4
- Q use of at least one closed manual valve or blind flange, j)gj
@pE Otherwise, be in at least HOT SHUTDOWN within the next 12 hours ano in COLD SHUTDOWN within the following 24 hours.
L. 2 pr wa wuo p gp D With one or more of the reactor instrumentation line excess ILA. O ] c flow check valvests m at1ea m table W .3-1]1noperable, opera- ] tion may continue canane__ provi swns vi SpeMiications X%3 4-
,fTnd'5 0.4 are rtR applicsMafprovided that within i 1. The inoperable valve is returned to OPERABLE status, or
- 2. n The Jent.i instrument s dethtred line is isolatedAnd the ah(iated 1Mtru /
inopihNLble. Otherwise, be in at least HOT SHUTDOWN within the next 12 l (k q w hours and in COLD SHUTDOWN within the following 24 hours. E Oafesce p AtT w F i h v m.) HATCH - UNIT 2 3/4 6 -15' 14 2f I _ _ _ _ _ _ _ _ l
n . 1
., CONTAINMENT SYSTEMS ;
b fed 9che) ' 3. I l'3 SURVEILLANCE REQUIREMENTS
)
A.6.3.1 Each primary containment isolation valve Tpecttiea in iaoie
~
W C3.6~~3~1 I
. shall be~ demonstrated 0PERABLE prior to returning the valve to i service after. maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verification (of specified isolation time. _
)
gg7M'l.6.3.2 Each primary containment automatic isolation valve Mbsctf-teNn 6Ih4a% 60 shall be demonstrated OPERABEEW ww snUTKnQ j h79 at least once per 18 months by verifying that on a containment ; isolation test signal i each automatic isolation valve actuates to its isolation position,
>g n # 4.6.3.3 The isolation / time of each power operated or automatic valve GTeafminNie M 3-D shall be determined to be within its limit w1en tested pursuant to Specification 4.0.5. !
I-
,3*I.6.3.4 Each reactor instrumentation line excess flow check valve shall be demonstrated OPERABLE at least once per 18 months by verifying that the lb valve stops excess flow. !
T!
-^-
?
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. l, Amendment No. 104 /
HATCH'- UNIT 2 J/4 6-17 . M 3.U5 ,
l X REACTOR COOLANT SYSTEM , 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES MM 34,I.3 LIMITING CONDITION FOR OPERATION b I J.(, . l .3 A .7 gco 3. wo ain Steam Line Isolation Valves (MSIVs) per main steam liny j hall be OPERABLE with closing times 2,3 and s 5 seconds.f ; APPLICABILITY: CONDITIONS 1, 2 and 3. ! ACTION: i With one or more MSIVs inoperable, operation may continug gp fprov4sions or snacification wa are not aponcable.Tfr'ov eu Tnat a p ,q m lin one naiv iMaintainec UVtKMQ in eaCn EJeCled main p pea ; i that oenfind either: in le valvbNQs restorbg0PERABLEhatus witig)-- l
- 2. The affected main steam line is isolated within 8 hours by use jg of FHiitetivated Qn the closed position. (g { .
'(7 Acmoa(E Otherwise, be in at least HOT SHUTDOWN within the next 12 hours ana l Cl in COLD SHUTDOWN within the following 24 hours. l Pgd W ! SURVEILLANCE RE0VIREMENTS . _ _ 6 4.4.7 Each of the above required MSIVs shall be demonstrated OPERA %E
,f l' bg 3 6 by verifying full closure between 3 and 5 seconds when tested put want ,
to Specification 4.0.5. O . l HATCH - UNIT 2 3/4 4-19 ! lieh( ;
)
i I l n 3/4.6 CONTAINMENT SYSTEMS , 3/4.6.1 PRIMARY CONTAINMENT 5FS 'Yd# 34l'3 PRIMARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION
.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained. h*c $ l APPLICABILITY: CONDITIONS 1, 2* and 3. (
%%J
"'s l
ACTION: iv % '
.5 m u.
Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within I hour or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. j l v l SURVEILLANCE RE0VIREMENTS I
\ !
\ \
1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstratD ('
- a. At least once per 31 days by verifying that; -
m wA m k u 343 ci.3 t-
- l. 11 penetrations
- not capable of being closed by OPERABLE 3.u.';.t ontainment automatic isolation valves and required to be t
[hti closed during accident conditions are closed by valves, lind flanaes.aor ceactivateo automatic vaives secured ina
/
kb45%ositson,gnd < gii c.nux v u w.m a . s w eg j . All equipment hatches are closed and seal (2. l hee Special Test Exception 3.10.1j j 'Except valves, blind flanges, and deactivated automatic valves which i are located inside the containment, and are locked, sealed or otherwise u.s3 3 secured in the closed position. These penetration. shall be verified (()':nkt ; closed during each COLD SHUTDOW except such verification need not be performed more often than once er 92 days. ~ L.7 p4g Sprisp%d% L-kg HATCH - UNIT 2 'M33 Q 3/4 6-1 2c42f
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT LEAKAGE
- ) p g , b h .f 3_
LitiLTING CONDITION FOR OPERATION 3.6.1.2 Primary containment leakage rates shall be limited to:
- a. An overall integrated leakage rate of:
- 1. sLa , 1.2 percent by weight of the containment air per 24 <
hours at P,, 57.5 psig, or f I
- 2. s L t, 0.849 percent by weight of the containment air per )
24 hours at a reduced pressure of Pt , 28.8 psig. krbuy 2rb.t.r, l
- b. A combined leakage rate of: pg
- l. s 0.60 La for all penetrations and valves, except for iMi4Q' N
4, ! main steam isolation valves, subject to Type B and C , y tests when pressurized to P., and j i 5(f3.t.t.3.l? 2. s 0.009 L, for the following penetrations *: ~lk
) Main steam conden te drain, penet tion 8; (b) Deleted (c) actor water cleanup, penetration 14; (d) Equ ment drain sump dis harge, penetrati 18; (e) Floor ain sump discharge, penetration 19; nd (f) Chemical ain sump discharge penetration 55;
-6 I'(g) Deleted ._
g aen .aduns s w p % m kV e Co\ N 6150ScN O "U *%i4 A mA W es, 5 g. 3 4 1 1 I! c. . )sef per hour for any one main steam isolation valve 4when tested at 28.8 psig.** d' AM" PLICABILITY: When PRIMARY CONTAINMENT INTEGRITY is required per Specification 3.6.1.1. ! r~N c' ~ Potential bypass leakaan nathe l 6ExemptdontoAppeodixJoK10CFA5
~
HATCH - UNIT 2 3/4 6-3 Amendment No. 49, 101 2)o5 z{
' CONTAINMENT SYSTEMS DfQhO 3.k t.)
O SURVEILLANCE RE0VIREMENTS (Continued) L)
- b. If any periodic Type A test fails to meet either .75 L,, or
.75 Lt , the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A-tests fail to meet either .75 L, or .75 L ,t a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet either .75aL or .75 Lt , at which time the above test schedule may be resumed.
- c. The accuracy of each Type A test shall be verified by a supplemental test which:
- 1. Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 La or 0.25 Lt '
- 2. Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test, and
{
- 3. Requires the quantity of gas injected into the contain-ment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the b]
total measured leakage at P,, 57.5 psig, or Pt , 28.8 psig. See h e ,A
. Type B and C tests
- shall be conducted at P,, 57.5 psig, during each reactor shutdown for refueling but in no case
$$hb at intervals greater than 2 years except for tests involving: ct J; w
- 1. Air locks, which shall be tested and demonstrated OPERABLE oer Surveillance Reauiramant a A LL and SN A 3 Il 2. Main steam line isolation valves, which shall be leak 7g li tested at 28.8 psig.
- e. All test leakage rates shall be calculated using observed data converted to absolute values. Error analyses shall be per-formed to select a balanced integrated leakage measurement -
system.
- f. The provisions of Specification 4.0.2 are not applicable. l
*All Type B and Type C Leakage Tests (i.e., Local Leak Rate Tests) that fail (i.e., test leakage is such that an LER would be required) during an outage ,
! shall be reported according to 10 CFR 50.73 by one 30-day written report that' is due within 30 days of the first leakage test failure in the outage. All l other leakage test failures discovered during the outage will be reported in (V a revision to the original report due within 30 days after the end of the age. I
' HATCH - UNIT 2 3/4 6-5 Amendment No. 86, 99 2M W
i C0r!TAINMENT SYSTEMS t)'
/ PRIMARY CONTAINMENT PURGE SYSTEM Spe n Abw %c.r.3 LIMITING CONDITION FOR OPERATION Lt o 3.c.t.3 _f 3.6.6.5.1J The drywell and suppression chamber 18-inch purge supply and Exhaust isolation valves shall be OPERABLE with:
- a. Each valve closed except for purge system operation for inerting, deinerting, and pressure control.
- b. A leakage rate such that the provisions of Specification 3.6.1.2 are met. }
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. ith an 18-inch drywell and suppression chamber purge supply and/or
> exhaust isolation valve (s) inoperable or open for other than 8g6 bg inerting, deinerting or pressure control. close the open 18-inch q
valve (s) orlotnerwise isolate the penetrations (sywithin 4 hours or ; be in at least HOT SHUIDOWN within the next 12 hours and in COLD [ SHUTDOWN within the following 24 hours. D} Asm.) E SURVEILLANCE REQUIREMENTS 4.6.6.5.1 The primary containment purge system shall be demonstrated OPERABLE: pM SnrW\
- a. In addition to the requirementsdf Specification 3.6.3, at least once per 31 days, when notWORGING and VENTIRID, by verifying that p k k3.s each 18-inch drywell and suppression chamber isolation valve is l closed.
