Text

Amend 33 to License NPF-5,providing Addl & Revised Trip Setpoints Reflecting Design Mods to Reduce Containment Loads from Plant Transients
	ML20076M410
Person / Time
Site:	Hatch
Issue date:	06/29/1983
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Georgia Power Co, Oglethorpe Power Corp, Municipal Electric Authority of Georgia, City of Dalton, GA
Shared Package
ML19268D962	List: ML20076M407; ML20076M410; ML20076M414;
References
	TAC 49989, TAC 51054, NPF-05-A-033 NUDOCS 8307200360
	Download: ML20076M410 (51)
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UNITED STATES

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t NUCLEAR REGULATORY COMMISSION g (, '

,k WASHINGTON. O. C. 20555 iA

.( 9/

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 33 License No. NPF-5 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Georgia Power Company, et al.,

(the licensee) dated February 23, 1983, as supplemented April 19, 1983, and application dated !! arch 30, 1983, as supplemented fiay 10, May 20, and May 26, 1983, comply with the standards and requirements of the Atomic Enerr)y Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chanter I; B.

The facility will operate in confamity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the ccmmon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-5 is hereby amended to read as follows:

(2) Technical Saecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.33, are hereby incorporated in the license. The licensee shall operate the facility in i

accordance with the Technical Specifications.

8307200360 830629 PDR ADOCK 05000366 P

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This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION U$

hn

. Stolz, Chief Opera ing Reactors Branch #4 ivi on of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

June 29,1983 i

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4 ATTACHMENT TO LICENSE AMENDMENT NO. 33 FACILITY OPERATING LICENSE NO. NPF 5 DOCKET NO. 50-366 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contai.n a vertical line indicating the area of change. The overleaf pages are provided to maintain document completeness.

Remove Insert 3/4 2-1 3/42-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4

- 3/4 2-4a 3/4 2-4A 3/4 2-4b 3/4 2-4B 3/4 2-4c 3/4 2-4d 3/4 2-4e 3/4 2-4f 3/4 2-6 3/4 2-6 3/4 2-7b 3/4 2-7b 3/4 2-7c 3/42-8 3/4 2-8 3/4 3-11 3/4 3-11 3/4 3-16 3/4 3-16 3/4 3-19 3/4 3/19 3/4 3 -21 3/4 3-21 N

- ?

2 Insert Remove 3/4 3-27 3/4 3-27 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-32 3/4 3-32 3/4 3 54 3/4 3-54 3/4 4-4 3/4 4 4 3/4 4 da B 3/4 1-2 B 3/4 1-2 B 3/4 2-1 B 3/4 2-1

~

B 3/4 2-4 8 3/4 2-4 1

j B 3/4 2-5 B 3/4 2-5 I

B 3/4 3-6 B 3/4 3-6 B 3/4 4-1 B 3/4 4-1 5-1 5-1 l

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d' 3/A.2 PCwER DISTRIBUTICN LIMITS 3/a.2.1 AVERAGE PLANAR LINEAR HEAT GENERATICN PATE LIMITIFC CONDITION FCR OPEAATION 3.2.1 All AVERAGE PLANAR LItEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed tne limits shown in Figures 3.2.1-1 thru 3.2.1-3.

l APPLICABILITY:

CONDITION 1, when THERMAL PCWER >. 25% of RATED THERMAL power.

ACTION:

With an APLHGR exceecing the limits of Figures 3.2.1-1 thru 3.2.1-8, initiate l

corrective action within 15 minutes and centinue corrective action so that APLKR is within the limit within 2 hcurs or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next a hours.

SURVEILLANCE AECUIREWNTS 4.2.1 All APLHGRs shall be verified to be ecual to or less than the applicaole limit cetermined frcm Figures 3.2.1-1 througn 3.2.1-8:

l a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , c.

Whenever THERMAL PGhER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITIBC CONTRQ. RCD PATTERN for APLHGR.

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HATCH - UNIT 2 3/4 2-1 Amercment No. 27, 23, 33

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ItATE (HAPLllGR) VERSUS AVIRAGE PLANAR EXPOSilRE flGilRI 3.2.1-1

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Ea is ro UNACCEPTABLE OPERATION E-WC 31 a

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RATE (HAPlllGR) VERSUS AVERAGE PLANAR EXPOSURE

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POWER DISTRIEUT10'i LIMITS 3ja.2.2 APRM SET:0!NTS LIMITING CONDITIO" FOR OPERATION 3.2.2 The AFRM flow referenced simulated thermal power scram trip set-point (5) anc control rod block trip setpoint (Sgg) shall be established

according to the following relationships:

l 5 ; (0.66W + 51%)

Spg ; (0.66W + 42L) where:

5 and S e a re in percent of RATED THERMAL POWER, and p

W = Loop' recirculation flow in percent of rated flow.

ADPLICABillTY:

CONDITION 1, when THERMAL POWER ; 25% of RATED THERMAL POUER.

C' ! C ':

kith 5 or 5-~ exceed 1nc the allowatie valuE. ici!1ste Correct 1Ve action v.itnin 15 mihutes and continut corrective action so tnat 5 and 5, are within the reouireo limits

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or recute THERMAL POW lR l

less than 25'. of RATED THERMAL POWE: within tr.e ne,.: 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLA.:CE PE0fi:EMENTS 4.2.2 The CMILPD thall be deternined and the 20DF flow referenced simulated thecta' power scram and control red block trip setecints or LORM readingt ac.usted, as required:

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , a.

r b.

Whenever THERMAL POWE: has been increased by at least 15! Of RATED THERMAL POWER ano steady stats

erating conciticns nave been established, and I

Initially and at least once cer 12 hcurs when the reactor is l

c.

operatinc with a CMFLPD > FRTP.

-t'i th CORE max !M.4 FRACTION OF L::T ING POWER DEN!'TY (CMFLPD) greate-than the fraction of RATED THERMAL ;0VE: ( FRT P ), z..f f 'h'-kp-[-fI-ff.,,3..

up to 95: of RATED THERMAL POWER, rather than adyustinc the ADRM that APRM rencines art setpoints, the APRM gain may be adjustec such the ac]us ac crea ter than or ecual to 100i times CMFLPD. crev tced tha t kPRM reading does not exceed 100L of RATED THERM;, T OUER anc tr e recuire:

gain adjustment increment does not exceed 10 of RATED THERM;L 00EE;.

I HATCH - UNIT 2 3/4 2-5 Amencrent No.

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N POWER DISTRIBUTION LIMITS t

3/4.2.3 MINIMLN CRITICAL POWER RATIO LIMITIFG CCNOITION FOR CFERATICN 3.2.3 The MINIMUM CRITICAL POWER RATIO (PCPR), as a function of average scram time and core flow, shall be ec;ual to or greater than shcwn in Figure 3.2.3-1, Figure. 3.2.3-2 or Figure 3.2.3-3 multiplied by the Kr shcwn in Figure 3.2.3-4, where:

( (. ave t8), whichever is greater,

^-

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t 0 or

=

T A -L8 T

= 1.096 sec (Specification.3.1.3.3. scram time limit to A

notch 36),

k T s = 0.834 1.65 [

Nt

] -(0.059 ),

+

, S. Ni n

co

^

E NE T.

.f ave =

'-1 t

d,Ni i. :.

n = number of surveillance tests performed to date in cycle, Ni = number of active control rods measured in the ith surveillance

test, Ti = average scram time to notch 36 of all rods measured in the ith surveillance test, and Ni = total number of active rods measured in 4.1.3.2.a.

