Text

Proposed Tech Specs Pages Revising TS to Implement Previously Approved Generic Changes
	ML20203C593
Person / Time
Site:	Hatch
Issue date:	02/05/1999
From:	; SOUTHERN NUCLEAR OPERATING CO.
To:	;
Shared Package
ML20203C589	List: HL-5591, Application for Amends to Licenses DPR-57 & NPF-5, Respectively.Amends Would Implement Some of Generic Changes to ITSs Previously Approved by NRC; ML20203C593;
References
	NUDOCS 9902120139
	Download: ML20203C593 (72)
	v • d • e

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l SLs l 2.0 l I

' 2.0- SAFETY LIMITS (SLs) l l

2.1 SLs !

l 2.1.1 Reactor Core Si i l

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2.1.1.1 With the reactor steam dome pressure < 785 psig or core l flow < 10% rated core flow: I THERMAL POWER shall be s 25% RTP. l l

2.1.1.2 With the reactor steam dome pressure a 785 psi; and core '

flow 210% rated core flow:

MCPR shall be a 1.10 for two recirculation loop operation or 21.12 for single recirculation loop operation.*

2.1.1.3 Reactor vessel water level shall be greater than the top l of active irradiated fuel.

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'2.1.2 Reactor Coolant System (RCS) Pressure SL Reactor steam dome pressure shall be 51325 psig.

The spacified limits are for Cycle 18 only.

2.2 SL Violations ;

l With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

l 2.2.1 Restore compliance with all SLs; and

! l 2.2.2 Insert all insertable control rods. '

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9902120139 990205 7 yDR ADOCK05000321h PDR [ ,

I HATCH UNIT 1 2.0-1 Amendment No. 3/25/98 i

LC0 Applicability 3.0 3.0 LCO APPLICABILITY i

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 3.0.4 is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3.

LC0 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY, 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 :esting.

LC0 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only thc 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, an evaluation is 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 LC0 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 Amer.dment No. 3/25/98 t-

Control Red OPERABILITY.

3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A; (continued) A.2 Perform SR 3.1.3.2 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from and SR 3.1.3.3 for discovery of each withdrawn Condition A OPERABLE control rod. concurrent with thermal power greater than the low power setpoint (LPSP) of the RWM.

AND A.3 Perform SR 3.1.1.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B. Two or more withdrawn B.1- Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I control rods stuck.

(continued)

HATCH UNIT 1 3.1-8 Amendment No. 3/25/98

RPS Instrumentation 3.3.1.1 Table 3.3.1.1 1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDIiIONS MODES OR REQUIRED REFERENCED OTHER CNANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VAWE

7. Scram Discherpe voltme Water Level-Nigh

a. Resistance 1,2 2 G SR 3.3.1.1.9 5 71 pallons Temperature Sk 3.3.1.1.13 Detector SR 3.3.1.1.15 5(*) 2 N SR 3.3.1.1.9 s 71 gallons SR 3.3.1.1.13 SR 3.3.1.1.15

b. F!ont Switch 1,2 2 G SR 3.3.1.1.13 s 71 salLons SR 3.3.1.1.15 5(*) 2 N SR 3.3.1.1.13 s 71 sattons SR 3.3.1.1.15

8. Turbine stop t 30% RTP 4 E SR 3.3.1.1.9 s 10% closed valve - Closure SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.15

9. Turbine control Valve t 30% RTP 2 E SR 3.3.1.1.9 t 600 psig fast Closure, Trip Oil SR 3.3.1.1.11 Pressure - Low SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.16

10. Reector Mode switch - 1,2 1 G SR 3.3.1.1.12 NA Shutdown Position SR 3.3.1.1.15 5(a) 1 N SR 3.3.1.1.12 NA SR 3.3.1 *.15

11. Manual Scram 1,2 1 G SR 3.3.1.1.5(b) gg SR 3.3.1.1.15 5(a) 1 N SR 3.3.1.1.5 NA SR 3.3.1.1.15 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Includes exercising the K14 auto scram relays. l l

HATCH UNIT 1 3.3-8 Amendment No. 3/25/98

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Prinary Containment Air Lock 3.6.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.2.1 ------------------NOTES------------------

1.. An inoperable air lock door does not ,

invalidate' the previous successful '

performance of the overall' air lock leakage test.

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2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.1.

Perform required primary containment air In accordance

lock leakage rate testing in accordance with the with the Primary Containment Leakage Rate Primary Testing Program. Containment Leakage Rate Testing Program

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! SR 3.6.1.2.2 Verify only one door in the primary 24 months containment air lock can be opened at a l time.

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HATCH UNIT 1 .

3.6-7 Amendment No. 3/25/98 ;

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t PCIVs 3.6.1.3 SURVEILLANCE RE0VIREMENTS SURVEILLANCE FREQUENCY ,

i SR 3.6.1.3.1 ------------------NOTE-------------------

Not required to be met when the 18 inch !

primary containment purge valves are open !

for inerting, de-inarting, pressure ;

control, ALARA, or cir quality :

considerations for personnel entry, or ;

Surveillances that require the valves to '

be open. ,

i Verify each 18 inch primary containment 31 days purge valve is closed.

SR 3.6.1.3.2 ------------------NOTE-------------------

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

l l 2. Not required to be met for PCIVs that l are open under administrative l controls. -

Verify each primary containment isolation 31 days ;

manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed.

(continued) l i

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i HATCH UNIT 1 3.6-12 Amendment No. 3/25/98 l 1

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PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

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 or 3 from not locked, sealed, or otherwise secured MODE 4 if and is required to be closed during primary accident conditions is closed. containment was de-inerted while in MODE 4, if not performed within the previous 92 days SR 3.6.1.3.4 Verify continuity of the traversing 31 days

! incore probe (TIP) shear isolation valve explosive charge.

SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated and each automatic PCIV, except with the for MSIVs, is within limits. Inservice Testing Program (continued)

HATCH UNIT 1 3.6-13 Amendment No. 3/25/98 1

l DC Sources - Operating

, 3.8.4 SURVEILLANCE REQUIREMENTS L -----------------------------------NOTE---------------------------------------

SR 3.8.4.1'through SR 3.8.4.8 are applicable only to the' Unit 1 DC sources.

u SR 3.8.4.9 is applicable only to the Unit 2 DC sources.

SURVEILLANCE FREQUENCY l SR 3.8.4.1 Verify battery terminal voltage is a: 125 V 7 days

on float charge.

l SR 3.8.4.2 Verify no visible corrosion at battery 92 days

terminals and connectors.

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i Verify battery connection resistance is within limits.

SR 3.8.4.3 Verify battery cells, cell' plates, and 18 months l racks show no visual indication _ of physical damage or abnormal deterioration that.could degrade battery performance.

.SR 3.8.4.4 Remove visible corrosion, and verify 18 months battery cell to cell and terminal connections are coated with anti-corrosion material.

SR 3.8.4.5 Verify battery connection resistance is 18 months within limits.

L (continued)

HATCH UNIT 1 3.8-30 Amendment No. 3/25/98

F Prrgra:s and Manuals -

5.5- !

5.0- ADMINISTRATIVE CONTROLS: i 5.5 . Programs and Manuals j i

The following programs and manuals shall be established, implemented, and - !

maintained. i i

5.5.1 Offsite Dose Calculation Manual (ODCML l

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from i radioactive gaseous and liquid effluents, in the calculation ,

of gaseous and liquid effluent monitoring alarm and trip j setpoints, and in the conduct of the radiological i environmental monitoring program; and ;

b. The ODCM shall also contain the radioactive effluent 1 '

controls and radiological environmental monitoring activities,.and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release reports required by Specification 5.6.2 and Specification 5.6.3, respectively.

Licensee initiated changes to the ODCM:

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a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. Sufficient information to support the change (s) and appropriate analyses or evaluations justifying the change (s),and

2. A determination that the change (s) maintain the levels of radioactive effluent control required by 10 CFR 20.106, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and does not adversely' impact the accuracy or reliability of effluent, dose, or I setpoint calculations. '

b. Shall become effective after approval of the plant manager; '

and (continued) !

HATCH UNIT 1 5.0-7 Amendment No. 3/25/98

SLs 2.0 .

2.0 SAFETY LIMITS (SLs) ;

2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core '

flow < 10% rated core flow:

THERMAL POWER shall be s: 25% RTP.

2.1.1.2 With the reactor steam dome pressure 2: 785 psig and core flow 2: 10% rated core flow:

MCPR shall be at 1.12 for two recirculation loop operation or a 1.14 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top !

of active irradiated fuel. I 2.1.2 Reactor Coolant System (RCS) Pressure SL Reactor steam dome pressure shall be s; 1325 psig.

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2.2 SL Violations j With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

l 2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

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HATCH UNIT 2 2.0-1 Amendment No. 3/25/98

l LCO Applicability 3.0 t

3.0. LCO APPLICABILITY i LCO 3.0.4 to comply with ACTIONS or that are part of a. shutdown of the ;

(continued)- unit.

l Exceptions to this Specification are stated in the !

individual Specifications. These exceptions allow entry l into MODES or other specified conditions in the i Applicability when the associated ACTIONS to be entered j allow unit operation in-the MODE or other specified ,

condition in the Applicability only for a limited period of !

time. 1 I

LCO 3.0.4 is only applicable for entry into a MODE or other !

specified condition in the Applicability in MODES 1, 2, j and 3.

s LCO 3.0.5 Equipment removed from service or-declared inoperable to ,

comply with ACTIONS may be returned to service under l administrative control solely to perform testing required to a demonstrate its OPERABILITY or the OPERABILITY of other l equipment. This is an exception to LC0 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 to a 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 LCO 3.0.2 for the supported system. -In this event, an evaluation is required in accordance with i 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.