- 5. At least once per 18 months by replacing the valve seat of each I 9 18-inch drywell and suppression chamber purge supply and exhaust 3(1 isolation valve having a resilient material '
seat _e g verit y tha) 31eakaA rate is ww.hin its lwt
/N -
HATCH - UNIT 2 3/4 6-46 Amendment No. ES, 108 2.T4 2(
~
CONTAINMENT SYSTEMS , PRIMARY CONTAINMENT PURGE SYSTEM 3p, & 3 ,, 3 LIMITING CONDITION FOR OPERATION l.t o h .i. 3 3.6.63 ? h > JThe drywell and suppression chamber 18 inch fast acting exces] (flowisolationdampersshallbeOPERABLE. -- APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. l ACTION:
- a. With an 18 inch drywell and suppression chamber excess flow geas isolation damper inoperable, close the open 18 inch drywell , i p6 and suppression chamber purge supply and exhaust isolation valves or fotherwise isolate the oenetration)within 4 hours or be in at ,
least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN .- D[ )[within the following 24 hours. ' SURVEILLANCE REQUIREMENTS The primary containment purge system excess flow isolation O ~4.6.6.5.2 dampers shall be demonstrated capable of performing their design functio .. Ma. At least once per operating cycle tha dampers will betvisual
. c-/. 3 4 ^ u ,
un5Detted anIvcycle yc v verity Ene camper > nave nu unmap n . . wn Cenaers snem incapaole of performing their design function. l O HATCH - UNIT 2 3/4 6-47 Amendment No. 58 LG4 Ef
. - . . _ _ . = - . -
1 I DISCUSSION OF CHANGES !p) ( ITS: SECTION 3.6.I.3 - PRIMARY CONTAINMENT ISOLATION VALVES i ADMINISTRATIVE (continued) A.6 The LC0 3.0.4 statement has been deleted since proposed LC0 3.0.4 provides 1 this allowance (i.e., this allowance has been moved to LC0 3.0.4). The j LC0 3.0.3 statement has been deleted since it is redundant to the "Otherwise" action. That is, LC0 3.0.3 is not applicable anyway since a : shutdown action has been provided. Therefore, deletion of these , allowances is administrative. 1 A.7 The current Technical Specifications repeat most of the requirements, provisions and actions for MSIVs, purge valves and excess flow dampers in Specifications separate from all other primary containment isolation ; valves. The proposed Technical Specifications . incorporate these requirements and associated restoration times into the primary containment ; isolation valve Specification. This is a presentation preference, except as noted by other comments. , A.8 This Surveillance has been deleted since it is redundant to current Specification 4.6.1.2.d (proposed SR 3.6.1.1.1). Since the requirement is l still maintained, this change is considered administrative. A.9 Proposed LC0 3.6.1.3 applies to each PCIV, except reactor building-to- .A' ( suppression chamber vacuum breakers. LC0 3.5.1.8 covers these vacuum A , breakers and thus, they do not need to be considered ~in this LCO. Since the requirement is still maintained, this change is considered administrative. A.10 This reference to NRC approved exemptions to 10 CFR 50 Appendix J was provided in the original Unit 2 TS, and provides background for the test pressure (2: 28.8 psig). This type of information is not needed in the TS . and is more appropriate for plant specific documents. Thus, this note has i been removed from this Specification. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 An additional Applicability has been added (i.e., when associated instrumentation is required to be OPERABLE per LCO- 3.3.6.1, " Primary Containment Isolation Instrumentation"), which effectively adds a MODE 4' and 5 requirement to the RHR Shutdown Cooling System isolation valves. , Appropriate ACTIONS have been added (proposed ACTION F) for when the j valves cannot be isolated or restored within the current 4 hour. limit (since the unit is already in MODE 4 or 5, the current shutdown action . would not provide any restriction). This change is an additional ! restriction .on plant operation. ] f7 b i HATCH UNIT 2 2 REVISION'D ;
l DISCUSSION OF CHANGES (3 f v
; ITS: SECTION 3.6.1.3 - PRIMARY CONTAINMENT ISOLATION VALVES ,
TECHNICAL CHANGE - MORE RESTRICTIVE i (continued) M.2 Two Surveillance Requirements have been added to verify the continuity of the TIP shear valve squibs every 31 days (proposed SR 3.6.1.3.4) and to test a TIP squib every 18 months on a STAGGERED TEST BASIS (proposed SR A i 3.6.1.3.9). These new SRs are additional restrictions on plant operation. I L6 M.3 The existing Action only restricts heating up reactor coolant above ! 212* F. This existing action would allow a startup and control rod l withdrawal from cold conditions (e.g., < 212 F). Should leakages above l limits be discovered while operating, the existing Action is non-specific as to the appropriate action to take. The aroposed ACTION D provides the appropriate operational restriction, which 's consistent in limitation and < i time to the existing LCO. If leakage is discovered while shutdown, proposed ACTION D does not allow continued operations, similar to the current requirement. The MSIV Actions would allow continued operation if the valves are closed, but only if overall primary containment leakage were within Specifications. This is consistent with the accident analysis. Therefore the proposed presentation and associated Actions for containment leakage rate beyond limits will result in establishing and maintaining the reactor in a cold shutdown, all-rods-in, condition until the leakage is corrected; resulting p ' in increased safety to the allowances of the existing Action. i M.4 An explicit listing of the types of isolation valves acceptable for use is A added, precluding other possible methods. This could be an additional la restriction on plant operation. TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 The list of PCIVs 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.
LA.2 Any time the operability of a system or component has been affected by repair, maintenance or replacement of a component, post maintenance testing is required to demonstrate operability of the system or component. 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 allowed as discussed in the Bases for proposed SR 3.0.1. LA.3 Details of the penetrations that are potential bypass leakage paths are relocated to the Bases and the Technical Recuirements Manual. Changes to the Bases will be controlled by the prov sions of the proposed Bases Control Process discussed in Chapter 5 of the Technical Specifications and changes to the Technical Requirements Manual will be controlled by the Q provisions of 10 CFR 50.59. () HATCH UNIT 2 3 REVISION D
l DISCUSSION OF CHANGES .,1 ( ITS: SECTION 3.6.1.3 - PRIMARY CONTAINMENT ISOLATION VALVES
'w./
TECHNICAL CHANGE - LESS RESTRICTIVE LA.4 Details of visual inspections of valves have been relocated to plant I procedures. This type of inspection is more appropriate for plant - l procedures. The valves are still required to be cycled, which should l ensure their operability. Any change to the procedures would be controlled by 10 CFR 50.59. LA.5 Details of isolation times for MSIVs have been relocated to plant procedures, similar to other PCIVs. Changes to the Bases will be controlled by the provisions of the proposed Bases Control Process ; described in Chapter 5 of the Technical Specifications. ' t
" Specific" L.1 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 V 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. 1 Many penetrations are designed with check valves as acceptable isolation barriers. With forward flow in the .line secured, a'. check valve is essentially equivalent to a closed manual valve. For those penetrations A designed with check valves.as acceptable isolation devices, this proposed change provides an equivalent level of safety. For penetrations not g designed with check valves for isolation, the proposed change does not affect the requirements to isolate with a closed deactivated automatic valve or closed manual valve. . ACTIONS allowing closed manual valves or check valves with flow secured also apply to isolating main steam lines, even though the design does not provide for these type of isolation devices. 'This change is simply a result of simplicity in providing a consistent presentation for all penetrations. While this apparent flexibility does not result in any actual technical change in the Technical Specifications, it is listed here for completeness. L.2 ~ In the event both valves in a penetration are inoperable, the existing ! Specification, which requires maintaining one isolation valve OPERABLE, 6 would not be met and an immediate shutdown is required. Proposed ACTION -() i B provides I hour prior to commencing a required shutdown. This proposed I l HATCH. UNIT 2 4 REVISION D j l j
f DISCUSSION OF CHANGES (m) ITS: SECTION 3.6.1.3 - PRIMARY CONTAINMENT ISOLATION VALVES TECHNICAL CHANGE - LESS RESTRICTIVE L.2 (continued) I hour period is consistent with the existing time allowed for conditions when the primary containment is inoperable. The proposed change will provide consistency in actions for these various containment degradations. L.3 In the event the inoperable valve is an uccess flow check valve, the proposed time to allow for restoration prior to requiring a shutdown is 12 hours. In this event, a limiting event would still be assumed to be within the bounds of the safety analysis (the excess flow lines contain orifices and are approximately 1/2 inch in diameter.) Allowing an extended restoration time, to potentially avoid a plant transient caused by the forced shutdown, is reasonable based on the probability of a EFCV line break event and does not represent a significant decrease in safety. L.4 The proposed surveillance (for a functional test of each primary containment and drywell isolation valve) does not include the restriction on plant conditions that requires the surveillance to be performed during cold shutdown or refueling. Some isolation valves could be adequately tested in other than cold shutdown or refueling, without jeopardizing safe l (N plant operations. The control of the plant conditions appropriate to (_) perform the test is an issue for procedures and scheduling, and has been determined by the NRC Staff to be unnecessary as a Technical Specification restriction. As indicated in Generic Letter 91-04, allowing this control is consistent with the vast majority of other Technical Specification surveillances that do not dictate plant conditions for the surveillance. p\ HATCH UNIT 2 [M REVISION D
,e] [0NTAINMENT SYSTEMS h/ec'Md^' '
LIMITING CONDITION FOR OPERATION (Continued) ACTION (Continued)
- d. With one suppression chamber - drywell vacuum breaker open, ACTp 6 operation may continueJprovided Surveillance Requiremerit (4.6.4.1.a is performed on the OPERABLE vacuum breakers and
. Surveillance Requirement 4.6.4.1.b is performed within 2 hours and at least onen nor ?? hmire thorpafter.fUt'fiETWise, be in W L, at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
Ee . With one p ' tion indicator of any uppression chamber drywell vacau reaker inoperable, op ation may continu LC . I in OPERATIONAL DITIONS 1, 2 and/or 3 til the next s COLD SHUTDOWN prov ed Surveillance Requir ent 4.6.4.1.b 1 performed within ours and at least once er 15 days the after. Otherwise, in at least HOT SHUT WN within the ne 12 hours and in C SHUTDOWN within the 11owing / Q 4 hours. , SURVEILLANCE RE0VIREMENTS A L) 4.6.4.1 Each suppression chamber - drywell vacuum breaker shall be ( Q J Ndemonstrated OPERABLE:
- a. At least once per 31 days and withinQ) hours after any dis SRp g.% charge of steam to tne suppression chamber from the safety / D relief valves, by cycling each vacuum breaker through at least ,
one test cycle and verifying that each vacuum breaker is closed. C Whenever vacuum breaker intheopenhition,byconduct7 bagatestt verifies tha the differentil ressure is -M1 maihtained >0.5 si for one ho ithout makeu s
- c. At least once per 18 months during shutdown by;
- 1. Verifying the opening setpoint, from the closed position,
@ z ,g .g.3 to be s 0.5 psid, i T. Performa e of a CHANNEL'QLIBRATION that ydrifies that s i each posit n indicator ind ates the vacuum eaker to c]D l e open if ti vacuum breaker es not satisfy e AP l t t in 4.6.4.1. and ,
J ' F3. Conducting a leak test at an initial differential pressure /i;ce t'
/ of 1 psi and verifying that the differential pressure ' /'&ovCD'#
\]/
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does not decrease by more than 0.25 inches of water per minute for a 10 minute period. W p,- TTS G 4 lJ l i
'~ %W % % dl ',
HATCH - UNIT 2 3/4 6-34 Amendment No. 13 14L
g DISCUSSION OF CHANGES ITS: SECTION 3.6.1.8 - SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS TECHNICAL CHANGE - LESS RESTRICTIVE (continued) L.2 The requirement to cycle the vacuum breakers after an S/RV lift has been revised from 2 hours after the lift, to 12 hours after the lift. The 1 operability of a vacuum breaker is not affected by an S/RV lift. Torus i modifications were completed in the' early 1980's. The new T-quenchers that were installed ensure that all steam is condensed in the suppression pool and do not increase the humidity in the suppression chamber air space - (this increased humidity is what is postulated to impact the OPERABILITY l of the vacuum breakers). In addition, a review of the vacuum breaker failure rate during the Surveillances performed after an S/RV lift shows that it is essentially the same as the failure rate during the routine 31 days Surveillance. Therefore any extension in the performance of this functional test following an SRV discharge is not safety significant. Furthermore, this change is recommended in Generic Letter 93-05, item 8.4. l l 1 m 4 l 1 (. HATCH UNIT 2 3 REVISION D
DISCUSSION OF CHANGES (,-w). ITS: SECTION 3.6.3.2 - PRIMARY CONTAINMENT OXYGEN CONCENTRATION v ADMINISTRATIVE A.1 This redundant requirement has been deleted. Current LC0 4.0.4 and proposed SR 3.0.4 require surveillances to be performed prior to entering the Applicability of an LCO. Therefore, this does not need to be repeated as a separate Surveillance Frequency and its deletion as considered administrative. l TECHNICAL CHANGE - MORE RESTRICTIVE 1 H.1 The time that this LC0 is applicable (after " THERMAL POWER is > 15% RTP . following startup" until " reducing THERMAL POWER to 15% RTP preliminary to ! a scheduled reactor shutdown") has been decreased from 72 hours to 24 hours. This is an additional restriction on plant operations. TECHNICAL CHANGE - LESS RESTRICTIVE
" Specific" L.1 Currently, no time is provided to restore oxygen concentration to within l
( the limit prior to requiring a plant shutdown. Proposed Required Action
'v] A.1 and associated Completion Time will allow 24 hours to restore oxygen to within the limit prior to requiring a plant shutdown. During this time, the hydrogen recombiners are normally still operable, thus, a means to control hydrogen exists. This new ' ACTION would possibly prevent an unnecessary shutdown and the increased potential for transients associated with each shutdown.
