Text

Proposed Tech Specs,Clarifying Applicability of Certain Surveillances Addressing Rvp & Temp Limits & Replacing Vessel Pressure & Temp Limit Curves W/New Curves
	ML20113F554
Person / Time
Site:	Hatch
Issue date:	09/19/1996
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20113F553	List: HL-5213, Application for Amends to Licenses DPR-57 & NPF-5,clarifying Applicability of Certain Surveillances Addressing Rvp & Temp Limits & Replacing Vessel Pressure & Temp Limit Curves W/New Curves; ML20113F554;
References
	NUDOCS 9609240351
	Download: ML20113F554 (87)
	v • d • e

- - .

?

Enclosure 3 1

Edwin I. Hatch Nuclear Plant Request to Revise Technical Specifications:

Pressure and Temperature Limits .,

Page Change Instructions ;

i Unit 1 a gg P.a. Instruction 3.4-22 Replace 3.4-23 Replace l 3.4-24 Replace 3.4-25 Replace :

3.4-26 Replace 3.4-27 Replace i Unit 2 ;

Eagg Instruction f 3.4-22 Replace I 3.4-23 Replace 3.4-24 Replace !

3.4-25 Replace j 3.4-26 Replace ]

3.4-27 Replace i l

l l

i l

9609240351 960919 PDR ADOCK 05000321 P PDR HL-5213 E3-1

RCS P/T Limits 3.4.9 ACTIONS (continued) t CONDITION REQUIRED ACTION COMPLETION TIME l

C. ---------NOTE--------- C.1 Initiate action to Immediately '

Required Action C.2 restore parameter (s) '

shall be completed if to within limits. l this condition is ;

entered. ANQ -

_ C.2 Determine RCS is Prior to Requirements of the acceptable for entering LCO not met in other operation. MODE 2 or 3 than MODES 1, 2, and 3.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 1

SR 3.4.9.1 Verify: 30 minutes )

I

a. RCS pressure and RCS temperature are

{

within the limits specified in !

Figures 3.4.9-1 and 3.4.9-2 during RCS I non-nuclear heatup and cooldown l operations, and RCS inservice leak and hydrostatic testing; and

b. RCS heatup and cooldown rates are s 100 F in any I hour period during RCS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing.

(continued)

HATCH UNIT 1 3.4-22 96 8/2/96

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued) l l

SURVEILLANCE FREQUENCY SR 3.4.9.2 --------------------NOTE-------------------

Only required to be met when the reactor is i critical and immediately prior to control !

rod withdrawal for the purpose of achieving l I

criticality.

Verify RCS pressure and RCS temperature are Once within 15 within the criticality limits specified in minutes prior Figure 3.4.9-3. to initial l control rod withdrawal for the purpose of achieving criticality SR 3.4.9.3 --------------------NOTE-------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the bottom Once within 15 head coolant temperature and the reactor minutes prior pressure vessel (RPV) coolant temperature to starting an is 5 145 F. idle recirculation pump SR 3.4.9.4 --------------------NOTE-------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the reactor Once within 15 coolant temperature in the recirculation minutes prior loop to be started and the RPV coolant to starting an temperature is s 50 F. idle recirculation pump (continued)

HATCH UNIT 1 3.4-23 96-23--8/2/96

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.5 -------

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

Only re uired to be met when tensioning /

detensi ning the reactor vessel head bolting studs.

Verify reactor vessel flange and head Once within flange temperatures are 176*F. 30 minutes prior to tensioning /

detensioning the reactor vessel head bolting studs and every 30 minutes thereafter SR 3.4.9.6 --------------------NOTE-------------------

Only required to be met when the reactor .

vessel head is tensioned.

i Verify reactor vessel flange and head Once within l flange temperatures are 176*F. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after !

RCS temperature ;

is s 106 F in i MODE 4, and I 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

thereafter i AND Once within 1 30 minutes after RCS temperature is s 86*F in MODE 4, and 30 minutes thereafter HATCH UNIT 1 3.4-24 96-23--8/2/96

.

l i

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- -BOTTOMMEAD l 106'F LIMITS

, = = UPPER VESSEL

, 1 LIMITS 200 , i i

, i BELTUNE AND '

soTTOM HEAD 6tT - ! k FLANGE REGION 76T '

O 0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

Figure 3.4.9-1 (page 1 Of 1)

Pressure / Temperature Limits for Inservice Hydrostatic and Inservice Leakage Tests HATCH UNIT 1 3.4-25 96 8/2/96

RCS P/T Limits 3.4.9 l

1600 !