I 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 applicabic Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

(continued)

HATCH UNIT 2 3.0-2 Amendment No. 3/25/98

J #8 Bi -_+e _p. ---JK A. .-AA4 =u.3 g m J p & --- Ju - 4 +A-%.h -. - J M,.* A _ a -,.J.-_h-.5,.4a6-J w- r Lu Control Rod OPERABILITY 3.1.3

- ACTIONS j CONDITION REQUIRED ACTION COMPLETION TIME j i

A. (continued) A.2 Perform SR 3.1.3.2 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from ,

and SR 3.1.3.3 for discovery of {

each withdrawn Condition A !

OPERABLE control rod, concurrent with ;

thermal power greater than the low power ,

setpoint (LPSP) <

of the RWM. l l

O A.3 Perform SR 3.1.1.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ,

l B. Two or more withdrawn B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l i control rods stuck. l (continued) i HATCH UNIT 2 3.1-8 Amendment No. 3/25/98

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Pritary Containment Air Lock I 3.6.1.2 h

't SURVEILLANCE' REQUIREMENTS .

1 SURVEILLANCE FREQUENCY !

i SR 3.6.1.2.1 ------------------NOTES----------------- I

1. An inoperable air lock door,does not ;

invalidate the previous successful 1 performance of the overall air lock j leakage test. '

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2. Results shall be evaluated against l acceptance criteria applicable to i SR 3.6.1.1'.1 Perform required primary containment air In accordance i lock leakage rate testing in accordance with the with the Primary Containment Leakage Rate Primary ,

Testing Program. Containment i Leakage Rate Testing Program SR 3.6.1.2.2 Verify only one door in the primary 24 months containment air lock can be opened at a ;

time.

HATCH UNIT 2 3.6-7 Amendment No. 3/25/98

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PCIVs 3.6.1.3 l

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.3.1 ------------------NOTE-------------------- I Not required to be met when the 18 inch primary containment' purge. valves are open ,

.for inerting, de-inerting,' pressure .

control,'ALARA, or air quality i considerations for personnel entry,.or i Surveillances that require the valves to be open.

Verify each 18 inch primary containment 31 days purge valve is closed.

SR 3.6.1.3.2 ------------------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 primaryL containment isolation 31 days 1 manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured

.and is required to be closed during ~

accident conditions is closed.

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i HATCH UNIT 2 3.6-12 Amendment No. 3/25/98 i

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! PCIVs i 3.6.1.3 l

StlRVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY r SR 3.6.1.3.3 ------------------NOTES------------------

1. Valves and blind flanges in high ,

radiation areas may be verified by.

use'of administrative means.

i i Not required to be met for PCIVs that 2.

are open under administrative i controls.

Verify each primary containment manual Prior to )

isolation valve and blind flange that is entering MODE 2 located inside primary containment and or 3 from i not locked, sealed, or otherwise secured MODE 4 if ;

and is required to be closed during primary ;

accident conditions is closed. containment was i de-inerted I while in !

MODE 4, if not 1 performed within the previous 92 days SR 3.6.1.3.4 e'erify continuity of the traversing 31 days ncore probe (TIP) shear isolation valve i explosive charge. l SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated and each automatic PCIV, except with the for MSIVs, is within limits. Inservice .

Testing Program I l

(continued) l i

HATCH UNIT 2 3.6-13 Amendment No. 3/25/98 l

DC Sources -- Operating .

, 3.8.4 SURVEILLANCE REQUIREMENTS l 1

NOTE------------------------------------- i SR 3.8.4.1~ through SR 3.8.4.8 are applicable only to the Unit 2 DC sources. '

SR 3.8.4.9.is applicable only to the Unit 1 DC sources. !

SURVEILLANCE FREQUENCY SR 3.8.4.1' Verify battery terminal voltage is 2: 125 V- 7 days on float charge. -i SR 3.8.4.2 Verify'no visible corrosion at battery 92 days terminals and connectors.

QB !

Verify battery connection resistance is j within limits. i i

.SR 3.8.4.3 Verify battery cells, cell plates, and 18 months !

racks show no visual indication of physical damage or' abnormal deterioration that could degrade battery performance. l SR 3.8.4.4 Remove visible _ corrosion, and ,terify -18 months battery cell to cell and termiral connections are coated with anti-corrosion material. l t

SR 3.8.4.5 Verify battery connection resistance is 18 months '

within limits.

(continued)

HATCH UNIT'2 3.8-30 Amendment No. 3/25/98 l

l

Prggrans and Manuals 5.5

, 5.0 ADMINISTRATIVE CONTROLS

~5.5 Programs and Manuals ;

l The following programs and manuals shall be established, . implemented, and l maintained. i 1

5.5.1 Offsite Dose Calculation Manual (ODCM) '

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from j radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b. The ODCM shall also contain the radioactive effluent

, controls and radiological environmental monitoring ,

activities, and descriptions of the information that should ,

be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release reports required

by Specification 5.6.2 and Specification 5.6.3, respectively.

Licensee initiated changes to the ODCM:

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a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. Sufficient information to support the change (s) and appropriate analyses or evaluations justifying the change (s),and

2. A determination that the change (s) maintain the levels !

of radioactive effluent control required by l 10 CFR 20.106, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and does not adversely impact ;

the accuracy or reliability of effluent, dose, or setpoint calculations.

b. Shall become effective after approval of the plant manager; l l and (continued)

HATCH UNIT 2 5.0-7 Amendment No. 3/25/98

l Enclosure 4 I Edwin I. Hatch Nuclear Plant Request to Revise Technical Specifications to Implement Previously Approved Generic Changes Bases Page Change Instructions

.U!!ill Eagg Instruction l l

B 2.0-4 Replace l B 2.0-5 Replace B 3.1-16 Replace B 3.1-17 Replace B 3.1-18 Replace ;

l B 3.1-19 Replace l l

B 3.1-20 Replace B 3.1-21 Replace

, B 3.6-13 Replace l B 3.6-24 Replace l l l ILaill Eggg Instruction 1

B 2.0-4 Replace B 2.0-5 Replace B 3.1-16 Replace B 3.1-17 Replace l B 3.1-18 Replace B 3.1-19 Replace

, B 3.1-20 Replace i i B 3.1-21 Replace l B 3.6-13 Replace l B 3.6-24 Replace 4

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HL-5591- E4-1

Reactor Core SLs l B 2.1.1 l BASES .

l APPLICABLE 2.1.1.3 Reactor Vesjiel Water level ;

SAFETY ANALYSES (continued) During MODES I and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the i active irradiated fuel to provide a reference point and to also provide adequate margin for effective action. r SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel -

water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT 2.2.1 and 2.2.2 VIOLATIONS Exceeding an SL may cause fuel damage and create a potential for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the operators take prompt remedial action and also~ ensures that the probability of an accident occurring during this period is minimal.

(continued) l HATCH UNIT 1 B 2.0-4 REVISION 3/25/98 J

. . ~ - . - - .- - . . . . . . - . .

Reacter Core SLs ,

B 2.1.1 i

BASES (continued) l REFERENCES 1. 10 CFR 50, Appendix A, GDC 10. l

2. NEDE-240ll-P-A, " General Electric Standard Application !

for Reactor Fuels," (revision specified in the COLR). i

3. 10 CFR 50.72.

4. 10 CFR 100.

5. 10 CFR 50.73.

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HATCH UNIT 1 B 2.0-5 REVISION 3/25/98 l

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Control R:d OPERABILITY B 3.1.3 BASES ACTIONS A.I. A.2. and A.3 (continued) to perform the Required Action in an orderly manner. The control rod must be isolated from both scram and normal ,

insert and withdraw pressure. Isolating the control rod !

from scram and normal insert and withdraw pressure prevents damage to the CRDM. The control rod should be isolated from I scram and normal insert and withdraw pressure, while maintaining cooling water to the CRD.

i Monitoring of the insertion capability of each withdrawn !

control rod must also be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of discovery of the stuck withdrawn control rod, concurrent with thermal power being greater than the low power setpoint (LPSP) of the RWM. SR 3.1.3.2 and SR 3.1.3.3 perform periodic tests of the control rod insertion capability of :

withdrawn control rods. Testing each withdrawn control rod ensures that a generic problem does not exist. The allowed Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests. Required Action A.2 is modified by a Note, which states that the requirement is not applicable when THERMAL POWER is less than or equal to ,

the actual low power setpoint (LPSP) of the RWM since the notch insertions may not be compatible with the requirements i of rod pattern control (LCO 3.1.6) and the RWM ;

(LCO 3.3.2.1).

To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Should a DBA or transient require a shutdown, to preserve the single failure criterion, an additional control rod would have to be assumed to fail to insert when required. Therefore, the original SDM demonstration may not !

be valid. The SDM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn.

The allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to verify SDM is 1 adequate, considering that with a single control rod stuck l

'in a withdrawn position, the remaining OPERABLE control rods )

are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an ;

additional control rod adjacent to the stuck control rod !

l (continued)

HATCH UNIT I B 3.1-16 REVISION 3/25/98

Centrol Rod OPERABILITY B 3.1.3 BASES ACTIONS A.1. A.2. and A.3 (continued)

(continued) also fails to insert during a required scram. Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach and maintain MODE 3 conditions (Ref. 5).

L1 l With two or more withdrawn control rods stuck, the stuck.

control rods must be isolated from scram pressure within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per Action A.1 and the plant brought to MODE 3 l within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The control rods must be isolated from 1 both scram and normal insert and withdraw pressure. '

Isolating the control rod from scram and normal insert and i withdraw pressure prevents damage to the CRDM. The control !

rod should be isolated from scram and normal insert and withdraw pressure, while maintaining cooling water to the CRD. The allowed Completion Time is acceptable, considering the low probability of a CRDA occurring during this interval. The occurrence of more than one control rod stuck at a withdrawn sosition increases the probability that the reactor cannot 3e shut down if required. Insertion of all !

insertable control rods eliminates the possibility of an I additional failure of a control rod to insert. The allowed l Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

C.1 and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and disarmed (electrically or hydraulically) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected.