n HATCH UNIT 2 1 REVISION D 4
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i CONTAINMENT SYSTEMS-( 1/4.6.5 SECONDARY CONTAINMENT {tkeikko 3.6.(.1 SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION usw3.6.5.1H Hatch-Unit 2 SECONDARY CONTAINMENT IfiTCORITY and Hatch-Unit 1 j secondary containment integrity shall be maintained.cP%. 4.i
" - 1 l
APPLICA'BILITY: CONDITIONS 1, 2, 3, h - See % ="d' # N'b**
-G :nv s ru.s ,manae l ACTION: k A.-J a 34u w M. J 4 % oper_Lo !
Without Hatch-Unit 2 SECONDARY CONTAINMENT INTEGRITY and/or without '" % A< 'a.,..
#*A atch-Unit I secondary containment integrity, restore Hatch-Unit 2 SECONDARY CONTAINMENT INTEGRITY and Hatch-Unit I secondary containment integrity within 4 hours or be in at least HOT SHUTDOWN within the next , po > D Il2 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- I strated by: I
- a. Verifying at least once per 31 days: 4 p.5,.4 u 1. All equipment hatches are closed and sealed, and h
~~
545 A.P 2. At leastcone doo'r31n each access to the secondary i containment is closed.
- b. Verifying at ast once per 92 ays that each seco ary contain- ^'
l t ventilatio stem automati solation damper i PERABLE or cured in the sed position p Soecification 3. . 2. !
- c. At'least once per 18 monthsdo >uSTabcubbrenEM ]
- 1. Verifying that one standby gas treatment subsystem will Q M g ,3 draw down the secondary containment to a 1/4 inch of vacuum water gauge in s 120 seconds, and
- 2. Operating'one standby gas treatment subsystem for one )
hour and maintaining h 1/4 inch of vacuum water gauge in i SO '"a.s g the secondary containment at a flow rate not exceeding J 4000 CFM. 58t 3 PM. 5.1.2 HaRh Unit i secondary containment integrity shall be demon / '
- S p.t.u.1,(strated per Hatch-Unit 1 Technical Specifications. _
' r) s
\
(*When performing inservice hydrostatic or leak testing with the react 6P-- t ennlant temoerature above 212*F. HATCH--' UNIT 2 3/4 6-36 Amendment No. 91
- )dl
m DISCUSSION OF CHANGES I) v ITS: SECTION 3.6.4.1 - SECONDARY CONTAINMENT - OPERATING 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 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. o V 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. M.2 The current Surveillance requires that one Unit 2 subsystem be tested every eighteen months. However, the same SGT subsystem could be tested every outage. The proposed Specification will now require both Unit 2 subsystems be tested in the course of two outages - as represented by the Staggered Test Basis requirement of the Frequency. This is an additional restriction on plant operation. M.3 The Unit 1 Secondary Containment Surveillances have been specifically written into this LC0 instead of providing a cross reference. The current Hatch Unit 1 Surveillances are written as proposed SR 3.6.4.1.4 and SR 3.6.4.1.6. A change was made to proposed SR 3.6.4.1.4 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 Unit 1 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. V I HATCH UNIT 2 1 REVISION D
~ .
' CONTAINMENT SYSTEMS SECONDARY CONTAINMENT AUTOMATIC ISOLATION DAMPERS 5 p % w x .4.4 LIMITING CONDITION FOR OPERATION p o '7.t M 3
M.6.5 2/ ihe secondary containment @ntilathn systh.automa115, isolatioiii
~ i
~
fs)(nown. m lable%6.5>2-Dshall APPLICABILITY: CONDITIONS 1, 2, 3, and *. be OPERABLE. Da , ACTION: - o j(r'sroxL At+ 2Q3 % gg.. u.1 gg , With one or more of tne secondary containment 6ent.11atioAsvstemwtomatTe>- (g - isolationfiEWs>doeeuied Nable .5 h2-Dinoperable. operation may 41 e fcontinue emLtha nrmmons os snart fication 3 n ure not a 12LicaDI L D provided.that)t one isolar amper 1 Maintained OP NAQLt in gach a W ted penetr ion that is op a and; no rab perisreston4toOPERABLEhgswithin8 E b. The affected penetration is isolated by use of a closed (Tamper. ( ywa3rea fd ~, within 8 hours, or b .S RY CONTA INTEGRITY monstrateo wi 8 hour and t . damper is r tored to OPE E_ status within days. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and 'i
. d# in COLD SHUTDOWN within the following 24 hours, or K .we
.R w
SURVEILLANCE RE0VIREMENTS : (63 , 4.6.5.2 Each secondary containmentJentWat< ion spheal autbmatgOLA1 ; isolation @cpec1TitThiaoie 5,i.53;L shall be demonstrated OPERABLE: p -- a $ pic d i a. At least once per 92 da by cycling each tomatic damF; y o .c.d M testaih during plant ope tion through at 1 t one complety cycle ofM11 travel. - s_.% 4 Tu %to f, L%, a H ks),k 3
^fhwyteyw d *,When performing inservice hydrostatic or leak testing with the reactor coolant temperature above 212'F. s HATCH - UNIT 2 3/4 6-37 Amendment No. 91 ,
1o0 l
n -] i
? -. DISCUSSION OF CHANGES ,
ITS: SECTION 3.6.4.4 - SECONDARY CONTAINMENT ISOLATION VALVES - OPERATING 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 pat:bre 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 i to 92 days, consistent with the BWR Standard Technical Specifications,. NUREG 1433. 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.
(7 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 C/ of the isolation dampers has been generically changed 'to secondary containment isolation valves. LA.2 This surveillance has been deleted since~ it is_duplicative of SR 3.6.4.4.2. The valve is still ~ cycled every 92 days via proposed SR 3.6.4.4.2, which also measures the stroke time. LA.3 Any time the operability of a system or component has been affected by. repair, maintenance or replacement of a component, post - maintenance testing is required to demonstrate operability of the' system or component. 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. n b HATCH-UNIT 2 2 REVISION D
1 i r ~s DISCUSSION OF CHANGES () -ITS: SECTION 3.6.4.4 - SECONDARY CONTAINMENT ISOLATION VALVES - OPERATING TECHNICAL CHANGE - LESS RESTRICTIVE (continued)
" 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 3.6.4.4.1 Note 2. Opening of secondary containment penetrations on a intermittent basis is required for many of the same reasons as primary containment penetrations and the potential impact on consequences is less significant.
L.2 Current ACTIONS list some, but not all, possible acceptable isolation devices that may be used to satisfy the need to isolate a penetration with an inoperable isolation valve. The proposed change provides a complete A list of acceptable isolation devices. Since the result of the ACTION /_B 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 (_)) k_ 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 Shutdown or Refueling. Some isolations could be adequately tested in other than Cold Shutdown or Refueling, without jeopardizing plant operations. The control of the plant conditions appropriate to perform the test is an issue for procedures and scheduling, and has been determined by the NRC Staff to be unnecessary as a Technical Specification restriction. As indicated in Generic Letter 91-04, allowing this control is consistent with the vast majority of other Technical Specification surveillances that do not dictate plant conditions for the surveillance. 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
'T demonstrated in either case since the SCIV itself cannot discriminate (V between " actual" or " simulated" signals.
HATCH UNIT 2 3 REVISION D
EEEMELING OPERATIONS
, SECONDARY CONTAINMENT AUTOMATIC ISOLATION DAMPERS DPCc A'*b M 9 LIMITING CONDITION FOR OPERATION V0%AG LA . \
3.9.5.2 The secondary containmenTFehlatioksy_ stem _ aWmati ion ( __ (shDwa_ in Tah 3.9.bx-D shall be OPERABLE. APPLICABILITY: @NDITIONS nd *. ACTION: Tened usC\A emw
# % c L % z u 3 A /rm _ .A<3
- a. With one or more of th_e secondary containment @tmationsystehD Visolatiorh aamr>ers spec 411ed in Mble 315.2-D inoperable, Q k N 30 @ operation may continue provided thatjat gast one ionisol damp P A.