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1400 .

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l FIANGE REGION 76*F ! !

l O l l li \ \ '

0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F) t Figure 3.4.9-2 (page 1 Of 1) l Pressure / Temperature Limits for Non-Nuclear Heatup, Low Power Physics Tests, and COOldown Following a Shutdown l

HATCH UNIT 1 3.4-26 96 8/2/96

RCS P/T Lisits 3.4.9 1600 l

tm 22 1400 '

I

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1 ! f 1000 3 o

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MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) '

Figure 3.4.9-3 (page 1 Of 1)

Pressure / Temperature Limits for Criticality l

HATCH UNIT 1 3.4-27 96 8/2/96 i

i i

RCS P/T Limits 3.4.9

. ACTIONS (continued) !

CONDITION REQUIRED ACTION COMPLETION TIME C. _--------NOTE--------- C.1 Initiate action to Immediately Required Action C.2 restore parameter (s) shall be completed if to within limits, this Condition is entered. ANQ

_ C.2 Determine RCS is Prior to Requirements of the acceptable for entering LCO not met in other operation. MODE 2 or 3 than MODES 1, 2, ;

and 3. -

I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

l SR 3.4.9.1 Verify: 30 minutes

a. RCS pressure and RCS temperature are within the limits specified in Figures 3.4.9-1 and 3.4.9-2 during RCS non-nuclear heatup and cooldown operations, and RCS inservice leak and hydrostatic testing; and

b. RCS heatup and cooldown rates are 1 100 F in any I hour period during RCS heatup and cooldown operations, and RCS inservice leak and hydrostatic testing.

l (continued)

HATCH UNIT 2 3.4-22 96 8/2/96 l

. _ _ _ _ _ _ . . _ . _ . _ .. _ _ _ .. .... _ _ _ .. _ . _ _ _. _ _ _ __m._ .

j RCS P/T Limits l 3.4.9 1

4 l

l SURVEILLANCE REQUIREMENTS (continued) I i

i SURVEILLANCE FREQUENCY )

l l

i SR 3.4.9.2 --------------------NOTE------------------- l 2 .Only required to be met when the reactor is l

, critical and immediately prior to control l

, rod withdrawal. for the purpose of achieving I j criticality.

a i

i Verify RCS pressure and RCS temperature are Once within 15-

[ within the criticality limits specified in minutes prior-

Figure 3.4.9-3. to initial l control rod i

withdrawal for-the purpose of 1 achieving criticality j l

SR 3.4.9.3 --------------------NOTE-------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Verify the difference between the bottom Once within 15 head coolant temperature and the reactor minutes prior pressure vessel (RPV) coolant temperature to starting an

.is s 145'F. idle recirculation pump 1

i SR 3.4.9.4 --------------------NOTE------------------- i Only required to be met in MODES 1, 2, 3, i and 4 during startup of a recirculation pump.

Verify the difference between the reactor Once within 15 coolant temperature in the recirculation minutes prior loop to be started and the RPV coolant to starting an temperature is s 50*F. idle recirculation pump (continued)

. HATCH UNIT-2 3.4-23 96 8/2/96

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.5 --------------------NOTE-------------------

Only required to be met when tensioning /

detensioning the reactor vessel head bolting studs.

Verify reactor vessel flange and head Once within flange temperatures are 190 F. 30 minutes prior to tensioning /

detensioning the reactor vessel head )

bolting studs and every 30 minutes thereafter SR 3.4.9.6 --------------------NOTE-------------------

Only required to be met when the reactor vessel head is tensioned.

Verify reactor vessel flange and head Once within flange temperatures are 190 F. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature is s 120 F in MODE 4, and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 1 thereafter l

AND Once within i 30 minutes l after RCS temperature is s 100 F in MODE 4, and l 30 minutes !

thereafter HATCH UNIT 2 3.4-24 96-23--8/2/96 i

RCS P/T Limits 3.4.9 1600 1400 ,

I I .l sji

- 1200 ,II I -

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5 z I I g 10m

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BELTUNE AND FLANGE REGION 90*F i g BOTTOM HEAD 68'F ,

e g l l, i :

0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

Figure 3.4.9-1 (page 1 Of 1)

Pressure / Temperature Limits for Inservice Hydrostatic and Inservice Leakage Tests HATCH UNIT 2 3.4-25 96 8/2/96

RCS P/T Lisits.