APPLICABILITY: CCNOITION 1, when THERMAL POWER 2 25% PATED THERMAL F3WER ACTICN:

With MCPR less than the applicable limit determined frcm Figure 3.2.3-1, Figure 3.2.3-2 or Figure 3.2.3-3, initiate corrective action within 15 l

minutes and continue corrective action so that MCPR is ecual to or greater than the applicable limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL P0hER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE PEOUIREMENTS 4.2.3 The tCPR limit at rated ficw shall be determined for each type of fuel (8X8R, P8X8R, and 7X7) from Figures 3.2.3-1, 3.2.3-2, and 3.2.3-3 using:

T = 1.0 prior to the initial scram time ceasure ents for the cycle a.

performeo in accordance with Specification 4.1.3.2.a, or HATCH - UNIT 2 3/42-6 A. endment No. 27, 33

=

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o 1.33 1.32 ACCEPTABLE CPERA" ION 1.31 1.30 1

1.29 E

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UNA:CEP"ABLE OPERCION 1.I7 1.26-1.25 1.24-1.23 0.0 0.2 0.4 0.6 0.2 1.0 T

MCPR LIMIT FOR 7X7 FUEL AT RATED FLOW FIGURE 3.2.3-3 i

HAICH - UNIT 2 3/4 2-7b Amendmen:.;o.32 1

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o' PChER DISTRIBUTICN LIMITS 3/4.2.4 LINEAR HEAT GENEPATICN RATE LIMITIM3 CONDITICN FOR OPERATION 3.2.4 All LIrEAR HEAT GENERATION RATES (LHGRs) shall not exceed 13.4 Kw/ft for 8X8R/P8X8R fuel or 18.0 Kw/ft for 7X7 fuel.

APPLICABILITY:

CONDITION 1, when THERMAL POWER 225% of RATED THERMAL POWER.

ACTION:

With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and continue corrective action so that the LHGP is within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL PO)ER to less than 25% of RATED' THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIPEWNTS 4.2.4 LHGRs shall be determined to be ec;ual to or less than the limit; a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Wnen THERMAL P0hER has been increased by at least 15% of RATED THERMAL POWER and steacy state cperating conditions have teen established, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating on a LIMITItG CONTRCL ROD PATTERN FOR LHGR.

HATCH - UNIT 2 3/4 2-8 Amencment No.

33

i.--

T_A__llt.E 3. 3. 2-1 IS01.ATION ACTllATION INSTHilHENTATION 2

D Q

VAI.VE CHOUPS HININUH NUMBEH APPLICAlsl.E OPEHATEI) IlY OPERABl.E CilANNELS OPERATitsNAl.

THIP FilNCT10N S I GNAI.( a )

PEH THIP SYSTEN(Is)(c) CONDITION ACTluN cx U

l.

PHitlARY CONTAINHENT ISOI.ATION a.

Heactor Vessel Water I.evel 1.

l.ow (2B21-N017 A,11, C,11) 2, 5, 6, 10, 2

1, 2, 3 20 11, 12

2.

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1, 2, 3 20 2B21-N025 A, B) 3.

l.ow-1.ow-Low (2B21-N024 A, B and 1

2 1, 2, 3 20 2821-N025 A, B) b.

D.ywell Pressure - liigh 2, 5, 6, 7, 10, 2

1, 2, 3 20 (2C71-N002 A, B, C,11)

II, 12, #,

c.

Hain Steam 1.ine m

3 1.

Radiat. ion - liigh 1, 12, #, (ii) 2 1, 2, 3 21 (2Dil-K603 A,11, C, D) m j

,8 2.

Pressure - Low I

2 1

22 (2il21-N015 A,11, C, D) 3.

Flow - liigh 1, #

2/line 1, 2, 3 21 (21121-N006 A,11, C, D)

(2il21-N007 A, B, C, D) i (2tl21-N008 A,11, C, 1))

(21121-N009 A, 11, C, D) i it. flain Steam 1.ine Tunne 1 liigh Temperat ure - liigh I

2/line I, 2 J 21 li (2H21-N010 A, 11, C, D) j k

( 21121 -Nu l l A, 11, C, 1))

g

( 21121 -N012 A, 11, C, D) g (21121-N013 A, 11, C, 11) 1, 2,(f; 1(1) 2:1 o

e.

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2 h

(21521-N056 A, 11, C, D) w l.

Turbine lluilding Area Temperature - liigh I

2 1, 2, 's 21 j

(.2l161-H00I, 2116 I-H002, 21161-l(003, j

2116 l-H004 )

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i TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION 7

VALVE GROUPS MINIMUM NUMBER APPLICABLE E

TRIP FUNCTION OPERATED BY OPERABLE CHANNELS OPERATIONAL SIGNAL (a)

PER TRIP SYSTEM (b)(c)

CONDITION ACTION

--4 ru 2.

SECONDARY CONTAINMENT ISOLATION 4

a.

Reactor Building Exhaust Radiation - High 6, 10, 12,

2 1,2,3,5 and**

24 (20ll-K609 A, B, C D) b.

Drywell Pressure - High 2, 5, 6, 7, 10, 2

1, 2, 3 24 (2C71-N002 A, B, C, D) 11, 12, #,

c.

Reactor Vessel Water Level - Low 2, 5, 6, 10,,11, 12,

2 1,2,3 24 i

t (2B21-N017 A, B, C, 0) d.

Refueling Floor Exhaust Radiation - High 6, 10,' 12, #,

2 1,2,3,5 and**

24 (2Dll-K611 A, B, C, 0)

I 3.

REACTOR WATER CLEANUP SYSTEM ISOLATION a.

A Flow - High (2G31-N603 A, B) 5 1

1,2,3 25 b.

Area Temperature - High 5

1 1,2,3 25 i

(2G31-N600 A, B, C, D, E, F)

I c.

Area Ventilat' ion a Temp. - liigh 5

1 1, 2, 3 25 (2G31-N602 A, B, C, D, E, F)

I9) d.

SLCS Initiation (NA)

S NA 1, 2, 3 25 e.

Reactor Vessel Water Level - Low 2, 5, 6, 10, ll, 12 2

1,2,3 PS (2B21-N017 A, B, C, D) 9

T AB'_E 3. 3. 2-1 (Conti nued )

ISOLAT!ON a:TL'AT!ON INSTRU.vENTATION ACTION Se in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN ACTION 20 within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Be in at least STARTUP with the main steam line isolation valves ACTION 21 closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l Be in at least STARTUP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , ACTION 22 Be in at least STARTUP with the Group 1 isolation valves closed ACTION 23 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l Establish SECONDARY CONTAINMENT INTEGRITY with the standby ACTION 24 gas treatment system operating within one hour.

Isolate the reactor water cleanup system.

ACTION 25 Close the affected system isolation valves and declare the ACTION 26 affected system inoperable.

Verify power availability to the bus at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ACTION 27 or close the affected system isolation valves and declare the affected system inoperable.

Close the shutdown ec511ng supply and reactor vessel head spray ACTION 28 isolation valves unless reactor steam dome pressure 1 135 psig.

NOTES Actuates operation of the main control room environmental control system in the pressurization mode of operation.

Actuates the standby gas treatment system.

When handling irradiated fuel in the secondary containment.

See Specification 3.6.3.1, Table 3.6.3.1-1 for valves in each valve group.

a.

A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for b.

required surveillance without placing the trip system in the trioced condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

With a design providing only one channel per trip system, an inoperable channel need not be placed in the tripped condition where this would cause c.

In these cases, the inoperable channel shall the Trio Function to occur.

be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION recuired by Table 3.3.2-1 for that Trip Function shall be taken.

d.

Trios the mechanical vacuum pumps, A channel is, OPERABLE if 2 of 4 instruments in that channel are OPERABLE.

e.

May be bypassed with reactor steam pressure < 1045 psig and all turbine f.

s:ce valves closed.

c' w enly 0.WC2 r:.: - isclation valve 2331-FO'4

-r on ly.

5 O;;.s tr ie u:: c 6:

nutes.

