ML20210J049
| ML20210J049 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/08/1997 |
| From: | SOUTHERN NUCLEAR OPERATING CO. |
| To: | |
| Shared Package | |
| ML19317C599 | List: |
| References | |
| NUDOCS 9708140316 | |
| Download: ML20210J049 (136) | |
Text
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Edwin I. IIntch Nuclear Plant j
Request for License Amendment Extended Power Uprate Operation L
Page Change Instructions t
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9700140316 970800 PDR ADOCK O$000321 P
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U Edwin I. Hatch Nuclear Plant
. Request for License Amendment Extended Power Uprate Operation 4
Pase Channe Instructions The proposed changes to the Plant Hatch Unit I and Unit 2 Facdity Operating Licenses and Technical Specifications are incorporated as follows:
UniLLOperating Lictnat s
has Inalrusilon 3
Replace i-
-Unit 1 Technisal Soecificatiorn Egge Instructlpa 1.1-5 Replace l (
3.3 2* --
Replace 3.3-5
- Replace 3.36*
Replace 3.38
. Replace 3.3-27 Replace r
3.3-28 Replace 3.3-29 Replace 3.4-25 "
Replace 3.4-26 "
Replace 3.4-27 "
Replace
[
. 5.0-16a Replace L' These pages reflect the revisions approved by the NRC in Amendment 205, Power Range Neutron Monitor Retrofit, which will be issued prior to the Unit 1 Fall 1997 Outage. -The
- corresponding Bases pages are also included.
t
. " The current P/T limit figures and the proposed P/T limit figures provided by letter dated J
April 29,1997, are included in the marked-up Technical Specifications pages The proposed
> figures will likely be approved and issued prior to approval and issuance of the extended power uprate proposed changes.
L.
IIL-5413i E4-1
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Egge Change Instructions
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Unit 2 Operating License Eagg Instruction 4
Replace 3_
Unit 2 Technical Specifiestions Eagg Instruction 1.1-5 Replace 3.32 Replace 3.3-5 Replace l
3.3-7 Replace 3.39 Replace 3.3-28 Replace l
3.3-29 Replace i'
3.3-30 Replace 3.4-25 Replace 3.4-26 Replace 3.4-27
- Replace 5.016a Replace i
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IIL-5413
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Unit !
Facility Operating License Proposed Change i
and 4
Technical Specifications Proposed Changes i
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1 the procedures and limitations set forth-in this license; and the Georgia Power Company, the Oglethorpe Power Corporation,:the Municipal Electric Authority of Georgia and the City of Dalton,.
Georgia.to possess but not operate-the facility in accordance with the procedures-and limitations set-forth in this license; (2) Southern Nuclear, pursuant to the-Act-and-10 CFR Part 70, to i
receive, possess and use at any time special nuclear material as j
reactor fuel, in accordance with the-limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Southern Nuclear, pursuant to the Act and 10 CFR Parts-30, 40 and 70, to receive, possess and use at any time any byproduct, source t
and special nuclear material as sealed neutron sources for reactor b
startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and'as fission detectors in amounts as required;
~ (4) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and n
70, to receive, possess and use in amounts as required any byproduct, source or special nuclear matcrial without restriction to chemical or physical form, for sample analysis or instrument L
calibration or associated with radioactive apparatus or components; i
(5) Southern Nuclear, pursuant t'., the-Act and 10 CFP. Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50-54 and 50-59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations,andordersoftheCommissignnoworhereafterineffect;and l
is subject to the additional conditions specified or incorporated below:
(1) tigimum Power Level 4
Southern Nuclear is authorized to operate toe i:
facility at steady state reactor core power levels not in excess.of 2763 megawatts thermal.
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3 The-original licensee authorized to possess, use and -operate the facility was Georgia Power Company (GPC).
Ccnsequently, certain-historical references to GPC-remain in the license conditions.
Amendment No.
x.
3.
= - -
Definitions 1.1
/O 1.1 Definitions (continued)
O PHYSICS TESTS.
PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.
These tests are:
a.
Described in Section 13.6, Startup and Power Test Program, of the FSAR; b.
Authorized under the provisions of 10 CFR 50.59; or c.
Otherwise approved by the Nuclear Regulatory Commission.
RATED THERMAL POWER RTP shall be a total reactor core heat transfer
- (RTP) rate to the reactor coolant of 2763 MWt.
l REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response ofme may be measured by p
means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that:
a.
The reactor is xenon free; b.
The moderator temperature is 68 F; and c.
All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
With control rods not capable of being fully inserted, the reactivity worth of these-control rods must be accounted for in the determination of SDM.
STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance (continued)
(]
V HATCH UNIT 1 1.1-5 EXTPWR - 7/24/97
RPS Instrumentation 4
3.3.1.1 c
ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION-TIME C. 'One or more Functions C.1 Rest' ore RPS trip 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with RPS trip capability.
- capability not-maintained.
D.
Required Action'and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, Table 3.3.1.1-1 for B, or C not met.
the channel, t
E.
As required by E.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Required Action D.1 to < 28% RTP.-
l and referenced in j
L
- Table 3.3.1.1-1.
- O F.
f As required by F.1 Be in MODE 2.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Required Action D.1 and referenced in
. Table 3.3.1.1-1.
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G.
As required by G.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- Required Action D 1 and referenced in Table 3.3.1.1-1. -
i H.--As required by
- 11. 1
-Initiate action to Immediately F
Required Action 0.1-fully insert all and referenced in insertable. control Tabl e 3.3.1.1-1.---
' rods in core cells-containing one or-more fuel assemblies.
.O HATCH UNIT-1 3.3-2
- AMEND.-205/EXTPWR - 7/24/97 j
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RPS Instrumentation 3.3.1.1
(T SURVEILLANCE REQUIREMENTS (continued)
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SURVEILLANCE FREQUENCY SR 3.3.1.1.11 Verify Turbine Stop Valve -- Closure and 184 days Turbine Control Valve fast Closure, Trip 011 Pressure - Low Functions are not bypassed when THERMAL POWER is a: 28% RTP.
l SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST.
18 months
'SR 3.3.1.1.13
..----------NOTES------------------
- 1. Neutron detectors are excluded.
- 2. For function 1. not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering It0DE 2.
N Perform CHANNEL CALIBRATION.
18 month:
(
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SR 3.3.1.1.14 (Notused.)
SR 3.3.1.1.15' Perform LOGIC SYSTEM FUNCTIONAL TEST.
18 months SR 3.3.1.1.16
NOTE-------------------
Neutroa detectors are excluded.
Verify the RPS RESPONSE TIME is within 18 months on a-limits.
STAGGERED TEST BASIS t
i HATCH UNIT l_
3.3-5 AMEND. 205/EXTPWR - 7/24/97
RPS Instrumentation 3.3.1.1
._ Table 3.3.1.1 1 (page 1 of 3)
Reactor Protection system Instrumentation 3Y APPilCASLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWASLE FUNCil0N' CONDifl0Ns sYsiEM ACTION D.1 REQUIRENEWis VALUE 1.
Intermodlete Range Monitor 2
3 G
st 3.3.1.1.1 s 120/125 i
st 3.3.1.1.4
-' divisions of M 3.3.1.1.6 full scale SR 3.3.1.1.7 st 3.3.1.1.13
- st 3.3.1.1.15 5(*)
3 N
sa 3.3.1.1.1 s 120/125 1
e SR-3.3.1.1.5 diviolone of SR 3.3.1.1.13 futt scale SR 3.3.1.1.15 b.
Insp 2
3 G
st 3.3.1.1.4 NA sa 3.3.1.1.15 5(*)
3 H
st 3.3.1.1.5 NA st 3.3.1.1.15 2.
Average Power Renee
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Monttor Neutronfkuu-Migh 2
3(8)
G st 3.3.1.1.1 s 20% RTP e.
t (setdown) st 3.3.1.1.7 st 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.13 b,
sinuteted Thornet 1
3(*)
F st 3.3.1.1.1 s 0.58 W +
Power - Nigh sa 3.3.1.1.?
5tX RTP Pnd l
s 115.M sa 3.3.1.1.8 RTP(b)
SR 3.3.1.1.10 SR 3.3.1.1.13 J
(continued)
(a) With any control rod withdrawn from a core cett contelning one or more fust asseabiles.
4 (b) 0.54 W + 5an
Operating."
(c) steh APRM channet providee inputs to both trip systems.
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HATCH UNIT 1 3.3-6 AMEND. 205/EXTPWR - 7/24/97 y
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Tebte 3.3.1.1 1 (page 3 of 3) 5 Reactor Protection system Instrumentation i
APPLICABLE.
CONDITIONS MODES OR REQUIRED REFERENCED 01NER CHANNELS FROM-SPECIFIED PER TRIP' REQUIRED
- SURVEILLANCE ALLOWASLE-FUNCil0N -
CONDITIONS SYSTEN ACfl0N D.1
-REQUIRENENTS VALUE 7.
seren Discherse Volume
- Water Level - Nigh e.
Reelstance 1,2 2
G sa 3.3.1.1.9 5 71 saitone Temperature-st 3.3.1.1.13 Detoctor SR 3.3.1.1.15 5(83
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st 3.3.1.1.9 5 71 pellone SR 3.3.1.1.13 st 3.3.1.1.15 b.
float sulsch 1,2 2
G st 3.3.1.1.13 s 71 settone SR 3.3.1.1.15 5(*)
2 N
st' 3.3.1.1.13 s 71 gettone SR 3.3.1.1.15
- 8. -Turbine stop t 28% RTP 4
E st-3.3.1.1.9 s 10% closed l
Velve - Cloeure st 3.3.1.1.11 st 3.3,1,1,13 SR 3.3.1.1.15 9.
Turbine Control Velve t 28% RTP 2
E st 3.3.1.1.9 2 600 pets l
feet closure, Trip oil SR 3.3.1.1.11 Pressure - Low st 3.3.1.1.13 -
st 3.3.1.1.15 54 3.3.1.1.16
(
- 10. Reactor mode switch -
1,2 1
G SR 3.3.1.1.12 NA shutdown PoeitIon sa 3.3.1.1.15 5(s) 1 M
st 3.3.1.1.12 NA st 3.3.1.1.15
- 11. Manuel scram 1,2 1
G sa 3.3.1.1.5 NA SR 3.3.1.1.15 5(*)
1 H
SR 3.3.1.1.5 NA -
SR 3.3.1.1.15 mm (e) 'With any control rod withdrawn from a core cell contelning one or more fuel essembtles, i '.-
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HATCH UNIT 1-3.3-8 EXTPWR - 7/24/97 rw--
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EOC-RPT Instrumentation 3.3.4.1 f
3.3 INSTRUMENTATION 3:.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation LCO 3.3.4.1 a.
Two channels per trip system for each EOC-RPT instrumentation function listed below shall be OPERABLE:
1.
Turbir.a Step Valve (TSV) - Closure; and 2.
Turbine Control Valve (TCV) Fast Closure, Trip 011 Pressure - Low.
0 11 b.
LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," limits for ino)erable EOC-RPT as specified in the COLR are made applica>1e.
APPLICABillTY:
illlRMAL POWER a: 28% RTP.
l ACTIONS
...................................--NOTE-------------------------------
Separate Condition entry is allowed for each channel.
I CONDITION REQUIRED ACTION COMPLETION TlHE A.
One or more channels A.1 Restore channel to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable.
OPERABLE status.
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A.2
NOTE---------
Not applicable if inoperable channel is the result of an inoperable breaker.
Place channel in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> J
trip.
1 (continued)
G HATCH UNIT 1 3.3-27 EXTPWR - 7/24/97
EOC-RPT Instrumentaticn 3.3.4.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B.
One or more functions B.1 Restore EOC-RPT trip 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with EOC-RPT trip capability, capability not maintained.
QB AND B.2 Apply the MCPR limit 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for inoperable MCPR limit for E0C-RPT as specified inoperable EOC RPT in the COLR.
not made applicable.
C.
Required Action and C.1 Remove the associated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion recirculation pump Time not met, from service.
QB C.2 Reduce THERHAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
/9 to < 28% RTP.
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SURVEILLANCE REQUIREMENTS
.-----------NOTE-------------------------------------
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains EOC-RPT trip capability.
SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL FUNCTIONAL TEST.
92 days (continued)
[
s HATCH UNIT'l 3.3-28 EXTPWR - 7/24/97
EOC-RPT Instrumentation 3.3.4.1 4
O SURVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE FREQUENCY SR 3.3.4.1.2 Verify TSV - Closure and TCV fast 184 days Closure, Trip 011 Pressure - Low functions are not bypassed when THERMAL POWER is k 28% RTP.
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SR 3.3.4.1.3 Perform CHANNEL CAtlBRATION, The 18 months Allowable Values shall be:
TSV - Closure: s 10% closed; and TCV fast Closure, Trip 011 Pressure - Low: a 600 psig.
SR 3.3.4.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.
HOTE-------------------
Braaker interruption time may be assumed from the most recent performance of SR 3.3.4.1.6.
Verify the EOC-RPT SYSTEM RESPONSE TIME 18 months on a is within limits.
STAGGERED TEST BASIS SR 3.3.4.1.6 Determine RPT breaker interruption time.
60 months O
- HATCH UNIT'l 3.1-29 EXTPWR - 7/24/97 e
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1400 I
I I INITIAL RTndt VALUES ARE A SYSTEM ff
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HYDRO 1EST UMIT 20'F FOR BELTUNE.
- 20 1300 WITH FUEL IN THE
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40*F FOR UPPER VESSEL, 24 VESSEL FOR ff f
AND HATCH 1 j
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32 1
1200 I
lj f
HEATUP/COOLDOWN 1100 i
IJ RATE 20'F/HR l
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BELTINE CURVES 1000 ADJUSTED AS SHOWN' h
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20 142 EFPY SKIFT (*F) n.
900 0
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BEL 11NE CURVES 800 ADJUSTED AS SHOWN-
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EFPY SHIF1 (*F)
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8 700 f
24 157 G
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mm mwS 300
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g ADJUSTED AS SHOWN f
l EFPY SHIFT ('F) g 28 167 a
500 E
I[
400 BELTINE CURVES
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ADJUSTED AS SHOWN n,
EFPY SHIFT ('F)
" "I )T2 P$lol-32 180 f
300
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FLANGE BELTINE UMITS BCLTLtNE 1
REOlON AND i
to r
- - BOTTOM HEAD 100 BOTT N LIMITS HEAD t
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... UPPER VESSEL O
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60 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)
[ ACADI r349ii 1 Figure 3.4.9-1 (Poge 1 of 1)
(m Pressure / Temperature Umits for
'(v}
inservice Hydrostotle and Inservice Lookoge Testa Hatch Unit 1 3.4-25 Proposed 8/97 i
t RCS P/T UWITS 3.4.9 1400 1
4
- 0. CORE NOT f
1300 CRITICAL UMIT
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^,f INITIAL RTndt VALUES ARE FOR HATCH 1 l
20'F FOR BELTLINE.
t200 l
40'F FOR UPPER VESSEL AND 10'F FOR BOTTOM HEAD 1100 y
l l
g HEATUP/COOLDOWN j
1000 j
j RATE 100*F/HR a
f g
900
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I a
I DELTINE CURVES h
700 ADJUSTED AS SHOWN:
U Y
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EFPY SHIFT ('F) s
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32 180 E
000
)
l 500
/
l oa 400 E
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a.
300
/
-BELTINE UMITS 200 estil.ine
~ = NW ND gyo j
UMITS BONOM 100
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...... UPPER VESSEL H.ANGE REGION UMITS l
,6*F 0
O 50 100 150 200 250 300 350 400
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MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)
[ ACA01 83492il Figure 3.4.9-2 (Poge 1 of 1)
Preasure/ Temperature Umits for Non-Huelear Hootup, low Power Physics Tests, and Cooldown Following a Shutdown Hotch Unit 1 5.4-26 Proposed 8/97
- - - - - - - - - - - - - - - ~ ~ - ~ ~
RCS P/T LlWITS 3.4.9
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1400 C. CORE 1300 CRITICAL LIMIT INITIAL RTndt VALUES ARE FOR HATCH 1 20'F FOR BELTLINE, 40'F FOR UPPER VESSEL, 1200 AND 10*F FOR BOTTOM HEAD 7
HEATUP/COOLDOWN RATE 100'F/HR g 1000 900 g
9 I
g 799 DELTINE CURVES g
ADJUSTED AS SHOWtt EFPY SHIFT ('F) 000 32 180 O\\
I 500 1
400 300
-BELTLINE AND NON-BELTLINE 200 Ligi7s Minimum cnticainy 100 Temperature 76 0
0 60 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)
D241 ruest 1 Figure 3.4.9-3 (Page 1 of 1)
Pressure / Temperature Limits for Crlficatify O
Hatch Unit 1 3.4-27 Proposed 8/97
Progra:s and Manuals 5.5
[
5.5 Programs and Manuals L J' 5.5.11 Technical Specifications (TS) Bases Control Proaram (continued) d.
Proposed changes that meet the criteria of b. above shall be reviewed and approved by the NRC prior to implementation.
Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
5.5.12 frimsrv Contgjnment Leakaae Rate Testina Proaram A program shall be established to implemert the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J. Option B, as modified by approved exemptions.
This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.
The peak calculated primary containment internal pressure for the design basis loss of coolant accident, P, is 50.5 psig.
l The maximum allowable primary containment leakage rate, L, at P,
/^.
is 1.2% of primary containment air weight per day.
Leakage rate acceptance criteria are:
a.
Primary containment overall leakage rate acceptance criterion is s 1.0 L,.
During the first unit startup following testing in accordance with this program, the combined Type B and Type C tests, and s 0.75 f.,for the leakage rate acceptance criteria are s 0.60 L for Type A tests; b.
Air lock testing acceptance criteria are:
1)
Overall air lock leakage rate is s 0.05 L, when tested at 1 P,
2) for each door, leakage rate is s 0.01 L when the gap between the door seals is pressurized t,o 2 10 psig for at least 15 minutes.
The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.
[j]
(continued)
L.
HATCH UNIT 1 5.0-16a EXTPWR - 7/24/97
4 l
.i Unit 2 Facility Operating License Proposed Cliange and Teclinical Speelf1 cations Proposed Clianges llL-5413
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' L C.
This license shall be deemed to contain and tw subject to the conditions specified in the following Commission regulations in r
-10 CFR Chapter 1: Part 20, Section 30.34 of Part 30. Section 40.41 of part 40, Sections 50.54 and 50.49 of Part 50, and tection 70.32 of Part 70; and is subject to all applicable provisions of the Act andtotherules, regulations,andordersoftheCommissionnowgr j
hereafter in effect; and is sybject to the additional conditions specified or incorporated below (1)
Maxi - Power Level Southern Nuclear is authorized to operate the facility i
at steady state reactor core power levels not in excess of 2763 megawatts thermal in accordance with the conditions l
specified herein and in Attachment I to this license. is an integral part of this license.-
l (2) lechnical Snacifications The Technical Specifications in Appendix A and the Environmental Protection Plan contained in Appendix 8, as revised through Amendment No.144 are hereby incorporated in the license.
Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
The Surveillance Requirements (SRs) contained in the Appendix A Technical Specifications and listed below are j
not required to be performed immediately upon implementation of Amendment No.135. The SRs listed below shall be successfully demonstrated prior to the time and condition specified below for each a)
SRs 3.3.1.1.15, 3.3.1.1.16 (for function g),
3.3.2.2.2, 3.3.2.2.3, 3.3.3.2.2, 3.3.6.1.6 (for i
function 1.f), 3.3.8.1.4, 3.7.7.2 and 3.7.7.3 shall be successfully demonstrated prior to entering MODE 2 on the first plant startup following the sixteenth refueling outage; j
and b)
SRs 3.8.1.8, 3.8.1.10, 3.8.1.12. 3.8.1.13,d at their=
i 3.8.1.18 shall be successfully demonstrate next regularly scheduled performance; c)
SRs 3.6.4.1.3 - and 3.6.4.1.4 will t>e met at implementation for the= secondary containment configuration in effect at-that time.
The SRs shall be-successfully demonstrated for the other secondary centainment-configuration prior to the plant entering the 1.C0 applicability for the configuration.
O The original licensee authorized to possess, use, and operate the 1
-facility was Georgia Power Company (GPC).
Consequently, certain historical references to GPC remain in the license conditions._
i Amendment No._
}
i Definitions 1.1 1.1 Definitions to cause some point in a)propriate correlation (s) boiling transition, MINIMUM CRITICAL POWER RATIO (MCPR)
(continued) tle assembly to experience divided by the actual assembly operating power.
r MODE A MODE shall correspond tc any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tersioning specified in Table 1.1-1 with fuel in the reactor vessel.
OPERABLE - OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABillTY when it is capable of serforming its specified safety function (s) and wien all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary uquipment that are required for the system, subsystem, division, component, or device to perform its specified safety function (s) are also capable of performing their related support function (s).
PHYSICS TESTS shall be those tests performed to O
PHYSICS TESTT, measure the fundamental nuclear characteristics of the reactor core and related instrumentation.
These tests are:
a.
Described in Chapte-14, initial Tests and i)peration, of the FSAR; b.
Authorized under the provisions of 10 CFR 50.59; or c.
Otherwise approved by the Nuclear Regulatory Commission.
RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 2763 MWt.
l REACTOR PR0'.cTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve (continued)
HATCH UNIT 2 1.1-5 EXTPWR - 7/24/97
RPS Instrumentaticn 3.3.1.1 3
ACTIONS (continued) x CON 01T10N REQUIRED ACT10N COMPl.ET10N TIME C.
One or more function 1 C.1 Restore RPS trip I hour with RPS trip capability.
capability not maintained.
D.
Required Action and 0.1 Enter the Condition immediately associated Completion referenced in Time of Condition A.
Table 3.3.1.1-1 for 8, or C not met.
the channel.
E.
As required by E.1 Reduce 1HERMAl. POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Required Action D.)
to < 28% RTP.
l and referenced in i
Table 3.3.1.1-1.
f.
As required by f.1 Be in MODE 2.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Required Action 0.1 and referenced in Table 3.3.1.1-1.
G.
As required by G.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Required Action 0.1 and referenced in Table 3.3.1.1-1.
H.