The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. The control rods can be l electrically disarmed by disconnecting power from all four

! directional control valve solenoids. Required Action C.1 is l modified by a Note, which allows the RWM to be bypassed if required to allow insertion of the inoperable control rods (continued) i HATCH UNIT 1 B 3.1-17 REVISION 3/25/98

Centrol Rod OPERABILITY B 3.1.3 BASES ACTIONS C.1 and C.2 (continued) and continued operation. LCO 3.3.2.1 provides additional requirements when the RWM-is bypassed to ensure compliance with the CRDA analysis.

The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems.

D.1 and D.2 Out of sequence control rods may increase the potential reactivity worth of a dropped control rod during a CRDA. At

s; 10% RTP, the generic licensing basis banked position withdrawal sequence (BPWS) analysis (Ref. 5) assus.es inserted control rods not in compliance with BPWS to be separated by at least two OPERABLE control rods in all directions, including the diagonal. Plant specific BPWS analysis may justify relaxed requirements on inoperable control rod. separability. Therefore, if two or more inoperable control-rods are not in compliance with BPWS (and not separated by at least two OPERABLE control rods, unless the plant specific analysis relaxes this requirement),

action must be taken to restore compliance with BPWS or restore the control rod (s) to OPERABLE status. Condition D is modified by a Note indicating that the Condition is not applicable when > 10% RTP, since the BPWS is not required to be followed under these conditions, as described in the Bases for LCO 3.1.6. The allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is acceptable, considering the low probability of a CRDA occurring.

L.1 l l

If any Required Action and associated Completion Time of l Condition A, C, or D are not met, or there are nine or more inoperable control rods, 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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , i This en ures all insertable control rods are inserted and places the reactor in a condition that does not require the !

active function (i.e., scram) of the control rods. The i 1

(continued) i HATCH UNIT 1 B 3.1-18 REVISION 3/25/98 l j I

Control Rod OPERABfLITY '

B 3.1.3 BASES ,

ACTIONS Ed (continued)-

number of control rods permitted to be inoperable when ,

operating above 10% RTP.(e.g., no CRDA considerations) could be more than the value specified, but the occurrence of a large number of inoperable control rods could be indicative of a generic problem, and investigation and resolution of.

the potential problem should be undertaken. The allowedCompletion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on ;

operating experience, to reach MODE 3 from full power in an ;

orderly manner and without challenging plant systems.

l SURVEILLANCE $R 3.1.3.1 i REQUIREMENTS . .

r The-position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining control rod

-OPERABILITY and controlling rod patterns. Control rod ;

position may be determined by the use of OPERABLE position i indicators, by moving control rods to a position with an ;

OPERABLE indicator, or by the use of other appropriate i methods. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of this SR is based on- 7 operating experience related to expected changes in control l rod position and the availability of control rod position ;

indications in the control room.

)

SR 3.1.3.2 and SR 3.1.3.3 I control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves.

The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal. These Surveillances are not required when THERMAL POWER is less than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the Banked Position Withdrawal Sequence (BPWS) (LCO 3.1.6) and the RWM (LCO 3.3.2.1). The 7 day frequency of SR 3.1.3.2 is based on operating experience related to the changes in CRD performance and the ease of performing notch testing for fully withdrawn control rods. Partially withdrawn control (continued)

HATCH UNIT 1 B 3.1-19 REVISION 3/25/98 l J

p Control' Rod OPERABILITY B 3.1.3 t

BASES SURVEILLANCE SR 3.1.3/2 and SR 3.1.3.3 (continued)

REQUIREMENTS rods are tested at a 31 day Frequency,' based on the potential power reduction required to allow the control rod movement and considering the large testing sample of SR 3.1.3.2. 'Furthermore, the 31 day Frequency takes into account operating experience related to changes in CR0 performance. At any time, if a control rod is immovable, a determination of that control rod's tripability (capable ofinsertion by scram, i.e., OPERABILITY) must be made and appropriate action taken.

These SRs are each modified by a Note that allows 7 days and 31 days, respectively, after withdrawal of the control rod and THERMAL POWER is greater than the LPSP to perform the Surveillance. This acknowledges that the control rod must first be withdrawn and THERMAL POWER must be greater than the LPSP before performance of the Surveillance, and therefore avoids potential conflicts with SR 3.0.3 and SR 3.0.4.

SR 3.1.3.4 Verifying that the scram time for each control rod to notch position 06 is s; 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function.

This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1, " Reactor Protection System (RPS)

Instrumentation," and the functional testing of SDV vent and drain valves in LCO 3.1.8, " Scram Discharge Volume (SDV)

Vent and Drain Valves," overlap this Surveillance to provide complete testing of the assumed safety function. The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.

1 (continued)

HATCH UNIT 1 B 3.1-20 REVISION 3/25/98 l

Control Rod OPERABILITY B 3.1.3 ;

l BASES ;

SURVEILLANCE. SR 3.1.3.5 REQUIREMENTS

-(continued) Coupling verification is performed to ensure the control rod is connected to the CRDM and will perform its intended function when necessary. The Surveillance requires >

verifying a control rod does not go to the withdrawn ;

overtravel. position. The overtravel position feature i provides a positive check on the coupling integrity since l only an uncoupled CRD can reach the overtravel position. ]

The verification is required to be performed any time a l control rod is withdrawn to the' full-out position (notch )

position 48) or prior to declaring the control rod OPERABLE :

after work on the control rod or CRD System that could (

affect coupling. . This includes control rods inserted one ;

notch and then returned to the full-out position during the i performance of SR 3.1.3.2. This Frequency is acceptable, l considering the low probability that a control rod will ;

become uncoupled when it is not being moved and operating- i experience related to uncoupling events.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 27, GDC 28, i and GDC 29. .

2. FSAR, Section 3.4.

3. FSAR, Appendix M.

4. FSAR, Sections 14.3 and 14.4.

5. NEDO-21231, " Banked Position Withdrawal Sequence,"

Section 7.2, January 1977.

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

HATCH UNIT 1 B 3.1-21 REVISION J/25/98 l

l Primary Containment Air Lock B 3.6.1.2 >

BASES

~

SURVEILLANCE SR 3.6.1.2.2 (continued)

REQUIREMENTS 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 inadvertently occur. With a frequency of once per :

24 months, the interlock will more than-likely only' be '

challenged during periods when it is not required to be operable,'since testing is only performed during a plant shutdown. During normal ingress and egress, one door is kept fully closed while opening the other door.

REFERENCES 1. FSAR, Section 5.2.3.4.5.

i

2. FSAR, Section 5.2.

3. Primary Containment Leakage Rate Testing Program. ;

l

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

Specification Improvements," July 23, 1993. '

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5 HATCH UNIT 1 B 3.6-13 REVISION 3/25/98

i PCIVs ;

B 3.6.1.3 t

BASES >

SURVEILLANCE SR 3.6.1.3.1 (continued) l REQUIREMENTS 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.

l SR 3.6.1.3.2 This SR verifies that each primary containment isolation i manual valve and blind flange that is located outside primary containment and is required to be closed during accident conditions and is not locked, sealed, or otherwise secured is closed. The SR helps to ensure that post accident leakage of-radioactive fluids or gases outside the !

l primary containment boundary is within design limits. i L This SR does not require any testing or valve manipulation. !

l Rather, it involves verification that those isolation !

devices outside primary containment, and capable of being i mispositioned, are in the correct position. Since ~

, verification of valve position for isolation devices outside ;

primary containment is relatively easy, the 31 day Frequency i

was chosen to provide added assurance that the isolation devices are in the correct positions.

I j 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 }

l proper position, is low. A second Note has been included to L clarify that PCIVs that are open under administrative l 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 i 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 l

}. (continued) f HATCH UNIT 1 B 3.6-24 REVISION 3/25/98

, l 2

, _, _. _- _ _. - _ . . . ..- I

Reactor Core SLs B 2.1.1 BASES.

APPLICABLE 2.1.1.3 Reactor Vessel Water Lt.y31 SAFETY ANALYSES (continued) During MODES 1 and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect '

of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the ever' that the water level becomes < 2/3 of the core height. The reactor t vessel water level SL has been established at the top of the active irradiated fuel to provide a reference point and to also provide adequate margin for effective action.

1 SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT 2.2.1 and 2.2.2 VIOLATIONS Exceeding an SL may cause fuel damage and create a potential for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

(continued) l HATCH UNIT 2 B 2.0-4 REVISION 3/25/98

Reactor Core SLs B 2.1.1

-BASES-(continued)

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

4

2. NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuels," (revision specified in the COLR). I

3. 10 CFR 50.72. I

4. 10 CFR 100.

5. 10 CFR 50.73.

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l HATCH UNIT 2 B 2.0-5 REVISION 3/25/9B t

Centrol R:d OPERABILITY B 3.1.3 BASES ACTIONS A.I. A.2. and A.3 (continued) to perform the Required Action in an orderly manner. The control rod must be isolated from both scram and normal insert and withdraw pressure. Isolating the control rod 1 from scram and normal insert and withdraw pressure prevents damage to the CRDM. The control rod should be isolated from scram and normal insert and withdraw pressure, while maintaining cooling water to the CRD.

Monitoring of the insertion capability of each withdrawn control rod must also be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of discovery of the stuck withdrawn control rod, concurrent with thermal power being greater than the low power setpoint (LPSP) of the RWM. SR 3.1.3.2 and SR 3.1.3.3 perform periodic tests of the control rod insertion capability of withdrawn control rods. Testing each withdrawn control rod ensures that a generic problem does not exist. The allowed Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests. Required Action A.2 is modified by a Note, which states that the requirement is not applicable when THERMAL POWER is less than or equal to the actual low power setpoint (LPSP) of the RWM since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.C) and the RWM (LCO 3.3.2.1).