3/a 3-15 Arencrent *!o. ;

.ATCH - UN!T :

TAlli.E.1. :1. 2-2 h

ISul.ATION ACTilATION INSThilflENTATION SETPolNTS nx Al.LOWAlil.E i

T. H. _i l.'. Fil.H CT I ON TRIP SETPolNT_

VAI.UE c:

a l.

PRINARY CONTAINNENT IS01.ATION sa a.

Reactor Vessel Water f.evel 1.

I.ow

> 12.5 inches *

> 12.5 inchesi-2.

I.ow Low 5 -38 inches

i -38 inches'*-

l 3.

l.ow I.ow I.ow

[-146'.5 inches *

[-146.5 inches a

h.

Drywell Pressure - liigh 1 2 psig

$ 2 psig

t.,

Haiti Steam Line 1.

Hadiation - liigh 5 3 x full power background 5 3 x full powe background 2.

Pressure - I.ow

> 825 psig

> 825 psig u

3 3.

Flow - liigh

{140%ofrptedflow

{140%ofratedI' low d.

Hain Steam 1.ine Tunne!

Temperature - liigh 1 200*F

$ 200"F e.

Cointenser Vacuum - 1.ow

> 7" lig vacuum

> 7" lig vacuum 1.

Tushine liuilding Area Temp.-liigh

< 200 F

< 200"F SI-;CONilARY CONTA...I.N.t.lE. N.T...I SOI. AT. I O._N.

P l$

l(eactor linilding Exhaust a.

E l(ad i at ion - lii gh

< 60 mr/hr**

< 60 mr/hr*-4 u

< 2 psig i

b.

lirywell Pressure - lin gh

< 2 psig 5

i Heaitor Vessel Watce 1.e v e l - 1.ow

> 12.5 inches *

> 12.5 inchesa l

g it.

l(elueIing Floo Exhan:.t Radi.it ion - liigh

$ 20 mr/lir**

$ 20 mr/hr**

AS c It a :.c r.

Figure !! 'l/4 :)-l.

i n i t i.s l :;et guai sal.

Finaal setposait t o be detein;ined diarisig startup lesting.

TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#

1.

PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level 1.

Low

< 13*

2.

Low Low I 13*

3.

Low Low Low, except MSIVs 513*

b.

Drywell Pressure - High 5 13*

c.

Main Steam Line 1.

Radiation - High***

1 1.0**

2.

Pressure - Low i 13*

3.

Flow - High 1 1.0**

4.

Reactor Vessel Water Level - Low Low Low 1 1.0**

I d.

Main Steam Line Tunnel Temperature - High 1 13*

e.

Condenser Vacuum - Low NA f.

Turbine Building Area Temperature - High NA 2.

SECONDARY CONTAINMENT ISOLATION a.

Reactor Building Exhaust Radiation - High***

1 13*

I b.

Drywell Pressure - High 1 13*

c.

Reactor Vessel Water Level - Low 1 13*

d.

Refueling Floor Exhaust Radiation - High***

1 13*

The isolation actuation instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME. Response time specified is diesel generator start delay time assumed in accident analysis.

- Isolation actuation instrumentation response time.

- - Radiation detectors are exe.mpt from response time testing.

Response

time shall be measured trem detector output or the input of the first electronic component in the channel.

r/ Times to be added to valve movement times shown in Tables 3.6.3-1, 3.6.5.2-1 and 3.9.5.2-1 to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.

33 HATCH - UNIT 2 3/4 3-19 Amendment No.

o TA8LE 3.3.2-3 (Continued)

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds) 4.

REACTOR WATER CLEANUP SYSTEM ISOLATION a.

a Flow - High

< 13*

b.

Area Temperature - High 7 13*

c.

Area Ventilation Temperature aT - High 7 13*

d.

SLCS Initiation NA e.

Reactor Vessel Water Level-Low Low 1 13*

4.

HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION a.

HPCI Steam Line Flow-High

< 13*

b.

HPCI Steam Supply Pressure - Low

-7 13*

c.

HPCI Turbine Exhaust Diaphragm Pressure - High NA d.

HPCI Equipment Room Temperature - High NA e.

Suppression Pool Area Ambient Temp. - High NA f.

Suppression Pool Area AT - High NA g.

Suppression Pool Area Temp. Timer Relays NA h.

Emergency Area Cooler Temperature - High NA 1.

Drywell Pressure - High 1 13*

j.

Logic Power Monitor NA S.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.

RCIC Steam Line Flow - High NA b.

RCIC Steam Supply Pressure - Low NA c.

RCIC Turbine Exhaust Diaphragm Pressure - High NA d.

' Emergency Area Cooler Temperature - High NA e.

Suppression Pool Area Ambient Temp. - High NA f.

Suppression Pool Area aT - High NA l

g.

Suppression Pool Area Temperature Timer Relays NA h.

Drywell Pressure - High 1 13*

1.

Logic Power Monitor NA 6.

SHUTDOWN COOLING SYSTEM ISOLATION a.

Reactor Vessel Water Level - Low NA b.

Reactor Steam Dome Pressure - High NA HATCH - UNIT 2 3/4 3-20 i

e--

-ee,.

.o p,e, m.-

-ev-e, e

a,e-

--n4.s e-

-m.

c,

~

TABLE 4.3.2-1 ISOLATION ACTUN110N INSTRUHENTATION SURVEILLANCE REQUIREMENTS M

9 CilANNEL OPERATIONAI.

CilANNEL FUNCTIONAL CilANNEL CONillTIONS IN WillCli TRIP FilNCTION CllECK TEST CALIBRATION SURVEILLANCE REQUIRED c

H 1.

PRIMARY CONTAINHENT ISOI.ATION t4 a.

Reactor Vessel Water Level 1.

Low D

H Q

1, 2, 3 2.* Low Low D

H Q

1, 2, 3 3.

Low Low Low D

H Q

1, 2, 3

,1 L

h.

Drywell Pressure - liigh NA H

Q 1, 2, 3 c.

Main Steam Line 1.

Radiation - liigh D

W" R

1, 2, 3 2.

Pressure - Low NA H

Q l

3.

Flow - liigh 1)

H Q

1, 2, 3 v

D o

d.

flain Steam Line Tunnel m

Temperature - liigh NA R

R 1, 2, 3 e.

Condenser Vacuum - Low NA H

Q 1, 2#, 3#

1.

Turbine Building Area Temp. -

lii gh NA H

R 1, 2, 3 i

2.

SECONI)ARY CONTAINHENT ISOI.ATION a.

Reactor Building Exhaust Radiation - liigh 1)

H" R

1, 2, 3, 5 and

F i

g h.

Drywell Pressure - liigh NA H

Q 1, 2, 3 O'

r.

Reactor Vessel Water f.evel -

g 1.ow I)

H Q

1, 2, 3 If j

d.

Relueling Floor Exhaust Radiat ion - liigh 1)

H("

Q 1, 2, 3, 5 and

d ww

When handling irradiated luel in the secondary containment.

//When reactoi steam pressure > 1045 psig and/or any turbine stop valve is open.

j alustrument alignment using a standard current source.

44

d TABLE 4.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL x

l h

CHANNEL FUNCTIONAL CliANNEL CONDITIONS IN WillCil x

TRIP FUNCTION CllECK TEST CALIBRATION SURVEILLANCE REQUIRED e

c 3.

REACTOR WATER CLEANUP SYSTEM ISOLATION 2

1 l;

[

a.

A Flow - High D

M R

1, 2, 3 b.

Area Temperature - High NA M

R 1, 2, 3

]

c.

Area Ventilation A Temperature - High NA M

R 1, 2, 3 d.

SLCS Initiation NA R

NA 1,2,3 e.

Reactor V'essel Water Level -

Low D

M i

Q 1, 2, 3 4.

IIIGH PRESSURE C00iANT INJECTION R

SYSTEM ISOLATION h

a.

HPCI Steam Line Flow-High NA M

Q 1, 2, 3 y

b.

HPCI Steam Supply Pressure-Low NA M

Q 1, 2, 3 f

c.