3.4.9 1600 1400 - ,

4 1200 '

g f 6 i

g 1000 I O

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BELTUNE AND MOM HEAD H*F 'l

{ j l l l l FLANGE REO!ON 90*F

' l l l 0 '

O.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

Figure 3.4.9-2 (page 1 Of 1)

Pressure / Temperature Limits for Non-Nuclear Heatup, low P0wer Physics Tests, and C00ldown F0110 wing a Shutdown HATCH UNIT 2 3.4-26 96 8/2/96

0 RCS P/T Limits 3.4.9 1600 1400 1200 l

I 5

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$ 800 u l 5

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AND N N-BELTLINE LIMITS ;

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/ I i l MINIMUM CRmCAUTY I l j

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i TEMPERATURE 90*F

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0 0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

Figure 3.4.9-3 (page 1 Of 1)

Pressure / Temperature Limits for Criticality HATCH UNIT 2 3.4-27 96-23 8/2/96

, o Attachment I to Enclosure 3 Technical Specifications Unit I and Unit 2 Marked-Up Pages l

RCS P/T Limits 3.4.9 ACTIONS (continued) [

CONDITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE--------- C.1 Initiate action to Immediately )

Required Action C 2 . restore parameter (s) shall be completed if to within limits.

this Condition is entered. AND C.2 Determine RCS is Prior to Requirements of the acceptable for entering LCO not met in other operation. MODE 2 or 3 than MODES 1, 2, i and 3. I 1

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

. (

SR 3.4.9.1 ---- ---------

NOTE- --------- ----- 1 0 equired be er ormed du ng RCS hea and co down o rations nd RC in er ce le and by r tati testin Verify: 30 minutes

a. RCS pressure and RCS temperature are gW g6 within the limits specified f Figures 3.4.9-1 and 3.4.9-2,A and

-[ 10n %tcject(

nh0 Ad

b. RCS heatup and cooldown rates are \ dO

~< 100 F in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months p A eriod NgA@uce

&uyjpsen %

C O m Res heu k u M

(' (c "t' = d)

Goc \ dada Ope 6fi a, sw u- J> r /<- ~E t /udfo$kchC

/L t 5 i 4 7 Aosr, Jus t- ths. 3 HATCH UNIT 1 3.4-22 E nd ent Nu. M

--@E _ _ __

Ont3 requimd 40 be met A l RCS P/T Limits 3.4.9 Qk ggg b gg

- - ~ ____ _ _

SURVEILLANCE QUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.2 4Verify RCS pressure and RCS temperature are1 ; Once within within the criticality limits specified in j 15 minutes Figure 3.4.9-3. / prior tointM

control rod withdrawal for j CLDd kYhmedicdal pricNo convot rod tagg$aof i

O k b fd(d(d [0 ( % p' ipO b o f criticality

, 6 diteAlinC CTMirrd tlti.

a J SR 3.4.9.3 --------------------NOTE-------------------

Only required to be met in MODES 1, 2, 3,

and 4 during startup of a recirculation

pump.

1 Verify the difference between the bottom Once x A n-

! 15 minutes j head coolant temperature and the reactor DCiOC 9 0 6 k N

esr ssel (RPV) coolant temperature A g yg g'gg

kDh

SR 3.4.9.4 -------------------NOTE--------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation

pump.

l l-Verify the difference between the reactor Dnteei4 15 minutes 1

I coolant temperature in the recirculation pW $D $kLdm GA i loop to be started and the RPV coolant g d k g g f L'

, temperature is s 50*F.

i

(continued) .

1 I

4 i

HATCH UNIT 1 3.4-23 -

.. sd-niWo.In i

j

a 1 RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued) b l SURVEILLANCE FREQUENCY OW o u a ; Hi *s SR 3.4.9.5 -------------------NOTE--- -----------------

Onlyrequiredtobefv$dhdwhen l 3 e i 's b P ".

l tensioningjghe reactor vessel head bolting studs.

4, ; m ; ,, M Q/],

-------------------------------- W ' ' D b A + i n$

/ddewoni@

Verify reactor vessel flange afid head st.c. .o .-y 30 minutes flange temperatures are 176*F. #4,. qq.

SR 3.4.9.6 ------

NOT ------- ----------

t re uired to be p fo ed u il n

ghg 30 i utes after RC temp at e :s; 86*F r

! I MOD 4. (dtthL 1

Q --- - ---------- ---------- -------- ----

MI V rify acto vessel fl ge an hea minute g ggg ange te atures are _. 76*F. -

. (

,] } _/ / / \ l y p .p .p ./ - -------

NOTE------- -- -----'---- f No re ired to be erfo ed ntil 12 hou s aft CS tempera re s 1 in MODE 4.