Os mawuined UFtKAtht in eachWected penMration that s one /, and:
,g operable W mper is restore 05 40 OPERABLE status % Qhi Q Q page) sd
- 2. The affected penetration is isolated by use of a closed %~F.
within 8 hours. ~ f-- d g p A . 2. Otherwise, suspend handling of irradiated fuel in the Hatch - het$ n Unit I secondary containment, and suspend Hatch - Unit 2 CORE ("'} ALTERATIONS gNtivines Inatbuid redxe the SHUTDOWN MARtHNg
~~
TleprovisionsofSpecifications3.0.3 . are not applicable. T ^' M d .g Atw o SVRVEILLANCE RE0VIREMENTS 4.9.5.2.1 Each secondary containment hu Hbystem'NtutomaSp isola-tion damDec. speciMd in Table SS.5.z-T shall be demonstrated OPERABLE: {f a. At least once perm duringIOD4HUTD'0WLor REFlJELIN' by: g.vtl. Cycling each autom damper through at least one complete
%y cycle of full travel and measuring the isolation time, and at c.chd
- 2. Verifying that on a secondary containment isolation test g(2 Wg.3 signal each automatic isolation damper actuates to its Lb isolation position.
Prior to retur' 'ng the damper to sePv ce after maintenance, epair or replace nt work is performe on the damper or its g as ciated actuator, ontrol or power ci it by performance of th q cling test an verification of iso tion time. - i i 8 () *When irradiated fuel is being handled in the Hatch - Unit I secondary l containment. HATCH - UNIT 2 3/4 9-8 I M 2-
DISCUSSION OF CHANGES O v ITS: SECTION 3.6.4.6 - SECONDARY CONTAINMENT ISOLATION VALVES - REFUELING TECHNICAL CHANGE - MORE RESTRICTIVE (continued) M.2 An additional Surveillance Requirement is included (proposed SR 3.6.4.6.1) 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.3 The isolation time test Frequency (proposed SR 3.6.4.6.2) has been reduced to 92 days, consistent with the BWR Standard Technical Specifications, NUREG 1433. 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.
(7 Any change to the Technical Requirements Manual will be controlled by the w/ provisions of 10 CFR 50.59. In addition, due to the relocation, the name of the isolation dampers has been generically changed to secondary containme .t isolation valves. LA.2 Any time tN operability of a system or component has been affected by repair, maintenance or replacement of a component, post maintenance testing is required to demonstrate OPERABILITY of the system or component. 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. S
" Specific" L.1 The Applicability has been modified to require the SCIVs only during CORE ALTERATIONS, not all the time while in MODE 5. (The movement of l irradiated fuel is unchanged). CORE ALTERATIONS and movement of l irradiated fuel are the only operations that are postulated to result in a fission product release requiring the Secondary Containment (hence the ,q 'need for the SCIVs). This assertion is supported by the fact that the ;
V- I I I HATCH UNIT 2 2 REVISION D
'm l DISCUSSION OF CHANGES ITS: SECTION 3.6.4.6 - SECONDARY CONTAINMENT ISOLATION VALVES - REFUELING v i l TECHNICAL CHANGE - LESS RESTRICTIVE l L.I (continued) current Actions only require these operations suspended (i.e., it does not require further actions to restore the SCIVs after the operations are i suspended.) L.2 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 in SR 3.6.4.6.1 Note 2. Opening of secondary containment penetrations on a intermittent basis is required for many of the same reasons as primary containment penetrations and the potential impact on consequences is less significant. L.3 Current ACTIONS list some, but not all, aossible acceptable isolation devices that may be used to satisfy the neec to isolate a penetration with an inoperable isolation valve. The proposed change provides a complete A list of acceptable isolation devices. Since the result of the ACTION /_D\ continues to be an acceptably isolated penetration for continued operation, the proposed change does not adversely affect safe operation. L.4 In the event both valves in a penetration are inoperable, the existing IT Specification, which requires maintaining one isolation valve OPERABLE, O 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 aroposed change will provide consistency in actions for these varnous secondary containment degradations. L.5 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 Shutdown or Refueling. Some isolations could be adequately tested in other than Cold Shutdown or Refueling, without jeopardizing safe plant o>erations. The control of the plant conditions appropriate to perform t ie test is an issue for procedures and scheduling, and has been determined by the NRC Staff to be unnecessary as a Technical Specification restriction. As indicated in Generic Letter 91-04, allowing this control is consistent with the vast majority of other Technical Specification surveillances that do not dictate plant conditions for the surveillance. L.6 The abrase " actual or," in reference to the automatic isolation signal, has acen added to the surveillance requirement for verifying that each SCIV actuates on an automatic isolation signal. This allows satisfactory autnmatic SCIV isolations for other than surveillance )urposes to be used i to fulfill the surveillance requirements. Operabi'ity is adequately : demonstrated in either case since the SCIV itself cannot discriminate I /O between " actual" or " simulated" signals. , G l l HATCH UNIT 2 3 REVISION D , 1
l Spec;pg k3.9. I ! ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) u , j b ot 8 543.8.t;g 4. Verifying the dieseV generator capability. to reject a load ~of g at least 2775 kW'wT1hout tripping. The generator voltage ' poF*g shall not exceed 4800 volts during and following the load rejection. y 7,g,g.q 5. Simulating a loss of offsite power by itself, and: a) Verifying de-energization of the emergency busses and load ; shedding from the emergency busses, b) Verifying the diesel starts on the auto-start signal, l l energizes the emergency busses with permanently connected i A7 loads in s 12 seconds, energizes. the auto-connected shutdown Toads through the load sequencer, operates l for 15 minutes while its generator is loaded with the shutdown (emergency) loads, and achieves and- ~ 1 maintains a steady-state voltage of 4160 (t 47D] volts and i a steady-state frequency of 60 1.2 Hz. ' 5 tt 2.g.t.1D 6. Verify that on an ECCS actuation test signal, without loss of offsite power, the diesel generator starts on the auto-start ' signal and operates on standby for 2 5 minutes. t raped m. ,% hgewi O Ad
- 7. (deleted) ' xn' 423.8 1.11 8. Simulating a loss of offsite power in conjunction with an ECCS 1 3 actuation test signal, and AIT ho a) Verifying de-energization of the emergency busses and load l shedding for the emergency busses.
j SO.B.I.11 b) Verifying the diesel starts on the auto-start signal, energizes the. emergency busses with permanently connected 1-loads in s 12 seconds, energizes the auto-connected shutdown ] (emergency) loads through the load sequencer, operates for ) 1 15 minutes while its generator is loaded with the emergency j ( loads and achieves and maintains a steady-state voltage of
\ 4160 and a steady-state frequency of 60'i 1.2 Hz.
3 J M. 98 1,,B l.li c) Verifying t a all diesel generator trips, except engine overspeed, low lube oil pressure and generator differential, are automatically bypassed upon loss of l voltage on the emergency bus concurrent with an ECCS actuation signal. q A4 3
- For.the 18 diesel generator a single full load rejection test every SR.3.B48 18 months will satisfy the requirements' of Unit 1 S g -
4.9.A.2.a.5 and Unit 2 Specification 4.8.1.1.2.d.4.pecification - j. HATCH-UNIT 2 3/4 8-4 Amendment No. N , 11g i hohI! ! 1
l DISCUSSION OF CHANGES
, ITS: SECTION 3.8.1 - AC SOURCES-0PERATING V
TECHNICAL CHANGE - MORE RESTRICTIVE M.1 Certain equipment needed to meet 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, the definition of OPERABILITY only requires one source, since proposed LC0 3.8.1 provides the proper ACTIONS to take if sources are inoperable. Therefore, the Unit I required 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. These changes, are administrative only. However, ACTIONS have also been provided (proposed ACTION A as it applies to a Unit 1 offsite circuit, ACTION C as it applies to a Unit 1 DG, and ACTION D as it applies to both Unit 2 offsite circuits and one Unit 1 offsite circuit) to add requirements not currently required by Unit 2 Technical Specifications. These ACTIONS are consistent with the requirement;, for a Unit 2 source, with the exception of the restoration time provided for a Unit 1 DG. The l time provided is 7 days, which is consistent with the restoration time ! provided for in the LCOs for the individual components powered from Unit I sources. In addition, the SRs are also applicable to the Unit 1
/m sources; thus, a note applicable to all SRs and SR 3.8.1.19 has been added
() to ensure Unit I sources are tested. Therefore, this change is considered more restrictive on plant operations. M.2 An additional Required Action has been added (Required Action G.2) requiring the unit to be placed in MODE 4 within 36 hours if both offsite circuits are not restored to OPERABLE status within 24 hours. Currently, a MODE 3 shutdown is all that is required. This is an additional restriction on plant operation. M.3 Limitations on the operating power factor are added to the full load rejection test and to the 24-hour run Surveillance. These limitations ensure the DG is conservatively tested at as close to accident conditions as reasonable, provided the power factor can be attained. A note is also added. Note 3 provides guidance for when the power factor cannot be l f g attained. M.4 As with all other DG start requirements, proposed SR 3.8.1.10 is proposed to add the acceptance criteria for voltage limits-(upper and lower) and speed / frequency upper limit (lower limit included in the existing Surveillance). These acceptance criteria are consistent with all other DG start acceptance criteria. In addition, a time requirement has also been added, consistent with the accident analysis. Proposed SR 3.8.1.18 is proposed to add the voltage acceptance criteria. 77 ( HATCH UNIT 2 3 REVISION A D ) l l j
I-DISCUSSION OF CHANGES
,, ITS: SECTION 3.8.1 - AC SOURCES-0PERATING l \
TECHNICAL CHANGE - LESS RESTRICTIVE : LA.3 (continued) Any change to the loads placed on the DG will be controlled by 10 CFR 50.59 (a design is required to change the loads). Additionally, the voltage range to be maintained during this test is not detailed in the ITS. Any change to the voltage acceptance criteria for the DG will be controlled by 10 CFR 50.59. I LA.4 Any time the OPERABILITY of a system or component has been affected by ; repair, maintenance, or replacement of a component, post maintenance j testing is required to demonstrate OPERABILITY of the system or component. 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 allowed as discussed in the Bases for SR 3.0.1. LA.5 The diesel generator accelerated test frequency requirements are relocated in their current licensing bases form to plant procedures, leaving the i Technical Specifications periodic surveillance frequency as 31 days. A plant procedure implements the requirements and responsibilities for tracking emergency DG failures for the determination and reporting of (O) reaching trigger values specified in NUMARC 87-00. These requirements are more restrictive than those specified in NUREG 1433. In addition, Generic Letter 94-01, " Removal of Accelerated Testing and Special Reporting Requirements for Diesel Generators," allows Licensees to request removal from TS of provisions for accelerated testing and special reporting requirements for EDGs. Hatch proposes relocation only with no relaxation in ITS conversion. The allowances of GL 94-01 will be addressed separately, post ITS implementation.