HPCI Turbine Exhaust Diaphragm Pressure-High NA M

Q 1, 2, 3 d.

IIPCI Equipnent Room Temperature - High NA M

R 1, 2, 3 e.

Suppression Pool Area Ambient Temp. - High NA M

R 1, 2, 3 f.

Suppression Pool Area AT -

liigh NA M

R 1, 2, 3 9

Suppression Pool Area Temp.

Timer Relays NA SA R

1, 2, 3 h.

Emergency Area Cooler Temp. -

liigh NA M

R 1, 2, 3 1.

Drywell Pressure - liigh NA M

Q 1, 2, 3

j. Logic Power Monitor NA R

NA 1, 2, 3 e

4 lAllt.E

1..l. l-1 (Contgnueil) g Etti.ljt.l.NfY 4:ol!I' Cool.t ht; SYSTEti ACTilAllesN INSTHutlEgl'ATION

.c HINIHllH NUNilEH A PI't. l CAlti l.

j OPEHAlli.E CilANNEl.S Ol'i'.HA 1 l oN3 g

IHil' l pNCTiflN PER THil' SYSTEH, CONul'Iluun is 3.

ll1611 PHESSultE Chol. ANT IN.!ECTioN SYSTEN Heactor Vessel Water I.evel - 1.ow l.ow (2tl21-NO31 A,B,C,D) 2 1, 2, I

a.

h.

firywell Pressure - liigh (2 Ell-N0ll A,B,C,D) 2 l'2' i

2((b W )

c.

Conitensate Storage Tank 1.evel-l.ow (2E41-N002, 2E41-N00's)

I, 2, I

b) h-)

al. Suppression Chamber Water f.evel-liigh (2E41-N015A,B) 2 1, 2, i

g e.

l.ogic Power Honitor (2E41-KI)

I 1, 2, i 1.

Heactor Vessel Wat er f.evel-lingt (2tl21-NOI ? II,D) 2 1, 2, 1

4.

AlliotlATIC I)El'l(ESSilitlZATION SYSTEtt

~

lirywell Pressur e - liigh (.: Ell-Noll A,B,C,1))

2 1,

2, i

a.

h.

Reactor Vessel Water I.evel - 1.ow Low Low

( 2il21 -NO J I A,II,l.',1))

2 1, 2, I

ADS Timer (2tl21-K5A,II)

I

!, 2,

'l a

r

.l.

Iteact or Vessel Water f.evel-l.ow (Permissive) (2B21-N042A,II) 1 I,2

'l Core Spray Pump I)ischange l'ressure - High (Permissive) e

( 2E21 -N008A,ll; 2 E21 -N 00') A,ll) 2 1,

2,

'l l.

Hillt ll.PCI HoDE) l' ump th :.s ha rge Pressure - liigli (Permissive)

( 2 E l l -N016 A, ls, C,11; JE l l - NO20A,ll,C,ll) 2/loog)

I,

2, i

g.

Cont r ol Power tion a t os ( 2ll2 8 -K l A,ll) l/ bus I,

2, 1

l.tsW l.oW SET S/l(V SYSTEN I

a.

Heart or St eam 11ome Picssuse - lla gh II'eimissive) g 12il21 - N t:20 A,lt, s., D i 2

1, 2, I

o S.

la) Alasm only.

When s nope t als l e, ve sly power availability to the inns at least once pe 12 3

lion s :. or.lcslate the system a nope rain i c 5

( i, j l's.v ole:. a:gnai ta lil'Ci pinnp sus t on va 1ver. on1y.

[j

(. I Wh..n cathei s h.enne l i.I the aut omat a r t ransles lugar is inopesable, align lil*Cl pump surt ion 3

to t he supps e: s ion pool.

6.

il llPCI asol AlJS age ne.t s espe i s cel t o be OPl.ItAltl.1{ wi th reas t ot steam slome picssure - l'in ps e g.

, y sj

l y.

nl

s Q

TAOL.E 3.3.3-2 EHEIlCEtCY CORE COOLING SYSTEM ACillATI0il INSTRtlMENTATION SEIPOINTS y

Al.l.0WNILE m

i o

TilIP FiltCTION 1 RIP SEIP0 INT val.uE 1.

COIE SPitAY SYSTEM a.

Reactor Vessel Water Level - Low Low Low 2 -146.5 inchese 2-146.5 inches

b.

Drywell Pressure - H1 h 12 psig 12 psig 0

c.,

Reactor Steam Dome Pressure - Low 1500 psig 1500 psig d.

Logic Power 2nitor HA NA 2.

LOW PilESSillE COOLANT INJECTION H0DE OF Rift SYSIFH a.

Drywell Pressure - liigh 12 psig 5 2 psig i

u' b.

Reactor Vessel Water Level - Low Low Low 2.-146.5 inches 2.-146.5 inches

a 2

c.

Heactor Vessel Shroud Level - N1 h y_-203.5 inches

2-203.5 inches

l 0

u d.

Reactor Steam Dome Pressure-Low

< 50(pps10

.5500 psig h

e.'

Reactor Steam Dome Pressure-Low

.5335 ps10

$_335 psig f.

filft Punp Start - T!me Delay Relay

1) Punps A, 8 and D 10 + 1 seconds 10 + 1 seconds

2) Punp C 0.5 0.5rseconds 0.5 + 0.5 seconds l

Lo0 c Power tenitor HA,~

~

9 NA l

1

See Oases FIgste B 3/4 3-1.

i i

r

]

e 8

9 en f

TABLE 3.3.3-2 (Continued) gi ENERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS M

m ALLOWABl.E TRIP FUNCTION TRIP SETPOINT VALUE M

3.

111011 PRESSURE C001. ANT INJECTION SYSTEN w

a.

Reactor Vessel Water Level - low Low 1 -38 inches

1 -38 inches

b.

Drywell Pressure-High 5 2 psig i 2 psig c.

Condensate Storage Tank Level - Low 3 0 inches **

> 0 inches **

d.

Suppression Chamber Water Level - High*

$ 151 inches 5 151 inches e.

Logic Power Monitor NA NA f.

Reactor Vessel Water Level-High 1 58 inches 5 58 inches 4.

AUTOMATIC DEPRESSUHlZATION SYSTEH i

t' a.

Drywell Pressure-liigh i 2 psig 5 2 psig b.

Reactor Vessel Water Level - I.ow Low Low 1 -146.5 inches

1 -146.5 inches

T' c.

ADS Timer

< 120 seconds

'd d.

Reactor Vessel Water Level-Low

[12.5 inches *

[12.5 inches

Core Spray Pump I)ischarge Pressure - liigh 1 100 psig 1 100 psig e.

f.

RilR (I.PCI Hol)E) Pump Discharge Pressure - liigh 1 100 psig 1 100 psig g.

Control Power Monitor NA NA 5.

LOW I.0W SET S/RV SYSTEN Reactor Steam I)ome Pressure - liigh 5 1054 psig

$ 1054 psig a.

E a

8-Br+

5 t

Ul

~

See Bases Figure B 3/4 3-1.

L'- Equivalent to 10,000 gallons of water in the CST.

'l of l

- b' TABLE 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES ECCS RESPONSE TIME (Seconds) 1.

CORE SPRAY SYSTEM i 27 2.

LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM 1 40 3.

HIGH PRESSURE COOLANT INJECTION SYSTEM 1 30 4.

AUTOMATIC DEPRESSURIZATION SYSTEM NA 5.

ARM LOW LOW SET SYSTEM i1 l

l a

l

(

l l

HATCH - UNIT 2 3 / J. 3-30 Amendment No. 33 t

I

-w.,

ev-4

__.---u

,-_.7

___.-p 9

TABLE 4.3.3-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS x

D9 CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED g

l 1.

CORE SPRAY SYSTEM N

a.