To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Should a DBA or transient require a shutdown, to preserve the single failure criterion, an additional control rod would have to be assumed to fail to insert when required. Therefore, the original SDM demonstration may not be valid. The SDM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn.

The allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to verify SDM is adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an additir.;nal control rod adjacent to the stuck control rod (continued)

HATCH UNIT 1 B 3.1-16 REVISION 3/25/98 L -

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS A.I. A.2. and A.3 (continued)

(continued) also fails to insert during a required scram. .Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach and maintain MODE 3 conditions (Ref. 5).

IL1 I With two or more withdrawn control rods stuck, the stuck control . rods must be isolated from scram pressure within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per Action A.1 and the plant brought to MODE 3 l within.12. hours. The control rods must be isolated from both scram and normal insert and withdraw pressure.

Isolating the control rod from. scram and normal insert and withdraw pressure prevent damage to the CRDM. The control rod should be isolated from scram and normal insert and withdraw pressure, while maintaining cooling water to the CRD. The allowed Completion Time is acceptable, considering the low probability of a CRDA occurring during this interval. The occurrence of more than one control rod stuck at a withdrawn position increases the probability that the reactor cannot be shut down if required. Insertion of all insertable control rods eliminates the possibility of an additional failure of a control rod to insert. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

C.1 and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and disarmed (electrically or hydraulically) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected.

The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. The control rods can be electrically disarmed by disconnecting power from all four directional control valve solenoids. Required Action C.1 is modified by a Note, which allows the RWM to be bypassed if required to allow insertion of the inoperable control rods (continued) i HATCH UNIT 2 B 3.1-17 REVISION 3/25/98 j l

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- ._ -~ - _ _4

r-Control Rod OPERABILITY B 3.1.3 BASES ACTIONS C.1 and C.2 (continued) and continued operation. LCO 3.3.2.1 provides additional requirements when the RWM is bypassed to ensure compliance i with the CRDA analysis.

l The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems.

l D.1 and D.2 Out of sequence control rods may increase the potential reactivity worth of a dropped control rod during a CRDA. At j s 10% RTP, the generic licensing basis banked position i withdrawal sequence (BPWS) analysis (Ref. 5) assumes '

inserted control rods not in compliance with BPWS to be separated by at least two OPERABLE control rods in all directions, including the diagonal. Plant specific BPWS analysis may justify relaxed requirements en inoperable control rod separability. Therefore if two or more inoperable control rods are not in compliance with BPWS (and not separated by at least two OPERABLE control rods, unless the plant specific analysis relaxes this requirement),

action must be taken to restore compliance with BPWS or restore the control rod (s) to OPERABLE status. Condition D is modified by a Note indicating that the Condition is not applicable when > 10% 'tTP, since the BPWS is not required to be followed under these conditions, as described in the Bases for LC0 3.1.6. The allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is acceptable, con idrring the low probability of a CRDA occurring.

L.1 1 If any Required Action and associated Completion Time of l Condition A, C, or D are not met, or there are nine or more i inoperable control rods, 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 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . j This ensures all insertable control rods are inserted and l places the reactor in a condition that does not require the (continued)

HATCH UNIT 2 B 3.1-18 REVISION 3/25/98 I R

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS E.J (continued) active function (i.e., scram) of the control rods. The number of control rods permitted to be inoperable when operating above 10% RTP (e.g., no CRDA considerations) could be more than the value specified, but the occurrence of a large number of inoperable control rods could be indicative of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, baced on operating experience, to reach MODE 3 from fuli Jower in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.3.1 REQUIREMENTS The position of each .ontrol rod must be determined to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of this SR is based on operating experience related to expected changes in control rod position and the availability of control rod position indications in the control room.

l l SR 3.1.3.2 and SR 3.1.3.3 Control rod insertion capability ls demonstrated by l inserting each partially or fully w;thdrawn control rod at

! iest one notch and observing that the control rod moves.

The control rod may then be returned to its original position. This ensures the control red is not stuck and is free to insert on a scram signal. These Surve111ances are not required when THERMAL POWER 4 less than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the Banked Position Withdrawal Sequence (BPWS) (LCO 3.1.6) and the RWM l

(LCO 3.3.2.1). The 7 day Frequency of SR 3.1.3.2 is based on oprating experience related to the changes in CRD prformance and the ease of performing notch testing for (continued)

HATCH UNIT 2 B 3.1-19 REVISION 3/25/98 l l

I Centrol Rod OPERABILITY B 3.1.3 BASES-l SURVEILLANCE' SR 3.l.3.2 and SR 3.1.3.3 (continued) l REQUIREMENTS l fully withdrawn control rods. Partially withdrawn control L~ rods are tested'at a 31 day Frequency, based on the .

potsr.tial power reduction required to allow the control rod movament and considering the large testing sample of SR 3.1.3.2. Furthermore, the 31 day Frequency takes into account operating experience related to changes in CR0 performance.. At any time, if a control rod.is immovable, a i determination of that control rod's trippability (capable of l insertion by scram, i.e., OPERABILITY) must be made and:

appropriate action taken.

These SRs are each modified by a Note that allows 7 'ays d and-31 days, respectively, after withdrawal of the control rod -

and THERMAL POWER is greater than the LPSP to perform the Surveillance. This acknowledges that the control rod must L first be withdrawn and THERMAL POWER must be greater than

. the LPSP before performance of the Surveillance, and therefore avoids potential conflicts with SR 3.0.3 and SR 3.0.4.

i SR 3.1.3.4 Verifying that the scram time for each control rod to notch position 06 is 5 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby' completing its shutdown function.

This SR is performed in conjunction with the control rod ,

scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, l and SR 3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1, " Reactor Protection System (RPS) i Instrumentation," and the functional testing of SDV vent and '

drain valves in LCO 3.1.8, " Scram Discharge Volume (SDV)

Vent and Drain Valves," overlap this Surveillance to provide l complete testing of the assumed safety function. The associated Frequencies are acceptable, considering the more I frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.

, i l (continued)

HATCH UNIT 2 B 3.1-20 REVISION 3/25/98 I a

e=

Control R:d OPERABILITY B 3.1.3 l

BASES SURVEILLANCE SR 3.1.3.5 I REQUIREMENTS l (continued) Coupling verification is performed to ensure the control rod )

is connected to the CRDM and will perform its. intended 1 function when necessary. The Surveillance requires verifying a control. rod does not go to the withdrawn overtravel. position. The overtravel position feature i i

provides a positive check on the coupling integrity since only an uncoupled CRD can reach the overtravel position.

The verification is required to be performed any time a !'

control rod is withdrawn to the full-out position (notch position 48) or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could !

affect coupling. This includes control rods inserted one notch and then returned to the fu1 N ui position during the performance of SR 3.1.3.2. This Freque1cy is acceptable, l considering the low probability that a iontrol rod will f become uncoupled when it is not being r.oved and operating i experience related to uncoupling events.

l REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 27, GDC 28, I and GDC 29. i

2. FSAR, Section 4.2.3.2. l

3. FSAR, Supplement 5A.4.3.

4. FSAR, Section 15.1.

5. NED0-21231, " Banked Position Withdrawal Sequence,"

Section 7.2, January 1977.

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

l HATCH UNIT 2 B 3.1-21 REVISION 3/25/S8 l

h

.Pr1Cary Containment Air Lock

.p B 3.6.1.2

! i BASES-

)

SURVEILLANCE SR 3.6.1.2.2 (continued)

REQUIREMENTS 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 inadvertently occur. With a frequency of once per .

24 months, the interlock will more than likely only. be !

~c hallenged during periods when it is not required to be OPERABLE, since testing is only. performed during a plant shutdown. During normal ingress and egress, one door is kept fully closed while opening the other door, i- .

)

REFERENCES 1. -- FSAR, Section 3.8.2.8.2.2.

2. FSAR, Section 6.2. I i 3. Primary Containment Leakage Rate Testing Program.

1

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

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I HATCH UNIT 2 B 3.6-13 REVISION 3/25/98

y PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.1 (continued)

REQUIREMENTS 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.

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 and is not locked, sealed, or otherwise secured 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. .

, This SR does not require any testing or valve manipulation.

L 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 1 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 l controls are not required to meet the SR during the time that the PCIVs are open.

1 SR 3.6.1.3.3 This SR verifies that each primary containment manual l isolation valve and blind flange that is located.inside i' primary conta;4 ment and is required to be closed during accident conditions is closed. The SR helps to ensure that l post accident leakage of radioactive fluids or gases outside (continued)

HATCH UNIT 2 B 3.6-24 REVISION 3/25/98

-y Enclosure 5 i

Edwin I. Hatch Nuclear Plant !

Request to Revise Technical Specifications to Implement Previously Approved Generic Changes ;

Hand-Marked Panes l

I l

i i

1 l

l I

i l

l 1

HL-5591 i

SLs 2.0 l l

'2.0 SAFETY LIMITS (SLs) '

l V i l

2.1 SLs 2.1.1 Reactor Core SLs t I

'2.1.1.1 With the reactor. steam dome pressure < 785 psig or core * .

- flow < 10% rated core flow. - ,J~

a-

. THERMAL POWER shall be s 25% RTP. -

~ 'j:

2.1.1.2 With the reactor steam dome pressure a 785 psig and core -

l flow a 10% rated core flow: _ _

't _

MCPR shall be 't 1.10 for two recirculation loop operation or a 1.12 for single recirculation loop operation.*

2.1.1.3 Reactor' vessel water level shall be greater than the top.

of active irradiated fuel. -

2.1.2 Reactor Coolant System (RCS) Pressure SL -

Reactor steam dome pressure shall be s 1325 psig.