As required by H.1 Initiate action to immediately Recuired Action 0.1 fully insert all anc referenced in insertable control Table 3.3.1.1-1.
rods in core cells containing one or more fuel assemblies.
ha HATCH UNIT 2 3.3-2 EXTPWR - 7/24/97-
I RPS Instrumentation 3.3.1.1 Il SURVEILLANCE REQUIREMENTS (continued)
\\')
SURVE!LLANCE FREQUENCY SR 3.3.1.1.11 Verify Turbine Stop Valve - Closure and 18 months Turbine Control Valve Fast Closure Trip 011 Pressure - Low Functions are not bypassed when THERHAL POWER is 1 28% RTP.
l SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST.
18 months SR 3.3.1.1.13
NOTES------------------
- 1. Neutron detectors are excluded.
- 2. For Function 1, not required to be performed when entering MODE 2 from H0DE 1 until 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> after entering MODE 2.
Perform CHANNEL CAllBRATION.
18 months
(
SR 3.3.1.1.14 (Notused.)
SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST.
18 months (continued)
/~'N HATCH UNIT 2 3.3 5 EXTPWR - 7/24/97
RPS Instrumentation 3.3.1.1 table 3.3.1.1 1 (pePe 1 of 3) teector Protection system Instrumentation
\\
APPLICABLE CouDiff0Ns NOD (s OR tieulttD RIfit(NCEO 01 Nit ChAme(Ls fa0M
$PEClfit0 PER TRIP REeuthED suRVilLLANCE ALLOW 4sLE PUNCil0K ConDITitus sisitM Acil0W D.1 REeulREMEWis VAltE!
1.
Intermedts..e tence Ne'titor a.
Woutten f tua. Nigh 2
3 o
sa 3.3.1.1.1 s 120/125 sa 3.3.1.1.6 divleinne of M 3.3.1.1.6 futt scale sa 3.3.1.1.7 i
sa 3.3.1.1.13 la 3.3.1.1.15
$(a) 3 N
sa 3.3.1.1.1 s 120/125 la 3.3.1.1.5 divlelone of SR 3.3.1.1.13 futt scale SR 3.3.1.1.11 b.
Inap 2
3 G
Sa 3.3.1.1.6 NA sa 3.3.1.1.15 5(e) 3 N
$R 3.3.1.1.5 PA la 3.3.1.1.15 2.
Average Power tange Monitor s.
Neutron flun - High 2
3(83 G
sa 3.3.1.1.1 s 20% RTP (setdown) la 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.10 sa 3.3.1.1.13 b,
slauteted thermel 1
M83 F
st 3.3.1.1.1 s 0.58 W +
Power. Nigh SR 3.3.1.1.2 58% RTP and s 115.5%
$4 3.3.1.1.8 RTP(D) l Sa 3.3.1.1.10 sa 3.3.1.1.13 (continued)
(a) With any control rod withdrawn from a core cent contelning one or more fust asedlies.
(b) 0.56 W e $85
- 0.58 &W RTP when reset for ainste loop operation per LC0 3.4.1, " Recirculation Loope Operating."
(c) tech APael chamel provideo IrpJts to both trip evetems.
O (s
HATCH UNIT 2 3.3-7 EXTPWR = 7/24/97
RPS Instrumentation 3.3.1.1
)
fable 3.3.1.1 1 (pape 3 of 3) acector Protection System Instrumentellen APPLsCASLE C')M0l f l0NS N0Dil OR REQUlktD afftatNCED OTHta CHANNELS FROM SPECl8tfD Pta falP ateulalD SuaVtlLLANCE Alt 0WABLE FUNC110N COND11:nNS SYlfEM AC110N D.1 atoulatMEWi$
VALUE 7.
Scram elecherpe Volume Water level. High I
e.
Reelstence 1,2 2
0 sa 3.3.1.1.9 s 57.15 settone temperature la 3.3.1.1.13 Detector la 3.3.1.1.15 S *I 2
N sa 3.3.1.1.9 s $1.15 settone I
sa 3.3.1.1.13 Sa 3.3.1.1.15 b.
float Switch 1,2 2
0 sa 3.3.1.1.13 s 57.1b pellone la 3.3.1.1.15 5(e) 2 N
$a 3.3.1.1.13
$ 57.15 gallons la 3.3.1.1.15 8.
Turbine stop t 28% RTP E
ta 3.3.1.1.9 s 10% closed l
Veive-Clocure sa 3.3.1.1.11 la 3.3.1.1.13 sa 3.3.1.1.15 la 3.3.1.1.16 9.
Turbine Control Velve t 281 atP 2
E la 3.3.1.1.9 t 600 pels l
Feet Closure, Trip Olt la 3.3.1.1.11 p
Pressure - Low da 3.3.1.1.13 la 3.3.1.1.15 gs sa 3.3.1.1.16
- 10. Reactor Mode Switch.
1,2 2
C sa 3.3.1.1.12 NA shutdown Position sa 3.3.1.1.15 5(8) 2 N
la 3.3.1.1.12 NA sa 3.3.1.1.15
- 11. Manuel scres 1,2 2
C sa 3.3.1.1.5 NA la 3.3.1.1.15 5(e) 2 N
SR 3.3.1.1.5 NA sa 3.3.1.1.15 (a) With any control rod withdrawn from a core cell contelning one or more fuel assedlies, lh HATCH UNIT 2 3.3-9 EXTPWR - 7/24/97 i
~.
E00-RPT Instrumentation 3.3.4.1
)
3.3 INSTRUMENTATION 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation LCO 3.3.4.1 a.
Two channels per trip system for each EOC-RPT instrumentation Function listed below shall be OPERABLE:
l t
1.
Turbint Stop Valve (TSV) -- Closure; and 2.
Turbine Control Valve (TCV) Fast Closure, Trip 011 Pressure -- Low.
QB b.
LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," limits for inoserable EOC-RPT as specified in the COLR are made applica)1e.
APPLICABILITY:
THERMAL POWER a 28% RTP.
l ACTIONS
...............................------NOTE-------------------------------------
Separate Condition entry is allowed for each channel.
i CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more channels A.1 Restore channel to-72 hours inoperable.
OPERABLE status.
QB A.2
NOTE---------
Not applicable if inoperable channel is the result of an inoperable breaker.
Place channel in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> trip.
(continued)
{x
<< HATCH UNIT 2 3.3-28 EXTPWR - 7/24/97
l EOC-RPT Instrumentation
-i 3.3.4.1 O
ACTIONS (continued)
\\"/
CONDITION REQUIRED ACTION COMPLETION TIME 8.
One or more Functions B.1 Restore EOC-RPT trip 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with EOC-RPT trip capability.
capability not maintained.
QB AND B.2 Apply the MCPR limit 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for inoperable MCPR limit for E0C-RPT as specified inoperable EOC-RPT in the COIR.
not made applicable.
C.
Required Action and 0.1 Remove the associated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion recirculation pump Time not met.
from service.
3 08 C.2 Reduce THERMAL PO!!ER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to < 28% RYP.
l SURVEILLANCE REQUIREMENTS
-NOTE-------------------------------------
When a channel is placed in an inopet gble status solely for performance of required Surveillances, ontry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provide.4 the associated function maintains EOC-RPT trip capability.
SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL FUNCTIONAL TEST.
92 days (continued) m HATCli UNIT 2 3.3-29 EXTPWR - 7/24/97
-l
EOC-RPT Instrumentaticn 3.3.4.1 SURVEILLANCE REQUIREM:ATS (continued)
\\
SURVElLtANCE FREQUENCY SR 3.3.4.1.2 Verify TSV - Closure and TCV fast 18 months Closure, Trip 011 Pressure - Low functions are not bypassed when THERMAL POWER is k 28% RTP.
l SR 3.3.4.1.3 Perfors CHANNEL CAllBRATION. The 18 months Allowable Values shall be TSV - Closure: s 10% closed and TCV fast closure, Trip 011 Pressure - Low: a 600 psig.
SR 3.3.4.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.
I
\\
NOTE-------------------
Breaker interruption time may be assumed from the most recent performance of SR 3.3.4.1.6.
Verif,v the EOC-RPT SYSTEM RESPONSE TIME 18 months on a is within limits.
STAGGERED TEST BASIS SR 3.3.4.1.6 Determine RPT breaker interruption time.
60 months t
i V
HATCH UNIT 2 3.3-30 EXTPWR - 7/24/97
f INiilAL RinCl VALUES AME 50*F FOR BELTLINE 1300 20*F FOR VI'PER VESSEL AND 50*F FOR BOTTOM HEAD BELTINE CURVES ADJUSTED AS SHOWN.
ijon EFPY SHIFT ('F)
/
32 127 1000 1
/-
e, 000 HEATUP/COOLDOWN S
/
RATE 20'F/HR f
FOR CURVE A 800 3
700 A'- CORE DELTUNE 000 A - NON-BELTUNE E
A - PRESSURE TEST WITH l
FUEL IN THE VESSEL
\\
5 600 1400 NON BELTLINE BELTI.lNE AT 32 300 1812 PS'Gb CURVE A' IS VAUD UP TO 32 FrPY 200 0F OPERATION.
OvuuP CURVE A 90*F IS VAUD UP TO 32 EFPY 100 0F OPERATION FOR BELTLINE AND EOL FOR NON-BLLTUNE.
0 0
60 100 160 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)
Igel rwi i Figure 3.4.9-1 (Poge 1 of 1)
Pressure / Temperature Umits for intervice Hydrostatic and inservice Lookoge Tests
[N tj Hoich unit 2 3.4-25 Propo.ed s/s7
RCS P/T UMITS 3.4.9 1400 B B' O
L INITIAL RTndt VAES ARE
~
50'F FOR BELTLINt?
26'F FOR UPPER VESSEL 1300 AND 50'F FOR BOTTOM HEAD i
BELTINE CURVES ADJUSTED AS SHOWN; 1100 EFPY SHIFT ('F) 32 127 1000 m
a.
000 HEATUP/COOLDOWN RATE 100*F/HR FOR CURVE B 800 t
b 700 t;
3
}
B'- CORE BELTJNE E
coa B - NON-BELTUNE
/
E f
B - NON-NUCLEAR k
l
/
HEATUP/COOLDOWN 500 CORE NOT CRITICAL 400
-NON-BELTUNE k
... BELTLINE AT 32 300 1312 PSlO!
~
CURVE B' IS VAUD UP TO 32 EFPY 200 0F OPERATION.
pou vr' CURVE B
- F IS VAUD UP TO 32 EFPY 100 2
OF OPERATION FOR BELTUNE AND EOL FOR NON-BrLTUNE.
0 0
50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)
[ AC@ l F3492 ]
Figure 3.4.9-2 (Page 1 of 1)
Pressure / Temperature Umits for Non-Nuclear Hedup.
Low Power Phples Tests, and Cooldown Following a Shutdown V
Hatch Unit 2 3.4-26 Proposed 8/97
RCS P/T UWITS 3.4.9
(
1400 C C' b
INITIAL RTndt VALUES ARE 50'F FOR BE(TLINE 1300 26*F FOR OPPER VESSEL l
AND
$0*F FOR BOTTOM HEAD 1200 E
BELTINE CURVES ADJUSTED AS SHOWN.
1 goo 1
EFPY SHIFT ('F) h 32 127 i
1000 0
0z 900 g
HEATUP/.
'JWN RATE 1 1R i
FOR
- 3 800 I
700 t;
g 000 C'- CORE BELTUNE
[
C - NON-BELTUNE C - NON-NUCLEAR
(
HEATUP/COOLDOWN 500 COR2 NOT CRITICAL v4 a
400
- NON BELTLINE h
BELTUNE AT 32 300 jaiz esial e
CURVE C' IS VAUD UP TO 32 EFPY 200 0F OPERATION.
",$.f CURVE C 100 IS VAUD UP TO 32 EFPY
/
OF OPERATION FOR BELTUNE AND f
EOL FOR NON-BELTLINE.
0 50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F) 1 ArAD I F3493 1 Figure 3.4.9-3 (Page 1 of 1)
Pressure / Temperature Umits for Criticality Notch Unit 2 3.4-27 Proposed 8/97
Prograns and Manuals 5.5 5.5 Prograrr.s and Manuals 5.5.11 Jechnical Specifications (TStann_(gntrol Proan!D (continued) d.
Proposed changes that meet the criteria of b. above shall be reviewed and approved by the NRC prior to implementation.
Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
5.5.12 Primary Containment Leakaae Rate Testina Proaram A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guicielines contained in Regulatory Guide 1.163, "Performan:e-Based Containment Leak-Test Program," dated September 1995.
The peak calculated primary containment internal pressure for the design basis loss of coolant accident P,, is 46.9 psig.
l The maximum allowable primary containment leakage rate, L,, at P, is 1.2% of primary containment air weight per day.
Q Leakage rate acceptance criteria are:
a.
Primary containment overall leakage rate acceptance criterion is & l.0 L,
During the first unit startup following testing in accordance with this prorJram, the leakage rate acceptance criteria are s 0.60 L for the combined Type B and Type C tests, and s 0.75 [, for Type A tests; b.
Air lock testing acceptance criteria are:
1)
Overall air lock leakage rate is s 0.05 L, when tested at s P.,
2)
For each door, leakage rate is s 0.01 L when the gap between the door seals is pressurized t,o 2 10 psig for at least 15 minutes.
The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.
/9 (continued)
Q;'
HATCH UNIT 2 5.0-16a EXTPWR - 7/24/97
_..______..._.,..__._____.__.___..__._._..__.._q l
l I
i f
Unit 1 Marked-Up Operating Licensing and Techr.ica! Specifications Pager 1
d
...,,er-.-
TIAcilEll{ Oft M T.D q LE(r @ 6
- the procedures and limitations set forth in this license; and the Georgia Power Company, the Oglethorpe Power Corporation, the Municipal Electric Authority of Georgia and the City of Dalton, Georgia to possess but not operate the facility in accordan;e with the procedures and limitations set forth in this license; (2) Southern Nuclear, pursuant to the Act and 10 CFR Part 70, to l
receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical-or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 and 70, l
O to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40 Sections 50-54 and 50-59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations,andordersoftheCommissignnoworhereafterineffect;and I
is subject to the additional conditions specified or incorporated below:
(1) Maximum Power level Southern Nuclear is authorized to operate-the l
facility at steady state It ctor core power levels not in excess o megawatts thermal.
BTLB
/ \\
3 The original licensee authorized to possess, use and operate the facility was Georgia Power Company (GPC).
Consequently, certain historical 4
references to GPC remain in the license conditions.
pTT.1 Amendment No. 203
Definitions 1.1 1.1 Definitions (continued)
PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.
These tests are:
i a.
Described in Section 13.6, Startup and Power Test Program, of the FSAR; b.
Authorized under the provisions of 10 CFR 50.59; or c.
Otherwise approved by the Nuclear Regulatory Conaiission.
RATED THERMAL POWER RTP shall be a total reactor core at transf (RTP) rate to the reactor coolant of MWt.
- 3 l
REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or
\\
total steps so that the entire response time is measured.
SDM shall be the amount of reactivity by which the reactor is suberitical or would be suberitical assuming that:
a.
The reactor is xenon free; b.
The moderator temperature is 68'F; and c.
All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.
With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDN.
STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, l
channels, or other designated components during the interval specified by the Surveillance (continued)
. HATCH UNIT 1 1.1-5 Amendment No. 197
RPS Instrumentation 3.3.1.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TlHE C.
One or more Functions C.)
Restore RPS trip 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with RPS trip capability, capability not maintained.
D.
Required Action and 0.1 Enter the Condition InNdiately 4ssociated Completion referenced in Time of Condition A, Table 3.3.1.1-1 for 8, or C not met, the channel.
E.
As required by E.1 Reduc HERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Required Action D.1 to <
RTP.
and referenced in i
Table 3.3.1.1-1.
I,9f;[
F.
As required by F.1 Be in MODE 2.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Required Action D.1 and referenced in Table 3.3.1.1-1.
I G.
As required by G.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Required Action 0.1 and referenced in Table 3.3.1.1-1.
(
H.
As required by H.1 Initiate action to Immediately Required Action D.1 fully insert all and referenced in insertable control T.
v 3.3.1.1-1.
rods in core cells containing one or more fuel assemblies.'
l
(^')
C/
HATCH UNIT 1 3.3-2 Amendment No. 205 l w
+
r-n
RPS Instrumentation 3.3.1.1
)
SURVEILLANCE REQUIREMENTS (continued)
\\s_,/
SURVEILLANCC FREQUENCY SR 3.3.1.1.11 Verify Turbine Stop Valve -- Closure and 184 days Turbine Control Valve Fast Closure, Trip 011 Pressure -- Low Functions are bypassed when THERMAL POWER is RTP.
1 SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST.
18 months SR 3.3.1.1.13
NOTES------------------
- 1. Neutron detectors are excluded.
- 2. For Function 1. not required to be perforwed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.
Perform CHANNEL CALIBRATION.
18 months SR 3.3.1.1.14 (Not used.)
l SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST.
18 months SR 3.3.1.1.16
NOTE-------------------
Neutron detectors are excluded.
Verify the RPS RESPONSE TIME is within 18 months on a limits.
STAGGERED TEST BASIS N s/
s HATCH UNIT 1 3.3-5 Amendment No. 205 4
RPS Instrumentation 3.3.1.1 febte 3.3.1.1 1 (pope 1 of 3)
Rootter Protection Breten Instrumentellen APPL 1CA8LI CouDtiIONS NDE5 Cat M euletD Mf(48httD 01utt tn4NetL$
fRCN SPECIFit0 Pts felP titultt0 EmVtlLLAstt ALLOW 48tt PW Cilon ta elfl0st tYlttet ACTION D.1 staulttatsis VAtut 1.
Intermediate tage monitor
- ~
2 3
s am 3.3.1.1.1 s 120/125 se 3.3.1.1.4 divlelene of a 3.3.1.1.6 fuit scale sa 3.3.1.1.7 a 3.3.1.1.13 at 3.3.1.1.15 S(e) 3 m
sa 3.3.1.1.1 s 120/125 at 3.3.1.1.5 divlelene of am 3.3.1.1.13 full ecale at 3.3.1.1.15 b.
Insp 2
3 6
. 38 3.3.1.1.4 hA GR 3.3.1.1.15 S(e) 3 u
sa 3.3.1.1.5 NA et 3.3.1.1.15 2.
Average Power G m ge mentter a.
Soutron Fl w = Wigh 2
3 'I I
8 Sk 3.3.1.1.1 8 30E RTP se 3.3.1.1.7 OR 3.3.1.1.8 3R 3.3.1.1.10 et 3.3.1.1.13 b,
eleAeted thennel 1
3(83 P
as 3.3.1.1.1 g 0.54 W
- l puuer - a f gh ta 3.3.1.1.2 625 aIP and
- s til. X l
sa 3.3.1.1.8 afp(b st 3.3.1.1.10 l
Yk; m 3.3.1.1.13 (continued)
(e) with any centrol red withdrom free a cers cett sentelnig one or more fuel seeeWise.
0.58 W *M 0.54 av afP iden reset fee eigte leap operstlen per LCD 3.4.1, Mectreute Ib)
OperetlN.*
(t) tech APen i provides ( m ute to both trfp systems.
l f
\\
HATCH UN11 1 3.3-6
^
--,,----e
RPS Instrumentation 3.3.1.1 Table 3.3.1.1+1 (pese 3 of 3)
Reactor Protection fyeten instrumentation APPLICABLE Couplflout MnDES OR DEEllRED REFERENCED OfMER CNAmeELS PROM 9PECIFIED PER TRIP ReeUIRED suRVilLLANCE ALLOWALE PUNCTION ConDIT!DNS SYSTEM Acil0N D.1 RNUIREMENIS VALUE r.
Scran Discherpe vetume Water Level = Nlph e.
Reelstance 1,3 2
8 3R 3.3.1.1.9 s 71 pottens Temperature SR 3.3.1.1.13 Detector M 3.3.1.1.15 S(*)
t N
st 3.3.1.1.9 s 71 peltens SR 3.3.1.1.13 3R 3.3.1.1.15 b.
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Tusine Centret Velve t Jef RTP 2
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- 10. Reester pode twitch -
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shutdeen Peeltlen et 3.3.1.1.15
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st 3.3.1.1.5 E
st 3.3.1.1.15 ai (e) With any control red withdream from a core cell contelning one or more fuel estembltes.
d HATCH UNIT 1 3.3-8 Amendmant No. 195
EOC-RPT-Instrumentaticn 3.3.4.1 i
O 3.3' INSTRUMENTATION 4 v
-3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation s
LCO 3.3.4.'.
a.
Two channels per trip system for each EOC-RPT.
instrumentation Function listed below shall be OPERABLE:
1.
Turbine Stop Valve (TSV) - Closure; and 2.
Turbine Control Valve (TCV) Fast Closurc, ' rip 011 Pressure - Low.
QR s
b.
LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," limits for inoperable EOC-RPT as specified in the COLR are made applicable.
EfW
-APPLICABILITY:
THERPA', POWER a:
P.
ACTIONS
NOTE-------------------------------------
(
Separate Condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more channels A.1 Restore channel to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable.
OPERABLE status.
98-A.2
NOT E--------
Not applicable if r
inoperable channel is i
the result nf an-inoperable breaker.
Place channel in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> trip.
I L
(continued) t HATCH UNIT 1 3.3-27 Amendment No. 195
~. - -...
EOC-RPT-Instrumentation
~
3.3.4.
- --,h.-p. ; -ACTIONS-(continued)
CONDITION REQUIRED ACTION COMPLETION TIME 8..
One or more Functions 8.1 Rest:;rs EOC-RPT trip 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with EOC-RPT trip capability, capability not k
c maintained.-
E I 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2
AE 8.2 Apply the MCPR limit for inoperable MCPR limit for_
E0C-RPT as specified inoperable EOC-RPT in the COLR.
not s.ade applicable.
-C.
Required Action and C.1 Remove the associated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion recirculation pump Time not met, from service.
E C.2 Reduc T ERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to TP.
i i
-SURVEILLANCE REQUIREMENTS
......................-NOTE-------------------------------------
Wt.en a channel is placed in an inoperable status solely for performance of i
ev. red Surveillances, entry _into associated Conditions and Required _ Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> _provided the associated Function maintains c
EOC-RPT trip capability.
SURVEILLANCE FREQUENCY SR -3,3.4.1.1-Perform CHANNEL FUNCTIONAL' TEST.
92 days-i (continued)
F N
_ HATCH UNIT'1 3.3-28_
Amendment.No.:195 L
.-s
-u
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EOC-RPT Instrumentation 3.3.4.1
{'~'}
SURVEILLANCE REQUIREMENTS (continued)
N- '
SURVEILLANCE FREQUENCY M.