Reactor Vessel Water Level -

Low Low Low D

M Q

1,2,3,4,5 b.

Drywell Pressure - High NA M

Q 1, 2, 3 c.

Reactor Steam Dome Pressure - Low NA M

Q 1,2,3,4,5 d.

Logic Power Monitor NA R

NA 1,2,3,4,5 2.

LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM a.

Drywell Pressure - High NA M

Q 1, 2, 3 10 b.

Reactor Vessel Water Level -

Low Low Low D

M Q

1, 2, 3, 4*, 5*

Y'

'c. Reactor Vessel Shroud Level -

E' High D

M Q

1, 2, 3, 4*, 5*

d.

Reactor Steam Dome Pressure - Low NA M

Q 1, 2, 3, 4*, 5*

e.

Reactor Steam Dome Pressure - Low NA M

Q 1, 2, 3, 4*, 5*

f.

RHR Pump Start-Time Delay Relay NA NA R

1, 2, 3, 4*, 5*

g.

Logic Power Monitor NA R

NA 1, 2, 3, 4*, 5*

Not applicable when two core spray system subsystems are OPERABLE per Specification 3.5.3.1.

I i

1

TAliLE 4. 3. 3-1 (Coutinued) g EHERGENCY CORE COOLING SYSTEM ACTilATION INSTRUMENTATION SURVEILLANCE REQUIREtlENTS M

c CilANNEL OPERATIONAL.

8 CllANNEL FUNCTIONAL CNANNEL CONDITIONS IN WillCil y

TRIP FUNCTION CllECK TEST CALIBRATION SURVEILLANCE REQUIREI)#

a E

').

tilGli PRESSURE COOLANT INJECTION SYSTEM a.

Reactor Vessel Water Level -

Low Low D

H Q

1, 2, 3 b.

Drywell Pressure-liigh NA H

Q 1, 2, 3

'I c.

Condensate Storage Tank I.evel -

Low NA H

Q 1, 2, 3 I

d.

Suppression Chamber Water Level - High NA H

Q 1, 2, 3 Logic Power Honitor NA R

NA 1, 2, 3 e.

l.

Reactor Vessel Water Level-liigh NA H

Q 1, 2, 3

,i 4.

AUTOHATIC DEPRESSUNIZATION SYSTEM y

a.

Drywel1 Pressure-liigh NA H

Q 1, 2, 3 b.

Reactor Vessel Water Level -

[

Low Low Low D

H Q

1, 2, 3 J

to c.

Alls Timer NA NA R

I, 2, 3 d.

Reactor Vessel Water I.evel - Low D H

Q 1, 2, 3 j

c.

Core Spray Ptunp Discharge Pressure - liigh NA H

Q 1, 2, 3 1.

RilR (i.PCI HODE) Pump Discharge I

Pressure - liigh NA H

Q I, 2, 3 g.

Control Power Honitor NA R

NA 1, 2, 3 E

5.

1_.0_W l_.0._W_SE_T___S/RV SYSTEN.

S' a.

Reactor Steam Dome Pressure -

lii gh S

H R

1, 2, 1

,9

o d

// lil Cl anil dliS a re'~not sequireil to be OPERAhl.E with reactor steam dome pressure < 150 psig.

i i

e o

9

INSTRUMENTATION POST-ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6.4 The post-accident monitoring instrumentation channels shown in Table 3.3.6.4-1 shall be OPERABLE.

APPLICABILITY: CONDITIONS 1 and 2.

ACTION:

With one or more of the above required post-accident monitoring a.

channels inoperable, either restore the inoperable channel (s) to OPERABLE status within 30 days or be in at le&st HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.6.4 Each of the above required post-accident monitoring instrumen-tation channels shall be demonstrated OPERABLE by perfonnance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown tn Table 4.3.6.4-1.

HATCH - UNIT 2 3/4 3-53

y.

TABLE 3.3.6.4-1 2

R h

POST-ACCIDENT HONITORING INSTRilHENTATION HINIMUM CHANNELS INSTRUNLNT OPERABLE 1.

Reactor Vessel Pressure (2C32-R605 A, B, C) 2 2.

Reactor Vessel Wter Level (21121-R610, 2B21-R615) 2 i

3.

Suppression Chamber W ter Level (2T48-R622 A, B) 2 4.

Suppression Chamber W ter Temperature (2T47-R626, 2T47-R627) 2 5.

Suppression Chamber Pressure (2T48-R608, 2T48-R609) 2 6.

Drywell Pressure (2T48-R608, 2T48-R609) 2 s

7.

Drywell Temperature (2T47-R626, 2T47-k627) 2 U

8.

Post-LOCA Gamma Radiation (2Dil-K622 A, B, C, D) 2 9.

Drywell 11

-0 Analyzer (2P33-R601 A, 8) 2 2

2 10.a)Salety/ Relief Valve Position Primary Indicator (2B21-N301 A-H and K-M) b) Safety /Reliel Valve Position Secor.d'ary Indicator (2il21-N004 A-H and K-M)

Ilg all either the primary ur secondary indication is inoperable, the torus temperature will be monitored q

at l e.as t once per shift to observe any unexplained temperature incre.ases which might be indicalive g

of an open SRV. With both the primary and secondary monitoring channels of an SRV inoperable, either verily that the S/RV is closed through monitoring the backup low low set logic position o

g indicators (2il21-N302 A-Il and K-ti) at least once per shif t or restore suf ficient inoperable channels such that no more than one SRV has both primary and secondary channels inoperable within 7 days or be in at least hot shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d m

O 4

REACTOR COOLANT SYSTEM IDLE RECIRCULATION LOOP STARTUP-LIMITING CONDITION FOR OPERATION 3.4.1.3 An idle recirculation loop shall not be started unless the temperature differential between the reactor coolant within the dome and the bottom head drain is < 145'F, and a.

The temperature differential between the reactor coolant I

within the idle loop to be started up and the coolant in the reactor pressure vessel is 1 50*F when both loops have been idle, or i

b.

The temperature differential between the reactor coolant within the idle and operating recirculation loops is < 50*F when only one loop has been idle, and the operating loop flow rate is 1 50% of rated loop flow.

APPLICABILITY: CONDITIONS 1, 2, 3 and 4.

ACTION:

With temperature differences, and[er flow rate exceeding the above limits,-

suspend startup of any idle recirculation loop.

SURVEILLANCE REQUIREMENTS 4.4.1.3 The temperature differential and flow rate shall be detemined to be within the limit within 30 minutes prior to startup of an idle l

recirculation loop.

I i

e i

HATU., UNIT 2 3/4 4-3 7i 3 gym 1..

REACTOR CCOLANT SYSTEM J/4.4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2.1 The safety valve' function of the following reactor coolant system safety / relief valves shall be OPERABLE with the mechanical lift settings within t 1% of the indicated pressures *.

4 Safety-relief valves @ 1090 psig.

4 Safety-relief valves @ 1100 psig**.

3 Safety-relief valves @ 1110 psig**.

APPLICABILITY: CONDITIONS 1, 2 and 3.

ACTION:

For low-low set valves, take the action required by Specification a.

3.4.2.2.

For ADS valves, take the action required by Specification 3.5.2.

b.

With one or more safety / relief valves stuck open, place the reactor mode switch in the Shutdown position, c.

With one or more S/RV tailpipe pressure switches of an S/RV declared l

inoperable and the associated S/RV(s) otherwise indicated to be open, place the reactor mode switch in the shutdown position.

d.

With one S/RV tailpipe pressure switch of an S/RV declared inoperable and the asacciated S/RV(s) otherwise indicated to be closed, plant operation may continue. Remove the function of that pressure switch i

from the low low set logic circuitry until the next COLD SHUTDOWN.

Upon COLD SHUTDOWN, restore the pressure switch (s) to OPERABLE status before STARTUP.