" Specific" '
L.1 The requested deletion involves the requirement to start the DGs under degraded offsite power conditions. The normal Technical Specification surveillance testing schedule provides adequate assurance that the OPERABLE DGs will be capable of performing their intended safety functions. The inoperability of an offsite AC source in no way affects the reliability of the OPERABLE DGs as previously demonstrated by their normal Technical Specification surveillance testing. In some circumstances, the inoperability of the AC sources will automatically start the associated DG. In these cases, the DG will already be supplying the safety bus. The reliability and availability of the DGs are not adversely affected solely as a result of the loss of offsite circuit (s) and the DG should not be required to be started if this condition exists. < Additionally, once the DG started to meet the existing ACTION, the DG m manufacturer recommends loading that DG prior to a return to standby status. l Iv) HATCH UNIT 2 5 REVISION g
)
1 DISCUSSION OF CHANGES ITS: SECTION 3.8.1 - AC SOURCES-0PERATING TECHNICAL CHANGE - LESS RESTRICTIVE L.1 (continued) The most probable cause of an offsite AC source becoming inoperable is severe weather or an off-normal grid condition. Severe weather or other off-normal grid conditions can also cause the loss-of a DG and leave its safety bus without AC power if the DG is tied to the offsite source when it becomes inoperable. NRC Information Notice ' 84-69 warns against operating DGs tied to offsite power when the unit's AC sources are O O HATCH-UNIT 2 [ A REVISION
l DISCUSSION OF CHANGES ITS: SECTION 3.8.2 - AC SOURCES-SHUTDOWN 7 b TECHNICAL CHANGE - MORE RESTRICTIVE M.1 (continued) Therefore, the Unit I required AC Sources have been added to this LCO. Since Unit 1 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 I 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 D 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 req"hed load centers, A MCC, buses, etc. are not capable of being powered via an offsite circuit, QD that circuit is inoperable. In this event it may not be necessary to suspend all CORE ALTERATIONS, irradiated fuel handling, and OPDRVs. p Conservative ACTIONS can be assured if all required equipment without d qualified offsite power availability is declared inoperable and the associated ACTIONS taken. D Therefore, along with the conservative additional requirements placed on the OPERABLE circuit, Required Action A.1 is also proposed. These additions represent restrictions consistent with implicit assumptions for operation in shutdown conditions; restrictions which are not currently imposed via the Technical Specifications. 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 OPEPABLE. 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 Unit 1 Secondary Containment. Since this could occur when the reactor is defueled (thus, not in MODE 4 or 5), this change is an additional restriction on plant l operation. l l / \ HATCH UNIT 2 2 REVISION \D
ELECTRICAL POWER SYSTEMS Sf'*U5"W* 3 E.lo SURVEILLANCE REQUIREMENTS (Continued) g.. faPosed Fo4e cp rb bh s.r.6-t @3ropsed v 3,g.c.,ibw L 4o ---g g a 3.g4 1 2. The pilot cell specific gravity,Ccorrected to 77'N is - 5R 3n.t 3. SMh The pilot cell voltage is 2
' 2.(3 volts, andl
]*
, Ng,$y . The overall battery voltage is 2120 volts)
\
- b. At least once per 92 days bvernying ro rov)snaw V
- Fm mw sR5.5.6 2 Oz,(3 - /
- g.g re i 1. The voltage of each connected cell is 2 d volts under 3,s. g _ , float charge jand has noi decrea>ed more inan D.17 voirs canc e16 ) from the vaTue observed during the original acceptance LAI 2.
L4 lolRS The specific gravit' , 9 totes [A key- & too to //"F1 of each con- l W "*_4 h I nected cell is 2 3 .. jand nas nos decrea>=d more G n ** 0 Q.02fromthevalueobservedduringtheprevioustest)Uy'PP1 and y g g,___ (, _. N Tod
- 3. The electrolyt M=.v.c #of each connected cell is between the minimum and maximum level indication marks. 7earavd Nok * )
O rQeast once per us montns oy vernying snew fo rable 3 84 *l }
- 1. The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration, A. Oscoder
- 2. d deps &
The cell-to-cell and terminal connections are clean, ir5 3.g.q,0 c. tight, free of corrosion and coated with anti-corrosion 1 material, and 5,, 3.opa,4 g3 3g;,
- 3. The battery charger will supply at least 400 amperes at a minimum of 129 volts for at least 4 hours.
- d. At least once per 18 months, during shutdown, by verifying that either:
- 1. The battery capacity is adequate to supply and maintain I
in OPERABLE status all of the actual emergency loads for 2 hours when the battery is subjected to a battery service test, or
- 2. The battery capacity is adequate to supply a dummy load of the applicable profile given in Figure 3.8.2.3-1 while n12taining the battery teminal voltaae k 105 volts.
c g_% sed nWC Mu c m6pM geropsd sR Lt4.3} HATCH - UNIT 2 3/4 8-14 Amendment No. 74 2oW
ELECTRICAL POWER SYSTEMS Speci fiadun 5.E.6 i\ - k,l SURVEILLANCE REQUIREMENTS (Continued) 4.8.1.1.3 Each diesel generator battery and battery charger shall be demonstrated OPERABLE: SR 3.fr.G.I ,&
- a. At least once per 7 days by verifying that: '
$Y f**bhe 3.i.$--l K sed rd d 1. The electrolyte level of each pilot cell is Ti tne Ji g ,g ,g minimum and m mum level indication marks, e / ~ 'fg
~
- 2. The p 11 specific ty, corrected to 2 5.;t059 yroposedr4 & c 4v n m 2.s.L-i 2.is
- 3. The pilot cell voltage is 2 volts, and (fu doc for (T3 3.t.4 The overall hatterv_ vnltage is 2120 voltsl froposed 2"* FM7uu,cq p j, 6R 1841 b. At least once per 92 days by verity 1ng that:
- 1. The voltage of each connected cell is 2 under float chargerano he:, nut uecreaseo more snan u.u voits W
- e value observed during the original acceptance
' b* -h h"7/,d4N2-k -h ,
A3 2. The specifiq gravity,(corrected to // B, of each connectec ( cell is 2 id2Sfr[ana nas not cecreaseo more Inan u.uc _ a.d.;,44 (from the value observed durino the Drevious test " nd ' glts QimsnJ e4o R c. +o r b u 3.y.4 -1 L2 _ . t g 3. The electrolyte level of each connected e 1 is between . [' g the minimum and maximum level indication marks. pmsd
#J e le. a do gj C At least once per 18 months by verity 1ng that: re 3.%.4-1 E N*M k 1. The cells, cell plates and battery racks show no visu 'C6 MNa}k indication of physical damage or abnormal deterioration, .Nas.opvamg in h 5ub - 2. The cell-to-cell and terminal connections are clean, tight, free of corrosion and coated with anti corrosion material, and
- 3. The battery charger will supply at least 100 amperes at a minimum of 129 volts for at least 4 hours.
- d. At least once per 60 months during shutdown by verifying that the battery capacity is at least 80% of the manufacturers rating when subjected to a performance discharge test. This performanc discharge test shall be performed subsequent to the satisfactory aN' e et w - tse re c e sstry :em, w ....
a LL &LGd Pro? mp,.,,j S(z 3,g.9y HATCH - UNIT 2 3/4 8-6
- 3 d f
DISCUSSION OF CHANGES 7 ITS: SECTION 3.8.6 - BATTERY CELL PARAMETERS
! F v
ADMINISTRATIVE A.1 The proposed Technical Specifications present the DG and station service battery cell parameters limits in a separate LC0 (proposed LC0 3.8.6). Thus, a new LC0 statement has been provided reflecting this. The appropriate ACTIONS and SRs have been moved to this LC0 also. Current Specification 4.8.2.4.2 is being deleted since its provisions only reference requirements in current Specification 4.8.2.3.2. Proposed LC0 3.8.6 contains these current provisions of 4.8.2.3.2 and thus no reference is necessary. A.2 The Applicability of this new LC0 is "when associated DC electrical power subsystem is required to be OPERABLE." This covers the current MODES 1, 2, 3, 4, and 5 and fuel handling requirements, and is actually more restrictive for the DC power subsystems since, a) the DG DC source Applicability has been changed (in proposed LC0 3.8.5) to include fuel handling (see Discussion of Changes for ITS: 3.8.5 for further - discussion), and b) more than one of the DGs and station service batteries may be required in MODES 4 and 5 since the DC sources Applicability has been changed (in proposed LCO 3.8.5, see Discussion of Changes for ITS: 3.8.5 for further discussion). However, since these restrictions are not discussed in this specification, these changes are considered administrative in nature. (O ,/ A.3 The battery cell parameter limits have been combined into one Table, (proposed Table 3.8.6-1), which provides the limits for each pilot all (Category A) and for each connected cell (Category B). Category C limits have also been added, as described in comment L.1 below. The proposed SRs (SR 3.8.6.1 and 3.8.6.2) are reworded to verify the appropriate limits (Category A or B) are met. No technical changes are made, unless described in the "M" or "L" comments below. TECHNICAL CHANGE - MORE RESTRICTIVE M.1 The individual cell voltage limit is being increased from 2.0 volts to 2.13 volts. This ensures the overall battery voltage is satisfactory. ' This is an additional restriction on plant operation. M.2 New Surveillances are being added, consistent with the BWR Standard Technical Specifications. A new Frequency is being added to proposed SR - 3.8.6.2 requiring all the cell parameters to be verified once within 24 hours after a battery overcharge > 150 V. Proposed SR 3.8.6.3 requires a verification that electrolyte temperature is 2 65*F for each station service battery and a 40 F for each DG battery every 92 days. This helps to ensure battery OPERABILITY. Proposed SR 3.8.6.1 and SR 3.8.6.2 have been added for the DG batteries when in MODE .4 or 5, or when handling irradiated fuel. Currently, these SRs are only required _ in MODES 1, 2, . fm and 3. A requirement for level correction of specific gravity is also D added to proposed Note (b). These are additional restrictions on plant (^) operation. HATCH UNIT 2 I REVISION /D
DISCUSSION OF CHANGES 1 ITS: SECTION 3.8.6 - BATTERY CELL PARAMETERS I gs i ) V TECHNICAL CHANGE - LESS RESTRICTIVE - (continued) L.3 An allowance to utilize charging current in lieu of specific gravity is I proposed following a battery recharge for a miximum of 7 days. This l allowance is consistent with the BWR Standard Technical Specification. In l conjunction with a requirement to measure actual specific gravity at the ! end of this period, this limited allowance will assure excessive reliance i on charging current is not made, while allowing a more accurate indication ; of return to full charge. Since this allowance is consistent with IEEE-450 recommendations, this change remains sufficiently conservative to assure continued battery 0PERABILITY. L.4 IEEE-450 working group recommendations to the NRC for appropriate Technical Specifications on battery electrolyte, as well as the NRC approved Bases for the BWR Standard Technical Specifications, provide specific gravity limits that are: 0.015 below the manufacturer's fully charged nominal specific 0 gravity for Category A limits on individual cells; 0.020 below the manufacturer's fully charged nominal specific gravity for Category B limits on 'ndividual cells; 0.010 below the manufacturer's fully charged nominal specific gravity for Category B limits on the average of all connected cells; ,A) ( - 0.020 between any individual cell and the average of all connected cells; and 0.020 below the manufacturer's fully charged nominal specific gravity for Category C limits on the average of all connected cells. Currently, the Hatch Unit 2 specific gravity limit of 1.205 is excessively conservative with respect to this guidance. The batteries utilized at Plant Hatch are nominal full charged specific gravity of 1.215, resulting in proposed limits that are consistent with the above guidance. Therefore the proposed revisions to specific gravity limits provides sufficient assurance of continued battery capability. \ ,Y HATCH UNIT 2 M3 REVISION Q
l ELECTRICAL POWER SYSTEMS l 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS Sp,4 l A.C. DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION heauol rpi .oh 's M ( tco3.8.2 3.s.".J.1 1 The following A.C. distribution Estem buse3, dnverter3)and _
.chr-cenewtor (MUNetsj shall be OPERABLE pun oreams open m ween reconqantj A- tmses a
\ ,
- s. 4160 volt E'ssential Buses 2E, 2F, and 2G, !