The specified limits are for Cycle 18 only. l 2.2 SL Violations- w 'dki s 2. 6 o With any SL violation, the following actions shall be complete I in I ho otif e y n,accetda ]nc I

2.2 Within 2 ours:

2.2.41 Restore compliance with all SLs; and V \

. 2.A,2 Insert all insertable control rods. 1

-~

2. . Within ours, notif pla na e cor ejtxeentivp' i the offsite /

i resp ible for ov p1 nuc1 safety, l

( r. ew committ .

l (continued) 1 i HATCH UNIT 1 2.0-1 Amendment No. 209 l l

. . ~. . - .. . - - - . . .

i SLs i 2.0 1

2.0 SLs s s

\

2.2 SL Violatio (conti ed) j 2.2.4 thin 30 ays, a Lie see Even .eport R) shall prepared j pursuan to 10 CFR .73. Th ER sha be submit d to the NRC, j

the of site revie committe , the pl. manager, nd the corporate <

exec ive respo ible for verall ant nuclear safety. j 2.2.5 peration f the uni shall n be resumed ntil authori by the '

NRC. ;

1

/ / / ( . / '

l l

l i

fL .~

4' t m i.

i HATCH UNIT 1 2.0-2 Amendment No. 195

q LCO Applicability 3.0 i

! 3'.0 .1.C0 APPLICABILITY !

? -

i

! LCO 3.0.4 to comply with ACTIONS or that are part of a shutdown of the (continued) unit. !

Exceptions t'o this Specification are stated in the !

i individual Specifications. These exceptions allow entry i into MODES or other specified conditions in the i 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 3.0.4 is only applicable for entry into a MODE or other

! specified condition in the Applicability in MODES 1, 2, and 3. ;

l 2

1- ,

j LCO 3.0.5 Equipment removed from service or declared inoperable to

{

- comply with ACTIONS may be returned to service under .

j administrative control solely to perform testing required to demonstrate its OPERABILITY, the OPERABILITY of other -

i i equipment. This is an exception to LCO 3.0.2 for the system j returned to service under administrative control to perform the required testing.

l' 1-LCO 3.0.6 When a supported system LCO is not met solely due .to a 1

support system LC0 not being met, the Conditions and-

Required Actions associated with this supported system are j not required-to be entered. Only the support system LCO j ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, ;

j- :dditL;.el evaluationA, d li:it:ti:n ny Qre utred in - Q accordance with Specification 5.5.10, " Safety Funcu an Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LC0 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 l' 3.0-2 Amendment No. 195

_me y .+i+, w u---- -w-- m -

es--*

. . _ . . . . . _ _ _ . _ . . . _ _ . . _ . _ . .. _ . ~ . . ____ - . .

Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 N--

N N

e70w power setpoint (LPSP) of the RWM.

N Perform SR 3.1.3.2 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 7ve,w Jacs ,,y and SR 3.1.3.3 for j g , ,; ,,, , , , , , ,

each withdrawn OPERABLE control rod. e;14 &-- t/ p*s

y u t.-

A.3 Perform SR 3.1.1.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B. Two or more withdrawn B.1 arm t t do control rods stuck. CRD. /ed s Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (continued).

HATCH UNIT 1- 3.1-8 Amendment No. 195 l

I i

I RPS Instrumentation 3.3.1.1 .<

2 Table 3.3.1.1 1 (pege 3 of 3)

Reactor Protection system ;nstrumentation ..

APPLICA8LE CONDITIONS MODES OR REQUIP O eEFERENCED !

"j OTHER CHANNELS FROM l SPECIFIED PER TRIP REQUIRED SURVE!LLANCE ALLOWABLE I FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMILITs VALUE

7. scram Discharge Volume Water Levet - High !

, l

s. Resistance 1,2 2 G sa 3.3.1.1.9 s 71 galtons i

i Temperature sa 3.3.1.1.13 Detector {

sa 3.3.1.1.15 l 5(a) 2 N st 3.3.1.1.9 s 71 estlons 34 3.3.1.1.13 sa 3.3.1.1.15

b. Ftost switch 1,2 2 G st 3.3.1.1.13 5 71 settons sa 3.3.1.1.15 5I *) 2 N sa 3.3.1.1.13 s 71 gattons ;

sa 3.3.1.1.15 1

8. Turbine stop 2 30% RTP .4 E sa 3.3.1.1.9 s 10% closed Velve - Ct osure SR 3.3.1.1.11 sa 3.3.1.1.13 j l

SR 3.3.1.1.15

, 9. Turbine'ControlValve a 30% RTP 2 E SR 3.3.1.1.9 2 600 psig i r

Fast closure, Trip 011 st 3.3.1.1.11 '

Pressure - Low SR 3.3.1.1.13 sa 3.3.1.1.15 i

st 3.3.1.1.16 .

l 10. Reactor Mode switch- 1,2 SR 3.3.1.1.12 shutdoom Position 1 G SR 3.3.1.1.15 NA

( l 5(*) 1 N sa 3.3.1.1.12 NA SR 3.3.1.1.15 l

11. Manuel Scram 1,2 1 0 sa 3.3.1.1.5 W NA st 3.3.1.1.15 5(a) 1 h sa 3.3.1.1.5 NA l st 3.3.1.1.15 l

(a) With any control rod withdrawn from a core cell conte.ning one or more fuel assemblies.

(e} In as kcs 4. =

cesisi g 9,_ *W a A- Jiva- r.t a y s k

l i

s HATCH UNIT 1 3.3 8 Amendment No. 195

=

. -. - - - . - ~ - -

Primary Cantt hment Air ick

3. A. t.2 SURVEILLANCE REQUIREMENTS

=

SURVEILLANCE FREQUENCY l

l SR 3.6.1.2.1 ------------------NOTES------------------

1. An inoperable air lock door does not invalidate the previous successful i performance of the overall air lock i

leakage test. !

i

2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.1. {

Perform required primary containment air In accordance lock leakage rate testing in accordance with the with the Primary Containment Leakage Rate Primary Testing Program. Containment Leakage Rate Testing Program

/ /

.SR 3.6.1.2.2 T --------- -- -

l On1 req ed a be erfo 'u e ry i o exi thr gh t pri yc ai nt * ;

ir 1 ck en t pri y to ain nt i J de- ert .

l Verify only one door in the primary la' h ys containment air lock can be opened at a time. p44 d

4 4

HATCH UNIT 1 3.6-7 Amendment No. 200

(

f

. _ ._ ... . . __ ._ . . . _ _ .._._ _ _ _ _ _ _ _ _ . . . . . _ . ~ . __ _ _ _

PCIVs .

3.6.1.3 l l

! SURVEILLANCE REOUIREMENTS SURVEILLANCE- FREQUEN':Y l

SR 3.6.1.3.1 ------------------NOTE-------------------

Not required to be met when the 18 inch primary containment purge valves are open I for inerting, de-inerting, pressure ,

control, ALARA, or air quality l considerations for personnel entry, or j Surveillances that require the valves to j l

i

'_be open.

l l . :

Verify each 18 inch primary containment 31 days i l

purge valve is closed.

i

?^ l SR 3.6.1.3.2 ------------------NOTE-------------------

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

2. Not required to be met for PCIVs that l are open under administrative control s.

1 Verify each primary containment isolation 31 days manual valve and blind flange that is located outside primary containment and is required to be closed during ac ident l conditions is closed. l 1

l 1

e l

~/

E (continued)

[a na ut l u ks S , .) c.tle D , o y.

t i

G ( F 4.T 34- [ C.CJ A A

l l

i i

I HATCH UNIT 1 3.6-12 Amendment No. 195 i

g.

/

PCIVs 3.6.1.3 I SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE Edh0ENCY

~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 N0DE 2 located inside primary containment and is or 3 from required to be closed during acci nt MODE 4 if conditions is closed, primary containment was de-inerted a n D- **I i 'l*4 ,.5< O 4 6 '~ while in MCDE 4, if not performed o%c,6 3c J ccs u 4 within the previous 92 days

.SR 3.6.1.3.4 Verify continuity of the traversing 31 days incore probe (TIP) shear isolation valve ,

explosive charge. !

SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated and each automatic PCIV, except with the for MSIVs, is within limits. Inservice Testing Program (continued) l i

HATCH UNIT 1 3.6-13 Amendment No. 195 1

. - . . .- .. . . - . . - - . _ - - . ~ - . . - . - . - - .. --

DC Sources - Operating i 3.8.4 r- . .

SURVEILLANCE REQUIREMENTS !

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

SR 3.8.4.1 through SR 3.8.4.8 are applicable only to the Unit 1 DC sources. ;

SR 3.8.4.9 is applicable only. to the Unit 2 DC sources. '

______________________________________________________________________________ t SURVEILLANCE FREQUENCY :

i 4

SR 3.8.4.1 Verify battery terminal voltage is a 125 V 7 days i on float charge.

t SR 3.8.4.2 Verify no visible corrosion at battery 92 days j terminals and connectors.

I

{

i

- 9.8 e

Verify battery connection resistance is l within limits.

1 t SR 3.8.4.3 Verify battery cells, cell plates, and 18 months

! racks show no visual indication of physical l

damage or abnormal deterioration.

NNg (44at < . . t .1 d e < an h tie-, - ..t., ,. b

... _ w SR 3.8.4.4 Remove visible corrosion, and verify 18 months battery cell to cell and terminal i

connections are coated with anti-corrosion I

material. I SR 3.8.4.5- Verify battery connection resistance is 18 months within limits.

(continued) i l

l

! I i !

c ;

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I !

HATCH UNIT'l 3.8-30 Amendment No. 195

. i

i Programs and Manuals 5.5 :

e 5.0 ADMINISTRATIVE CONTROLS 5.5- Programs and Manuals The following. programs and manuals shall be established, implemented, and maintained. '

5.5.1 offsite Dose Calculation Manual (ODCM) !