With both S/RV tailpipe pressure switches of an S/RV declared inop-e.

erable and the associated S/RV(s) otherwise indicated to be closed, restore at least one inoperable switch to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and l

in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.2.1 The tail-pipe pressure switches of each safety / relief valve shall be l

demonstrated OPERABLE by performance of:

i a.

CHANNEL FUNCTIONAL TEST:

1.

At least once per 31 days, except that all portions of the channel inside the primary containment may be excluded from the CHANNEL FUNCTIONAL TEST, and 2.

At each scheduled outage of greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during which entry is made into the primary containment, if not performed within the previous 31 days.

b.

CHANNEL CALIBRATION and verifying the setpoint to be SS pstg, with an allowable tolerance of +15 psig and -5 psig, at least once per 18 months.

The Itft setting pressure shall correspond to ambient conditions of the 4

valves at nominal operating temperature and pressure.

- Up to two inoperable valves may be repl. iced with spare OPERABLE valves with lower setpoints of 1090 and 1100 ps tg, respectively, untti the next refueling outage.

4 HATCH - UNIT 2 3/4 4-4 Amendnent Co. 33

- *

f go ***

, - - _ _ _.~

,---n,

-+

REACTOR COOLANT SYSTLM

$AFETY/ RELIEF '.'ALVES LOW-LOW SET FtNCTION LIMITING CONDITION FOR OPERATION 3.4.2.2 The relief valve function and the low-low set function of the following reactor coolant system safety /reli'ef valves shall be OPERABLE with the following low-low set function lift settings:

Low Low Set Allowable Value (psia)*

Valve Function Open Close Low

< 1010

< 360 Medium Low I 1025 I 875 Medium High 51040 5890 High 1 1050 1 900 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3 ACTION:

s.

With the relief valve function andyor the low-low set function of one of the above required reactor coolant system safety / relief valves inoperable, restore the inoperable relief valve function'and low-low set function to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With the relief valve function and/or the low-lov set function of more than one of the above required reactor coolant system safety / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIRE."ENTS i

l 4.4.2.2 The low-low set relief valve function and the low-low set function pressure actuation instrumentation shall be demonstrated OPERABLE by performance of a:

a.

CHANNEL FUNCTIONAL TEST, including calibration of the trip unit and the dedicated high steam dome pressure channels **. at least once per month.

i l

b.

CHANNEL CALIBRATION, LOGIC SYSTEM FUNCTIONAL TEST and simuisted auto:aat:c l

operation of the entire system at least once per refueling outage.

"The 1 tit setting pressure of the valves is defined ta subsectica 3/4 3.4.2.1 l

The accuracy of the low-low set setpoints is deitnea to,e tne ac:uracy of tne instrumentation controlling the setpotnts of the ic -iaw set valves.

i

- The setpotat for dedicated high steam dome pressure channels is ;ess enan or j

equal to 1054 psig.

RATCH - UNIT 2 J/a 4 a Amendment No. 33

d 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold xenon-free condition and shall show the core to be subcritical by at least R + 0.28% aK or R + 0.38%aK, as appropriate.

The value of R in units of %4K is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of R must be positive or zero and must be datermined for each fuel loading cycle.

Satisfaction of this limitation can be best demonstrated at the time of fuel loading but the margin must be determined anytime a control rod is incapable of insertion. This reactivity characteristic has been a basic assumption in the analysis of plant performance.

Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN. The nighest worth rod may be determined analytically or by test. The SHUTDOWN MARGIN is demonstrated by an insequence control rod withdrawal at the beginning of life fuel cycle and, if necessary, at any future time in the cycle if the first demonstration indicates that the margin could be reduced as a function of exposure. Observation of subcriticality in this condition assuret subcriticality with.the most reactive control rod fully witndrawn.

3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is small, a careful check on actual conditions to the predicted conditions is necessary, and the changes in reactivity can be inferred from these comparisons of rod patterns. Since the comparisons are easily done, frequent checks are not an imposition on normal operations. A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated. A change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated transients.

HATCH - UNIT 2

- B 3/41-1

.-7.

-m-.

4 W

-4 PEACTIVITY CCNTROL SYSTEYS EASES 3/4.1.3 CONTROL RCDS The specifications of this secticn ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the ceintrol red insertion times are consistent with those used in the accicent analysis, and (3) the potential effects of the red drop accicent are limited.

The ACTION statements permit variaticns from the basic recuirements, but at the same time imose more restrictive criteria for centinued cceration.

A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shace will De kept to a minimum.

The recuirements for the various scramtime measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rec drive mechanism coulc be a generic problem; therefore, with a centrol roc immovable because of excessive friction or mecnanical interference, cperation of the reacter is limited tc a time period whicn is reascnacle to determine the cause of the inoperability and at the same time. prevent cperation with a large number cf incceracle centrol rods.

Control rods tnat are inceerable for other reascns are permittec te ce taken out of service provicec that those in the nonfully-insertec cesiticn are censistent with the SHUTDCWN MARGIN requirements.

The numoer of centrol rocs permittec to be incoeracle coulc be more tha-the eight allowec ey the scecification, but the occurrerce of eignt incperable rocs coulc be incicative of a generic prcblem and the reacter must be shutcown fcr investigaticn and resolution of the procier.

The centrol rod system is analyzed to bring the reacter succritical at a rate fast encugn to prevent the MCPR frem becoming less than 1.07 curirq tna limiting power transient analy:ec in Secticn 15 of the FSAR.

This analysis shows tnat tne negative reactivity rates resulting frcm the scram with the average rescense of all the crives as given in the scecifications provice l

the recuired protection and MCPR remains greater than 1.07.

The occurrence l

of scram times longer than those scecified shoula be viewec as an incicaticn of a systematic proolem with the roc drives and therefore the surveillance interval is reduced in order to prevent coeration of the reacter for long periods of time with a potentially serious problem.

Control rods with incoerable accumulators are declared inecerable anc Specification 3.1.3.1 then applies.

This prevents a pattern of incceracle accumulators that would result in less reactivity insertion en a scram HATCH - UNIT 2 E 3/4 1-2 Amercrent Nc. 33

.o 1

3/4.2 POER DISTRIBUTION LI!!ITS BASES The specifications of this section assure that the peak clacding tencerature following the postulated design basis loss-of-coolant accicent will not exceed the 2200cF limit specified in the Final Acceptance Criteria (FAC) issued in June 1971 considering the postulated effects of fuel pellet densification.

3/4.2.1 AVERAGE PLANAR LI TAR HEAT GEERATION RATE This specification assures that the peak cladding temocrature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50, Appendix K.

The peak cladding tengerature (PCT) following a

postulatec loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is cependent only secondarily on the rod-to-rod power distribution within an assembly.

The peak clad temperature is calculated assuming an LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for censification.

This LHGR times 1.02 is used in the heatuo coce along with the exposure dependent steacy state gap conductance and rod-to-roc local peaking factor.

The Technical Specification APLHGR is this LHCR of the highest powered rod divided by its local peaking factor.

The limiting value for APLHGR is shown in the figures in Technical Specification 3/4.2.1.

The calculational procedure used to establish the APLHGR shown in the figures in Tecnnical Specification 3/4.2.1, is based on a loss-of-coolant accicent analysis.

The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code emolayed in the analysis is presented in Reference 1.

Differences in this analysis comparec to previous analyses performed with Reference 1 are:

(1) the analysis assumes a fuel assembly planar power consistent with 102% of the MAPLHGR l

shown in the figures in Technical Specification 3/4.2.1; (2) fission procuct l

l decay is computed assuming an energy release rate of 200 EV/ fission; (3) pool boiling is assumed after nucleate boiling is lost during the ficw stagnation period; and (4) the effects of core spray entrainment and counter-current flow limitiation as described in Reference 2, are inclucec in the Itflooding calculations.

A list of the significant plant input parameters to the loss-of-coolant accident analysis presented in bases Table B 3.2.1-1^.