SR 3.3.4.1.2 Verify TSV -- Closure and TCV Fast 184 days Closure, Trip Oil Pressure -- Low Functions are not bypassed when THERMAL POWER is 2:
RTP.
Lt'lej SR 3.3.4.1.3 Perform CHANNEL CALIBRATION, The 18 months Allowable Values shall be:
TSV.- Closure: s 10% closed; and TCV Fast Closure, Trip 011 Pressure -- Low: 2: 600 psig.
SR 3.3.4.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.
(
NOTE-------------------
Breaker interruption time may be assumed from the most recent performance of SR 3.3.4.1.6.
Verify the EOC-RPT SYSTEM RESPONSE TIME 18 months on a is within limits.
STAGGERED TEST BASIS SR 3.3.4.1.6 Determine RPT breaker interruption time.
60 months
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Pressure / Temperature Limits for i
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. _._. ~ _ _ _._. _.. _. _..___
Programs and Manuals
-5.5 5.5 Programs and Manuals 5.5.11 Technical Snacifications (TS) Bases Control Proaram (continued) d.
. Proposed changes that meet the criteria of b. above shall be reviewed and approved by the NRC prior to implementation.
Changes to the Bases implemented without prior NRC approval 4
shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
5.5.12 Primary Containment Leakaae Rate Testina Proaram A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.
The peak calculated primary containment inte Lpressure for the design basis loss of coolant accident, P., is psig. g The maximum allewable primary containment leakage rate, L,, at P, is 1.2% of primary containment air weight per day.
Leakage rate acceptance criteria are:
a.
Primary containment overall leakage rate acceptance criterion is s 1.0 L,.
During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 L combined Type B and Type C tests, anj s 0.75 f.,for the for Type A i
tests; l-b.
Air lock testing acceptance criter'a are:
1)
Overall air lock leakage ra'.e is s 0.08 L, when tested at 2 P,
2)
For each door, ledage rate is s 0.01 L when the gap between the door seals is pressurized t,o 2 10 psig for i
at least 15 minutes.
The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary containment Leakage Rate Testing Program.
(continued)
HATCH UNIT'l' 5.0-16a Amendment No. 200 l
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Unit 2 Facility Operating License Prop <, sed Change and Technical Specifications Proposed Changes j
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d O
FNILITO tSemuca LiteKw i
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C.
This license shall be deemed to contain and is subject to the conditions specified in the following Comission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.5g of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Comission now pr hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level Southern Nuclear is authorized to operate the facility I
at steady state reactor core power levels not in excess of
@ ified herein and in Attachment 2 to this license.
mefawatts thermal in accordance with the conditions s_
spec
. is an integral part of this license.
(2)
Technical Snecifications The Technical Specifications in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.144 are hereby incorporated in the license. Southern Nuclear shall operate the l
facility in accordance with the Technical Specifications and the Environmental Protection Plan, Os The Surveillance Requirements (SRs) contained in the Appendix A Technical Specifications and listed below are not required to be performed immediately upon implementation of Amendment No. 135.
The SRs listed below shall be successfully demonstrated prior to the time and condition spectfied below for each:
a)
SRs 3.3.1.1.15, 3.3.1.1.16 (for function 9),
3.3.2.2.2, 3.3.2.2.3, 3.3.3.2.2, 3.3.6.1.6 (for function 1.f), 3.3.8.1.4. 3.7.7.2 and 3.7.7.3 shall be successfully demonstrated prior to entering MODE 2 on the first plant startup following the sixteenth refueling outage; b)
SRs 3.8.1.8, 3.8.1.10, 3.8.1.12, 3.8.1.13, and 3.8.1.18 shall be successfully demonstrated at their next regularly scheduled performance; c)
SRs 3.6.4.1.3 and 3.6.4.1.4 will be met at implementation for the secondary containment configuration in effect at that time.
The SRs shall be successfully demonstrated for the other secondary containment configuration prior to the plant entering the LCO applicability for the configuration.
8 The original licensee authorized to possess, use, and operate the facility was Georgia Power Company (GPC).
Consequently, certain historical references to GPC remain in the license conditions.
LL O I T d k. Amendment No. 144
Definitions 1.1 1.1 Definitions i
MINIMUM CRITICAL POWER a>propriate correlation (s) to cause some point in RATIO (MCPR)
(continued) tie assembly to experience boiling transition, divided by the actual assembly cperating power.
MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
I OPERABLE - OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function (s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function (s) are also capable of performing their related support function (s).
O PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.
These tests are:
a.
Described in Chapter 14. Initial Tests and Operation, of the FSAR; b.
Authorized un'ier the provisions of 10 CFR 50.59; or OtNrwise approved by the Nuclear Regulatory c.
Commission.
RATED THERMAL POWER RTP shall be a total reactor core-heat tr fer (RTP) rate to the reactor coolant o(g i
l REACTOR PROTECTION
-The RPS RESPONSE TIME shall be that ime i terval SYSTEM (RPS) RESPONSE-from when the monitored parameter excee t' RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve im
(
(continued)
HATCH UNIT 2 1.1-5 Amendment No. 138
RPS Instrumentation 3.3.1.1 OACTIONS
~
COMPLET10N. TIME (continued) j
'COND! TION REQUIRED ACTION
~
C.
One or more Functions-C.1 Restore RPS trip I hour i
with RPS trip.
capability.
i capability not t
maintained.
D.
Required Action and 0.1.
. Enter the Condition Immediately i'
associated Completion referenced in
?
-Time of Condition A, Table 3.3.1.1-1 for -
B, or C not met.
'the channel.
i E.
As required by E.1 Reduce _ THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Recluired Action D.1 to (7 3GR R
' anc referenced in F
Table 3.3.1.1-1.
0 7*
i
.F.
As required by F.1 Be in MODE 2.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
~
Required Action D.1 and referenced in Table 3.3.1.1-1.
l G.-
As required by G.1 Be in M0DE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Required Action D.1 i
and referenced in Table 3.3.1.1-1.
4 i
.H.
As required by H.1 Initiate action to Immediately Recuired Action D 1 fully insert all anc referenced in-insertable control
. Table 3.3.1.1-1.
rods in core cells containing one or-more fuel assemblies..
- O HATCH UNIT 2' 3.3-2 Amendment No. 146 l
r
RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY 3.3.'.1.11 Verify Turbine Stop Valve - Closure and 18 months SR 1
Turbine Control Valve fast Closure, Trip 011 Pressure - Low Functions are bypassed when THERMAL POWER is 7.(.
T
-)
SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST.
18 months SR 3.3.1.1.13
NOTES------------------
- 1. Neutron detectors are excluded.
- 2. For Function 1 not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.
O Perform CHANNEL CALIBRATION.
18 months SR 3.3.1.1.14 (Not used.)
l J
SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST.
18 months (continued)
HATCH UNIT 2 3.3-5 Amendment No. 146
RPS Instrumentation I
3.3.1.1 1.
l f able 3.3.1.1 1 (pose 1 of 3)
Reector Protectf un system instrumentation I
APPIiCABLE CONDIT10Ns
"" DR REeulRED REFERENCED
- 4..it C*iANNELs FROM SPECIFIED PER TRIP REsulRED SURVEILLANCE ALLOW 48LE FUNCil0N Coelflows
$YSTEM ACTION D.1 REGUIREMENis VALUE 5.
Intereediate Renee Monitor a.
Woutron FLum - 4:Ish 2
3 G
st 3.3.1.1.1 s 120/125 et 3.3.1.1.4 divisions of SR 3.3.1.1.6 full scete SR 3.3.1.1.7 st 3.3.1.1.13 sa 3.3.1.1.15 5(*I 3
N sa 3.3.1.1.1 s 120/125 SR 3.3.1.1.5 divisions of sa 3.3.1.1.13 futt scale sa 3.3.1.1.15 b.
Inop 2
3 G
Sa 3.3.1.1.4 hA sa 3.3.1.1.15 5(a) 3 N
sa 3.3.1.1.5 h4 at 3.3.1.1.15 2.
Average Power Range Monttor
- a. - Neutron Flum - Mish 2
3(8)
G sa 3.3.1.1.1 s 201 RTP (setaoim) 3 sa 3.3.1.1.7 sa 3.3.1.1.8 SR 3.3.1.1.10 68 g/c sa 3.3.1.1.13 i
b.
flautated Therent 1
3(CI F
sa 3.3.1.1.1 LQ.58 U +
l Power - Mish st 3.3.1.1.2 6ER)RTP and
'115.5%
l SR 3.3.1.1.8 RTP(b) sa 3.3.1.1.10 sa 3.3.1.1.13 (continued)
(a) With td withdrawn from a core cett containing one or more fuel assemblies.
0.58 AW RTP when reset for single loop operation per LCD 3.4.1, Meelrculation Loops ~
(b) 0.58 +
/
opera (c) Each APRM chamel Vovides irpets to both trip systems.
l s
HATCH UNIT 2 3.3-7 Amendment No. 146 4-d.
y 5
y e
.mm m
w,.
---u
i RPS Instrumentation 1
3.3.1.1 Table 3.3.1.1 1 (page 3 of 3) g Reactor Protection system instruwntation APPLICABLE CONDifl0Ns MODES OR REQUIRED REFERENCED OTHER CNANNELS FROM SPECIFIED PER TPIP REculRED SURVE!LLANCE ALLOWA8LE FUNCfl0N CONDITIONS
$YstEM Acil0N D.1 REQUIREMENis VALUE 7.
scram Discharge volume Water Levet - Nigh a.
Resistance 1,2 2
G sa 3.3.1.1.9 s 57.15 saltons Temperature sa 3.3.1.1.13 Detector sa 3.3.1.1.15 5(*)
2 N
sa 3.3.1.1.9 s $7.15 sellons sa 3.3.1.1.13 st 3.3.1.1.15
~
b.
float switch 1,2 2
G SR 3.3.1.1.13 s 57.15 saltons
$R 3.3.1.1.15 5(a) 2 N
SR 3.3.1.1.13 s 57.15 pattons sa 3.3.1.1.15 8.
Turbine stop t
RTP 4
E sa 3.3.1.1.9 s 101 eiosed valve - Closure SR 3.3.1.1.11 st 3.3.1.1.13 sa 3.3.1.1.1%
gJb st 3.3.1.1.16 9.
Turbine Controt Vatye t SOS RTP 2
E sa 3.3.1.1.9 2 600 psig Fast closure, trip 011 sk 3.3.1.1.11 Pressure - Low SR 3.3.1.1.13 sa 3.3.1.1.15
'N st 3.3.1.1.16
- 10. Reactor Mode switch -
1,2 2
G sa 3.3.1.1.12 NA
~
shutdown Position st 3.3.1.1.15 5(a) 2 N
st 3.3.1.1.12 NA 54 3.3.1.1.15
- 11. Manual scram 1,2 2
C sa 3.3.1.1.5 NA sa 3.3.1.1.15 5(*)
2 N
sa 3.3.1.1.5 u
SR 3.3.1.1.15 (a) With Pny contret rod withdrsun from a core cent containing one or mort fust asseetles.
V(
HATCH UNIT 2 3.3-9 Amendment No. 135 1
i
EOC-RPT Instrumentation 3.3.4.1
}.3.3 INSTRUMENTATION.
3.3.4.1 End of Cycle Recirculation Pump. Trip. (EOC-RPT) Instrumentation n
LCO 3.3.4.1-a.
Two channels per trip system for each EOC-RPT instrumentation Function listed below shall be OPERABLE:
1.
Turbine Stop Valve (TSV) -- Closure; and 2.
Turbine Control 1 Valve (TCV) Fast Closure, Trip.
Oil Pressure -- Low.
t QB b.
LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR),". limits for inoperable EOC-RPT as specified in the COLR are made s
applicable.
APPLICABILITY:
THERMAL POWE R
i-ACTIONS
...........................----------NOTE-------------------------------------
Separate Condition entry is allowed for each channel.
b CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more channels A.1 Restore channel to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable.
OPERABLE status.
QR A.2
NOTE---------
.Not applicable if L
inoperable channel is the result of an inoperable breaker.
+
Place channel in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
trip.
(continued) j.
i HATCH UNIT 2 3.3-28 Amendment No. 135 4
+4 w
awwt's ez ase.>,-*--Ew-e eF1-
EOC-RPT Instrumentation 3.3.4.1 ACTIONS -(continued)-
CONDITION ~
REQUIRED ACTION COMPLETION TIME 1
B.
One or more Functions B.)
Restore EOC-RPT trip 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s-with EOC-RPT trip capability.-
capability not maintained.
QB AND B.2 Apply the MCPR limit 2 houis for inoperable 2
MCPR limit for EOC-RPT as specified inoperable EOC-RPT in the COLR.
not made applicable.
-C.
Required Action and C.1 Remove the associated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion recirculation pump Time not met.
from service.
QB -
C.2 Reduc ERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
- to <
TR.
t% )
v i
SURVEILLANCE REQUIREMENTS
...................................... NOTE-------------------------------------
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains EOC-RPT1 trip capability.
4 SURVEILLANCE FREQUENCY SR 3.3.4.1.1
' Perform CHANNEL FUNCTIONAL TEST.
92 days (continued) i 4
HATCH UNIT.2 3.3-29 Amendment-No. 135 i
4
-.. ~, -
-n v-,,v,-
.,.,anL-<
ve
,n.,
s 0 -
e --
~~~,
,m
EOC-RPT Instrumentation 3.3.4.1
(
SURVEILLANCE REOUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.4.1.2 Verify TSV - Closure and TCV Fast 18 months
-Closure. Trip Oil Pressure - Low Functions ot bypassed when THERMAL POWER is TR.
11.}
SR 3.3.4.1.3 Perform CHANNEL CALIBRATION, The 18 months-Allowable Values shall be:
TSV - Closure: s 10% closed; and TCV Fast Closure, Trip 011 4
Pressure - Low: 2: 600 psig.
SR 3.3.4.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.
NOTE-------------------
Breaker interruption time may be assumed from the most recent performance of SR 3.3.4.1.6.
Verify the E0C-RPT SYSTEM RESPONSE TIME 18 months on a is within limits.
STAGGERED TEST BASIS SR 3.3.4.1.6 Determine-RPT breaker interruption-time.
60 months
-G HATCH UNIT 2 3.3-30 Amendment No. 135
N i
I Y~ 3DLLk6 M M 7
M b
NPk ("TClLd6 r
["
- M _I(@tCRTF).
- te 6
\\L M (D 1400 INITIAL RTndt VALUES ARE l
l
-50*F FOR BELTLINE t
26*F FOR UPPER VESSEL, 1300 AND
'l 50*F FOR BOTTOM HEAD BELTINE CURVES 1200 i
I ADJUSTED AS SHOWN:
EFPY SHIFT (*F)
I il 1100 32 127
.P l
HEATUP/COOLDOWN
$ 1000 l
RATE C3 5
20'F/HR FOR CURVE A, 4
i 100*F/HR FOR CURVES B&C 900 I
A', B', C' CORE BELTLINE t-d I
A, B, C NON-BELTLINE 800 g
l Oy l
A - PRESSURE TEST WITH g
FUEL IN THEVESSEL o
700 fi l
i i
I i
l B NON NUCLEAR b
i HEATUP/COOLDOWN h
600 i
i 1
CORE NOT CRITICAL g
C - NON-NUCLEAR h
~
HEATUP/COOLDOWN CORE CRITICAL E
400 NON-BELTLINE E
5 A
i l
BELTLINE AT 32 300 1312 PSIGl EFPY I
CURVES A', B*,C' l
ARE VAUD UP TO 32 EFPY 200 OF OPERATION CURVES A. B, C j
l l
ARE VAUD UP TO 28 EFPY i
100 OF OPERATION
}
FOR BELTUNE AND EOL FOR NON-BELTUNE O
O 50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)
- v Figure 1.1. P-T Curve for Hat alid to 32 EFPY
RCS P/T Limits
[
3.4.9 1
O musce1,,3uscotus %,.
1600 f
\\
/
\\
/
A' A s
1400
\\
'l
/
~
1 N !/
/
- 1200-
/
h
\\
/
I r 1000
! l l
/
x /
x CORE BELTUNE w
mR _ E -Em 800 sHwT MOW AN INmAL WELD RTeenOF -50*F i
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A SY5 TEM HYDROTEST UWIT g 600 WTH FUEL IN VE3SEL p
t:
3
[
4
- vessel DISCONTINUffY f400 uWeg
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= = CORE BELTLINE WTTH g
=Em rr g
m.
E 2M soLTu/
CURVE N 15 NOT UumWC i
so*r/
FOR INFORMATION ONLY\\
/
CURVE A 15 VAUD FOR
\\
M EFPY OF OPERATION
\\
i
.0
'O 100 200 300 400 500 600 WNIMUM REACTOR VESSEL METAL TEMPERATURE fr)
Figure 3.4.9-1 (page 1 of 1)
Pressure / Temperature Limits for.
Inservice Hydrostatic and Inservice Leakage Tests 1
HATCH UNIT 2 3.4-25 Amendment No. 135
F L
RCS P/T Limits f
3.4,9 p
54ttW (dYTR-b b 609/6 AVACfN.
V
\\
\\
\\
B' B
/
1400
/
1
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ll i
- 1200 il
/
s 8
\\
=
x.,
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e
/
s' - CORE sELTUNE N
/
ArrER AssuutD s2 Em 500 N
shirt FROM AN INftML g
WELD RTa7 -50*F a
NoN HEATUP/
600 n
/
3
- VE55 DSCONnNUffY w
uuris g 400
}
s"
~
E E
su rse E
/ /
200
[
CURVE s' is Not UM so 30, FOR INFORMATION ONLY CURVE B 15 VAUD FOR 0
4 0
100 200 300 400 500 600 14NIMUM REACTOR VE5SEL METAL TEMPERATURE fri i
Figure 3.4.9-2 (page 1 of 1)
Pressure / Temperature Limits for Non-Nuclear Heatup, Q]
[
Low Power Physics Tests, and Cooldown Following a Shutdown 4
-~
HATCH UNIT 2 3.4-26 Amendment No. 135 5
_/
i RCS P/T Liaits 3.4.9 t')E k l W O N k b l
ObdVE ATI Acth'D, 1600
\\
s C' C 1400
\\
/,
~
N
/
1200'
'/dfhk '
\\
l
/
l
\\
l h1000 xl
,e.
/
l
,j
\\ /
c' - core BtLTUNE n
N AFTER ASSUMED S2 EFPY g 800 sNirt rROM AN INTTML
,/
wtto %Or -soar E 600
'uu I#"'# " )
N e
E
--- Vts5EL DSCONTINUffY 8.a uurts N h.i
/_
l
= = CORE BELTLINE WTTH m pse
/
J s2 Erry sarT B
/
x 0
I CURVE C' 5 NOT UMm*;
gog
/
FOR INrORETION ONLY a
so r
/
CURVE c s VAuo rOR
/
32 EFPY OF OPERATION O
^
e e
0 100 200 300 400 500 600 WNIMUM REACTOR VE3SEL METAL TEMPERATURE fr)
Figure 3.4.9-3 (page 1 of 1)
Pressure / Temperature Limits for Criticality t
's HATCH UNIT 2 3.4-27 Amendment No. 135
/
Pr;gra s and Manuals 5.5 (J) 5.5 Programs and Manuals 5.5.11 Technical Soecificationi (TS) Bases Control Proaram (continued) d.
Proposed changes that meet the criteria of b. above shall be reviewed and approved by the HRC prior to implementation.
Changes to the Bases impicmented without prior NRC approval shall be provided to the NRC on a fraquency consistent with 10 CFR 50.71(e).
5.5.12 Primary Containment Leakaae Rate Testina Proaryl A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appenoix J, Option B, as modified by approved exemptions.
This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.
The peak calculated primary containment inte essure for the design basis loss of coolant accident, P,, is The maximum allowable primary containment leakage ra e, L,, at P, is 1.2% of primary containment air weight per day.
{,)
Leakage rate acceptance criteria are:
a.
Primary containment overall leakage rate acceptance criterion is s 1.0 L,.
During the first unit startup following testing in accordance with this program, the leakags rate acceptance criteria are s 0.60 L for the combined Type B and Type C tests, and s 0.75 [, for Type A tests; b.
Air lock testing acceptance criteria are:
1)
Overall air loa leakage rate is s 0.05 L, when tested at s P,,
2)
For each door, leakage rate is s 0.01 L, when the gap between the door seals is prsssurized to 2 10 psig for at least 15 minutes.
The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.
[
(continued)
Li HATCH UNIT 2 5.0-16a Amendment No. 141 l
l l
Unit 1 l
Hevised 11ases Pages and Corresponding A1arked-Up Pages i
i d
4 A
1 r
,. ~ -
,y,
--n-,
,ew
,,.,,,,,u+,,-.,.---,.-,...-,,--,-,..,n.,
. - - -, _. ~ _ _ _ _.. - - _. _. _. _ _ _ _
i RPS Instrumentation B 3.3.1.1 O
BASES V
APPLICABLE The trip setpoints are then determined accounting for the SAFETY ANALYSES, remaininginstrumenterrors(e.g., drift).
The trip LCO, and setpoints derived in this manner provide adequate protection APPLICABILITY because instrumentation uncertainties, process effects, (continued) calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.
The OPERABILITY of scram pilot valves and associated solenoids, backup scram valves, and SDV valves, described in the Background section, are not addressed by this LCO.
The individual functions are required to be OPERABLE in the MODES or other specified conditions specified in the Table, which may require an RPS trip to mitigate the consequences of a design basis accident or transient.
To ensure a reliable scram function, a combination of functions are required in each MODE to provide primary and diverse initiation signals.
The only MODES specified in Table 3.3.1.1-1 are MODES 1 (which encompasses a: 28% RTP) l and 2, and MODE 5 with any control rod withdrawn from a core O
cell containing one or more fuel assemblies.
No RPS function is required in MODES 3 and 4 since all control rods are fully inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (LCO 3.3.2.1) does not allow any control rod to be withdrawn.
In MODE 5, control rods withdrawn from a core cell containing no fuoi assemblies do not affect the reactivity of the core and, therefore, are not required to have the capability to scram.
Provided all other control rods remain inserted, no RPS Function is required, in this condition, the required SDM (LCO 3.1.1) and refuel position one-rod-out interlock (LC0 3.9.2) ensure that no event requiring RPS will occur.