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation I of gaseous and liquid effluent monitoring' alarm and trip '

setpoints, and in the conduct of the radiological !

environmental monitoring program; and

b. The ODCM shall also contain the radioactive effluent !'

controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release reports required '

by Specification 5.6.2 and Specification 5.6.3, respectively. .

Licensee initiated changes to the ODCH: t I

a. Shall be documented and records of reviews performed shall ;

be retained. This documentation shall contain.

1. Sufficient information to support the change (s) and appropriate analyses or evaluations justifying the i change (s),and

2. A determination that the change (s) maintain the levels of radioactive effluent control required by 10 CFR 20.106, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and does not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b. Shall become effective after re'et:r ::d ::::;;tein.= Li a=

l r.;it: ce.ie ;;xitte: ::d tha- approval of the plant manager; and l

(continued) ,

1 f

' i HATCH UNIT 1 5.0-7 Amendment No. 195

SLs 2.0 2.0 SAFETY LIMITS (SLs) '

-2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam done pressure < 785 psig or core I

flow < 10% rated core flow:

THERMAL POWER shall be s 25% RTP.

2.1.1.2 With the reactor steam done pressure a 785 psig and core flow = 10% rated core flow:

MCPR shall be = 1.12 for two recirculation loop operation !

or a 1.14 for single recirculation loop operation.

j 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel, ,

t 2.1.2 Reactor Coolant System (RCS) Pressure SL I

Reactor steam done pressure shall be s 1325 psig.

l 2.2 SL Violations At in t kwa j

With any SL violation, the following actions shall be completed:f Within u not OperAt4 ens K y oanse ]

.2.2 Within hours: /

V

'2. 2 X.1 Restore compliance with all SLs; and V

L2.24.2 Insert all insertable control rods.

?2 . Within hours, no y the plan nager, corporat exec

~

l ve r nsible fo erall pl uclear ety, and off e eview co tee.

(continued) i HATCH UNIT 2 2.0-1 Amendment No. 149

r SLs 2.0 2.0 SLs

, s / ,

, 2.2 Violation (contin 2.2.4 hin 30 s, a Licensee ent Report R) shall be p ared ursuant o 10 CFR 50.73 The LER sha be submitted the NRC, the of te review c ttee, the pl t manager, an the corporate exec ive responsibl for overall ant nuclear ety.

2.2.5 peration of t unit shall t be resumed til authori by e NRC.

,/

/

/ / /

.1

)

l 1

4 i

I I i )

l l

i HATCH UNIT 2 2.0-2 Amendment No. 135 l j

.- .. . . . . - - - -~ -- . - . . - - . - . . . - . . - - _ - - . _ . _ . . . -

LCO Applicability l 3.0 3.0 LCO APPLICABILITY I LCO. 3.0.4 .

to comply with ACTIONS or that are part of a shutdown of the (continued) -unit.

t Exceptions to this Specification are state'd in the !

individual Specifications. These exceptions allow entry into MODES or other specified conditions in the -!

Applicability when the associated ACTIONS to be entered l

' allow unit operation in the MODE or other specified !

condition in the Applicability only for a limited period of time. l

/!

LCO 3.0.4 is only applicable for entry into a MODE or other I specified condition in the Applicability in MODES 1, 2, l and 3. - '

l l

LCO 3.0.5 Equipment. removed from service or declared inoperable to )

comply with ACTIONS may be returned to service under J

administrative control solely to perform testing required to 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 required testing.

LCO 3.0.6 When a supported system LCO is not met solely due to a 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 LCO 3.0.2 for the supported system. In this event, -d9 add 4+ione4-evaluatiorg er,d ifettath r; Arequired in accordance with Specification 5.5.10, " Safety FuncIlon j3 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) i HATCH UNIT 2 .. 3.0-2 Amendment No. 135 .

'~. O . .

w ~ -- , , , , . _ n .

__ , a .;

[

Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 m = N^! h Not epl-;cabi: J en less t;;.r. Or ; wal to l

lthe iow(LPSP setpoint powe m) of}

Lthe RWM. _ ,

A __ ___ __ _

Perform SR 3.1.3.2 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months fv== Jome1 and SR 3.1.3.3 for each withdrawn * ' * * * ' U ' ^'"""#

OPERABLE control rod.(

er * %<< ma nu.-

\ > n e &"

A.3 Perform SR 3.1.1.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months trol r s c C i

[ Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (continued) l l

J l

)

l 1

HATCH UNIT 2 3.1-8 Amendment No. 135

Primary CCntainment Air Lock 3.6.1.2 i

SURVEILLANCE REQUIRENENTS SURVEILLANCE FREQUENCY SR 3.6.1.2.1 --------------- NOT E S----------------- 1

, 1. An inoperable air lock door does not !

invalidate the previous successful l performance of the overall air lock leakage test.

' 2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.1 l

Perform required primary containment air In accordance i lock leakage rate testing in accordance with the <

] with the Primary Containment Leakage Rate Primary

Testing Program. Containment !

Leakage Rate !

! Testing Program i l,

\

SR 3.6.1.2.2 --- '

- /----- -NOT --- d - -------

fn[yr uir ob per rue ponjnfry i i

/ or it oug e imar cont (neent p lo who the imar con noen s 1

,.cp_.rted [

3 l Verify only one door in the primary 164 days-containment air lock can be opened at a time.

HATCH UNIT 2 3.6-7 Amendment No. 141 1

, . _ _ . _ _ _ _ _ _ . ~ - . __..._ _._.._ _ -. _.._ _ -. _ . - _ _ - _ . - - - . - . _ _ .

V PCIVs 3.6.1.3 l

l SURVEILLANCE REQUIREMENTS i

' SURVEILLANCE FREQUENCY

{

\

SR 3.6.1.3.1 ------------------NOTE-------------------

! Not required to be met when the 18 inch primary containment' purge valves are open

! for inerting, de-inerting, pressure control, ALARA, or air quality l considerations for personnel entry, or Surveillances that require the valves to l be open.

l Verify each 18 inch primary containment 31 days purge valve is closed.

SR 3.6.1.3.2 --'---------------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 edmini..L. tive controls.

i _________________________________________

Verify each primary containment isolation 31 days manual valve and blind flange that is located outside primary containment and isrequiredtobeclosedduringaccident conditions is closed.

l (continued) l l l

a wa. nJ t cLR, Sulif or o % rJss .r w , J l

f l

l HATCH UNIT 2 3.6-12 Amendment No. 135 L

f i.

PCIVs 3.6.1.3 4

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY r

SR 3.6.1.3.3 -----------------NOT E S-----------------

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.

s Verify each primary containment m;"Ja1 Prior to.

isolation valve and blind flange that is entering MODE 2 located inside primary containeen and is or 3 from required to be closed during acci nt MODE 4 if conditions is closed. primary containment was

! de-inerted while in 2"A "'t io &<.A '

r c. a l <1 MODE 4, if not performed j or eW< wha .r e u ~ 4 " the ylou

!

92 days

! s s

)'

SR 3.6.1.3.4 Verify continuity of the traversing 31 days incore probe (TIP) shear isolation valve ,

1

. explosive charge.

j.

'SR 3.6.1.3.5 Verify the isolation time of each power In accordance operated and each automatic PCIV, except with the for MSIVs, is-within limits. Inservice Testing Program ;

l (continued)

HATCH UNIT 2 3.6-13 Amendment No. 135 I

- , _ , ,. .e* -. - *

. - . _ . . . . . . _ - _ _ _ _ _ _ _ _ _ . . _ - ~ = - __ --_- - _ . . . _ -

1 DC Sources - Operating 3.8.4 :

l SURVEILLANCE REQUIREMENTS

- ----------------------NOTE------------------------------------- !

SR 3.8.4.1 through SR 3.8.4.8 are applicable only to the Unit 2 DC sources. i

'_SR 3.8.4.9 is applicable only to the Unit 1 DC sources.

i SURVEILLANCE FREQUENCY SR. 3.8.4.1 Verify battery terminal voltage is 2: 125 V 7 days on float charge.

i SR 3.8.4.2 Verify no visible corrosion at battery 92 days I l terminals and connectors. l i

D.8 l l

Verify battery connection resistance is ;

!. within limits. ;

l l SR 3.8.4.3 Verify battery cells, cell plates, and 18 months racks show no visutt iuJication of physical c y ioration. .

Q& -c ,w .c u a s .0 c b s H c ~, s ,.< & m m c e7

-SR 3.8.4.4 Remove visible corrosion, and verify 18 months battery cell to cell and terminal connections are coated with anti-corrosion material .

SR 3.8.4.5 Verify battery connection resistance is 18 months within limits.

(continued) 3 4

J HATCH UNIT 2 3.8-30 Amendment No. 135 f'

Programs and Manuals 5.5 '

J

. 5.0 ' ADMINISTRATIVE CONTROLS i

2 5.5 Programs and Manuals l t

The following programs and manuals shall be established, implemented, and 1 maintained. l 4

t 5.5.1 Offsite Dose Calculation Manual (00CM) i j a. The ODCM shall contain the methodology and parameters used I

i

! in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and 1

b.

! The ODCM shall also contain the radioactive effluent :

controls _ and radiological environmental monitoring

- activities, and descriptions of the information that should be included in the Annual Radiological Environmental i

4 Operating and Radioactive Effluent Release reports required !

by Specification 5.6.2 and Specification 5.6.3, l

respectively. '

Licensee initiated changes to the ODCM: l a.- Shall be documented and records of reviews performed shall 4 4 be retained. This documentation shall contain:

a

1. Sufficient information to support the change (s) and appropriate analyses or evaluations justifying the :

i change (s),and 4

2. A determination that the change (s) maintain the levels of 'radioactive effluent control requiref by i 10 CFR 20.106, 40 CFR 190,10 CFR 50.3/4 , and j

10 CFR 50, Appendix I, and does not aftersely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b. Shall become effective after +evi;a : d ac::pt:a;; by the ea;ite r;;t;; cer. ittes ;ad the approval of the plant manager; and i

t 3

(continued)

HATCH UNIT 2 5.0-7 Amendment No. 135 1

Reacter Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.3 Reactor Vessel Water Level SAFETY ANALYSES (continued) During MODES 1 and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel (Wing this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a reference point and to also provide adequate margin for effective action.