HATCH - LNIT 2 8 3/4 2-1 Amencment No. 27, l$, 33

o Bases Table B 3.'2.1-1 SIGNIFICANT INPUT PARANETERS TO THE -

l LOSS-OF-COOLANT ACCIDENT ANALYSIS FOR HATCH-UNIT 2 Plant Parameters:

Core Thermal Power...............

2531 Mwt which corresponds to 105% of license core power

0 Vessel S team Output..............

10.96 x 10 lbm/h which corresponds to 105% of rated steam flow Vessel Steam Oome Pressure.......

1055 psia Design Basis Recirculation Line Break Area For:

2 a.

Large Breaks............ 4.0, 2.4, 2.0, 2.1 and 1.0 ft

~

2 b.

Small Breaks............ 1.0, 0.9, 0.4 and 0.07 ft Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kw/ft)

FACTOR RATIO Initial Core 8x8 13.4 1.4 1.18 A more detailed list of input to each model and its source is presented in Section II of Reference 1 and subsection 6.3.3 of the FSAR.

This power level meets the Appendix K requirement of 102%.

The core heatup calculation assumes t bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification

~

linear heat generation rate limit.

HATC'd - UNIT 2 3 3/4 2-2 4

"m t

erem=m ww s

e>

ne,

,s_

= - - - -

m.

v

--.-----r

--.-m-p------------+c----

ev.i.+

r"-

.g [.

B

'

POWER DISTRIBUTION LI!!!TS SASES 3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Lim'its of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER. The scram setting and rod block functions of the APRM instru-ments or APRM readings must be adjusted to ensure that the MCPR does not become less than 1.0 in the degraded situation. The scram settings and rod block settings or APRM readings are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and CMFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.07, and an analysis of abnormal I

operational transients.

For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady l

state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting as given in Specification 2.2.1.

To assure that the fuel cladding integrity Safety Limits are not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to detennine which results in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity l

insertion, and coolant temperature decrease.

The limiting transient which determines the required steady state MCPR limit is the load rejection trip with failure of the turbine bypass.

l This transient yields the largest a CPR. When added to the Safety Limit l

MCPR of 1.07 the required minimum operating limit MCPR of Soecification 3.2.3 is obtained.

HATCH - UNIT 2 8 3/4 2-3 Amendment No. J*, 21 Y

Y

?

)

.o POWER DISTRIBUTICN LIMITS

?n,AE?m?

MINIMJM CRITICAL POWER RATIO (Ccntinued)

The evaluaticn of a given transient begins with the system initial parameters shown in FSAR Table 15.1-6 that are ircut to a GE-core dynamic behavior transient computer program described in NEDO-10802(3).

Also, the voic reactivity coefficients that were ircut to the transient calculaticnal procecure are based on a new methed of calculation termed NEV which provices a better agreement between the calculated anc plant instrument power distributions.

The cutputs of this program clong with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic SCAT code cescribec in NEDO-20566(1).

The princical result of this evaluation is the recucticn in MCPR causec by the transient.

The purpose of the Kf factor is to define cperating limits at other than related flow conditions. At less than 10C% of rated flow the recuire MCPR is the product of the o0I! rating limit MCPR and the Kr facter.

Specifically, the Kr facter provides the recuired thermal margin to protect against a flow ircrease transient.

The most limiting transient initiatec from less than rated flow conditions is the recirculation puma sceec w caused by a meter-generator speec control failure.

Fcr coeration in the automatic flow centrcl mece, the Kf factors assure that the operating limit MCPR of Specification 3.2.3 will not be violated shculd the mest limitino transient occur at less than rated flow.

In the manual flow control mcce, the Kr factors assure that the Safety Limit MCPR will not be violatec thould the most limiting transient.cccur at less than rated flow.

The Kr facter values snown in Figure 3.2.3-4 were cevelecee l

l generically and are acolicacle to all EWR/2, swr /3 anc EWR/4 reacters.

The l

Kf factcIs wre deriven using the ficw cent:01 line ccrrescencing tc RATED THERMAL POWER at ratec core flow.

For the manual flow control moce, the Kf factors were calculatec Suen that the maximum flow rate, as limited by the pump scoco tube set point anc the cer esponcing THERMAL POWER along the rated flow centrol line, the limiting bundle's relative pcwer, was adjusted until the MCPR was sligntly above the Safety Limit.

Using this relative bundle power, the MCPRs were calculated at different points along the rated ficw contr:1 line i

cor:esconding to different core ficw.

The ratio of the MCPR calcula,tec at a given point of core flow, diviced by the ocerating limit MCPR, Cetermines l

the Kr.

l l

l HATCH - UNIT 2 E 3/A 2-4 AmenCrant NC. 33 l

O PCWER DISTRIBUTION LIMITS EASES MINIMUM CRITICAL POWER RATIC (Continued)

For cperation in the automatic flow control mode, the same procecure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at RATED THERMAL POWER and ratec flow.

The Kr factors shown in Figure 3.2.3-4 are conservative for the l

General Electric Plant operation because the operating limit MCPRs cf Specification 3.2.3 are greater than the original 1.20 operating limit MCPP used for the generic derivation of Kr.

At THERMAL POWER levels less than or equal to 25% of RATED THERf'".

POWER, the reactor will be operating at minimum recirculation pumo sceed anc the moderator void content will be very small.

For all cesignatec control roc patterns which may be emoloyed at this point, operating plant experience inoicated that the resulting MCPR value is in excess of recuirements cy a consicerable margin.

With this icw void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. Curing initial startte testing of the plant., an MCPR evaluation will be mace at 25% of RATED THERMAL PO)ER with mininum recirculation pumo sceec.

The MCPR margin will thus be cemonstrated such that future MCPR evaluaticn belcw this Acwer level will be shown to be unnecessary.

The cally recuirement for calculating MCPR above 25% of RATED THERMAL PCwER is sufficient since power distribution shifts are very slow when there have nct been significant power or control rod changes.

The recuirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power snace, regardless of magnitude that could place operation at a thermal limit.

3/4.2.4 LItEAR HEAT GENERATION RATE The LHGR sCecification assures that the linear heat generation rate ir any rod is less than the cesign linear heat generation even if fuel pellet densification is pcstulated.

I HATCH - UNIT 2 8 3/4 2-5 Amencment No. 33 l

l l

l

[

o F

p0WER DISTRIBUTION LIMITS BASES

References:

1.

General Electric Company Analytical Model for loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K NEDO-20566 (Draft), August 1974.

2.

General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to USAEC by letter, G. L. Gyorey to V. Stallo, Jr., dated December 20, 1974.

3.

R. B. Linford, Anslytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NEDO-10802).

l l

HATCH - UNIT 2 8 3/4 2-6 m

i t

?

NS"':U*ENT ATION EASE 3 K.NI~.F*NG INS RudENTAT!CN (C:ntinuec) r!:E E:c_7ICN INS :UaENTATION (C:ntintec)

In :ne event tnat a ;::::icn of tne fire catecti:n inst: nentati:n is inc: era:1e, increasing tne f:ec' ency of fi:e ;:st::1s in :ne affe :ac areas is rem 4 ec :: ;=vice :etecti:n cacacility until :ne ineceracle ins:=.;mentati:n is rests:sc :: CPERASILITY.

3/4.3.7' 71.REINE CVE"s*PEC MCTECTICN SYST94 This s ecifi:stien is 'pr= viced to ensure that the tu: cine eversceed

=:ecrien system instrmientatien and tne ::==ine scoed cent::1 valves are CPERAELE acc will pr: tac: :ne ::c=ine f=m exesssive everscesc.

P=tacti:n f=m ::::ine excassive evers:eec is recuirac since exesssive evers:eec of tne tu=ine c=uld generata potentially camaging missiles wni:n -'u incact and camage safety-relatec'c=mocnents, M'%t or stremas.