The specific Applicable Safety Analyses LCO, and Applicability discussions are listed below on a Function by function basis.
Intermediate Ranae Monitor (IRM)
-1.a.
Intermediate Ranae Monitor Neutron Flux - Hiah The IRMs monitor neutron flux levels from the upper range of the source range monitor (SRM) to the lower range of the I
(Continued)
HATCH UNIT 1 B 3.3-4 PROREV EXTPWR - 7/24/97 e-e
,y
-y-em,.
g e
-,imer+=n-%ww.,w-e,a-vy-p-
+-,i.--t'**'
CN'-M ' - ' * = -
- --?
?
-v 1"5 v-w--'
RPS Instrumentation B 3.3.1.1 V]
(
BASES APPLICABLE 7.a. and 7.b.
Scram Discharae Volume Water level - Hiah SAFETY ANALYSES, (continued)
LCO, and APPLICABillTY four channels of each type of Scram Discharge Volume Water Level - High function, with two channels of each type in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from these functions on a valid signal.
These functions are required in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
At all other times, this function may be bypassed.
8.
Turbine Stoo Valve - Closure Closure of the TSVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.
Therefore, a reactor scram is initiated on a TSV-Closure signal before the TSVs are completely closed in anticipation of the transients that would result from the closure of these valves.
The Turbine Stop Valve - Closure function is the primary scram signal v
for the turbine trip event analyzed in Reference 2.
For this event, the reactor scram reduces the amount of energy recuired to be absorbed and, along with the actions of the Enc of Cycle Recirculation Pump Trip (E00-RPT) System, ensures that the MCPR SL is not exceeded.
Turbine Stop Valve - Closure signals are initiated from position switches located on each of the four TSVs. Two independent position switches are associated with each stop valve. One of the two switches provides input to RPS trip system A; the other, to RPS trip system B.
Thus, each RPS trip system receives an input from four Turbine Stop Valve - Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure function is such that three or more TSVs must be closed to produce a scram.
In addition, certain combinations of two valves closed will result in a half-scram.
This function must be enabled at THERMAL POWER a 28% RTP, This is l
normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this function.
(continued)
()
HATCH UNIT 1 B 3.3-16 PROREV AMEND. 205/EXTPWR - 7/24/97
RPS Instrumentation B 3.3.1.1 O
BASES Q
APPLICADLE 8.
Turbine Sten Valve - Closure (continued)
SAFETY ANALYSES, LCO, and.
The Turbine Stop Valve - Closure Allowable Value is selected APPLICABILITY to be high enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient.
i Eight channels of Turbine Stop Valve - Closure function, with four channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this function if the TSVs should close.
This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is a 28% RTP.
This function is not required when THERMAL POWER is < 28% RTP since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High Functions are adequate to maintain the necessary safety margins.
9.
Turbine Corft$k,b;e Fast Closure. Trio 011 Pressure - Low fast closure of the TCVs results in the loss of a heat sink O
that produces reactor pressure, neutron flux, and heat flux transients that must be limited.
Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves.
The Turbine Control Valve Fast Closure, Trip 011 Pressure - Low function is the primary scram signal for the generator load rejection event analyzed in Reference 2.
For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the EOC-RPT System, ensures that the MCPR SL is not exceeded.
Turbine Control Valve Fast Closure Trip 011 Pressure - Low signals are initiated by the electrohydraulic control (EHC) fluid pressure at each control valve.
One pressure transmitter is associated with each control valve, and the signal from each transmitter is assigned to a separate RPS logic channel. This function must be enabled at THERMAL POWER h 28% RTP.
This is normally accomplished l
automatically by )ressure transmitters sensing turbine first
-stage pressure; t1erefore, opening of the turbine bypass valves may affect this function.
O (continued)
O HATCH UNIT 1 B 3.3-17 PROREV AMEND. 205/EXTPWR - 7/24/97
I RPS Instrumentation 8 3.3.1.1
?
BASES l
APPLICABLE 9.
Turbine Control Valve Fast Closure. Trio 011 i
SAFETY ANALYSES, Pressure - Low (continued) i LCO, and j
APPLICA8ILITY The Turbine Control Valve Fast Closure. Trip 011 l
' Pressure - Low Allowable Value is selected high enough to detect tuninent TCV fast closure.
Four channels of Turbine Control Valve Fast closurc, Trip 4
011 Pressure - Low Function with two channels in each trip system arranged in a one-out-of-two logic are required to be f
OPERABLE to ensure that no single instrument failure will i
preclude a scram from this function on a valid signal.
This Function is required, consistent with the analysis
-assumptions, whenever THERMAL POWER is a 28% RTP. This i
Function is not required when THERMAL POWER is < 28% RTP, i-since the Reactor Vessel Steam Dome Pressure - High and the
-Average Power Range Monitor Neutron Flux - High Functions are adequate to maintain the necesst.ry safety margins.
i 10.
Reactor Mode Switch - Sh'stdown Position The Reactor Mode Switch.- Shutdown Position Function O
provides signals, via th) manual scram logic channels, 4
directly to the scram pilot solenoid power circuits.
These manual scram logic channels are redundant to the automatic i
protective instrumentation channels and provide manual reactor trip capability. This function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required 4
by the NRC approved licensing basis.
The reactor mode switch is a single switch with two channels, each of which provides input into one of the RPS manual scram logic channels.
There is no Allowable Value for this function, since the channels are mechanically actuated based solely on re,ctor mode switch position.
Two channels of Reactor Mode Switch - Shutdown Position Function, with one channel in each-manual scram--trip system, are available and required to be OPERABLE.- The Reactor Mode r
l (continued)
HATCH UNIT 1 8 3.3-18 PROREV ANEND. 205/EXTPWR - 7/24/97
- _ _ -..-==_
RPS Instrumentation B 3.3.1.1 i
BASES SURVEILLANCE SR 3.3.1.1.11 REQUIREMENTS (continued)
This SR ensures that scrams initiated from the Turbine Stop Valve - Closure and Turbine Control Valve fast Closure, Trip 011 Pressure - Low functions will not be inadvertently bypassed when THERMAL POWER is a 28% RTP.
This involves l
calibration of the bypass channels.
Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint.
Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is i
derived from turbine first stage pressure), the main turbine bypass valves must remain closed during the calibration at THERMAL POWER m 28% RTP to ensure that the calibration is l
valid.
If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at 2 28% RTP, either due l
to open main turbine bypass valve (s or other reasons), then theaffectedTurbineStopValve-C)losureandTurbine Control Valve fast Closure, Trip 011 Pressure - Low functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass),
ifplacedinthenonbypasscondition(Turbine g
Stop Valve - Closure and Turbine Control Valve Fast Closure, s
Trip 011 Pressure - Low functions are enabled), this SR is met and the channel is considered OPERABLE.
The Frequency of 184 days is based on engineering judgment and reliability of the components.
SR 3.3.1.1.13 A CHANNEL CAllBRATION is a complete check of the instrument loop and the sensor.
This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CAllBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology.
For MSIV - Closure,.SDV Water Level - High (float Switch), and TSV - Closure Functions, this SR also includes a physical inspection and actuation of the switches.
For the APRM Simulated Thermal Power - High function, this SR also includes calibrating the associated recirculation loop flow channel.
O (continued)
\\
HATCH UNIT 1 B 3.3-29 PROREV AMEND. 250/EXTPWR - 7/24/97
feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 i
B 3.3 INSTRUMENTATION B 3.3.2.2 feedwater and Main Turbine High Water Level Trip Instrumentation BASES 1
BACKGROUND The feedwater and main turbine high water level trip instrumentation is designed to detect a potential failure of i
the feedwater Level Control System that causes excessive feedwater flow.
With excessive feedwater flow, the water level in the reactor vessel rises toward the high water level setpoint, i
causing the trip of the two feedwater pump turbines and the I
main turbind.
Reactor Vessel Water Level - High signals are provided by level sensors that sense the difference between the pressure 1
due to a constant column of water (reference leg) and the pressure due to the actual water level in the reactor vessel (variableleg). Three channels of Reactor Vessel Water Level - High instrumentation are provided as input to a two-out-of-three initiation logic that trips the two O
feedwater pump turbines and the main turbine.
The channels kj include electronic equipment (e.g., trip relays) that t
compare measured input signals with pre-established 3
setpoints. When the set)oint is exceeded, the channel output relay actuates, witch then outputs a main feedwater and turbine trip signal to the trip logic.
A trip of the feedwater pump turbines limits further increase in reactor vessel water level by limiting further addition of feedwater to the reactor vessel. A trip of the main turbine and closure of the stop vaives protects the turbine from damage due to water entering the turbine.
APPLICABLE-The feedwater and main turbine high water level trip SAFETY ANALYSES instrumentation is assumed to be capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref.1).
The high level trip indirectly initiates a reactor scram from the main turbine trip.(above 28% RTP) and trias the l
feedwater pumps, thereby terminating the event.
T1e reactor scram mitigates the reduction in MCPR.
f3 (continued)
Q,)
HATCH UNIT 1 B 3.3-54 PROREV EXTPWR - 7/24/97
l EOC-RPT Instrumentation B 3.3.4.1 O
BASES V
BACKGROUND both recirculation pumps will trip.
There are two E0C-RPT (continued) breakers in series per recirculation pump. One trip system trips one of the two EOC-RPT breakers for each recirculation pump, and the second trip system trips the other E0C-RPT breaker for each recirculation pump.
APPLICABLE The TSV - Closure and the TCV Fast Closure, Tri) 011 SAFETY ANALYSES, Pressure - Low functions are designed to trip tie LCO, and recirculation pumps in the event of a turbine trip or APPLICABILITY generator load rejection to mitigate the increase in neutron flux, heat flux, and reactor pressure, and to increase the margin to the MCPR SL.
The analytical methods and assumptions used in evaluating the turbine trip and generator load rejection are summarized in References 2 and 3.
To mitigate pressurization transient effects, the E0C-RPT must trip the recirculation > umps after initiation of closu:e movement of either tie TSVs or the TCVs. The combined effects of this trip and a scram reduce fuel bundle O
power more rapidly than a scram alone, resulting in an increased margin to the MCPR SL. Alternatively, MCPR limits for an inoperable EOC-RPT, as specified in the COLR, are sufficient to prevent violation of the MCPR Safety Limit.
The EOC-RPT function is automatically disabled when turbine first stage pressure is < 28% RTP.
l EOC-RPT instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref. 6).
The OPERABILITY of the E0C-RPT is dependent on the OPERABILITY of the individual instrumentation channel functions.
Each Function must have a required number of OPERABLE channels in each trip system, with their setpoints within the specified Allowable Value of SR 3.3.4.1.3.
The set)oint is calibrated consistent with applicable setpoint met 1odology assumptions (nominal trip setpoint).
Channel OPERABILITY also includes the associated EOC-RPT breakers.
Each channel (including the associated EOC-RPT breakers) must also respond within its assumed response time.
O (continued)
U HATCH UNIT 1 B 3.3-80 PR0REV EXTPWR - 7/24/97
l EOC-RPT Instrumentation B 3.3.4.1 BASES V
APPLICABLE Turbine Ston Valve - Closure SAFETY ANALYSES, t
LCO. and Closure of the TSVs and a main turbine trip result in the APPLICABillTY loss of a heat sink and increases reactor pressure, neutron (continued) flux, and heat flux that must be limited.
Therefore, an RPT is initiated on a TSV - Closure signal before the TSVs are completely closed in anticipation of the effects that would result from closure of these valves.
[00-RPT decreases reactor powcr and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient.
Closure of the TSVs is determined by measuring the position of each valve. While there are two separate position switches associated with each stop valve, only the signal from one switch for each TSV is used, with each of the four channels being assigned to a separate trip channel.
The logic for the TSV - Closure function is such that two or more TSVs must be closed to produce an EOC-RPT.
This function must be enabled at THERMAL POWER a 28% RTP.
This l
i is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect
/3 this function.
Four channels of TSV - Closure, with two V
channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this function on a valid signal.
The TSV - Closure Allowable Value is selected to detect imminent TSV closure.
This protection is required, consistent with the safety analysis assumptions, whenever THERMAL POWER is a 28% RTP.
Below 28% RTP, the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor (APRM) Neutron Flux -
High functions of the Reactor Protection System (RPS) are adequate to maintain the necessary margin to the MCPR Safety Limit.
luthipe Control Valve Fast Closure. Trio Oil Pressure -low fast closure of the TCVs during a generator load rejection results in the loss of a heat sink that produces reactor presscre, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on-TCV Fast Closure. Tri) Oil Pressure - Low in anticipation of the transients t1at would result from the closure of these (continued)
HATCH UNIT 1 B 3.3-82 PROREV AMEND. 205/EXTPWR - 7/24/97
EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE lurbine Control Valve Fast Closure. Trin Oil Pressure - Low SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY valves. The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient.
Fast closure of the TCVs is determined by measuring the electrohydraulic control fluid pressure at each control valve. There is one pressure transmitter associated with each control valve, and the signal from each transmitter is assigned to a separate trip channel.
The logic for the TCV Fast Closure. Trip 011 Pressure - Low Function is such that two or more TCVs must be closed (pressure transmitter trips) to produce an EOC-RPT.
This Function must be enabled at THERMAL POWER a: 28% RTP.
This is normally accompitshed l
automatically by )ressure transmitters sensing turbine first stage pressure; t1erefore, opening of the turbine bypass valves may affect this function.
Four channels of TCV Fast Closure, Trip 011 Pressure - Low, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an s
EOC-RPT from this ' unction on a valid signal.
The TCV Fast Closure. Trip 011 Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure.
This protection is required consistent with the safety analysis whenever THERMAL POWER is 2 28% RTP.
Below 28% RTP, the Reactor Vessel Steam Dome Pressure - High and the APRM Neutron Flux - High Functions of the-RPS are adequate to maintain the necessary margin to the MCPR Safety Limit.
ACTIONS A Note has been provided to modify the ACTIONS related to EOC-RPT instrumentation channels.
Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section-l.3 also specifies that Required
- Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable EOC-RPT instrumentation channels provide (d
(continued)
HATCH UNIT 1 B 3.3-83 PROREV AMEND. 205/EXTPWR - 7/24/97
E0C-RPT Instrumentation B 3.3.4.1 BASES ACTIONS B.1 and B.2 (continued)
]
capability when sufficient channels are OPERABLE or in trip, such that the EOC-RPT System will generate a trip signal from the given function on a valid signal and both recirculation pumps can be tripped. Alternately, Required Action B.2 requires the MCPR limit for inoperable EOC-RPT, as specified in the COLR, to be applied.
This also restores the margin to MCPR assumed in the safety analysis.
The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Com)1etion Time is sufficient time for the o)erator to ta(e corrective action, and takes into account 11e likelihood of an event requiring actuation of the E0C-RPT instrumentation during this period, it is also consistent with the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time provided in LCO 3.2.2 for Required Action A.1, since this instrumentation's purpose is to preclude a MCPR violation.
C.1 and C.2 With any Required Action and associated Completion Time not met, THERMAL POWER must be reduced to < 28% RTP within l
{]N 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Alternately, the associated recirculation pump may be removed from service, since this performs the intended function of the instrumentation.
The allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience, to reduce THERMAL POWER to < 28% RTP from full l
power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains E0C-RPT trip capability, Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 4) assumption of the average time required to perform channel Surveillance.
That (continued)
C HATCH UNIT 1 B 3.3-85 PROREV EXTPWR - 7/24/97
l l
EOC-RPT Instrumentation B 3.3.4.1
[]
BASES U
SURVEILLANCE analysis demonstrated that the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> testing allowance REQUIREMENTS does not significantly reduce the probability that the (continued) recirculation pumps will trip when necessary.
SR 3.3.4.1.1 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Frequency of 92 days is based on reliability analysis of Reference 4.
SR 3.3.4.1.2 This SR ensures that an EOC-RPT initiated from the TSV - Closure and TCV Fast Closure, Trip Oil Pressure - Low functions will not be inadvertently bypassed when THERMAL r
POWER is a 28% RTP.
This involves calibration of the bypass l
())
channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint.
Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from first stage pressure) the main turbine bypass valves must remain closed during the calibration at THERMAL POWER m 28% RTP to ensure l
that the calibration is valid, if any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at a 28% RTP, either due to open main turbine
{
bypass valves or other reasons), the affected TSV - Closure and TCV Fast Closure, Trip 011 Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass),
if placed in the nonbypass condition (Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure-Low Functions are enabled), this SR is met with the channel considered OPERABLE.
The Frequency of 184 days is based on engineering judgment and reliability of the components.
(3 (continued)
Q)
~
HATCH UNIT 1 B 3.3-86 PR0REV EXTPWR - 7/24/97
{
Pri::ary Containment Isolation Instrumentation B 3.3.6.1 O
BASES G
APPLICABLE 1.c.
Main Steam line Flow - Hiah (continued)
SAFETY ANALYSES, LCO, and The MSL flow signals are initiated from 16 transmitters that APPLICABILITY are connected to the four MSLs.
The transmitters are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to detect the high flow.
Four channels of Main Steam Line flow-High function for each unisolated MSL (two channels per trip system) are available and are required to be OPERABLE so that no single instrument failure will preclude detecting a break in any individual MSL.
The Allowable Value is chosen to ensure that offsite doso limits are not exceeded due to the break.
The Allowable Value corresponds to s 150 psid, which is the parameter l
monitored on control room Instruments.
This function isolates the Group 1 valves.
j bd.
Condenser Vacuum - Low The Condenser Vacuum - Low Function is provided to prevent O
overpressurization of the main condenser in the event of a loss of the main condenser vacuum.
Since the integrity of the condenser is an assumption in offsite dose calculations, the Condenser Vacuum - Low function is assumed to be OPERABLE and capable of initiating closure of the MSIVs.
The closure of the MSIVs is initiated to prevent the addition of steam that would lead to additional condenser pressurization and possible rupture of the diaphragm installed to protect the turbine exhaust hood, thereby preventing a potential radiation leakage path following an accident.
Condenser vacuum pressure signals are derived from four pressure transmitters that sense the pressure in the condenser.
Four channels of Condenser Vacuum - Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
The Allowable Value is chosen to prevent damage to the condenser due to pressurization, thereby ensuring its integrity for offsite dose analysis. As noted (footnote (a) to Table 3.3.6.1-1), the channels are not required to be (continued) 3 HATCH UNIT 1 B 3.3-155 PR0REV EXTPWR - 7/24/97
Pritary Containment i
8 3.6.1.1 BASES (continued)
(O APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.
The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA.
In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.
Analytical methods and assumptions involving the primary containment are presented in References 1 and 2.
The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABillTY of the prit ary containment ensures that the leakage rate assumed in the safety analyses is not exceeded.
The maximum allowable leakage rate for the primary Q
containment (L ) is 1.2% by weight of the containment air per24hoursalthedesignbasisLOCAmaximumpeak
!,Ly containment pressure (P,) of 50.5 psig (Ref.1).
l Primary containment satisfies Criterion 3 of the NRC Policy Statement (Ref. 4).
LCO Primary containment OPERABILITY is maintained by limiting performing a r, except prior to the first startup after leakage to s L equired Primary Containment Leakage Rate Testing Program (Ref. 5) leakage test. At this time, applicable leakage limits specified in the Primary Containment Leakage Rate Testing Program must be met.
Compliance with this LC0 will ensure a arimary containment configuration, including equipment hateles, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses.
Individual leakage rates specified for the primary coritainment air lock are addressed in LCO 3.6.1.2.
)
(continued) w/
HATCH UNIT 1 B 3.6-2 PROREV EXTPWR - 7/24/97
Primary Containment Air Lock B 3.6.1.2 BASES I
BACKGROUNO containment leakage rate to within limits in the event of a (continued)
DBA.
Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the unit safety analysis.
APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA.
In the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled (
the rate of primary containment leakage.
The primary containment is dtsigned with a maximum allowable leakage rate (L ) of 1.2%
b.rweightofthecontainmentairper24hoursatthe calculated design basis LOCA maximum peak containment pressure (P ) of 50.5 psig (Ref. 2). This allowable leakage l
rateformsthebasisfortheacceptancecriteriaimposedon the SRs associated with the air lock.
Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and O
pressurire the secondary containment.
The primary containment air lock satisfies Criterion 3 of the NRC Policy Statement (Ref. 4).
LCO As part of primary containment, the air lock's safety function is related to control of containment leakage rates following a DBA.
Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.
The primary containment air lock is required to be OPERABLE.
For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Ty)e B air lock leakage test, and both air lock doors must
)e OPERABLE.
The interlock allows only one air lock door to be opened at a time.
This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be r
(continued)
HATCH UNIT 1 B 3.6-7 PROREV EXTPWR - 7/24/97
Drywell Pressure B 3.6.1.4
[')
B 3.6 CONTAINMENT SYSTEMS B 3.6.1.4 Drywell Pressure BASES BACKGROUND The drywell pressure is limited during normal operations to preserve the initial conditions assumed in the accident analysis for a Design Basis Accident (DBA) or loss of coolant accident (LOCA).
APPLICABLE Primary containment performance is evaluated for the entire SAfLTY ANALYSES spectrum of break sizes for postulated LOCAs (Ref. 1).
Among the inputs to the DBA is the initial primary containment internal pressure (Ref. 1).
Analyses assume an initial drywell pressure of 1.75 psig.
This limitation ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that the peak LOCA drywell internal pre'isure does not exceed the maximum allowable of 62 psig.
The maximum calculated drywell pressure occurs during the (Q'/
reactor blowdown phase of the DBA, which assumes an instantaneous recirculation line break.
The calculated peak drywell pressure for this limiting event is 50.5 psig l
(Ref. 1).
Drywell pressure satisfies Criterion 2 of the NRC Policy Statement (Ref. 2).
LCO in the event of a DBA, with an initial drywell pressure
- s; 1.75 psig, the resultant peak drywell accident pressure will be maintained below the drywell design pressure.
APPLICABillTY In MODES 1, 2, and 3, a DF.A could cause a release of radioactive material to primary containment.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining drywell pressure within limits is not required in MODE 4 or 5.
,m
(
)
(continued) q,/
HATCH UNIT 1 B 3.6-29 PROREV EXTPWR - 7/24/97
Main Condenser Offgas B 3.7.6 BASES LCO with this requirement (2436 MWt x 100 Ci/MWt-second =
(continued) 240 mci /second).