- b. 400 volt Essential Busey 2C and 2D,
- c. 120/208 volt Essential Cabinet:s 2A and 2B,
- d. 120/208 voHC J.nstrument. Buses 2A and 2B, and P('P6'S ttoF67b
@ A.C. inverters 2R44-5002 and 2R44-500 -
mm A % w 5.r.) APPLICABILITY: CONDITIONS 1, 2 and 3 ACTION: ;g g
- a. With one of the inverters in 3.8.2.1.e inoperable, restore the inverter l pg y/ F to an OPERABLE status within a period not to exceed seven (7) consecutive days or be in at least HOT SHUTDOWN within the next 12 hours and be in COLD SHUTDOWN within the following 24 hours.
- b. With one of the above requf red A.C. distribution ditem buse A Nm' C- inoperable, restore the inoperable bus to OPERABLE status within 8 !
hours *0r be in at least HOT SHUTDOWN within the next 12 hours and in
- p. WE COLD SHUTDOWN within the following 24 hours.
"E WL+
- p. c. Eith two or more of tne aoove requi red A.L.. cistributiok Cvstem bused j
[sW or inverters inoperable, restore at least all except one of The
*7 m
p t,.1' inoperable buses and inverters to OPERABLE status within 2 hour or 5. ! Ac w in at least HOT SHUTDOWN within the next 12 hours and in C0l ,HUTDOWN P within the following 24 hours. ~ w SURVEILLANCE REQUIREMENTS 4.8.2.1 The above recuired A.C. distribution system buses (nd inverter) shall '*
- M4
. be determined OPERABLE:
(g LLO kes << b. C. d ; .
- a. At least once per 7 days by verifying correct breaker alignment
\
*g' 2 O Nacaten p nvte rNLya 1 1 a nJ 1 t v . any g# 4 c- moveddo LC O 3.5. i - /A D
- b. At least once per 31 d$ by determining that the 250 volt DC/600 volt AC inverters 2R44-5002 and 2R44-5003 are OPERABLE by verifying inverter
,m output voltage of 600 volts + 5% while supplying their respective N.] & nch %% 54l HATCH - UNIT 2 3/4 8-10 Amendment No. 23, 36 J ot 2_ 1
Ipee,Ted,ex k L 6.0 ADMINISTRATIVE CONTROLS V f, 2 . l . c) d. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.
,f,2 2 6.2.2 UNIT STAFF
- a. \Each on duty shi shallbecomposehofatleastthemikimumshift O'I Krewcompositionsownintable6.2.E-1. \
p,.7,y, b b. At least one licensed Operator shall be in the control room for each reactor containing fuel. ht) b 2l' y . east two licensed Operators shall be present in the control room for each reactor in the process of start-up, scheduled reactor shutdown and during recovery from reactor trips. g* g,.gJ d. An individual qualified to implement radiation protection procedures shall be on site when fuel is in either reactor. [ A 1 CORE ALTERATIONS hall be directly s ervised by either a Q,1. li nsed Senior React Operator or Senior Reactor Operator imited to el Handling who ha no other concurren responsibilities during
^'N this peration.
(Q
- f. A Fir Team of at least fi members shall be m intained onsite t)
'3 all tim s. The Fire Team sh 11 not include the inimum shift cre. !
necessar for safe shutdown o Units 1 and 2 or a ' personnel required f other essential fug tions during a fi emergency.
- g. Administrative procedures shall be developed and implemented to gabg
- 4 limit the workino hours of Unit staff who perform safety-related functions; g., senior rea tor operators, gactor opera rs, }
~
fauxistaryopeators, health ysicists, and ktg maintenanc g,l personnel. Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a nominal 40-hour week while the plant is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used or during extended periods of shutdown for refueling, major maintenance, or major , plant modifications, the following guidelines shall be followed on i a temporary basis: , / (1) An individual should not be permitted to work more than 16 hours straight, excluding shift turnover _ time.. i ( , ce , ud o ra r 1, t o rv/ epeahntc , fa a q beala F h / s.cc A L.e r, Q__) f,y,go n , pocam e pnsonMAmendment i oh - ) No. 57, 94 HATCH - UNIT 2 'I 6-2 2 cf (,
c DISCUSSION OF CHANGES l ( ITS: SECTION 5.2 - ORGANIZATION V) l ADMINISTRATIVE I 1 A.1 The current Technical Specification provides examples of the Unit staff I positions who perform safety-related functions and whose working hours are ! limited. Since these examples may not include all positions that could be i limited and since these positions may change, the examples have been j generalized. The modification of these examples clarifies present E i requirements and thus is an administrative change. TECHNICAL CHANGES - MORE RESTRICTIVE M.1 CTS 6.2.2.c requires that at least two licensed operators be present in the control room for each reactor in the process of startup, scheduled reactor shutdown, and during recovery from reactor trips. The ITS is more restrictive by requiring a Senior Reactor Operator to be present in the control room while the unit is in MODES 1, 2, or 3, in addition to at least one licensed Reactor Operator. ! M.2 New requirements are being added in the ITS to specify the function of the Shift Technical Advisor (STA). The STA shall provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, (3 reactor engineering, and plant analysis with regard to the safe operation () of the unit. TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" LA.1 Details of the minimum shift crew requirements located in current Technical Specifications Table 6.2.2-1 are relocated to plant procedures.
The minimum shift crew requirements for licensed operators and senior operators contained in 10 CFR 50.54 (k), (1), and (m) and do not need to be repeated in the ITS. The minimum shift crew requirements for non- , licensed plant equipment operators transferred from present Table 6.2.2-1 to ITS 5.2.2.a. In addition, ITS 5.1.5 contains requirements for the - control room command function, ITS 5.2.2.c contains minimum requirements for licensed Reactor Operators and Senior Operators to be present in the control room, and ITS 5.2.2.g contains STA requirements. The relocation of the details of the minimum shift crew requirements to plant procedures is acceptable considering the controls provided by regulations, the remaining requirements in the ITS, and plant procedure change control by the 10 CFR 50.59 process. a p
/
HATCH UNIT 2 1 REVISIONkp
fgce,T,'cate'on 55.4 6.18 RADI0 ACTIVE EFFLUENTS CONTROLS PROGRAM (Continued) b
- 2) Limitations aat all
/S material rel(eayo in liquid effluents to UNRESTRICTED AREASti 3 on the
()f.f,4, d conforming to 10 M 5s the concentrations stated in 10 CFR Part 20 Table 2, Column,Wppenoix 2, B (to paragraphs 20.1001 - 20.2401),
- 3) sampling MonitoringInaccordan,cewith10CFR20.1302and andanalysis with theofmethodology radioactive liquid and gI i
f.f.4'C effluents and parameters in the ODCM, l
- 4) Limitations on the annual and quarterly doses or dose commitment to
- f. f. 4. 5 a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50, g,5 y' e 5) Determination of cumulative and projected dose contributions from j radioactive effluents for the current calendar quarter and current '
calendar year in accordance with the methodology and parameters in 1 the ODCM t least every 31 days, j g*g
- 9, p 6) Limitations on theT0PIRABILITY]and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment confo ing to Appendix I to 10 CFR Part 50, f' g ,y' O 7) Limitations on the dose rate resultina from radioactive mat rial fO released in gaseous effluentsQrom the site to areas at anSbeyond the SITE BOUNDARY as follows:
- a. For noble gases, less than or equal to a dose rate of 500 mrem dose rate /ofyear 3000tomrem the total body
/ year andskin, to the lessand than or equal to a 1
- b. For Iodine-131, tritium and all radionuclides in particulate formIodine-133, with half- lives gre,ater than 8 days, less than 4 i
or equal to a dose rate of 1500 mrem / year to any organ. , I E g,q., b 8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas ; beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, f, g,9, 2 9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131 Iodine-133, tritium and all radionuclides ; in particulate form with half-lives greater,than 8 days in gaseous ' effluents released from each unit to P.reas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, and p g *y J 10) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from ; uranium fuel cycle sources conforming to 40 CFR Part 190. g See wcw Av i 6.19 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM l A program shall be established, implemented, and maintained to monitor k c5c . (..h, io the radiation and radionuclides in the environs of the plant. The l % s e,,s. ;) program shall provide: in the highest potential [1)exposure representative pathways,measurements and of radioactivity r) i v the accuracy of the effluent monitoring program a(2) environmental exposure pathways. The program shall: in the ODCM Part 50, an verification nd modeling of of uidance of Appendix (, to 10 CFR1) be contained conformtothegowing: include the foi HATCH - UNIT 2 6-22a Amendment No. 129 2 ei L
DISCUSSION OF CHANGES O'- ITS: SECTION 5.5.4 - RADI0 ACTIVE EFFLUENT CONTROLS PROGRAM ADMINISTRATIVE A.1 Comment Number not used. A.2 Consistent with NUREG 1433, the phrase "at all times" and "from the site to areas at and" have been deleted. The intent of the requirement remains the same and is considered administrative. TECHNICAL CHANGES - MORE RESTRICTIVE None D ' TECHNICAL CHANGE - LESS RESTRICTIVE
" Specific" i L.1 The present TS uses the term " operability" when referring to radioactive liquid and gaseous monitoring instrumentation and treatment systems. The proposed TS uses the term " functional capability." The proposed change is necessary because the Radioactive Effluent Controls Program is located outside the TS in - the ODCM. Use of the TS term operability can be confusing when used in programs which are not in the TS. The term A
C functional capability means that the component or system is capable of performing its design function. Since it is not a TS defined term, the use of the " functional capability" is considered less restrictive than the ~ use of the TS term " operability." ,y k HATCH UNIT 2 1 .REVISIONhp ;
-'$p o'cde'Oh Se $o b See bisnwuof cL9es ;
Gr STS. 3 0;tcwa sa APPLICABILITY /)q /,i g;/, g ,s g u h a 7 0
.i M: SURVEILLANCE REOUIREMENTS (Continued)
~
4.0.3 Performance of a Surveillance Requirement within the specified time interval shall constitute compliance with OPERABILITY requirements- . for a Limiting Condition for Operation and associated ACTION statements unless otherwise r.equired by the specification. Surveillance require-ments do not have to be performed on inoperable equipment. . 4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable ,
, state shall not be made unless the. Surveillance Requirement (s) associated '
with the Limiting Condition for Operation have been performed within , the applicable surveillance interval or as otherwise specified. '
- f. 5, e, LA. I T0.5 Surveillance Requirements for Enbervice ihnectio1 and testing of ASME Code Class 1, 2, & 3 components snai s be app' icable as follows:
- a. [In'farvice inspe'et. ion of ASMNode C1 ass h' 2. and 3 combonentV Cu,l _ n n m v.ce te ins vi mm. coae alves shall sass 1, c, ana a wump.