3/A.3.3 CEGACC STAT!CN VCLTAGE P8vm.ICN INS"MU(NTAT!CN The uncervcitate relays small aut:matica11y initite the disw.e.;;.icn of cffsi:a ::cwer scu=es vnenever the vcitage set;:cint anc time celay 1'*d ts have

een ex:secac.

This action small p= vise voltage p= tecti:n fer =e eme=;ency

cwer systems t:y ;::eventing sustained cegraced voltage ::::nci-*-r s cue t= the offsits ;::wer scu== and interacti:n t:stween the offsite anc cnsita emergency

wer systems.

The uncervoltage :alays have a time celay enarsetaris ic :nat e

= vices

: =tacti:n against !:cth a 1 css of voltage and ceg:scec voltage

nciti n and tnus sin **4 = the effect.cf snc:: curati:n cistu :ances witncut exceecing tne maxi..ium time celay, inclucing mar.,in, taa is assumac in :ne eSAR ac=1:ent analyses.

l

(

HATCH-UNIT 2 3 3/4 3-5 Ame c=ent No. 27 i

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+58 TARWAY (7) 559

+42 C/ PAC (4) 549

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(3) 529.5

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(2) 479 38 TARWAY (LL2)

=

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(1) 370.5

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(0) 313.5

207.5 TARWAY

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- 577

+60 -

+60 --

+60 - -

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(5)

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MPCI &

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"CIO (4)- 32 La A W M TRIPS

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DRTER SEIRT ERO lllr 0

0--. REacMR 517 -

0

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500 - =

TEED N

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MAIN MSI?E SPRAY

=

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TRIPS W,

450 -

PtHrs.

goo---

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352.56 START CIESEL.

35 7 d

CLost "SIV's AAV CONTRIBUTE *O A.D.S.

g3

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PERMIS3 DT 300- -

ACTn't FttL 25 0 - =

1P 317 " "

200" E RECIRC.

j

=L 78.56 -DISCHARG RECIRC.

Notz;E

$UCTTCN-161.5-nozzzz 150--

o 1*~

REACTOR VESSEL WATER LEVEL 50 --

Bases Figure B 3/4 3-1 o --

HATCH - UNIT 2 3 3/4 3-6 Amendment No. 30, 33

B A

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM Operation for longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with a reactor core coolant recir-culation loop inoperable is prohibited until an evaluation of the perform-ance of the ECCS during one loop operation has been performed, evaluated and determined to be acceptable.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does present a hazard in case of a design basis accident by increasing the blowdowa area and elimi-nating the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.

In order to prevent undue stress on the vessel nozzles and bottom head region the recirculation loop temperatures shall be within 50 F of each other prior to startup of an idle loop. Since the coolant in the bottom of the vessel is at a lower temperature than the water in the upper regions of the core, undue stress on the vessel would result if the temperature difference were greater than 145*F.

The loop temperature must be within 50*F cf the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.

3/4.4.2.1 SAFETY / RELIEF VALVES l

The reactor coolant system safety valve function of the safety-relief valves operate to prevent the system from being pressurized above the Safety Limit of l' 25 psig. The system is designed to meet the requirements a

of the ASME Boiler and Pressure Vessel Code,Section III, for the pressure vessel, and ANSI B31.1, 1975 Code, for the reactor coolant system piping.

The capacity of the safety-relief valves is based on the full MSIV closure transient with failed trip scram, position switches, as described in Supplement 5.A of the FSAR, Section 5.A.6.

Demonstration of the safety-relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.

3/4.4.2.2 LOW-LOW SET SYSTEM The low-low set (LLS) system lowers the opening and closing setpoints on four preselected safety / relief valves (S/RVs). The LLS system lowers the setpoints after any S/RV has opened at its normal steam pilot setpoint when a concurrent high reactor vessel steam dome pressure scram signal is present. The purpose of the LLS is to mitigate the induced high frequency loads on the contatn-.

ment and thrust loads on the SRV discharge line. The LLS system increases the amount of reactor depressurization during an S/RV blowdown because the lowered LLS setpoints keep the four selected LLS S/RVs open for a longer time.

The high reactor vessel steam dome pressure signal for the LLS logic is provided by the exclusive analog trip channels. The purpose of installing special de-7 dicated steam dome pressure channels is to maintain separation from the RPS l'

high pressure scram functions.

HATCll-UNIT 2 3 1 4 4-1 Amendment No. 33

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REACTOR COOLANT SYSTEM BASES 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems are provided to monitor and detect leakage from the reactor coolant pressure boundary.

3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been

~

based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered.

The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for uniden-tified leakage the probability is small that the imperfection or crack associated with such leakage would grow rapidly.

However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE the reactor will be shutdown to allow further investigation and corrective action. Service sensitive reactor coolant system Type 304 and 316 austenitic stainless steel piping; i.e., those that are. subject to high stress or that certain relatively stagnant, intermittent, or low flow fluids, requires addi-tional surveillance and leakage limits.

3/4.4.4 CHEMISTRY The water chemistry limits of the reactor coolant system are estab-lished to prevent damage to the reactor materials in contact with the coolant.

Chloride limits are specified to prevent stress corrosion cracking of the stainless steel. The effect of chloride is not as great when the oxygen concentration in the coolant is low; thus the higher limit on chlorides is permitted during full power operation.

During shutdown and refueling operations the temperature necessary for stress corrosion to occur is not present.

Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits. With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.

The surveillance requirements provide adequate assurance that concen-trations in excess of the limits will be detected in sufficient time to take corrective action.

HATCH-UNIT 2 B 3/4 4-2 w.

5.0 CESIGN rEATUPES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.

LOW POPULATION ZONE 5.1.2 The icw population zone coincices with the exclusion area and is aisc shown in Figure 5.1.1-1.

5.2 CONTAIPNENT CONFIGURATION 5.2.1 The primary containment is a steel structure composed of a series of vertical right cylinders and truncated cones whicn form a drywell.

This drywell is attached to' a suppression chamber through a series of vents. The suppression chamber is a steel pressure vessel in the shape of a torus. The primary coltainment has a total minimum free air volume of 255,978 cubic feet.

DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:

a.

Maximum design internal pressure 56 psig.

b.

Maximum allowable internal pressure 62 psig.

c.

Maximum internal temperature 340er, c.

Maximun external pressure 2 psig.

5.3 REACTOR CORE FUEL ASSEMGLIES 5.3.1 The initial core shall contain 560 fuel assemblies with each fuel

. assembly containing 62 fuel rods and 2 water rods clad with Zircaloy

-2.

I Each fuel rod shall have a nominal active fuel length of 150 inches anc l

contain a maximum total weight of 3341 grams uranium.

The initial core l

l loading shall have a maximum average enrichment of 1.87 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loacing and shall have a maximum average enrichment of 2.90 weight percent U-235.

7X7 fuel containing 49 fuel rods and no water rods may also be i

i insertec.

I HATCH - UNIT 2 5-1 Amendment No. 27, 33 l

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.t J.. W o =, =.= W t: w 4 l'bl. ' N p .h. s ; ; s ~ s. ~ g.m 5.. .4 I / 2 in.r \\.. 4l. = a if .i .ji 3 l p,.1.. . hi .J. 1:.:: s .~+ - ~. _ _ ... e.1

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' ' n: y .c s. ig v7 i; I b b b k k w.) h=2h; =.k l'e=k k l EXCLUSION AREA AND LOW POPULATION ZONE FIGURE 5.1.1-1 HATCH - UNIT 2 5-2 ~1 -}}

v t e Letter Sequence Other
TAC:49989, (Open) TAC:51054, (Approved, Closed)
Initiation Request, Request, Request Acceptance... Administration Questions Answers Results Approval... Other: ML20023C229, ML20072T045, ML20072T049, ML20074A867, ML20076E075, ML20076E981, ML20076M410
MONTHYEAR1983-05-12 [Table View]

ML20076M410

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