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT ashi VIOLATIONS o if-ry SL isviah+-d; ne MDF gg=2tinne ranter BEG-t-be not'fif, (o... WithTri 1 livur,4n-accorder.ca w+th-10 CFR 50.72 (continued)

HATCH UNIT 1 B 2.0-4 REVISION 0

- - . - -. - . - - - - - - .- - - . - -. - - ~_ - .-. - -

l l

l i

Reactor Core SLs l B 2.1.1 1 i

BASES '

7 . t.. t uJQ i

{

SAFETY LIMIT dI 2 2.2 VIOLATIONS '

(continued) Exceeding an SL may cause fuel damage and create a potential !

for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). Therefore, it is required l to insert all insertable control rods and restore compliance )

with the SLs within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time '

ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring j during this period is minimal.

w

-if any 5L is violeted, ii.e 3esi:r rits;- :-t Of th: ::: lear pl::t :nd the ut-?tity. =ad the S f:t, Re.iew ;oerd (SRS) shall be acti'!:d olihin ^4 h::r:. Tk- 2a hane p=-i:d pr::id:: ti;; fer pl.nie . v.ivi, _;d staf# te t:h: the immedi1ste action ana assess tne conaition of Riw v..it h:fere repe.tinf to Ine senior manag.. .i..

M tr ...j :L i: vi:1:t:d, : Lie::::: E:::: R:::rt :h.11 be orenared and suhaitted witki- 3a d y: t: t ie unc 4-i arterd::; aith 10 CFR 5v.73 (nef. 5). A cupy vi th: 7:; rt )

hell .ise be previd:d te the ..nie. .T.e.. ;;;;:t er + ke

' nuclear pl.nt end the-etility, and the SRB , ,

i li l

AstT

~. n vioi46 u, . . . . . . _. v. 6. .

L,...,. .n .. ,,. . . . . ..

. unii.

c_ -.ce unti1 autherized bj tii nRC. inis requi.... .i en: r:: th: "R0 th.t ell ...sei.. y .v,ews, :::ly ::, :nd !

Oti:C: :Te ee""'.eted 50f0f0 the unit b;;in: it re; tert te )

n;.e.e1 :;:r:tica. !

1 J

(continued) i HATCli UNIT 1 B 2.0-5 REVISION 0 l

y Control Rod OPERABILITY B 3.1.3 f

5.

BASES l

1

~

j ACTIONS' A.J. A.2. and A.3 (continued) j to perfr.3 the Required Action in an orderly manner. The j

control rod must be isolated from both scram and normal i insert and withdraw pressure. Isolating the control rod

]s +

! from scram and normal insert and withdraw pressure prevents e

damage to the CRDM. The control rod should be isolated from j scram and normal insert and withdraw pressure, while

{ ) ,

)p g ntaining cooling water to_the_CRD

!]$

( r k

y k Monitoring of the insertion capability of each w thdrawn control rod must also be performed within 24 hourf'.

3

! u g j SR 3.1.3.2 and SR 3.1.3.3 perform periodic tests of the

i. g j control rod insertion capability of withdrawn control rods.

] p Testing each withdrawn control rod ensures that a generic j j '; }

problem does not exist. The allowed Completion Time of

x % 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides a reasonable time to test the control 4-(

-c -

i rods, considering the potential for a need to reduce power to perform the tests. Required Action A.2 is modified by a i k Note, which states that the requirement is not applicable 1 1 a 0. when THERMAL POWER is less than or equal to the actual low j 4 power setpoint (LPSP) of the RWM since the notch insertions a \ Y may not be compatible with the reytirements of rod pattern L

control (LCO 3.1.6) and the RWM (Lt/) 3.3.2.1).

lI i D r

T o

f To allow continued operation with a withdrawn control rod ,

t stuck, an evaluation of adequate SDM is also required within j

! 4 4K 4 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Should a DBA or transient require a shutdown, to

k+ "

preserve the single failure criterion, an additional control

! g rod would have to be assumed to fail to insert when i o ) required. Therefore, the original SDM demonstration may not

, Z t

! be valid. The SDM must therefore be evaluated (by t A "

+

measurement or analysis) with the stuck control rod at its i ; D 1 stuck position and the highest worth OPERABLE control rod j

} ( assumed to be fully withdrawn.

! ? + The allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to verify SDM is

' N : 2 adequate, considering that with a single control rod stuck f f in a withdrawn position, the remaining OPERABLE control rods g 2 are capable of providing the required scram and shutdown j o l A reactivity. Failure to reach MODE 4 is only likely if an

<- e J / additional control rod adjacent to the stuck control rod

! V E also fails to insert during a required scram. Even with the j postulated additional single failure of an adjacer,t control

! rod to insert, sufficient reactivity control remains to ;

j reach and maintain MODE 3 conditions (Ref. 5). )

i

} (continued) j HATCH UNIT 1 B 3.1-16 REVISION 0 l l

Control R:d OPERABILITY l

! B 3.1.3 j 2

1 l BASES i

l l 1

ACTIONS B.1 and & e (continued) /u .4 c :rv o a ,,4, /

i With two more withdrawn control rods stuck, the stuck ,

control 's must be isolated from scram pressure within '

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months nd the plant brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4 -

The control rods must be isolated from both scram and norea'.

! insert and withdraw pressure. Isolating the control rod i from scram and normal insert and withdraw pressure prevents j damage to the CROM. The control rod should be isolated from 1' scram and normal insert and withdraw pressure, while i maintaining cooling water to the CRD. The allowed Completion Time is acceptable, considering the low probability of a CRDA occurring during this interval. The i occurrence of more than one control rod stuck at a withdrawn 5

position increases the probability that the reactor cannot be shut down if required. Insertion of all insertable control rods eliminates the possibility of an additional 3 failure of a control rod to insert. The allowed Completion j Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating i experience, to reach MODE 3 from full power conditions in an

orderly manner and without challenging plant systems, j i

j C.1 and C.2 i

With one or more control rods inoperable fer reasons other than being stuck in the withdrawn pesition, operation may continue, provided the control rods a? e fully inserted witMn 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and disarmed (electrics 11y or hydraulically) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Inserting a control roa viwures the shutdown and scram capabilities are not adversely affected.

The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. The control rods can be electrically disarmed by disconnecting power from all four directional control valve solenoids. Required Action C.1 is modified by a Note, which allows the RWM to be bypassed if required to allow insertion of the inoperable control rods and continued operation. LCO 3.3.2.1 provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.

The allowed Completion Times are reasonable, considering the mall number of allwed inoperable control rods, and provide l

(continued)

I HATCH UNIT 1 B 3.1-17 REVISION O l l

J

i Primary Containment Air Leck '

B 3.6.1.2 e

BASES '

SURVEILLANCE SR 3.6.1.2.2 (continued)

REQUIREMENTS 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 inadvertently 4...i.._._. occur.

15. 4..,. . . . . __2 _m -Ore te the p" rely .o :;;h::i:;l

___t..

r,..te.e er

, .. v . . .. . i .t _ .... 6..... ....i. . . _ . . . _ _ _ . . . . . . . -

0:ly th:ll::;;d h_r. th; pris.ry ;;;t:t. ::t sie leek deer t

+

.64.., .__.

u.. __.___2,

., _. _ .... 2.._ . . . n. ,._,_

4. ,a +

-- -- s. .._ r _--- - > . ._ _ _ -

t:rir.s ur exiting tne primary sun e 7 ::t ai- leck, k"+ is edP - 20t r:; fred ::r: fr:;:::tly th:2 I"' days -'-- ari==ry

/uJes,4,_

r raata!- ::t 1: de-irert:d. Th: 1*' d:y Fre-;2erry is h===d. l

er;in::rir.s jud-; i.t end is ;;;;id:r:d ide;"=*= 4" "teek-

-:f th;r edeir.iitietire controis such .3 tr.di::t'::: ef-int:rl;-k ;;;heni..i si tua, .eeil:bl: te eper:ti;;; -

________t r......-...

REFERENCES 1. FSAR, Section 5.2.3.4.5. I

2. FSAR, Section 5.2. i

3. Primary Containment Leakage Rate Testing Program. l

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

l l

8 l

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4 J

k i

l I

hatch' UNIT 1 B 3.6-13 REVISION 5 4

i

/p/ 3 FnT /

I t

With a frequency of once per 24 months, the interlock will mon: than likely only be ;.

challenged during periods when it is not required to be operable, since testing is only t done during a plant shutdown. During normal ingress and egress, one door is kept fully closed while opening the other door.

i f

l l

I

d PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.1 (continued)

! REQUIRENENTS J 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.

SR 3.6.1.3.2 y This SR verifies that nach primary containment isolation i

N

, manual valve and blind flange that is. located outside

primary containment anal is required to be closed during Y

accident conditions 5fs's closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside g the primary containment boundary is within design limits.

N

) ) This SR does not require any testing or valve manipulation.

j 3 Rather, it involves verification that those isolation N { devices outside primary containment, and capable of being mispositioned, are in the correct position. Since verification of valve position for isolation devices outside 4 4 primary containment is relatively easy, the 31 day Frequency

? was chosen to provide added assurance that the isolation devices are in the correct positions.