performed in ac rdance with Sectf n XI-of th ASME Boiler d Pressure Vesse Code and applica le g'y . Addenda s required b 10 CFR-50, Secti XI 50.55a(g), : except w e specific itten relief has en granted by t '
'Copgjts11on ursuant to 1 FR !io. Section 55a(g)(6)(1). 1 K.o. Surveillance intervals specified in Sectiori XI of the ASME' /
Boiler and Pressure Vessel Code and applicattle Addenda.for the f inservice (iftgectioNan#testirty activities required by the ASME ; Boilet and Pressure vessel Code ud applicable Addenda shall.be g- , applicable as follows in these Technical Specifications
- j -l 1
4 Lh,1 . :
\
I l a LO- .- HATCH - UNIT 2 3/4 0-2 Amendment No. M 7, 117-lot ' 1
~. . . _ . . _ . - - .-. _.
3/4.0 APPLICABILITY R 't" .i L SURVEILLANCE REOUIREMENTS (Continued) l ASHE Boiler and Pressure Vessel Required frequencieb : Code and applicable Addenda for performing inservice 4 terminology for. inservice inspection and testing-insoection and testina activities activities : W6ekly At least once per 7 days D Monthly At least once per 31 days Quarterly or eery 3 months At least once per 92 days Semiannually or every 6 months At'least once per 184 days Yearly or annually At 1.aut once per 366 days y'I*g^g X. The provisions of Specificatio are applicable to the id t above required frequencies orming[1nM rvice y spe q1oy and testing activities. , f
- d. Per ance of i.h buve mser e 6nsoNetionNn@te - ny A, 2, - activi es shall b n addition o other specified Su llan
[_ Require ts.
- e. No_ thing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical ,
OI* fed *d Specification. ; f The Inservice hspection Program for p ing identified in NRC 1 Generic Letter 8 -01 shall be performed n accordance with the-taff position on chedule, methods and p sonnel, and sample . { g ,g e ansion included thp generic letter, e cpt where specific wr t,tten relief has be granted by the Commis on' 5.5.L C % povlsions d SR 3.0.3 arx *) by $ ,& : to in ,,r v,'u test la) acWh'esJ ed e
..if.,
[ ) ' Ka\w,\techsp\hia,4 0.2 HATCH - UNIT 2 3/4 0-3 Amendment No. 117 l 1ef L
ADMINISTRATIVE CONTROLS
) l l f,f.y MONTHLY OPERATING REPORT l 6.9.1.10 Routine reports of operating statistics and shutdown ex]erience shall be submitted on a monthly basisito the Director, Office of Management l find Program Analysis, U. 5. Nuclear Regulatory Commission, Washington, l ,0. C. 20555, with a copy to the Regional Office of Inspection and i M Enforcemantfno later tnan tne ndi vi teacn montn tollowing the calendar Tonth covered by the report.
i f f,y CORE OPERATING LIMITS REPORT L l f.t.5,.6.9.1.11.a. Core operating limits shall be established and documented in l the CORE OPERATING LIMITS REPORT before each reload cycle or i any remaining part of a reload cycle for the following: 5 6. I. a 0 (1) Control Rod Program Controls - Rod Block Monitor for Specification 3.1.4.3, f,f. f , q.2.) (2) The Average Planar Linear Heat Generation Rate for l Specification 3.2.1 and Surveillance Requirement 4.2.1, ' f,4,5, 4 3) (3) The Minimum Critical Power Ratio for Specifications 3.1.4.3 and 3.2.3 and Surveillance Requirement 4.2.3,
,q , and v -
g*y ( The Linear eat Generatio ate for Speci nd Surveill ce Requiremen 4.2.4. ation 3.2.4 f 6, 3. /f. b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the following documents. I
.E6.5.4.D (1) NEDE-240ll-P-A, " General Electric Standard Application for l Reactor Fuel," (applicable amendment specified in the CORE l OPERATING LIMITS REPORT). I
{ g , 6. .f. 4. r.) (2) " Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendmerd Nos.151 and 89 to Facility Operating Licenses DPR-57 and NPF-5," dated January 22, 1988. f, f . 5. c. c . The core operating limite shall ba detarmined so that all appliccble limits (e.g., tuel thermab mechanical limits, core thermal-nydraulic lim 1 As, ECCS limits, nuclear limits such as g shutdown margin. and transient and accident analysis limit O of the safety analysis are met. 5,6. 5, ) d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be 3rovided upon ItJSERT 5 issuance, for each reload cycle, to the NR{fDocument Contro AJJ PTLR esk with copies to tne RegionaT'Aaministrator and Resident A,l Inspector.
~
HATCH - UNIT 2 6-14d Amendment No. 48, 86, 406, 129 SoM
P DISCUSSION OF CilANGES ITS: SECTION 5.6 - REPORTING REQUIREMENTS A ' TECHNICAL CHANGES - MORE RESTRICTIVE M. I' The current TS requirement in 6.9.1.5.b to submit an annual report for all challenges to safety / relief valves has been moved to proposed'ITS 5.6.1.4 for monthly reports. Since the report is required on a monthly basis instead of the current annual basis, this change is more restrictive in-nature. M.2 .This change details the information to be included in the report. These details are necessary to assure the reports are provided with similar content and format for comparison with other plants and with prior reports. M.3 A new report is required in conjunction with the changes described in Section 3.4 for' the reactor coolant system pressure and temperature , limits. In addition, requirements are included for methods used to determine such limits and for submitting the report to the'NRC. TECHNICAL CHANGE - LESS RESTRICTIVE
" Generic" ,
LA.1 The details associated with CTS 6.9.1.1, 6.9.1.2, and 6.9.1.3, " Start-Up Report," are proposed to be relocated to the FSAR. The Start-Up Report ; provides the NRC a mechanism to review the appropriateness of licensee ~ activities 'after-the-fact, but provides no regulatory authority.once the i report is submitted (i.e., no requirement for NRC approval). The Quality , Assurance requirements of 10 CFR- 50, Appendix B, and 'the Startup Test . Program provisions contained in the FSAR provide assurance the listed activities will be adequately performed and that appropriate corrective ! actions, if required, are taken. The placement of these CTS requirements ! in the FSAR also ensures. that change control is ' performed in accordance with 10 CFR 50.59. l O HATCH UNIT 2 3 REVISION K p
4 e bc.lo,,4 [&cn: us,,o./ a Jd- (/ ADMINISTRATIVE CONTROLSA , b c ;Owh. lh I [RECORDRETENTION(Continued) (Q< d c. Records of radiation exposure for all individuals entering radiation control areas.
- d. Records of gaseous and liquid radioactive material released to the environs i k e. Records i identifiedofintransient or o0erational Table 5.7.1 1. cycles for those unit components {-
i
- f. Records of reactor tests and experiments.
{ i
- g. Records of training and qualification for current members of the unit staff,
- h. Records of in-service ins Technical Specifications.pections performed pursuant to these
- i. Records of Quality Assurance activities required by the QA Manual.
J. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
- k. Records of meetings of the PRB and the SRB.
- 1. Records for Environmental Qualification which are covered under the provisions of paragraph 6.15.
- m. Records of analyses required by the Radiological Environmental
! Monitoring Program.
9>a n. Records of the service lives of all safety-related hydraulic and .\ ( - mechanical snubbers including the date at which the service life U\ pg5 commencesandassocIatedinstallationandmaintenancerecords. fwcd:c.n o. Records of reviews performed for changes made to the OFFSITE DOSE g ~ % ssu w,. CALCULATION MANUAL and the PROCESS CONTROL PP.0 GRAM. L RADIATION PROTECIIUN VHUGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved maintained Ladhered to for all operations involving personnel radiation exposure.and A 12 HIGH RADIATION AREA . 6.17 1 In lieu of the " control device" br " alarm signal" required by paragraph 20.16 radiati 1(a)is greater than 100 mrem /hr but less than 1000 mrem /hr** shall beof 10 CF barrica and conspicuously posted as a high radiation area and entrance thereto s 11 be controlled by requiring issuakce of a Radiation Work Permit *. Any individoel or group of individuals p ted Te cu c io eis 1 or
- Health Physich per::nn:;;;;rd;;;e 4th -:p r:/ dd re r:: d bh. , be
- ,=r personnel
- h; Health esc Physics exempt personnel from the RWP 4a issuance requirenteht during tEency per formance of their kssigned radiation protection duties provided they comply with ap proceduresforenf.rfintohighradiationareas.provedradiationprotection
& ** Measured at 30 cm frds the radiation source or from any sbtface that the radiation penetrates.