The 240 mci /second limit is conservative for a rated core thermal power of 2763 MWt.
l APPLICABILITY The LCO is applicable when steam is being exhausted to the main condenser and the resulting noncondensables are being processed via the Main Condenser Offgas System. This occurs during MODE 1, and during MODES 2 and 3 with any main steam line not isolated and the SJAE in operation, in MODES 4 and 5 steam is not being exhausted to the main condenser and the requirements are not applicable.
ACTIONS Ad if the offgas radioactivity rate limit is exceeded, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> i
is allowed to restore the gross gamma activity rate to within the limit. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on engineering judgment, the time required to complete the Required Action, the large margins O
associated with permissible dose and exposure limits, and the low probability of a Main Condenser Offgas System i
rupture.
1 B.I. B.2. B.3.1. and B.3.2 1
If th6 gross gamma activity rate is not restored to within the limits in the associated Completion Time, all main steam lines or the SJAE must be isolated.
This isolates the Main Condenser Offgas System from the source of the radioactive steam. The main steam lines are considered isolated if at least one main steam isolation valve in each main steam line is closed, and at least one main = steam line drain valve in the drain line is closed. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time is reasonable, based on operating experience, to perform the actions from full power conditions in an orderly manner and without challenging unit systems.
An alternative to Required Actions B.1 and B.2 is to place the unit in a MODE in which the LCO does not apply.
To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The (continued)
HATCH UNIT l-B 3.7-34 PROREV EXTPWR - 7/24/97
Main Turbine Bypass Systea 6 3.7.7 8 3.7 PLANT SYSTEMS 8 3.7.7 Main Turbine Bypass System BASES BACKGROUND The Main Turbine Bypass System is designed to control steam pressure when reactor steam generation exceeds turbine requirements during unit startup, sudden load reduction, and i
i cooldown.
It allows excess steam flow from the reactor to the condenser without going through the turbine.
The bypass capacity of the system is approximately 21% of the turbine l
design steam flow.
Sudden load reductions within the capacity of the steam bypass can be accommodated without reactor scram. The Main Turbine Bypass System consists of three valves connected to the main steam lines between the main steam isolation valves and the turbine stop valves.
Each of these three valves is operated by hydraulic cylinders. The bypass valves are controlled by the pressure regulation function of the Turbine Electrohydraulic Control System, as discussed in the FSAR, Scf ion 7.11 (Ref. 1).
The bypast valves are normally closed, and the )ressure regulator controls the turbine control valves t1at direct all steam flow to the turbine.
if the speed governor or the O
load limiter restricts steam flow to the turbine, the pressure regulator controls the system pressure by opening the bypass valves. When the bypass valves open, the steam flows from the by) ass chest, through connecting piping, to the pressure brea(down assemblies, where a series of orifices are used to further reduce the steam pressure before the steam enters the condenser.
APPLICABLE The Main Turbine Bypass System is assumed to function during SAFETY ANALYSES the feedwater controller failure to maximum flow demand as discussed in the FSAR Section 14.3.2.1 (Ref. 2). Opening the bypass valves during the pressurization event (subsequent to the resulting main turbine tria) mitigates the increase in reactor vessel pressure, whic1 affects the MCPR during the event. An inoperable Main Turbine Bypass System may result in an MCOR penalty.
The Main Turbine Bypass System satisfies Criterion 3 of:the NRC Policy Statement (Ref. 4).
1
(
(continued)
HATCH UNIT 1 B 3.7-36 PROREV EXTPWR - 7/24/97 y
..,._.,y.
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..,,_m..
,,.y
- - ~. _ _,.., _. -,,,
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE The trip setpoints are then determined accounting for the SAFETY ANALYSES, remaining instrument errors (e.g., drift). The trip LCO, and setpoints derived in this manner provide adequate protection APPLICABILITY because instrumentation uncertainties, process effects, (continued) cal 1Mation tolerances, instrument drift, and severe envir mental effects (for channels that must function in harsh avironments as defined by 10 CFR 50.49) are accounted for.
The OPERABILITY of scram pilot valves and associated solenoids, backup scram valves, and SDV valves, described in the Background section, are not addressed by this LCO.
The individual Functions are required to be OPERABLE in the MODES or other specified conditions specified in the Table, which may require an RPS trip to mitigate the consequences of a design basis accident or transient.
To ensure a reliable scram function, a combination of Functions are required in each MODE to provide primary and diverse 3
initiation signals. The only MODES specified in rfTo Table 3.3.1.1-1 are MODES 1 (which encompasses P) and 2, and MODE S with any control rod withdrawn rom a core cell containing one or more fuel assemblies.
No RPS
- O Function is required in MODES 3 and 4 since all control rods are fully inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (LCO 3.3.2.1) does not allow any control rod to be withdrawn.
In MODE 5, control rods withdrawn from a core cell containing no fuel assemblics do not affect the reactivity of the core and, therefore, are not required to have the capability to scram.
Provided all other cor. trol rods remain inserted, no R/S Function is required.
In this condition, the required SDN (LCO 3.1.1) and refuel position one-rod-out interlock (LCO 3.9.2) ensure that no event requiring RPS will occur.
The specific Applicable Safety Analyses LCO Applicability discussions are listed below on, and a Function by Function basis.
Intermediate Ranoe Monitor (IRM)
La.
Intermediate Ranae Monitor Neutron Flur - Hioh The IRMs monitor neutron flux levels from the upper range of the source range moniter (SRM) to the lower range of the (continued)
HATCH UNIT 1 8 3.3-4 REVISION 0
l l
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 7.a. and 7.b.
Scram Discharae Volume Water Level - Hioh SAFETY ANALYSES, (continued)
LCO, and APPLICABillTY Four channels of each type of Scram Olscharge Volume Water Levol - High Function, with two channels of each type in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from these Functions on a valid signal.
These Functions are required in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified condittor.s when control rods are withdrawn.
At all other times, this Function may be bypassed.
8.
Turbine Ston Valve - Closure Closure of the TSVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.
Therefore, a reactor scram is initiated on a TSV-Closure signal before the TSVs are
'O completely closed in anticipation of the transients that would result from the closure of these valves.
The Turbine Stop Valve -Closure function is the primary scram signal for the turbine trip event analyzed in Reference 2.
For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Trip (EOC-RPT) System, ensures that the MCPR SL is not exceeded.
Turbine Stop Valve - Closure signals are initiated from position switches located on each of the four TSVs. Two independent position switches are associated with each stop valve. One of the two switches provides input to RPS trip system A; the other, to RPS trip system B.
Thus, each RPS trip s Valve ystem receives an input from four Turbine Stop Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure Function is such that three or more TSVs must be closed to produce a scram.
In addition, certain combinations of two valves closed will result in a hal ram. This function must be enabled at THERMAL POWER RTP. This is normally accomplished automaticall y pressure transmitters sensing turbine first stage press et therefore, openiq r/
the turbine bypass valves may aff ct this Furic%tn.
(
ko V
(continued)
HATCH UNIT I B 3.3-16 PROPe9EfHtEVt5iGN i 16/v6 l
/
bM@M N
l RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 8.
Turbine Ston Valve - Closure (continued)
SAFETY ANALYSES, LCO, and The Turbine Stop Valve - Closure Allowable Value is selected APPLICABILITY to be high enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient.
Ei ht channels of Turbine Stop Valve - Closure Function, wi h four channels in each tri system, are required to be OPERABLE to ensure that no sin le instrument failure will preclude a scram from this function if th SVs should close.
This function is required, cor, st with analysis assumptions, whenever THERMAL POWER 11 k.
is function is not re utred when THERMAL
< )t$ TP since the Reactor essel Steam Dome Pressu e and the Average Power Range Monitor Neutron Flux -
gh Func ons l
are adequate to maintain the necessary safet r ins.
9.
Turbine Control Valve Fast Closure. Trio 011 Pressure - Low =
p Fast closure of the TCVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.
Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves.
The Turbine Control Valve Fast Closure, Trip 011 Pressure - Low function is the primary scram signal for the generator load rejection event analyzed in Reference 2.
For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the EOC-RPT System, ensures that the MCPR SL is not exceeded.
Turbine Control Valve Fast Closure, Vrip 011 Pressure - Low signals are initiated by the electrehydraulic control (EHC) fluid pressure at each control valvs.
One pressure transmitter is associated with each control valve, and the signal from each transmitter is assigned to a separate RPS logic ch e.el.
This function must be enabled at THERMAL POWER =
RTP.
This is normally accomplished automati ly by pressure transmitters sensing turbine first stage pre sure; therefore, opening of the turbine bypass valves m affect this function.
(continued)
HATCH 9 NIT 1 B 3.3 17 naaaasEc an!s!= 7/1 mins hMCND M
RPS Instrumentation B 3.3.1.1
(
BASES APPLICABLE 9.
Turbine Control Valve Fast Closurg. Trio 011 SAFETY ANALYSES, Pressure - Lost (continued)
LCO, and APPLICA8ILITY The Turbine Control Valve Fast Closure Tri) 011 Pressure - Low Allowable Value is selected ligh enough to detect imminent TCV fast closure.
Four channels of Turbine Control Valve Fast closure. Trip 011 Pressure - Low Function with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this function on a valid signal.
This function is required, consistent with the ialysis assumptions, whenever THERMAL POWER is a:l lRTP his Function is not re uired when THERMAL P06 s
- RTP, since the Reactor essel Steam Dome Pressure H
and the Average Power Range Monitor Neutron Flux - Hig Fu ctions l
are adequate to maintain the necessary safety m ns.
10.
Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position function V) provides signals, via the manual scram logic channels, directly to the scram pilot solenoid power circuits. These manual scram logic channels are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability.
This function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
The reactor mode switch is a single switch with two channels, each of which provides input into one of the RPS manual scram logic channels.
There is no Allowable Value for this function, since the channels are mechanically actuated based solely on reactor mode switch position.
Two channels of Reactor Mode Switch - Shutdown Position Function, with one channel in each manual scram trip system, are available and required to be OPERABLE. The Reactor Mode
'O (continued)
HATCH (JNIT 1 8 3.3-18 tROPOSE9-M VIS!" 7/!'/CA-Arne)0D209 erd $f
RPS Instrumentation 8 3.3.1.1 BASES APPLICABLE 9.
Turbine Control Valve Fast Closure. Trio 011 SAFETY ANALYSES, Pressure - Low (continued)
LCO, and i
APPLICABILITY transients that must be limited. Therefore, a reactor scram j
is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these 1
valves. The Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Function is the primary scram signal for the i
generator load rejection erant analyzed in Reference 2.
F6r this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of tlw E0C-RPT System, ensures that the MCPR SL is not exceeded.
Turbine Control Valve Fast Closure, Trip 011 Pressure - Low signals are initiated by the electrohydraulic control (EHC) fluiri pressure at each control valve. One pressure transmitter is associated with each control valve, and the signal from each transmitter is assigned to a separate RPS logic ch 1.
This function must be enabled at THERMAL POWER RTP.
This is normally accomplished automit y by pressure transmitters sensi turbine first 4
g-stage pressure; therefore, operiing of the tur ine bypass valves may affect this Function.
4 The Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Allowable Value is solected high enough to detect imminent TCV fast closure.
Four channels of Turbine Control Valve Fast closure, Trip 011 Pressure - Low Function with two channels in each trip system Arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.
This Function is required, consistent with lysis assumptions, whenever THERMAL POWER is RTP.
is Function is not required when THERMAL is <
- RTP, THee the Reactor vessei steam Dome Pressure - H vn and the Average Power Range Monitor Fixed Neutron Flux -High v..
Functions are adequate to maintain the necessary safety
- margins, 10.
Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position Function provides signals, via the manual scram logic channels, (continued)
HATCH UNIT 1 8 3.3-19 REVISION 0
,-ry,
- -,. -my 3, -. -
97,,ey
,y-%.. = ~~ v.r
. w s.
--mr..,-gewn--.----.=
-.i.w.w-c s.-m 7.,-....e-ve-w...,-a.~,-
-e
RPS Instrumentation B 3.3.1.1 BASES l
SURVEILLANCE SR 3.3.1.1.11 REQUIREMENTS (continued)
This SR ensures that scrans initiated from the Turbine Stop Valve - Closure and Turbine Control Valve Fast closure. Trip Oil Pressure - Low Functions will be inadvertently bypassed when THERMAL POWER is a RTP.
This involves calibration of the bypass channel.
Adequate margins for the instrument setpoint methodologie are incorporated into j-the actual setpoint.
Because main tu ine bypass flow can i
affect this setpoint nonconservatively THERMAL POWER is derived from turbine first stage pressu
), the main turbine bypass valves au emain closed during to calibration at THERMAL POWER k RTP to ensure that th p calibration is gg valid.
if any bypass channel's setpoint is non servative (i.e., the Functions are bypassed at a oRTP, either due to open main turbine bypass valve (s) o.
her reasons), then the affected Turbine Stop Valve - Closure and Turbine i
i Control Valve Fast closure, Trip 011 Pressure - Low Functions are considered inoperable.
Alternatively, the bypass channel can be placed in the conservative condition (nonbypass).
If placed in the nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast closure, i
Trip Oil Pressure - Low functions are enabled), this SR is met and the channel is considered OPERABLE.
j The Frequency of 184 days is based on engineering judgment and reliability of the components.
SR 3.3.1.1.13 l
i A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary 1
range and accuracy.
CHANNEL CAllBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology. For MSIV - Closure, SDV Water Level - High (Float Switch), and TSV - Closure Functions, this SR also includes a physical inspection and actuation of the switches.- For the APRM Simulated Thermal Power - High Function,- this SR also includes calibrating the associated recirculation loop flow channel.
i (continued)
HATCH UNIT'1 g 3.3-29 PROPOSED-AEVIS10N4/16/96^
ME@%CkTdd6
Feedwater and Main Turbine High Water Level Trip Instrumentaticn i
B 3.3.2.2 4
B 3.3
!NSTRUMENTATION B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation BASES BACKGROUND Ihe feedwater and main turbine high water level trip instrumentation is designed to detect a potential failure of the feedwater Level Control System that causes excessive feedwater flow.
J With excessive feedwater flow, the water level in the reactor vessel rises toward the high water level setpoint, causing the trip of the two feedwater pump turbines and the main turbine.
Reactor Vessel Water Level - High signals are provided by t
level sensors that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level in the reactor vessel (variable leg). Three channels of Reactor Vessel Water Level - High instrumentation are provided as input to a two-out-of-three initiation logic that trips the two i
feedwater pump turNoos and the main turbine. The channels include electronic squipment (e.g., trip relays) that compare measured input signals with pre-established setpoints. When the set >oint is exceeded, the channel j
output relay actuates, wtich then outputs a main feedwater and turbine trip signal to the trip logic.
A trip of the feedwater pump turbines limits further increase in reactor vessel water level by limiting further addition of feedwater to the reactor vessel. A trip of the sain turbine and closure of the stop valves protects the turbine from damage due to water entering the turbine.
APPLICABLE The feedwater and main turbine high water level trip SAFETY ANALYSES instrumentation is assumed to be' capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref.1).
1 The high level trip indirectly in teates a reactor scram from the main turbine trip (abov RTP) and trips the feedwater pumps, thereby teminat the event.
The reactor i
scram mitigates the reduction in MC (continued) i HATCH UNIT 1' B 3.3-54' REVISION 0
E0C-RPT Instrumentation l
B 3.3.4.1 BASES BACKGROUND both recirculation pumps will trip. There are two E0C-RPT (continued) breakers in series per recirculation pump. One trip system trips one of the two E0C-RPT breakers for each recirculation pump, and the second trip system trips the other EOC-RPT breaker for each recirculation pump.
APPLICABLE The TSV - Closure and the TCV Fast Closure, Trip 011 SAFETY ANALYSES, Pressure - Low Functions are designed to trip the 4
LCO, and recirculation pumps in the event of a turbine trip or APPLICA8ILITY generator load rejection to mitigate the increase in neutron flux, heat flux, and reactor pressure, and to increase the margin to the MCPR SL. The analytical methods and assumptions used in evaluating the turbine trip and generator load rejection are summarized in References 2 and 3.
To mitigate pressurization transient effects, the E0C-RPT must trip the recirculation pumps after initiation of closure movement of either the TSVs or the TCVs. The combined effects of this wrip and a scram reduce fuel bundle power more rapidly than a scram alone, resulting in an increased margin to the MCPR SL. Alternatively, MCPR limits for an inoperable E0C-RPT, as specified in the COLR, are sufficient to prevent violation of the MCPR Safet) Limit.
The EOC-RPT function is automatically disabled when turbine first stage pressure is < g RTP.
E0C-RPT instrumentation satisfies Cruerion 3 of the NRC PolicyStatement(Ref.6).
The OPERABILITY of the E0C-RPT is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have a required number of OPERABLE channels in each trip system, with their setpoints within the specified Allowable Value of SR 3.3.4.1.3.
The setpoint is calibrated consistent with applicable setpoint mettodology assumptions (nominal trip setpoint). Channel OPERABILI"Y also includes the associated E0C-RPT breakers.
Each channel (including the associated E0C-RPT breakers) must also respond within its assumed response time.
HATCH UNIT 1 8 3.3-80 REVISION 1
EOC-RPT Instrumentation B 3.3.4.1 O
BASES APPLICABLE Turbine Ston Valve - Closun SAFETY ANALYSES, LCO, and Closure of the TSVs and a main turbine trip result in the APPLICABILITY loss of a heat sink and increases reactor pressure, neutron (continued) flux, and heat flux that must be limited. Therefore, an RPT is initiated on a TSV - Closure signal before the TSVs are completely closed in anticipation of the effects that would result from closure of these valves.
EOC-RPT decreases reactor power and aids the reactor scram in ensuring that-the MCPR SL is not exceeded during the worst case transient.
Closure of the TSVs is determined by measuring the position cf each valve. While there are two separate position switches associated with each stop valve, only the signal from one switch for each TSV is used, with each of the four channels being assigned to a separate trip channel. The logic for the TSV - Closure Function is such that two or more TSVs must be closed to produce an EOC-RP
. This Function must be enabled at THERMAL POWER a RTP. This
{
is normally accomplished automatically by pr su transmitters sensing turbine first stage pressure;
(~
therefore, opening of the turbine bypass valves may affect t
this Function.
Four channels of TSV - Closure, with two V
channels in each trip system, are available and required to be OPERABLE to ensure that no single ir.ctrument failure will preclude an EOC-RPT from this Function on a valid signal.
The TSV - Closure Allowable Valus is selected to detect imi..ent TSV closure.
This protection is required, consistent with the s analys ssumptions, whenever THERMAL POWER is a RTP.
Below RTP, the Reactor Vessel Steam Dome Press
- High and verage Power Range Monitor (APRM) Neutron Flux --
l gg High Functions of the Reactor Protection System (RPS) are adequate to maintain the necessary margin to the MCPR Safety Limit.
Turbine Contrp1 Valve Fast tlosure. Trio 011 Pressure -low Fast closure of the TCVs during a generator load rejection results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on TCV Fast closure, Trip 011 Pressure - Low in anticipation of the transients that would result from the closure of these
/N (continued)
HATCH UNIT I B 3.3-82 PROPOSED REVISION 7/16/96 (UtT4M kb
l'
[
EOC-RPT Instrumentation B 3.3.4.1
- BASES APPLICABLE Turbine Control Valve Fast Closure. Trin Oil Pressure - Low SAFETY ANALYSES, (continued)
LCO, and APPLICA8ILITY valves.
The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient.
Fast closure of the TCVs is determined by measrts the electrohydraulic control fluid pressure at eact renrol valve.
There is one pressure transmitter assoc m ed with each control valve, and the signal from each transmitter is assigned to a separate trip channel.
The logic for the TCV Fast Closure. Trip 011 Pressure - Low Function is such that two or more TCVs must be closed (pressure transmitter trips) to produce _an E0C T.
This Function must be enabled at
. /nq -
Tiu.mmL POWE RTP. This is normally accomplished CKJ 6 automatically b
>ressure transmitters sensing turbine fk it stage pressure; tierefore, opening of the turbine bypass valves may affect this Function.
Four channels of TCV Fast closure, Trip 011 Pres are - Low, with two channels in each trip system, are availa011 and required to be OPERA 8LE to ensure that no single instrument failure will-preclude an O'
EOC-RPT from this Function on a valid signal. The TCV Fast closure, Trip 011 Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure, k
This protection is required consisten the safety lysis whenever THERMAL POWER is ;t RTP.
Below
'RTP, the Reactor Vessel Steam Dome )ressure - High and dNo APRM Neutron Flux - High Functions of the RPS are l
adequate to maintain the necessary margin to the MCPR Safety limit.
ACTIONS A Note has been provided to modify the ACTIONS related to EOC-RPT instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables-expressed in the Condition, discovered to be inoperable or not-within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable EOC-RPT inctrumentation channels provide (continued) l HATCH UNIT I B 3.3-83
- PmSED REVISIOKJ/16/96-
EOC-RPT Instrumentation B 3.3.4.1 BASES ACTIONS B.1 and B.2 (continued) capability when sufficient channels are OPERABLE or in trip, such that the EOC-RPT System will generate a trip signal from the given Function on a valid signal and both recirculation pumps can be tripped.
Alternately, Required Action B.2 requires the MCPR limit for inoperable EOC-RPT, as specified in the COLR, to be applied. This also restores the margin to MCPR assumed in the safety analysis.
The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Com)letion Time is sufficient time for the operator to tace corrective action, and takes into account the likelihood of an event requiring actuation of the EOC-RPT instrumentation during this period.
It is also consistent with the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time provided in LCO 3.2.2 for Required Action A.1, since this instrumentation's purpose is to preclude a MCPR violation.
C.1 and C.2 Lfb With any Required Action and associatadi letion Time not met, THERMAL POWER must be reduced to <
RTP within O
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Alternately, the associated rec rculation pump may be removed from service, since this performs the intended function of the instrumentation. The allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on o_perating i
experience, to reduce THERMAL POWER to <
P from full l
l power conditions in an orderly manner and'Wthou 2 W-challenging plant systems.
SURVEILLANCE The Surveillances are modified by a Note to indicate that REQUIREMENTS when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated i
l Conditions and Required Actions may be delayed for up to I
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the al.sociated Function maintains EOC-RPT l
trip capability.
Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
This Note is based on the reliability analysis (Ref. 4) assumption of the average l
time required to perform channel Surveillance. That
/%
(continued)
U HATCH UNIT 1 B 3.3-85 REVISION 1
E0C-RPT Instrumentation B 3.3.4.1 O
BASES 1
SURVEILLANCE analysis ~ demonstrated that the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> testing allowance REQUIREMENTS does not significantly reduce the probability that the
_(continued) recirculation pumps will trip when necessary.
SR 3.3.4.1.1 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant j
specific setpoint methodology.
I The Frequency of 92 days is based on reliability analysis of Reference 4.
2M SR 3.3.4.1.2 This SR ensure hat an E0C-RPT initiated from the TSV - Closu and TCV Fast Closure, Trip 011 Pressure - Low Functions nat be inadvertently bypassed when THERMAL
' POWER is RTP. This involves calibration of the bypass channels.-
equate margins for the instrument setpoint f
methodologies are incorporated into the actual setpoint.
Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from first stage pressure) the main turbine bypass valves must remain closed during the calibration at THERNAL POWER a(JEC, RTP to ensure that the calibration is valid.
If any bypn s channel's 28 7 setpoint is nonco.nservative (i.e., the Functions are bynasted at M3BORTP, either due to open main turbine i
ypass valves oFother reasons), the affected TSV - Closure and TCV Fast closure, Trip 011 Pressure - Low Functions are 2.P/*
considered inoperable. Alternatively, the bypass charnel can be placed in the conservative condition (nonbypass).
If
- placed in the nonbypass condition (Turbine Stop Valve-Closure and Turbine Control Valve Fast Clostre, Trip 011 Pressure-Low Functions are enabled), this SR 1s met with the channel considered OPERABLE.
The Frequency of 184 days is based on engineering judgment L
'and reliability of the components.
(continued)
HATCH UNIT 1 B 3.3-86 REVISION 1
Pri::ary Containment Isolation Instrumentation
(
B 3.3.6.1
. BASES APPLICABLE 1.c. ~ Main Steam Line Flow - Hiah (continued)
SAFETY ANALYSES, LCO, and The MSL flow signals are initiated from 16 transmitters that APPLICABILITY are connected to the four MSLs. The transmitters are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to detect the high flow.
Four channels of Main Steam Line Flow-High Function for each unisolated MSL (two channels i
per trip system) are available and are required to be OPERABLE so that no single-instrument failure will preclude detecting a break in any individual.MSL.
The Allowable Value is chosen to ensure that offsite dose limits are not exceeded a to the break. The Allowable Value corresponds to s sid, which is the parameter
-l monitored on control ro inst t
l This Function isolates the Group 1 valves.
1.d.
Condenser Vacuum - Low l
The condenser Vacuus - Low Function is provided to prevent overpressurization of the main condenser in the event of a loss of the main condenser vacuum.
Since the integrity of the condenser is an assumption in offsite dose calculations, the Condenser Vacuus - Low Function is assumed to be OPERABLE and capable of initiating closure of the MSIVs.-
The closure of the MSIVs is initiated to prevent the addition of steam that would lead to additional condenser pressurization and possible rupture of the diaphragm.
installed to protect the turbine exhaust hood, thereby 1
preventing a potential radiation leakage path following an
- accident, Condenser vacuum pressure signals are derived from four a
pressure transmitters that sense the pressure in the 3-condenser. Four channels of Condenser Vacuum - Low Function I
are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
The Allowable Value is chosen to prevent damage to the condenser due to pressurization,_thereby ensuring its integrity for offsite dose analysis. As noted (footnote (a) to Table 3.3.6.1-1), the channels are not required to be (Continued)
HATCH UNIT 1 8 3.3-155 REVISION 6
Pri::ary containment B 3.6.1.1 O
BASES (continued)
APPLICABLE The-safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures and temperatures of the 4
limiting DBA without exceeding the design leakage rate.
The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA.
In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.
Analytical methods and assumptions involving the primary 4
containment are presented in References 1 and 2.
The safety t
analyses assume a nonnechanistic fission product release j
following a DBA, which forms the basis for determination of i
offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded.
The maximum allowable leakage rate for the primary 2
containment (L ) is 1.2% by weight of the containment air O
per24hoursaI.thedesign OCA maximum peak l
containment pressure (P,) of ig (Ref. 1).
Primary containment satisfies C rion 3 of the NRC Policy l-Statement (Ref. 4).
LCO Primary containment OPERABILITY is maintained by limiting i
leakage to 1 L except prior to the first startup after perfoming a r,, quired Primary Containment Leakage Rate e
Testing Program (Ref. 5) leakage test. At this time, j
applicable ' eakage limits specified in the Primary i.
- Containment Leakage Rate Testing Program must be met.
Compliance with this LCO will ensure a primary containment configuration, including equipment hatches, that is structurally sound knd that will limit leakage to those leakage-rates assumed in the safety analyses.
Individual leakage rates specified for the primary containment air lock are addressed in LCO 3.6.1.2.
(continued)
HATCH UNIT I B 3.6-2 REVISION 5 3
s n
w-4 n-rr w
kN
---e m
-e
,,----w,----,-,,v ww-,w,-v~-
~
Pri::ary Containment Air Lock B 3.6.1.2
' BASES BACKGROUND containment leakage rate to within limits in the event of a (continued)
DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the unit safety analysis.-
APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA.
In the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage.
The primary containment is designed with a maximum allowable leakage rate (L ) of 1.2%
byweightofthecontainmentairper24hoursatIhe 4
1 calculated desig s4s LOCA maximum peak containment l
pressure (PJ o 49:tr psig Ref. 2). This allowable leakage (rate forms tte bas s for th(e acceptance criteria imposed on the SRs asociated with the air lock.
.i N'
i Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize the secondary containment.
i The primary containment air lock satisfies criterion 3 of the NRC Policy Statement (Ref. 4).
i LCO As part of primary containment, the air lock's safety function is related to control of containment leakage rates following a DBA. Thus, the air lock's structural integrity j
and leak tightness are essential to the successful mitigation of such an event.
The primary containment air lock is required to be OPERABLE.
For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type 8 air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not. exist when primary containment is required to be (continued)
HATCH UNIT 1 B 3.6-7 REVISION 5
i
-Drywell Pressure B 3.6.1.4 O - B -3.6.1.4 Drywell. Pressure -
B 3.6-CONTAINMENT SYSTEMS
' BASES BACKGROUND The drywell pressure is limited during normal operations to -
preserve the initial conditions assumed in the accident analysis for a Design Basis Accident (DBA) or loss of' coolant accident (LOCA).
L 4
i APPLICABLE Primary containment perfomance is evaluated for the entire SAFETY ANALYSES spectrum of break sizes for postulated LOCAs (Ref. 1).
Among the inputs to the DBA is the initial primary-containment internal pressure (Ref.1).
Analyses assume an 4
initial drywell pressure of 1.75 psig. This limitation c?.wres that the safety analysis remains valid by maintaining the expected initial conditions and ensures that the peak LOCA drywell internal pressure does not ex::eed the maximum allowable of 62 psig.
4 The maximum calculated drywell pressure occurs during the
~ O Mactor blowdown phase of the DBA, which assumes an instantaneous recirculation line break. Th culated peak drywell pressure for this limiting event is 45fdpsi l
l (Ref.1).
f
'O Drywell pressure satisfies Critarion 2 of the NRC Policy Statement (Ref.2).
LC0 In the event of a 08A, with an initial drywell pressure 4'
- s 1.75 psig, the resultant peak drywell accident pressure will be maintained below the drywell design pressure.
i
' APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment.
In MDDES 4 and-5, the probability and consequences of these events are reduced.due to the pressure and temperature limitations of these MODES. - Therefore, maintaining drywell pressure within limits is not required in N0DE.4 or 5.
p 4
(continued)
. HATCH UNIT 1 B 3.6-29
' REVISION 2 c
Main Condenser Offgas B 3.7.6 p
BASES O
LCO with this requirement (2436 MWt x 100 Ci/MWt-second -
(continued) 240 mci /second). The 240 mci /seco d limit is conservative for a rated core thermal power of HWt>
2H,3
~
APPLICABILITY The LCO is applicable when steam is being exhausted to the main condenser and the resulting noncondensables are being processed via the Main Condenser Offgas System.
This occurs during MODE 1, and during MODES 2 and 3 with any main steam line not isolated and the SJAE in operation.
In MODES 4 and 5, steam is not being exhausted to the main condenser and the requirements are not applicable.
ACTIONS Ad If the offgas radioactivity rate limit is exceeded, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to restore the gross gama activity rate to within the limit. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on engineering judgment, the time required p
to complete the Required Action, the large margins associated with permissible dose and exposure limits, and
(
g i
U the low probability of a Main Condenser Offgas System rupture.
B.1. B.2. B.3.1. and B.3.2 If the gross gamma activity rate is not restored to within the limits in the associated Completion Time, all main steam lines or the SJAE must be isolated.
This isolates the Main Condenser Offgas System from the source of the radioactive steam. The main steam lines are considered isolated if at least one main steam isolation valve in each main steam line is closed, and at least one main steam line drain valve in the drain line is closed. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time is reasonable, based on operating experience, to perform the actions from full power conditions in an orderly manner and without challenging unit systems.
An alternative to Required Actions B.1 and B.2 is to place the unit in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The (continued)
V HATCH UNIT 1 B 3.7-34 REVISION 6
Main Turbine Bypass System B 3.7.7 B 3.7. PL/AT SYSTEMS
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B S.7.7 _ Main Turbine Bypass System A
BASES BACKGROUND The Main Turbine _ Bypass System is designed to contral steam pressure when reactor steam generation exceeds turbine requirements during unit startup, sudden load reduction, and
- cooldown, it allows excess steam flow from the reactor to the condenser without going _through the turbine.
The bypass a ppu,3; _ y p "
capacity of the system ijf@$ of the turbine design steam flow.
Sudden load reductions within the capacity of the 21*/*
steam by) ass can be accommodated without reactor scram.
The Main Turaine Bypass System consists of three valves connected to the main steam lines between the main steam isolation valves and the turbine stop valves.
Each of these three valves is operated by hydraulic cylinders.
The bypass valves are controlled by the pressure regulation function of the Turbine Electrohydraulic Control System, as discussed in the FSAR, Section 7.11 (Ref. 1).
The bypass valves are normally closed, and the pressure regulator controls the turbine control valves that direct all steam flow to the turbine.
If the speed governor or the load limiter
(
restricts steam flow to the turbine, the pressure regulator controls the system pressure by opening the bypass valves.
When the bypass valves open, the steam flows from the bypass chest, through connecting piping, to the pressure breakdown assemblies, where a series of orifices are used to further reduce the steam pressure before the steam enters the e
condenser.
APPLICABLE The Main Turbine Bypass System is assumed to function during SAFETY ANALYSES the feedwater controller failure to maximum flow demand as discussed in the FSAR, Section 14.3.2.1 (Ref. 2). Opening the bypass valves during the pressurization event (subsequent to the resulting main turbine trip) mitigates the increase in reactor vessel pressure, which affects the MCPR during the event. An inoperable Main Turbine Bypass System may result in an MCPR penalty.
1 The Main Turbine Bypass System satisfies Criterion 3 of the NRC Policy Statement (Ref. 4).
p (continued)
HATCH UNIT 1:
B 3.7-36 REVISION 1
l i
I f
J 1
i Unit 2 Revised Bases Pages and Corresponding Marked-Up Pages i
I l
P l
n i
1 4
4 5
9
l RPS Instrumentation B 3.3.1.1 O
BASES V
APPLICABLE The trip setpoints are then determined accounting for the SAFETY ANALYSES, remaining instrument errors (e.g., drift).
The trip LCO, and setpoints derived in this manner provide adequate protection APPLICABILITY because instrumentation uncertainties, process effects, (continued) calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accountej for.
The OPERABILITY of scram pilot valve and associated solenoids, backup scram valves, and si)V valves, described in the Background section, are not addressed by this LCO.
The individual Functions are required to be OPERABLE in the MODES or other specified conditions specified in the Table, which may require an RPS trip to mitigate the consequences of a design basis accident or transient.
To ensure a reliable scram function, a combination of Functions are required in each MODE to provide primary and diverse initiation signals.
The only MODES specified in Table 3.3.1.1-1 are MODES 1 (which encompasses 2: 28% RTP) l and 2, and MODE 5 with any control rod withdrawn from a core O
cell containing one or more fuel assemblies.
No RPS
(
Function is required in MODES 3 and 4 since all control rods are fully inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (LCO 3.3.2.1) does not 1
allow any control rod to be withdrawn.
In MODE 5, control rods withdrawn from a core cell containing no fuel assemblies do not affect the reactivity of the core and, therefore, are not required to have the capability to scram.
Provided all other control rods remain inserted, no RPS Function is required.
In this condition, the required SDM (LC0 3.1.1) and refuel position one-rod-out interlock (LC0 3.9.2) ensure that no event requiring RPS will occur.
The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by function basis.
Intermediate Ranae Monitor (IRM) 1.a.
Intermediate Rance Monitor Neutron Flux - Hiah The IRMs monitor neutron flux levels from the upper range of the source range monitor (SRM) to the lower range of the (continued)
HATCH UNIT 2 B 3.3-4 PR0REV EXTPWR - 7/24/97
RPS Instrumentation B 3.3.1.1 V]
(
BASES APPLICABLE 7.a. and 7.b.
Scram Discharoe Volume Water Level - Hioh SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY Four channels of each type of Scram Discharge Volume Water Level - High function, with two channels of each type in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from these functions on a valid signal. These Functions are required in MODES I and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
At all other times, this function may be bypassed.
8.
Turbine Ston Valve - Closure Closure of the TSVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.
Therefore, a reactor scram
)
is initiated on a TSV-Closure signal before the TSVs are completely closed in anticipation of the transients that would result from the closure of these valves.
The Turbine p
i Stop Valve - Closure function is the primary scram signal l()
for the turbine trip event analyzed in Reference 2.
For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Trip (E0C-RPT) System, ensures that the MCPR SL is not exceeded.
Turbine Stop Valve - Closure signals are initiated from position switches located on each of the four TSVs.
Two independent position switches are associated with each stop valve. One of the two switches provides input to RPS trip system A; the other, to RPS trip system B.
Thus, each RP5 trip system receives an input from four Turbine Stop Valve - Closure channels, each consisting of one position switch.
The logic for the Turbine Stop Valve - Closure Function is such that three or more TSVs must be closed to produce a scram.
In addition, certain combinations of two valves closed will result in a half-scram. This Function must be enabled at THERMAL POWER 2 28% RTP.
This is l
normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function, (continued) n
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HATCH UNIT 2 B 3.3-16 PROREV EXTPWR - 7/24/97
RPS Instrumentation B 3.3.1.1 LBASES
't
- APPLICABLEl.
- 8; Turbine Ston Val've - diosure (continued)
SAFETY ANALYSES, LCO and~
The Turbine Stop Valve - Closure Allowable Value is melected APPLICABILITY to be high enough'to detect imminent TSV closure, thereby reducing:the severity of the subsequent pressure transient.
Eight channels of Turbine Stop Valve - Closure-Function, with four channels in each trip e,ystem, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this iunction if the TSVs should close. This function is required, consistent with analysis assumptions, whenever THERMAL POWER is k 28% RTP, This Function is not required when THERMAL POWER is < 28% RTP since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High Functions are adequate to maintain the necessary safety margins.
d 9.
Turbine Control Valve Fast Closure. Trio 011 Pressure - Low Fast closure of the TCVs results in the loss of a heat sink O
that produces reactor pressure, neutron flux,- and heat flux transients that must be limited. Therefore, a reactor scram-is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves.- The Turbine Control Valve Fast Closure, Trip 011 Pressure-Low Function is the primary scram signal for the generator load rejection event analyzed in Reference 2. - For j
this event,1the reactor scram reduces-the amount of energy required to be absorbed and, along with the actions of the EOC-RPT System, ensures that the MCPR SL is not exceeded.
Turbine Control Valve Fast Closure, Trip Oil Pressure - Low signals are initiated by the electrohydraulic control (EHC) fluid pressure at each control valve. One pressure transmitter is associated with each control valve, and the signal'from each transmitter is _ assigned to a separate RPS logic channel. This Function must be-enabled at THERMAL POWER a 28% RTP. This is normally; accomplished--
l automatically by pressure transmitters sensing turbine first stage. pressure;.therefore, opening of the turbine bypass
-valves may affect this Function.
5 L.
I (continued)
HATCH UNIT 2 B 3.3-17 PR0REV EXTPWR - 7/24/97
.mm
RPS Instrumentation B 3.3.1.1 i}
f BASES APPLICABLE 9.
Turbine Control Valve Fast Closure. Trio Oil SAFETY ANALYSES, Pressure - Low (continued)
LCO, and APPLICABILITY The Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure.
Four channels of Turbine Control Valve Fast Closure, Trip 011 Pressure - Low function with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This Function is required, consistent with the analysis assumptions, whenever THERMAL POWER is 2: 28% RTP, This function is not required when THERMAL POWER is < 28% RTP, since the Roactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High Functions are adequate to maintain the necessary safety margins.
10.
Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position Function v
provides signals, via the manual scram logic channels, to each of the four RPS logic channels, which are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
The reactor mode switch is a single switch with four channels, each of which provides input into one of the RPS logic channels.
There is no Allowable Value for this Function, since the channels are mechanically actuated based solely on reactor mode switch position.
Four channels of Reactor Mode Switch - Shutdown Position Function, with two channels in each trip system, are available and required to be OPERABLE.
The Reactor Mode
,m
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(continued)
(d HATCH UNIT 2 6 3.3-18 PR0REV EXTPWR - 7/24/97
/
RPS Instrumentation B 3.3.1.1
/^T t
BASES SURVEILLANCE SR 3.3.1.1.11 REQUIREMENTS (continued)
This SR ensures that scrams initiated from the Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is a 28% RTP. This involves l
calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint.
Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during the calibration at THERMAL POWER a 28% RTP to ensure that the calibration is l
valid.
If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at 2 28% RTP, either due l
to open main turbine bypass valve (s) or other reasons), then i
the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition h'V (nonbypass).
If placed in the nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low functions are enabled), this SR is met and the channel is considered OPERABLE.
The Frequency of 18 months is based on engineering judgment and reliability of the components.
SR 3.3.1.1.13 A CHANNEL CAllBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology.
For MSIV - Closure, SDV Water Level - High (Float Switch), and TSV - Closure Functions, this SR also includes a physical inspection and actuation of the switches. For the APRM Simulated Thermal Power - High Function, this SR also includes calibrating the associated recirculation loop flow channel.
(continued)
(O HATCH UNIT 2 8 3.3-29 PR0REV EXTPWR - 7/24/97
1Feedwaterfand Main Turbine High-Water-Level Trip Instrumentation i
B 3.3.2.2
-1 j
--8'3.3 INSTRUMENTATION-8_3.3.2.2J_Feedwater and Main Turbine _High Water Level Trip Instrumentation 1
BASES k
i e
~ BACKGROUND-The feedwater and main turbine high water level trip instrumentation is designed to detect _a potential failure of the Feedwater Level Control System that causes excessive feedwater flow.
With excessive feedwater flow, the water level in the reactor vessel rises toward the high water level setpoint, causing the trip of the two feedwater pump-turbines and the main turbine.
Reactor: Vessel Water Level - High signals are provided by
. level sensors that sense the difference between the pressure due to a constant-column of water (reference leg) and the -
pressure due to the actual water level in the' reactor vessel (variable -leg). Three channels of Reactor Vessel Water Level - High instrumentation are provided as input to a l
two-out-of-three initiation logic that trips the two O-feedwater pump turbines and the main turbine.
The channels include electronic equipment (e.g., trip relays) that F
compare measured input signals with pre-established setpoints. When the set >oint is exceeded, the channel L
j output relay actuates,-w11ch then outputs a main feedwater i
and turbine trip signal _ to the trip logic.
A trip of the feedwater pump turbines limits further increase in reactor vessel-water level by limiting further addition of feedwater to the reactor vessel. A trip of the
'i main turbine and closure of the stop valves protects the turbine from. damage due to water entering-the turbine.
APPLICABLE.
The feedwater and main turbine high water level trip SAFETY ANALYSES instrumentation ~is assumed to be capable of providing_a:
turbine trip in the design basis transient analysis for a-feedwater controller: failure,: maximum demand event (Ref.1).
.The high: level trip. indirectly initiates a reactor' scram.
- from the main turbine-trip-(above 28% RTP) and trips the l
feedwater pumps,Lthereby terminating the event.- The reactor.
scram mitigates.the reduction in MCPR.
, S
' HATCH UNIT.2-B 3.3-54 PROREV EXTPWR - 7/24/97
..m
~
C EOC-RPT Instrumentation-B 3.3.4.1 BASES 4
BACKGROUND
-both recirculation pumps will trip.
There are two EOC-RPT (continued) breakers in series per recirculation pump. One trip system trips one of the two EOC-RPT breakers for each recirculation i
pump,_and the second trip system trips the other EOC-RPT breaker for each recirculation pump.
^
APPLICABLE.
The TSV - Closure and the TCV Fast Closure. Tri) Oil SAFETY ANALYSES, Pressure - Low Functions are designed to trip tie LCO, and recirculation pumps in the event of a turbine trip or APPLICABILITY generator load rejection to mitigate the increase in neutron flux, heat flux, and reactor pressure, and to increase the margin to the MCPR SL. The analytical methods and assumptions used in evaluating-the turbine trip and generator load rejection are summarized in References 2 i
and 3.
/
To mitigate pressurization transient effects, the EOC-RPT 4
must trip the recirculation sumps after initiation of closure movement of either tie TSVs or the TCVs. The i
combined effects of this trip and a scram reduce fuel bundle power more rapidly than a scram alone, resulting in an increased margin to the MCPR SL. Alternatively, MCPR limits for an inoperable EOC-RPT, as specified in the COLR, are sufficient to prevent violation of the MCPR Safety Limit.
The EOC-RPT function is automatically disabled when turbine g
first stage pressure is < 28% RTP.
l EOC-RPT instrumentation satisfies Criterion 3 of the NRC L
-Policy Statement (Ref. 6).
The OPERABILITY of the EOC-RPT is dependent on the OPERABILITY of the individual instrumentation channel Functions.
Each function must have a required number of OPERABLE channels in each trip system, with their setpoints within the specified Allowable Value of SR 3.3.4.1.3.
The setpoint is calibrated consistent with applicable setpoint methodology assumptions-(nominal trip-setpoint).
Channel OPERABill"Y also includes the associated E0C-RPT breakers.