4 E Two Notes have been added to this SR. The first Note allows y

or valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing j L verification by administrative controls is considered D 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 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 accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside (continued)

HATCH UNIT 1 B 3.6-24 REVISION 1

t Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.3 Reactor Vessel Water level SAFETY ANALYSES ,

j (continued) During MODES I and 2, the reactor vessel water level is i required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a reference point and to also provide adequate margin for effective action.

SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs. SL 2.1.1.1 and .

SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel )

water level is greater than the top of the active irradiated ,

~

fuel in order to prevent elevated clad temperatures and resultant clad perforations.

t l

l APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES. l 1

l SAFETY LIMIT Z?CI !

VIOLATIONS If ry SL is Jicl:tt the "RC 0;cr: tion: Center =::t be na+ m ad "4%4n-1 heur, in :: r-d re: with 10 CFR 50.72 (R-f. ).

I (continued)

HATCH UNIT 2 B 2.0-4 REVISION 0 J

Reactor Core SLs B 2.1.1 BASES r~

i i

al . 7. 4 2d _

SAFETY LIM T 2.2.2 VIOLATIONS (continued)

Exceeding an SL may cause fuel damage and create a potential for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and ~ restore compliance with the SLs within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

If :n,and SL ii ".ielat^d, th= Sani;r :: ;;;. ..t Of- th :-le.r nlant +k.. .a. 41 4. o.,.

- '^ '

a nu' ..'..*.,u n -

n.-.>... .n..-

<e s.nos-., -

. ih a 11 he .'. '. '. 3. . 2 - i 6 n' i n .a ', . ' . -.- . _ . a' '

T..'.-. .. ....-- . r.--3.

. . .2 orovidar tiee fee plaat eper ter: cr.d st.if ^6 tsk: ths

=rar:p:d=+e 4:::di:t: actien .nd :sse:: th: ccr.ditic: Of th:

. unit-bsfere . - ar+ 4n; t: th; .an . .. ;;;;n:;:- :7 t.

mu.s.

(( :T.y CL-if yd-lated; a t ienngne [yent -Da sert ghg)) b; 9 .p..ed

_,,,__2.. n6a ivu' mitt d .with,ia ?^ d:y: t; t ::_. ___".. aC _,in-l

. . . <6m i.n. e.r.

.....o____

a

.n . , . gnus . a). n u vy, ..

.u.

.__ .____t

..r..

i

v. v.vvia au 1

e 1

_L. .r

. _ s. s 1. .v u -onc ,..niur manag .. .. __i

u. ,.L, 1

.u. '

.nuchar pl:.nt#he .;tiW:nd 11..-52 I i

l s.6.a WI b Yi^'?+=d, -

-astert Of th: :;;it shall nat cr-^ ara unt41 suthurizeu oy the ='C. Thi.-requi.;,.;nt

_^ asures-.the NRC that all nece:: ry--reyi = , 2n:1ycog, and uticas ara ca plJLtad_hefara the unit b:; ins it; re: tert to M Gycrab.ia!L -

I (continued)

HATCH UNIT 2 B 2.0-5 REVISION 0 t

CCntrol Rod OPERABILITY B 3.1.3

. BASES ACTIONS A.I. A.2. and A.3 (continued) to perfom the Required Action in an orderly manner. The control rod must be isolated from both scram and normal s insert and withdraw pressure. Isolating the control rod -

(/9* ^ *, from scram and nomal insert and withdraw pressure prevents damage to the CRDM. The control rod should be isolated from i scram and normal insert and withdraw pressure, while

~

, maintaining cooling water to the CRD.

h

? .:

Monitoring of the insertion capability of each withdrawn l, control rod must also be performed within 24 hourf

U SR 3.1.3.2 and SR 3.1.3.3 perform periodic tests of the kl control rod insertion capability of withdrawn control rods.

, Testing each withdrawn control rod ensures that a generic

/

j e / problem does not exist. The allowed Completion Time of o

% v 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides a reasonable time to test the control j rods, considering the potential for a need to reduce power D 3 ,' to perfom the tests. Required Action A.2 is modified by a Z .

Note, which states that the requirement is not applicable

/ s J when THERMAL POWER is less than or equal to the actual low i d l power setpoint (LPSP) of the RW since the notch insertions

may not be compatible with the requirements of rod pattern

{ )

/ control (LCO 3.1.6) and the RM (LC0 3.3.2.1).

[A

kI To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within <

{

'b R 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Should a DBA or transient require a shutdown, to '

l ,

I '

v preserve the single failure criterion, an additional control v 0 rod would have to be assumed to fall to insert when

{ s ) required. Therefore, the original SOM demonstration may not 4

3 [ be valid. The SOM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its o If h* stuck position and the highest worth OPERABLE control rod

' A- \ assumed to be fully withdrawn.

K '

D.

The allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to verify SOM is

} 4 adequate, considering that with a single control rod stuck Ns? p D in a withdrawn position, the remaining OPERABLE control rods 3 are capable of providing the required scram and shutdown 3

( reactivity. Failure to reach MODE 4 is only likely if an b [ additional control rod adjacent to the stuck control rod o s' also fails to insert during a required scram. Even with the o b postulated additional single failure of an adjacent control N rod to insert, sufficient reactivity control remains to i

reach and maintain MODE 3 conditions (Ref. 5).

(continued)

HATCH UNIT 2 B 3.1-16 REVISION 0 w ,- , - , - - - - -

e C:ntrol Rod OPERABILITY B 3.1.3 BASES inued) /4 A */

-With two more withdrawn control rods stuck, the stuck control s must be isolated from scram pressure within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and the plant brought to MODE.3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The control rods must be isolated from both scram and normal insert and withdraw pressure. Isolating the control rod from scram and normal insert and withdraw pressure prevents damage to the CROM. The control rod should be isolated from scram and normal insert and withdraw pressure, while maintaining cooling water to the CRD. The allowed Completion Time is accepteble, considering the low probability of a CLA occurring during this interval. The occurrence of more than one control rod stuck at a withdrawn position increases the probability that the reactor cannot be shut down if required. Insertion of all insertable control rods eliminates the possibility of an additional failure of a control rod to insert. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

C.1 and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and disarmed (electrically or hydraulically) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected.

The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. The control rods can be electrically disarmed by disconnecting power from all four directional control valve solenoids. Required Action C.1 is modified by a Note, which allows the RWM to be bypassed if required to allow insertion of the inoperable control rods and continued operation. LCO 3.3.2.1 provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.

The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide (continued)

HATCH UNIT 2 B 3.1-17 REVISION 0 I

Pritary Centainment Air Lock B 3.6.1.2 BASES SURVEILLANCE SR 3.6.1.2.2 (continued)

REQUIREMENTS air lock is being used for personnel transit in and out of I the containment. Periodic testing of this interlock i demonstrates that the interlock will function as designed l and that simultaneous inner and outer door opening will not inadvertently occur. 0;;; te the W ely scher.ic;l :: tere :f thi: ' .terlech, trd gi'fer thtt th: int:rl=k =;h=i= is 48[ [r,r.ly chell;;; d "r- the primry :=t:ir. .,.r.; air im.i dee,is c;: = d, th i

, ..+.% -itt ; th; giury ce,,teir._.r.; ;ir l=h, h;t ir

// 5 t.# aet r;;; ired =r.; fre;;.ertly i;,en,184-days 9 : ari e y

..2... c--....,.. a.. u .~..

,. . . s . s -- _ ~ . -

.> n. ,

a uw in., ... . . . .

cr. =; inn rin; jef--:-t s-d is ra rid r-d ede;eate i vi::

ef ;ther edrir.istr;tiv; nr.trel: :tch 5: indienti=; er 1.ter!ech :::h nire. t:ter, :=iletle te ogr:ti;= -

g ree..i.el.

REFERENCES 1. FSAR, Section 3.8.2.8.2.2.

2. FSAR, Section 6.2.

3. Primary Containment Leakage Rate Testing Program. l

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

HATCH UNIT 2 B 3.6 13 REVISION 7

. . _ _ .. . .._ .._._.._...._ _. _ _ .__ _ __._..._._.______ ___. _m.. .__m__>__....

/4 J F6T-E :

i 1

- With a frequency of once per 24 months, the interlock will more than likely only be challenged during periods when it is not required to be operable, since testing is only doce during a plant shutdown.. During normal ingress and egress, one door is kept fully closed while opening the other door.

l I

I i

l .

i f

I l

l I

f I

i 1

t 1

1 l

I 4 ,

i

, .. - - ..-.. . - - . - ., - . . - . - . - - - - . - . - - . . - . - . - . . - - - - . - .~_ -

PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.1 (continued)

REQUIREMENTS 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.

SR 3.6.1.3.2 Thi t each primary containment isolation

.g 5 nual valve and b1 flange that is located outside primary containment nd is required to be closed during

. i accident conditions s closed. The SR helps to ensure that v

j post accident leakage of radioactive fluids or gases outside s

9 the primary containment boundary is within design limits.

I j This SR does not require any testing or valve manipulation.

1 v 5 Rather, it involves verification that those isolation 4 devices outside primary containment, and capable of being

' \ j mispositioned, are in the correct position. Since

/ j verification of valve position for isolation devices outside 1

1 4 primary containment is relatively easy, the 31 day Frequency 4 .

i was chosen to provide added assurance that the isolation

)'* ,,t- devices are in the correct positions.

" '3 Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas to e be verified by use of administrative controls. Allowing

.i j verification by administrative controls is considered

'W s acceptable since access to these areas is typically

( 0 (' restricted during MODES 1, 2, and 3 for ALARA reasons.

Q' 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 that the PCIVs are open.

SR 3.6.1.3.3 - l I

This SR verifies that each primary containment manual I 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 (continued)

I HATCH UNIT 2 8 3.6-24 REVISION 1 l I

_ _ _ _ _ _ _ _ _ _ ,

ML20203C593

Text

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