HATCH - UNIT 2 6-18 Ordee 4t4 ' a m 'on 1 ca s ;,n d o. ls p,7(, ,/ ;,. N AmendmentNo.48[II,I26 rod.aM preke+;e- pvw/f yy oy .-
l 5pedAdou 5.} ADMINISTRATIVE CONTROLS - enter such areas shall be provided witn or acbqmpanied uy one or more of ' the 11owing:
- a. A radiation monitoring device which continuouMy indicates the radLation dose rate in the area.
- b. A radiathn monitoring device which continuously intbqrates the radiation d&se rate in the area and alarms when a presht integrated dose is receivetk. Entry into such areas with this monitbring device may be made (ter the dose rate level in the area has been -i established and persortne.1 have been made knowledgeable of thesq.
- c. An individual qualified in radiation protection procedures who ib equipped.with a radiation dose rate monitoring device. This individual shall be responsible t&r providing positive control over the activities within the area and siall perform periodic radiation surveillance at the frequency specified the facility Health Physics supervision in the Radiation Work it.
6.12.2 Th's requirements of 6.12.1, above, shall also apply, to each g radiation area in which the intensity of radiation is greater tharF<W +* 1000 mrem /hr* But less than 500 rads in I hour.** In addition, locked doors shall be provided' to prevent unauthorized entry into such areas arid .the kc)s N shall be maintained.under the administrative control of the Shift Sup'e(visor {V on duty andjor tha La'horatory Foreman on duty. ' [6.13 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT The licensee shall implement a program to reduce leakage from systems outside- D containment that would or could contain highly radioactive fluids-during a serious transient or accident to as low as practical levels. This program shall include the following:
- 1) Provisions establishing preventive maintenance and periodic visual- D inspection requirements, and _i i
2)- System leakage test-requirements, to the extent permitted by system j design and radiological conditions, for each system at a frequency not to exceed refueling cycle intervals. The systems subject to this testing are (1) Residual Heat Removal, (2) Core Spray, (3) Reactor Water Cleanup, , (4) HPCI, and (5) RCIC. See bassico dbyes 4 (
'T5' M 2, w % s .Ses w . /
l asurement made at 30 centime rs from the radiation s rce or from any su ace that the radiation penet tes. ' **Measur ent made at 1 meter from th radiation source or fro _ that th radiation penetrates. anysurfa)ce-HATCH - UNIT 2 6-19 Amendment No. M , 47, 129 - ML
l
,. 3 DISCUSSION OF CHANGES
( ) ITS: SECTION 5.7 - HIGH RADIATION AREA ADMINISTRATIVE j l A.1 Consistent with NUREG 1433, a phrase allowing individuals who are qualified in radiation protection procedures has been added. No change in intent is made. Only personnel qualified would be permitted to perform the function. A.2 Consistent with NUREG 1433, the phrase "in accordance with approved emergency procedures" has been deleted. The intent of this footnote was to allow performance of assigned duties (e.g., surveys) to be performed g without creating an undue administrative burden. The NUREG 1433 wording conveys the intent accurately, therefore this change is considered administrative. A.3 Consistent with NUREG 1433 or equal to is added, such that the range of the radiation intensity is covered. This change is considered administrative.
,y I 1 %)
. \ (T$ HATCH UNIT 2 %1 REVISION [
I 1 O UNIT 2 NO SIGNIFICANT HAZARDS DETERMINATION O I O
l l
,_ N0 SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.1.3 - PRIMARY CONTAINMENT ISOLATION VALVES (v)
L.1 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 proposed 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 a new or different kind of accident from any accident previously evaluated?
Since the result of the ACTION continues to be an acceptably isolated (^% penetration for continued operation, the proposed change does not V adversely affect safe 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 margin of safety considered in determining the required compensatory action is based on providing the single active failure proof boundary. 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. Therefore, the change does not involve a significant reduction in the margin of safety.
/h k% ))
HATCH UNIT 2 1 REVISION D
-% NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.6.1.8 - SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS
'(b') 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 change extended the requirement to cycle the vacuum breakers after an S/RV lift from 2 hours to 12 hours. The vacuum breakers are not assumed to be an initiator of any previously analyzed accident. Therefore, 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 vacuum breakers remain operable to mitigate the assumed accidents.
- 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 modification to the plant. Therefore, it does-
/~ not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The OPERABILITY of the vacuum breakers is not affected by an S/RV lift, since Georgia Power Company has completed the torus modifications for Plant Hatch Unit 2. The installed T-quenchers ensure that all steam is condensed in the suppression pool and will not significantly increase the humidity in the suppression chamber air space (this increased humidity is postulated to negatively impact the OPERABILITY of the Vacuum breakers). In addition, a review of the vacuum breaker failure rate during the surveillance performed after an S/RV lift shows that it is essentially the same as the failure rate during the routine 31 day Surveillance. Furthermore, this extension in the performance of the vacuum breaker functional test is supported by the NRC in Generic Letter 93-05, item 8.4. Therefore, extension of this requirement will not involve a significant reduction in a margin of safety. p V HATCH UNIT 2 3 REVISION D
-s NO SIGNIFICANT HAZARDS DETERMINATION (t) ITS: SECTION 3.6.4.4 - SECONDARY CONTAINMENT ISOLATION VALVES - OPERATING L.2 CHANG.1 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 considerW as an initiator of any previously analyzed accident. Therefore, this change does not significantly increase the probability of such accidei.t .. The proposed 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 (~T penetration for continued operation, the proposed change does not Q 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 significant reduction in a margin of safety?
The margin of safety considered in determining the required compensatory action is based on providing the single active failure proof boundary. 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. Therefore, the change does not involve a significant reduction in the margin of safety. /m HATCH UNIT 2 2 REVISION D
N0 SIGNIFICANT HAZARDS DETERMINATION -(v o) ITS: SECTION 3.6.4.6 - SECONDARY CONTAINMENT ISOLATION VALVES - REFUELING 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?
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 proposed 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 a new or different kind of accident from any accident previously evaluated?
Since the result of the ACTION continues to be an acceptably isolated
/~\ penetration for continued operation, the proposed change does not C) adversely affect safe 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 margin of safety considered in determining the required compensatory action is based on providing the single active failure proof boundary. 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. Therefore, the change does not involve a significant reduction in the margin of safety. l m
\
HATCH UNIT 2 3 REVISION D
- N NO SIGNIFICANT HAZARDS DETERMINATION ITS: SECTION 3.8.6 - BATTERY CELL PARAMETERS L.4 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 DC electrical power sources are used to support mitigation of the consequences of an accident; however, they are not considered the initiator of any previously analyzed accident. As such, revising the acceptance criteria for specific gravity measurements will not increase the probability of any accident preciously evaluated. The proposed LCO and SRs continue to provide adequate assurance of OPERABLE batteries since the change continues to ensure the battery state-of-charge does not affect l the battery's capability to perform its required function. Therefore, the l proposed change does not involve an increase in the consequences of any accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of i accident from any accident previously evaluated?
The proposed change does not introduce a new mode of plant operation and does not involve physical modification to the plant. Therefore, it does l not create the possibility of a new or different kind of accident from any accident previously evaluated. l
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed specific gravity limits are consistent with those recommended by the IEEE-450 Battery Working Group, and approved for Technical Specification application by the NRC. New limits on the average of all connected cells is also added with this change, with the value of the limit the same as the current individual cell acceptance criteria. As long as the average specific gravity is adequate (which is essentially the same as the current requirement assuming all cells at the specific gravity limit), deviation of individual cells will not reduce the net capacity of the battery. Therefore, this change does not involve a significant reduction in the margin of safety.
-HATCH UNIT 2 4 REVISION D
)
7 NUREG 1433 COMPARISON DOCUMENT - SPECIFICATIONS O i i O w
,LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LCO 3.0.7.
LCO .3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as ' providedyin LCO 3.0.6.
- e. If the LCO is met or is no longer applicable rior to WO3 expiration of the specified Completion Time (s , completion (f A - of the Required Action (s) is not required, unless otherwise
. stated. l l
LCO 3.0.3 When an LCO is not met and the associated ACTIONS'are not I
# metj g an associated ACTION is not provided, the unit shall.
be placed in a MODE or other specified condition in which
;gC yrh gb the LCO is not applicable. Action shall be initiated.within ia 1 hour to place the unit, as applicable, in:
p( O I A a. MODE 2 within 7 hours; M {Wpt Nh b. MODE 3 within 13 hours; and- [ A 2-
- c. MODE 4 within 37 hours.
Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that pensit operation in accordance with the LC0 or ACTIONS, completion of the actions required by LCO 3.0.3 is' not required. LC0 3.0.3 i a MODES 1, 2, and 3. vw LCO 3.0.4 When an LC0 is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued-operation in the MODE or other specified condition in the Applicability for an unlimited period'of time. This (continued) O BWR/4 STS 3.0-1 Rev. O, 09/28/92
f./ LCO Applicability 3.0 LCO APPLICABILITY g oCM r tha+ M8# #sre f r t o f a. LCO 3.0.4 Specification shall not prevent changes in MODES or other (continued) specified conditions i the Applicability that are required to comply with ACTIONSy Exceptions to this Specification are stated in the individual Specifications. These exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered allow unit operation in the MODE or other specified condition in the Applicability only for a limited period of time. LCO 3.0.5 Equipment ramoved from service or declared inoperable to ! comply with ACTIONS may be returned to service under ' administrative control solely to perform testing required to D demonstrate its OPERABILITY or the OPERABILITY of other equipment,. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. LCO 3.0.6 When a supported system LCO is not met solely due to a O support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to 1 LCO 3.0.2 for the supported system. In this event, p v% , additional evaluations and 1 itations may be required in (w) accordince witn specifi catic &:t, " Safety Function { pt Determination Program (SFDP)." If a loss of safety function j l is determined to exist by this program, the appropriate i Conditions and Required Actions of the LCO in which the loss l Ji) of safety function exists are required to be entered.
,) When a support system's Required Action directs a supported j system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the j applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.
{ (continued) ! BWR/4 STS 3.0-2 Rev. O, 09/28/92
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