Each channel (including the associated EOC-RPT breakers) must also respond within its assumed response time.
1 o'
(continued)
HATCH UNIT 2-B 3.3-80 PROREV EXTPWR - 7/24/97-
EOC-RPT Instrumentation B 3.3.4.1 O
BASES APPLICABLE Turbine Stoo Valve - Closure SAFETY ANALYSES, LCO, and Closure of the TSVs and a main turbine trip result in the 4
APPLICABILITY loss of a heat sink and increases reactor pressure, neutron (continued) flux, and heat flux that must be limited.
Therefore, an RPT is initiated on a TSV - Closure signal before the TSVs are.
1 l
completely closed in anticipation of the effects that would result from closure of these valves.
FOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient.
Closure of the TSVs is determined by measuring the position of each valve. While there are two separate position switches associkted with each stop valve, only the signal from one switch for each TSV is used, with each of the four channels being assigned to a separate trip channel.
The logic for the TSV - Closure Function is such that two or more TSVs must be closed to produce an E0C-RPT.
This Function must be enabled at THERMAL POWER a 28% RTP.
This l
1s normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this function.
Four channels of TSV - Closure, with two O
channels in each trip system, are available and required to
,V be OPERABLE to ensure that no single instrument failure will preclude an E0C-RPT from this Function on a valid signal.
The TSV - Closure Allowable Value is selected to detect imminent TSV closure.
This protection is required, consistent with the safety analysis assumptions, whenever THERMAL POWER is a 28% RTP.
Below 28% RTP, the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor (APRM) Neutron Flux -
High Functions of the Reactor Protection System (RPS) are adequate to maintain the necessary margin to the MCPR Safety Limit.
Turbine Control Valve Fast Closure. Trio 011 Pressure - Low Fast closure of the TCVs during a generator load rejection results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on TCV Fast closure, Tri) Oil Pressure - Low in anticipation of the transients t1at would result from the closure of these (continued)
,q HATCH UNIT 2
-B 3.3-82 PR0REV EXTPWR - 7/24/97
EOC-RPT Instrumentation B 3.3.4.1 BASES
-APPLICABLE Turbine Control Valve Fast Closure. Trio 011 Pressure -- Low SAFETY ANALYSES, (continued)
~
LCO, and APPLICABILITY.
valves. The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient.
Fast closure of the TCVs is determined by measuring the electrohydraulic control fluid pressure at each control valve. There is one pressure transmitter associated with each control valve, and the signal from each transmitter is assigned to a separate trip channel. The logic for the TCV Fast Closure, Trip 011 Pressure - Low Function is such that two or more TCVs must be closed (pressure transmitter trips) to produce an E0C-RPT. This Function must be enabled at THERMAL POWER a: 28% RTP.
This is normally accomplished l
automatically by aressure transmitters sensing turbine first stage pressure; t1erefore, opening of the turbine bypass valves may affect this Function.
Four channels of TCV Fast Closure, Trip 011 Pressure - Low, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an E0C-RPT from this Function on a valid signal.
The TCV Fast O'
Closure, Trip 011 Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure.
This protection is required consistent with the safety analysis whenever THERMAL POWER is 2 28% RTP.
Below 28% RTP, the Reactor Vessel Steam Dome Pressure - High and the APRM Neutron Flux - High Functions of the RPS are adequate to maintain the'necessary margin to the MCPR Safety Limit.
ACTIONS A Note has ovided to modify the ACTIONS related to E0C-RPT in' u r stion channels.
Section 1.3, Completion Times, spei #?
- at once a Condition has been entered, subsequent m
subsystems, components, or variables expressed in *7
' tion, discovered to be inoperable or not within 11..
I not result in separate entry into-the Condition. L aan 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable E0C-RPT instrumentation channels provide (continued)
HATCH. UNIT 2 8 3.3-83 PROREV EXTPWR - 7/24/97
E0C-RPT Instrumentation B 3.3.4.1
[
BASES SURVEILLANCE analysis demonstrated that the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> testing allowance does REQUIREMENTS not significantly reduce the probability that the (continued) recirculation pumps will trip when necessary.
SR 3.3.4.1.1 A CHANNEL FUFCTIONAL TEST is performed on each required channel to enwre that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Frequency of 92 days is based on reliability analysis of Reference 4.
This SR ensures that an E0C-RPT initiated from the TSV - Closure and TCV Fast Closure, Trip 011 Pressure - Low Functions will not be inadvertently bypassed when THERMAL
/
POWER is a 28% RTP. This involves calibration of the bypass l
' Q]
channels.
Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint.
Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from first stage
{
pressure) the main turbine bypass valves must remain closed during the calibration at THERMAL POWER 2 28% RTP to ensure l
that the calibration is valid. -If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at a 28% RTP, either due to open main turbine l
bypass valves or other reasons), the affected TSV - Closure and TCV Fast Closure, Trip 011 Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass).
If placed in the nonbypass condition (Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure-Low Functions are enabled), this SR is met with the channel' considered OPERABLE.
The Frequency of 18 months is based on engineering judgment and the reliability of the components.
(continued)
HATCH UNIT 2 B 3.3-86 PR0REV EXTPWR - 7/24/97 m
Primary Containment Isolation Instrumentation B 3.3.6.1 O
BASES V
APPLICABLE 1.c.
Main Steam line Flow - Hiah (continued)
SAFETY ANALYSES, LCO, and The MSL flow signals are initiated from 16 transmitters that APPLICABILITY are connected to the four HSLs. The transmitters are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to detect the high flow.
Four channels of Main Steam Line flow - High Function for each unisolated MSL (two channels per trip system) are available and are required to be OPERABLE so that no single instrument failure will preclude detecting a break in any individual MSL.
The Allowable Value is chosen to ensure that offsite dose limits are not exceeded due to the break. The Allowable Value corresponds to s 183 psid, which is the parameter l
monitored on control room instruments.
This Function isolates the Group 1 valves, l.d.
Condenser Vacuum - Low The Condenser Vacuum - Low function is provided to prevent O)
(
overpressurization of the main condenser in the event of a v
loss of the main condenser vacuum.
Since the integrity of the condenser is an assumption in offsite dose calculations, the Condenser Vacuum - Low Function is assumed to be OPERABLE and capable of initiating closure of the MSIVs.
The closure of the MSIVs is initiated to prevent the addition of steam that would lead to additional condenser pressurization and possible rupture of the diaphragm installed to protect the turbine exhaust hood, thereby preventing a potential radiation leakage path following an accident.
Condenser vacuum pressure signals are derived from four pressure transmitters that sense the pressure in the condenser.
Four channels of Condenser vacuum - Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
The Allowable Value is chosen to prevent damage to the condenser due to pressurization, thereby ensuring its integrity for offsite dose analysis.
As noted (footnote (a) to Table 3.3.6.1-1), the channels are not required to be f 's (continued)
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HATCH UNIT 2 B 3.3-155 PR0REV EXTPWR - 7/24/97
Primary Containment B 3.6.1.1 BASES-(continued)
APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.
The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA.
In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.
Analytical methods and assumptions involving the primary containment are presented in References 1 and 2.
The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses.
The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not f
4 exceeded.
The maximum allowable leakage rate for the primary Q]
per24hoursalthedesignbasisLOCAmaximumpeak t'
containment (L ) is 1.2% by weight of the containment air containment pressure (P,) of 46.9 psig (Ref.1).
l Primary containment satisfies Criterion 3 of the NRC Policy Statement (Ref. 4).
LCO Primary containment OPERABILITY is maintained by limiting performing a r, except prior to the first startup after leakage to s L equired Primary Containment Leakage Rate Testing Program (Ref. 5) leakage test. At this time, applicable leakage limits specified in the Primary Containment Leakage Rate Testing Program must be met.
Compliance with this LCO will ensure a primary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses.
Individual leakage rates specified for the primary containment air lock are addressed in LCO 3.6.1.2.
(continued)
HATCH UNIT 2 B 3.6-2 PROREV EXTPWR - 7/24/97
Prirary Containment Air Lock 8 3.6.1.2
-/ O BASES
-V BACKGROUND containment leakage rate to within limits in the event of a (continued)
DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the unit safety analysis.
APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA.
In the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (L ) of 1.2%
byweightofthecontainmentairper24hoursatIhe calculated design basis LOCA maximum peak containment pressure (P ) of 46.9 psig (Ref. 2).
This allowable leakage l
rateformsthebasisfortheacceptancecriteriaimposedon the SRs associated with the air lock.
j Primary containment air lock OPERABILITY is also required to siinimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and g
pressurize the secondary containment.
The primary containment air lock satisfies Criterion 3 of the NRC Policy Statement (Ref. 4).
LC0 As part of primary containment, the air lock's safety function is related to control of containment leakage rates following a DBA. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.
The primary containment air lock is required to be OPERABLE.
For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time.
This provision ensures that a gross breach of prinary containment does not exist when primary containment is required to be (continued)
HATCH UNIT 2 B 3.6-7 PROREV EXTPWR - 7/24/97
Drywell - Pressure B 3.6.1.4 i
8 3.6 ' CONTAINMENT-SYSTEMS B-3.6.1.4 Drywell Pressure BASES BACKGROUN0' The drywell pressure is limited during normal operations to preserve the initial conditions assumed in the accident analysis for a Design Basis Accident (DBA) or loss of coolant accident (LOCA).
APPLICABLE Primary containment performance is evaluated for the entire SAFETY ANALYSES spectrum of break sizes for postulated LOCAs (Ref. 1).
Among the inputs to the DBA is the initial primary containment internal pressure (Ref.1). Analyses assume an initial drywell pressure of 1.75 psig. This limitation ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that the peak LOCA drywell internal pressure does not exceed the maximum allowable of 62 psig.
The maximum calculated drywell pressure occurs during the reactor blowdown phase of the DBA, which assumes an instantaneous recirculation line break. The calculated peak drywell pressure for this limiting event is 46.9 psig l
(Ref. 1).
Drywell pressure satisfies Criterion 2 of the NRC Policy Statement (Ref. 2).
LCO In the event of_ a DEA, with an initial drywell pressure s 1.75 psig, the resultant peak drywell accident pressure will be maintained below the drywell design pressure.
l APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment.
In MODES 4 and 5, the probability and consequences of these. events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining drywell pressure'within lia.its is not required in MODE 4 or 5.
l (continued)
HATCH UNIT 2 B 3.6-30 PR0REV EXTPWR - 7/24/97 s
i
Main Condenser Offgas B 3.7.6 BASES v
LCO with this requirement (2436 MWt x 100 Ci/MWt-second -
(continued) 240 mci /second). The 240 mci /second limit is conservative for a rated core thermal power of 2763 MWt.
l APPLICABILITY The LCO is applicable when steam is being exhausted to the main condenser and the resulting noncondensibles are being processe:i via the Main Condenser Offgas System.
This occurs during MODE 1, and during MODES 2 and 3 with any main steam line not isolated and the SJAE in operation.
In MODES 4 and 5, steam is not being exhausted to the main condenser and the requirements are not applicable.
ACTIONS L1 If the offgas radioactivity rate limit is exceeded, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to restore the gross gamma activity rate to within the limit.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on engineering judgment, the time required to complete the Required Action, the large margins O'
\\
associated with permissible dose and exposure limits, and the low probability of a Main Condenser Offgas System rupture.
B.I. B.2. B.3.1. and B.3.2 If the gross gamma activity rate is not restored to within the limits in the associated completion Time, all main steam lines or the SJAE must be isolated. This isolates the Main Condenser Offgas System from the source of the radioactive steam. The main steam lines are considered isolated if at least one main steam isolation valve in each main steam line is closed, and at least one main steam line drain valve in the drain line is closed. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time is reasonable, based on operating experience, to perform the actions from full power conditions in an orderly manner and without challenging unit systems.
An alternative to Required Actions B.1 and B.2 is to place the unit in a MODE in which the LC0 does not apply.
To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The (continued) b/
HATCH UNIT 2 B 3.7-34 PROREV EXTPWR - 7/24/97
_-._.___.______..___.g Main Turbine Bypass System-B 3.7.7' ii
- B 3.7. PLANT SYSTEMS 1B 3.7.7J Main' Turbine Bypass System _
l BASES j
f The Main Turbine Bypass System is; designed to control steam i
BACKGROUND
_pressure when reactor-steam generation exceeds turbine requirements during unit startup, sudden load reduction, and I
cooldown.
It allows excess steam flow from the reactor to-the condenser without going through the turbine.
The bypass capacity of the system is approximi.tely 21% of the turbine
.l design steam flow. Sudden load reductions within the capacity of the steam bypass can be, accommodated without reactor scram. The Main Turbine Bypass System consists of 1
three valves connected to the main steam lines between the main steam isolation valves and the turbine stop valves..
3 Each of these three valves is operated by hydraulic cylinders. The bypass valves are controlled by the pressure regulation function of the Turbine Electrohydraulic Control System, as discussed in the FSAR, Section 7.7.4 (Ref. 1).
The bypass valves are normally closed, and the )ressure regulator controls the turbine control valves t1at direct all steam flow to the turbine.
If the speed governor or the load _ limiter restricts steam flow to the turbine, the pressure regulator controls the system pressure by opening the bypass. valves. When the bypass valves open, the steam flows from the bypass chest, through connecting piping, to the pressure breakdown assemblies, where a series of orifices are used to further reduce the steam' pressure before the steam-enters the condenser.
APPLICA8LE' The Main Turbine Bypass System is assumed to function during j
SAFETY ANALYSES the feedwater controller failure to maximum flow demand as discussed in~the FSAR, Section 15.1.7 (Ref. 2).
Opening the i
bypass valves-during the pressurization event (subsequent to i
the resulting main-turbine trip) mitigates.the increase in reactor vessel pressure, which affects the MCPR during the event. An inoperable Main Turbine-Bypass System may result in an-MCPR penalty.
The Main Turbine Bypass System satisfies Criterion'3 of the NRC Policy Statement (Ref. 4).
j o
j h.O (continued)
HATCH _ UNIT 2l 8 3.7 PROREV EXTPWR - 7/24/97
~,
w w,
...1
- --s'w&41.
i-RPS Instruneatation i
8 3.3.1.1 s..
0 8ASES APPLICA8LE The trip setpoints are then determined accounting for the SAFETY ANALYSES, remaining instrument errors (e.g., drift).
The trip LCO, and setpoints derived in this manner provide adequate protection APPLICABILITY because instrumentation uncertainties,-process effects,~
l (continued) calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted-
. for.
.d-l 2 ~~
The OPERABILITY of scram pilot valves and associated solenoids, backup scram valves, and SDV valves, described in j
the Background section, are not addressed by this LC0,.,
The individual Functions are required to be OPERABLE in the MODES or other specified conditions specified in the Table which may require an RPS trip to mitigate the consequences, of a design basis accident or transient.
To ensure a2 reliable scram function, a combination of Functions are required in each MODE to provide primary and diverse initiation signals. The only MODES specified in do/c Table 3.3.1.1-1 are MODES 1 (which encompasses a: 393) RTP) and 2, and MODE 5 with any control rod withdrawn a core cell containing one or more fuel assemblies. No RPS Function is required in MODES 3 and 4 since all control rods are fully inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (LC0 3.3.2.1) does not allow any control rod to be withdrawn.
In MODE 5, control rods withdrawn from a core cell containing no fuel assemblies do not affect the reactivity of the core and, therefore, are not required to have the capability to scram.
Provided all other control rods remain inserted, no RPS Function is required.
In this condition, the required SOM (LCO 3.1.1) and refuel position one-rod-out interlock (LC0 3.9.2) ensure that no event requiring RPS will occur.
The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis, t
Intermediate Ranae Monitor (IRM) l.a.
Intemediate Rance Monitor Neutron Flux - Hiah The IRMs monitor neutron flux levels from the upper range of the source range monitor (SRM) to the lower range of the 1
HATCH UNIT 2 B 3.3-4 REVISION 0 i
RPS Instrumentaticn 8 3.3.1.1
/
BASES
\\
APPLICABLE 7.a. and 7.b.
Scram Discharae Vol"= Water Level - Hiah SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY Four channels of each type of Scram Discharge Volume Water Level - High Function, with two channels of each type in each trip system, are required to be O'?RABLE to ensure that no single instrument failure will preclude a scram from these Functions on a valid signal.
These Functions are required in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel._ ~
assemblies, since these are the MODES and other specified conditions when control reds are withdrawn.
At all other times, this function may be bypassed.
8.
Turbine Ston Valve - Closure closure of the TSYs results in the loss of a heat sink that neutron flux, and heat flux produces reactor pressure transientsthatmustbellaited.
Thereface, a reactor scram is initiated on a TSV-Closure signal before the TSys are completely closed in anticipation of the transients that would result from the closure of these valves.
The Turbine
(
Stop Valve - Closure Function is the primary scram signal i
for the turbine trip event analyzed in Reference 2.
For 4
this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Trip (E0C-RPT) System, ensures that the MCPR SL is not exceeded.
Turbine Stop Valve - Closure signals are initiated from potition switches located on each of the four TSVs.
Two independent position switches are associated with each stop valve. One of the two switches provides input to RPS trip
,,.. system At the other, to RPS trip system B.
Thus, each RPS ""
1 trip system receives an input from four Turbine Stop Valve - Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure Function is such that three or more TSVs must be closed to produce a scram.
In addition, certain combinations of two valves closed will result in a hal -stram.
This Function must be enabled at THERMAL POWER RTP. This is nomally accomplished automatica11 ypressuretransmitters(IF%
sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function.
(continued)
HATCH UNIT 2 B 3.3-16 REVISION 14 l
RPS Instrumentation B 3.3.1.1 8ASES O
l APPLICABLE L Turbine stoo Valve - Closure (continued)
SAFETY ANALYSES, LCO and The Turbine Stop Valve - Closure Allowable Value is selected APPLICABILITY to be high enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure trantient.
Eight channels of Turbine Stop Valve - Closure Function, with four channels in each tri system, are required to be OPERA 8LE to ensure that no sin le instrument failure will' ' " -
preclude a scram from this Fun tis.1 if the TSVs should close. This function is required, cons t46tt_with analysisf--%
assumptions, whenever THERMAL POWER is RTP This 2.&
Function is not required when THERMAL is<(30%
since the Reactor Vessel Steam Dome Pressure - Hf and the Average Power Range Monitor Neutron Flux - High functions l
are adequate to maintain the necessary safvty margins.
9.
Turbine Control Valve Fast Closure. Trio 011 i
Pressura - tow Fast closure of the TCVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux O.
transients that Wst be limited. Therefore, a reactor scram is initiated on TCV fast closure in anticioation of the transients that would result from the closure of these valves. The Turbine Control Valve Fast Closure, Trip 011 Pressure - Low function is the primary scram signal for the generator load rejection ever,t analyzed in Reference 2.
For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of tl1e EOC-RPT System, ensures that the MCPR SL is not exceeded.
Turbine Control Valve Fast closure, Trip 011 Pressure - Low signals are initiated by the electrohydraulic control (EHC) fluid pressure at each control valve. One pressure
' v.
transmitter is associated with each control valve, and the signal from each transmitter is assigned to a separate RPS logic channel. This function must be enabled at THERMAL POWER MtID RTP. This is normally accomplishad 1Ur5iiia(lically by)ressure transmitters sanning turbine first stage pressure; tierefore, opening of the turbine bypass 22%
valves may affect this Function.
p (continued)
L)
HATCH UNIT 2 B 3.3-17 REVISION 14
RPS Instrumentation B 3.3.1.1 4
4 APPLICABLE t.
Turbine Contrni Valve Fast Closure. Trio 011 SAFETY ANALYSES, Pressure - Low (continued)
LC0, and i
APPLICABILITY The Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Allowable Value is selected high enough to detect ieminent TCV fast closure.
i Four channels of Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Function with two channels in each trip system arranged in a one-out-of-two logic are required to be a
OPERA 8LE to ensure that no single instrument failure will preclude a scram from this function on a valid signal. This function is required, consistent with th lusis_
(d,b i
assumptions, whenever THERMAL POWER is
'R ' r.
is Function is not required when THERNAL is <
4 since the Reactor Vessel Steam Dome Pressure - H a
the
- Average Power Range Monitor Neutrea Flux - High Functions l
are adequate to maintain the nn.essary safety margins.
'e 10.
Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position Function provides signals, via the manual scram logic channels, to each of the four RPS logic channels, whica are redundant to the automatic protective instrumentation channels and provide manual reactor trip cepability.
This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
The reactor mode switch is a single switch kith four channels, each of which provides input into one of the RPS logic channels.
A%;p e
since the T
P There is no Allowable Value for this Function channelsaremechanicallyactuatedbasedsolelyonreactor mode switch position.
Four channels of Reactor Mode Switch - Shutdown Position Function, with two channels in each trip system, are available and required to be OPERABLE.
The Reactor Mode e
f G
(continued)
O 4
HATCH UNIT 2 B 3.3-18 REVISION 14
....-...r
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.._,,.,,_.,....,.,,,.,_..,.._..____,_,y_
,.,..__.,,,,,,,,_,_.,,w,,.
RPS Instrumentation B 3.3.1.1 4
O BASES SURVEILLANCE 1R 3.3.1.1.11 REQUIREMENTS (continued)
This SR ensures that scrans initiated from the Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 4
011 Pressure - Low Functions wil not be inadvertently bypassed when THERMAL POWER is M$.RTP.
This involves
. g.4 uitbratiM of the bypass channe s.
Adequate margins for i
M instruneht setpoint methodologies are incorporated into in.a actual setpoint. Because main turbine bypass flow can~
- u.
affect this setpoint nonconservatively (THERMAL POWER is'FA derived from turbine first stage pressure), the main turbine i
bypass valves must-tenain closed during the calibration at~
THERMAL POWER 3052RTP to ensure that the calibration is valid.
].g Ub l
al) setpoint is nonconservativ If any b(i.e.,tbasschFunctions are bypassed at afpe57RTP, either D toopenmainturbinebypassvalve(s?osuetherreasons),then of the affected Turbine Stop Valve - C re and Turbine Control Valve Fast Closure Trip 011 Pressure - Low Functions are considered incperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypast).
If placed in the nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions are enabled), this SR is met and the channel is considered OPERABLE.
4 The Frequency of 18 months is based on engineering judgment and reliability of the components.
SR 3.3.1.1.13 l
~
A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channe1~
u.
responds to the meatiwod parameter within the necessary T "
range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology.
For MSIV - Closure, 50V Water Level - High (FloatSwitch),andTSV-ClosureFut.ctions,thisSRalso includes a physical inspection and actuation of the switches.
For the APRM Simulated Thermal Power - High Function, this SR also includes calibrating the associated recirculation loop flow channel.
(continued)
HATCH UNIT 2 B 3.3-29 REVISION 14
.r%...--er.m.---,-we*~-v--om+,*---v-v. 4-w---we, ev.s--~-
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w-.ip-------w--~*--
--~.-+-+,------a-.---ev.-
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l l
Feedwater and Main Turbir.e High Water Level Trip Instrumentation 1
B 3.3.2.2 l
l gs i
B 3.3 INSTRUMENTATION
\\
l B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation I
BASES l
BACKGROUND The feedwater and main turbine high water level trip 1
instrumentation is designed to detect a potential failure of the Feedwater Level Control System that causes excessive foodwater flow.
e
,c g
.q With excessive feedwater flow, the water level in the
~,
reactor vessel rises toward the high water level setpoint, causing the trip of the two feedwater pump turbines and the sain turbine.
Reactor Vessel Water Level - High signals are provided by i
level sensors that sense the difference between the pressure i
due to a constant column of water (reference leg) and the pressure due to the actual water level in the reactor vessel (variableleg). Three channels of Reactor Vessel Water Level - High instrumentation are provided as input to a two-out-of-three initiation logic that trips the two feedwater pump turbines and the main turbine. The channels include electronic equipment that compare measured input signals (e.g., trip relays)hed
.with pre-establis setpoints. When the setpoint is exceeded, the channel output relay actuates, wtich then outputs a main feedwater and turbine trip signal to the trip logic.
A trip of the feedwater pump turbines limits further increase in reactor vessel water level by limiting further.
addition of feedwater to the reactor vessel. A trip of the main turbine and closure of the stop valves protects the j
turbine from damage due to water entering the turbine.
i nW s
s i
APPLICABLE The feedwater and main turbine high water level trip SAFETY ANALYSES instrumentation is assumed to be capable of providing a
~
turbir.e trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref.1).
The high level trip indirectly initiates a reactor scram from the main turbine trip (above RTP) and triis the feedwater pumps, thereby teminati the event.
T5e reactor scram mitigates the reduction in MC R.
2f%
(continued)
HATCH UNIT 2 8 3.3-54 REVISION 0 1
v
.,,-.v.
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_.~,...rm,,..._,,,,..,-.,_m.
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,.,m.-
l E0C-RPT Instrumentation B 3.3.4.1 BASES BACKGROUND both recirculation pumps will trip. There are two E0C-RPT (continued) breakers in series per recirculation pump. One trip system trips one of the two E0C-RPT breakers for each recirculation pump, and the second trip system trips the other E0C-RPT breaker for each recirculation pump.
APPLICABLE The TSV - Closure and the TCV Fast Closure, Trip 011 M@c s
SAFETY ANALYSES, Pressure - Low Functions are designed'to trip the 1..
1 LCO, and recirculation pumps in the event of a turbine trip or l
APPLICABILITY generator load rejection to mitigate the increase in neutron flux, heat flux, and rnctor pressure and to increase the margin to the MCPR SL. The analytical methods and assumptions used in evaluating the turbine trip and generator load rejection are summarized in References 2 and 3.
To mitigate pressurization transient effects, the E0C-RPT must trip the recirculation pumps after initiation of closure movement of either the TSVs or the TCVs. The combined effects'of this trip and a scram reduce fuel bundle i
power more rapidly than a scram alone, resulting in an
(
increased margin to the MCPR SL. Alternktively, MCPR limits 4
i for an inoperable E0C-RPT, as specified in the COLR, are sufficient to prevent violation of the MCPR Safety Limit.
The E0C-RPT function is ap ically disabled when turbine first stage pressure is 6 TP ' %
2P i
E0C-RPT instrumentation satisfie ri ion 3 of the NRC
~
l PolicyStatement(Ref.6).
The OPERABILITY of the E0C-RPT is dependent on the OPERABILITV'of the individual instrumentation channel Functions. Each Function must have a required number of OPERABLE channels in each trip system, with their setpoints 4
within the specified Allowable Value of SR 3.3.4.1.3.
The setpoint is calibrated consistent with applicable setpoint methodolog'y assumptions (nominal trip setpoint).
Channel OPERABILI Y also includes the associated EOC-RPT breakers.
Each channel (including the associated E0C-RPT breakers) must also respond within its assumed response time.
(continued)
HATCH UNIT 2 8 3.3 REVISION 1
EOC-RPT Instrumentation G 3.3.4.1 E
O APPLICABLE Turbine sten Valve - Closure l
l SAFETY ANALYSES, LCO, and Closure of the TSVs and a main turbine trip result in the i
APPLICA81LITY loss of a heat sink and increases reactor pressure, neutron (continued) flux, and heat flux that must be limited.
Therefore, an RPT is initiated on a TSV - Closure signal before the TSVs are completely closed in anticipation of the effects that would result from closure of these valves.
EOC-RPT decreases
' "the MCPR SL is not exceeded during the worst case transient.,,;g,p" reactor power and aids the reactor scram in ensuring that m-j T.T.@ pee of each valve. While there are two separate osition
^
T!
Closure of the TSVs is detemined by measuring the position
+
switches associated with each stop valve, on1 the si nel from one switch for each TSV is used, with each of th four channels being assigned to a separate trip channel. The logic for the TSV - Closure Function is such that two or more TSVs must be closed to produce an E0C-PTm Thi
$ ? 'A Function must be enabled at THERMAL POWER This is normally accomplished automatically by p ure transmitters sensing turbine first stage pressure; i
therefore, ion. opening of the turbine bypass valves may affect this Funct Four channels of TSV - Closure, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an E0C-RPT from this Function on a valid signal.
i The TSV - Closure Allowable Value is selected to detect isminent TSV closure.
D This protection is required, consistent with the safat analy is-assumptions, whenever THERMAL POWER is at TP.
Belor RTP, the Reactor Vessel Steam Dome Pressure - High and t verage Power Range Monitor (APRM) Neutron Flux -
l High Functions of the Reactor Protection System (RPS) are 2.g y.
_,+ adequate to maintain the necessary margin to the MCPR Safety u m..
Limit.
7 Turbine Control Valve Fast closure. Trin 011 Pressure - Low Fast closure of the TCVs during a generator load rejection results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on TCV Fast i
_ Closure, Trip 011 Pressure - Low in anticipation of the transients that would result from the closure of these (continued)
(
HATCH UNIT 2 B 3.3-82 REVISION 14 l
l
i E0C-RPT Instrumentation
)
B 3.3.4.1 i
4 4
~
8ASES i
APPLICABLE Igthine Control Valve Fast Closure. Trio 011 Pressure - Law SAFETY ANAL)SES,
-(continued)
LCO, and APPLICA8ILITY valves. The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded i
during the worst case transient.
. Fast closure of the TCVs is detensined by measuring the ~ D.,
- c
~ gk.C^
. electrohydraulic control fluid pressure at each contro' e-4
,,G valve. There is one pressure transmitter associated w9 1
~
, each control valve, and the signal from each transmittel s.., 8,0'
+
assigned to a separate trip channel. The logic for the TCV i
Fast Closure. Trip 011 Pressure - Low Function is such that 2'
two or more TCV: sust be closed (pressure transmitter trips) to produce an E0C.RPT. This Function must be enabled at THE POWER k 2RTP. This is normally accomplished
- 1 auf.omat ca fessure transmitters sensing turbine first i
J.g'.
stage pressure; tierefore, opening of the turbine bypass ?.l.,
valves may gffect this Function.
Four channels of TCV Fast Closure, Trip 011 Pressure - Low, with two channels in each
~
trip system, are available and required to be OPERA 8LE to ensure that no single instrument failure will preclude an E0C-RPT from this Function on a valid signal. The TCV Fast j
Closure, Trip 011 Pressure - Low Allowable she s selected i
high enough to detect imminent TCV fast osure.
2.s Y.
g i
Jhavsafety 1
h l
This protection is required consistent wi anal sis whenever THERMAL POWER is g(qp RTP.
Selow t/
29%
TP, the Reactor Vessel Steam Dome Pressure - High and APRM Neutron Flux - High Functions of the RPS are l
adequate to maintain the necessary margin to the MCPR Safety Limit.
.R&;
z f.,
.f,, -
- cet - -tv E0C-RPT instrumentation channels.
Section 1.3, Cnapletion Times, specifies that once a Condition has been entered subsequentdivisions, subsystems, components,orvariables i
expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable E0C-RPT instrumentation channels provide e
(continued)
HATCH UNIT 2 B 3.3-83' REVISION 14-s n~.+
,,- - -, - ~,
nr-r
-r
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--~~.,-,-,n--
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,,,-,-...,,-----.-n,.~+
EOC-RPT Instrumentation B 3.3.4.1 O
BASES SURVEILLANCE analysis demonstrated that the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> testing allowance does REQUIREMENTS not significantly reduce the probability that the (continued) recirculation pumps will trip when necessary.
SR 3.3.4.1.1 A CHANNEL FUNCTIONAL TEST is perfomed on each required channel to ensure that the entire channel will perfons the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
The Frequency of 92 days is based on reliability knalysis of J
Reference 4.
SR 3.3.4.1.2 This SR ensures that an EOC-RPT initiated from the TSV - Closure and TCV Fast closure, Trip 011 Pressure - Low Functions ill not be inadvertently bypassed when THERMAL D
RTP. This involves calibration of the bypass 1
channels.
equate margins for the instrument setpoint i
e%
28 methodologies are incorporated into the actual setpoint.
Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from first stage pressure) the main turbine bypass valves sus m aain closed during the calibration at THERMAL POWER lt"P to ensure = y Ya that the calibration is valid.
If any bypass channel's setpoint is nonconservative (i.e., the Functions are bynssed at rMJTP, either due to open main turbine
'ypass valves r other reasons, the affected TSV - Closure and TCV Fast Closure, Trip 011) Pressure - Low functions are 2F7 considered inoperable. Alternatively, the bypass channel canbeplacedintheconservativecondition(nonbypass).
If placed in the nonbypass condition Turbine Stop Valve-Closure and Turbine Control (Valve Fast closure, Trip
-011 Pressure-Low functions are enabled), this SR is met with the channel considered OPERABLE.
The Frequency of 18 months is based on engineerteg judgment and-the reliability of the components.
(continued)
HATCH UNIT 2 8 3.3-86 REVISION 1 i
Primary Containment Isolation Instrumentaticn 8 3.3.6.1 j
b SASES i
APPLICA8LE 1.c.
Main Steam Line Flow - Hlah (continued)
SAFETY ANALYSES, LCO, and The MSL flow signals are initiated from 16 transmitters that APPLICA81LITY are connected to the four MSLs. The transmitters are i
arranged such that, even though physically separated from each other, all four connected to one MSL would be able to i
detect the high flow.
Four channels of Main Steam Line i
Flow - High Function for each unisolated MSL (d to be two channels per trip system) are available and are require
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-j OPERA 8LE so that.no single instrument failure will preclude
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detecting a break in any individual MSL.
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l The Allowable Value is chosen to ensure that offsite dose limits are not exceedoc to the break. The Allowable Value corresponds to psid, which is the parameter l
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monitored on contro nstruments.
ThisfunctionisolatesIheGro valves.
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Condenser Vacuum - Low The Condenser Vacuus - Low Function is provided to prevent 3-overpressurization of the main condenser in the event of a t
l loss of the main condenser vacuum.
Since the integrity of the condenser is an assumption in offsite dose calculations, i
the Condenser Vacuum - Low Function is assumed to be 0FERABLE and capable of initisting closure of the MSIVs.
1 i
The closure of the MSIVs is initiated to prevent the i-addition of steam that would lead to additional condenser pressurization and possible rupture of the diaphragm installed to protect the turbine exhaust hood, thereby preventing a potential radiation leakage path following an 4
accident.
+)..
Condenservacuumpris'sures..ignals are derived from four
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+
n' pressure transmitters that sense the pressure in the e
condenser. Four channels-of Condenser Vacuum -Low Function are.available and are required to be OPERABLE to ensure that no single instrument-failure can preclude the isolation function.
The Allowable-Value is chosen to prevent damage to the condenser due to pressurization, thereby ensuring its integrity for offsite dose analysis. As noted (footnote (a) i to Table 3.3.6.1-1), the channels are not required to be 4 :
j (continued)
HATCH UNIT 2 8 3.3-155 REVISION 5
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B 3.6.1.1 BASES (continued)
APPLICABLE The safety design basis for the primary containment is that
'i SAFETY ANALYSES it must withstand the pressures and temperatures of the s -
limiting 08A without exceeding the design leakage rate.
The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA.
In the
+
1 analysis of this accident, it is assumed that primary i
containment is OPERABLE such that release of fission i
products to the environment is controlled by the rate of primary containment leakage.
Analytical methods and assumptions involving the primary containment are presented in References 1-and 2.
The safety
- analyses assume a nonnechanistic fission product release i
following a DBA, which forms the basis for determination of I
offsite doses. The fission product release is, in turn, l
based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures t
i that the leakage rate assumed in the safety analyses is not exceeded.
The maximum allowable leakage rate for the primary containment (L ) is 1.2% b weight of the containment air t
per24hoursalthedesig si LOCA maximum peak l
l containment pressure (P,) of ig (Ref.1).
I Primary containment satisfies r'terion 3 of the NRC Policy Statement (Ref.4).
i LC0 Primary containment OPERABILITY is maintained by limiting leakage to 1 L performing a r, except prior to the first startup after equired Primary Containment Leakage Rate Testi Program (Ref. $ leakage test. At this time, applic ble sakege limi s s lfied in the Primary Containment Leakage Rate To ting Program must be met.
Compliance with this LC0 will ensure a rima containment configuration, including equipment hate es, t at is structurally sound and that will limit leakage to tho,se leakage rates assumed.in the safety analyses.
1 Individual leakage rates specified for the primary 3
- containment air lock are addressed in LCO 3.6.1.2.
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HATCH UNIT 2 B 3.6-2 REVISION 7 i
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Primary Containment Air Lock B 3.6.1.2 BASES BACKGROUND containment leakage rate to within limits in the event of a (continued)
DBA. Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the unit safety analysis.
APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA.
In the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (L ) of 1.2%
byweightofthecontainmentairper24hoursatthe calculated desi s LOCA maximum peak containment l
sig Ref. 2).
This allowable leakage
[ pressure (P'ratrf5Ni' Tie b i or th(e acceptance criteria imposed on the SRs associated with the air lock.
Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and O
pressurize the secondary containment.
The primary containment air lock satisfies Criterion 3 of the NRC Policy Statement (Ref. 4).
LC0 As part of primary containment, the air lock's safety function is related to control of containment leakage rates following a DBA. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.
The primary containment air lock is required to be OPERABLE.
For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE.
The interlock allows only one air lock door to be opened at a time.
This provision entures that a gross breach of primary containment does not exist when primary containment is required to be (continued) a HATCH UNIT 2 B 3.6-7 REVISION 7
Drywell Pressure
[
B 3.6.1.4 B 3'.6' CONTAlletENT SYSTEMS l
l B 3.6.1.4 Drywell Pressure t
BASES i
1 BACKGROUND The drywell pressure is limited during normal operations to t
preserve the initial conditions assumed in the accident analysis for a Design Basis Accident (DBA) or loss of coolantaccident(LOCA).
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APPLICABLE Primary containment perfomance is evaluated for the entire i
4 SAFETY ANALYSES spectrum of break sizes for postulated LOCAs (Ref. 1).
l Among the inputs to the DBA is ties initial primary containment internal pressure (Aef.1). Analyses assume an i
initial drywell pressure of 1.75 psig.
This limitation ensures that the safety analysis remains valid by 1
maintaining the expected initial conditions and ensures that j
the peak LOCA drywell internal pressure does not exceed the maximum allowable of 62 psig.
1 L
The maximum calculated drywell pressure occurs during the reactor blowdown phase of the DBA, which assumes an i
instantaneous recirculation line break.
T calculated peak drywell pressure for this limiting event i 11 l
(Ref.1).
Drywell pressure satisfies Criterion 2 of the NRC Policy Statement (Ref.2).
L LCO In the event of a 08A with an initial drywell pressure s 1.75 psig, the resultant peak drywell accident pressure will be maintained below the drywell design pressure.
APPLICABILITY In MODES 1, 2, and 3, a 08A could cause a release of radioactive material to primary containment.
In M00ES 4 and 5, the probability and consequences of these events are i
reduced due to--the pressure and temperature limitations-of these MODES. Therefore, maintaining drywell pressure within limits is not required in MODE 4 or 5.
(continued)
HATCH UNIT 2 B 3.6-30 REVISION 2
Main Condenser Offgas
' 8 3.7.6 1
BASES l
LC0 with this requirement (2436 MWt x 100 pCl/MWt-second =
(continued) 240 mci /second). The 240 sci /sec ad-1 mit is conservative for a rated core thermal power of 763 l
APPLICABILITY The LCO is applicable when steam is being exhausted to the main condenser and the resulting noncondensibles are being processed via the Main Condenser Offgas System.
This occurs i
during MODE 1, and during MODES 2 and 3 with any main steam line not isolated and the SJAE in operation.
In MODES 4 and 5, steam is not being exhausted to the main condenser and the requirements are not applicable.
2 ACTIONS L1 1'
If the offgas radioactivity rate limit is exceeded, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to restore the gross gamma activity rate to
- within the limit. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on engineering judgment, the time required to complete the Required Action, the large margins I
i associated with permissible dose and exposure inits, and the low probability of a Main Condenser Offgas System rupture.
i B 1. B.2. B.3.1. and B.3.2 If the gross gassia activity rate is not restored to within the limits in the associated Completion Time, all main steam lines or the SJAE must be isolated.
This isolates the Main Condenser Offgas System from the source of the radioactive steam. The main steam lines are considered isolated if at least one main steam isolation valve in each main steam line is closed, and at least one main steam line drain valve in the drain line is closed. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time is e
reasonabl6., based.on operating experience, to perform the actions from full power conditions in an orderly manner and without challenging unit systems.
- An alternativ9 to-Required Actions 8.1 and 8.2 is to place the unit in a MODE in which the LC0 does not apply. To achieve this status, the unit must be placed in at least M00E 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and:in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The j
(continued)-
HATCH UNIT 2-8 3.7-34 REVISION 5
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Main lurbine Bypass Systen B 3.7.7 p
B 3.7 PLANT SYSTEMS 5
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B 3.7.7 Main Turbine Bypass System BASES DACKGROUND The Main Turbine Bypass System is designed to control steam pressure when reactor steam generation exceeds turbine requirements during unit startup, sudden load reduction, and cooldown.
It allows excess steam flow from the reactor to the condenser without goin rough the turbine. The bypass
-- f apacity of the_ system is of the turbine design steam Cngpe.$r.dtl c
3 2/Y.
low.
Sudden loaa reductt s within the capacity of the steam bypass can be accommodated without reactor scram. The Main Turbine Bypass System consists of three valves connected to the main steam lines between the main steam isolation valves and the turbine stop valves.
Each of these three valves is operated by hydraulic cylinders.
The bypass valves are controlled by the pressure regulation function of the Turbine Electrohydraulic Control System, as discussed in the FSAR, Section 7.7.4 (Ref. 1).
The bypass valves are normally closed, and the pressure regulator controls the turbine control valves that direct all steam flow to the I
turbine.
If the speed governor or the load limiter
(
restricts steam flow to the turbine, the
)ressure regulator controls the system pressure by opening t1e bypass valves.
When the bypass valves open, the steam flows from the bypass chest, through connecting piping, to the pressure breakdown assemblies, where a series of orifices are used to further reduce the steam pressure before the steam enters the condenser.
APPLICABLE The Main Turbine Bypass System is assumed to function during SAFETY ANALYSES the feedwater controller failure to maximum flow demand as discussed in the FSAR, Section 15.1.7 (Ref. 2).
Opening the bypass valves during the pressurization event (subsequent to
.the resulting main turbine trip) mitigates the increase in reactor vessel pressure, which e.ffects the MCPR during the event. An inoperable Main Turbine Bypass System may result in an MCPR penalty.
The Main Turbine Bypass System satisfies Criterion 3 of the NRC Policy Statement (Ref. 4).
,. m
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(Continued)
HATCH UNIT 2 B 3.7-36 REVISION 1 i
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Edwin 1. Ilatch Nuclear Plant I
Request for I,1 cense Amendment Extended Power Uprate Operation Summary of Plant Modifications i O 1
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Edwin 1. Ilatch Nuclear Plant Request for License Amendment j
Extended Power Uprate Operation Summary of Plant Modifications i
I The list of plant design changes shown below indicates the Plant flatch extended power uprate I
program does not involve a material alteration to the facility (i.e., no construction permit is required.) Most of the modifications are not safety related and are not considered a commitment to the Nuclear Regulatory Commission.
1.
Replace or modify stages of the Unit I and Unit 2 high pressure turbine to increase steam flow capacity, i
4 2.
Modify the Unit I and Unit 2 main generator stator water cooling system to enhance cooling j
capacity.
1 3.
Modify the Unit 2 main generator isophase bus cooling system to enhance cooling capacity.
(Unit I changes are very minor.)
4.
Modify the Unit 1 'and Unit 2 condensate demineralizer system to decrease resin usage and system differential pressure.
5.
Modify the Unit I and Unit 2 condensate /feedwater system to increase suction pressure to the condensate booster pumps (CDPs). Changes (in addition to item no. 4) are not finalized but may include either trimming or replacing the condensate pump or CBP impellers, resetting the CDP suction trips, and installing digital programmable logic controllers at pump r
suction and discharge locations.
6.
Increase main condenser tube bundles staking at selected locations.
7; Install an enhanced temperature monitoring system on the Unit 2 main transformer.
i i
i O
IIL-5413 E51
Request for Licenso Amendment:
m
(
'Irtended Power Uprate Operation vSummary of Plant Modifications 8.
Perform adjustments to installed plant and switchyard instrumentation as necessary. Examples include the following:
Main steam line high flow.'
11ypass for turbine stop valve closure and turbine control bypass valve faster closure.'
APRM simulated thermal power scram.'
Main generator and switchyard protective devices.
9.
Process computer sollware and data changes.
Proposed Technical Specifications changes (considered an NRC commitment).
v rx IIL-5413 ES 2
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