V ify re to vessel ange an e hours f ange temp atures a e 2 76*F.

i l

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! HATCH UNIT 1- 3.4-24 'oK I

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t

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4 3 D"E LT i , y 3.4 e 4 I

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- - NOTE - - - - - - -

g unn acquired to be met when the reactor vessel head is tensioned ,I i Verify reactor vessel flange and head flange temperatures are 2 76*F.

i I

I M

Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature is 5106'F in MODE 4, hours thereafter. l l

l AND l

l- ' Once within 30 minutes after RCS temperature is s 86 F in MODE 4, and 30 minutes thereafter.

l

--]- g g g 3. + c7 S 1

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j VERTICAL LIMI INE FOR PRESSURE

} ASOVE 20% HYO TEST (312 W,

- 400 #

BASED ON tocFRs0 NOlX G i REQUIREMENT OF IR DT + 88*EI' j FLANGE REGION RTND

1W i SOLT PRELOAD TEMPERATU OF 78'F BASED ON RECORWENDE 200 / , tRTNoy + 000F) FOR 0.24-IN. FLA IN CLOSURE FLANGE REGION, RTNOT

18'F 0

100 200 300 RPV METAL TEMPERATURE (OF)

}bCL (L) b Figure 3.4.9-1 (page 1 of 1) gig Q Pressure / Temperature Limits for Inservice Hydrostatic and Inservice Leakage Tests 6t'"'

HATCH UNIT 1 3.4-25 a rc + =at Nc. 195

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RCS P/T Limits l 3.4.9 bpkcx ewurd Egolo3,4A-ICI"hcl F9 l

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Pressure / Temperature Limits for Inservice Hydrostatic and inservice Leakage Tests HATCH UNIT 1 3.4-25 96 8/2/96 i

1 j RCS P/T Limits l j 3.4.9 l

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Figure 3.4.9-2 (page 1 of 1)

Pressure / Temperature Limits for Non-Nuclear Heatup, Low Power Physics Tests, and Cooldown Following a Shutdown HATCH UNIT 1 3.4-26 ^-^ d int Nv. 195

f l

I RCS P/T Limits l 3.4.9 k b bM4s 4 3, ~~h bi % JLpd4 l

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BELTLINE AND / 9 g -

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0 l l l l !

0.0 100.0 200.0 300.0 400.0 500.0 600.0 i

MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) l Figure 3.4.9-2 (page 1 Of 1)

Pressure / Temperature Limits for NOn-Nuclear Heatup, Low Power Physics Tests, and C00ldown Following a Shutdown HATCH UNIT 1 3.4-26 96 8/2/96

i . o RCS P/T Limits 3.4.9 I

\

1600 l VALlO TO 16 EFFECTIVE FU L POWER YEARS OF OPERAT N 1400 I E 1200 3 l 6

I O \

>= I

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I

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/ \

NWb FEEoWATER NOZZLE TEMeERAruR M -

LIMIT FOR 1/4 T FLAW (8WR/6 RESULTS ADJUSTED TO 40'F RTNDTI 200 j

MINIMUM OPER ATING TEMPERATURE LIMIT OF 76*F FROM 10CFR50 APPENDlX G REQUIREMENT THAT (TMIN = RTNOT + 60*F),

FLANGE RTNOT = 16*F 0 g 100 200 300 400 500 600 MINIMUM VESSEL METAL TEMPERATURE (OF) 4, t

Figure 3.4.9-3 (page 1 of 1) l Pressure / Temperature Limits for Criticality l.

l HATCH UNIT 1 3.4-27 Amend =nt No.195 -

1 i

RCS P/T Lirsits !

3.4.9 )

l

% ' ig M

+

! L l

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? E g 1000

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LIMITS -

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l Tw 76Y I l i

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l l l .

l 0.0 100.0 200.0 300.0 400.0 500.0 600.0 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F) figure 3.4.9-3 (page 1- Of 1)

Pressure / Temperature Limits for Criticality HATCH UNIT 1 3.4-27 96 8/2/96

O 4 RCS P/T Limits 3.4.9 ACTIONS (continued) h CONDITION REQUIF.ED ACTION COMPLETION TIME j C. ---------NOTE--------- C.1 Initiate action to Immediately Required Action C.2 restore parameter (s) shall be completed if to within limits.

this Condition is entered. AND C.2 Determine RCS is Prior to Requirements of the acceptable for entering LCO not met in other operation. MODE 2 or 3 than MODES 1, 2, and 3.

l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 / ~

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

ly equire to b per ormed du ng C he up and oldown rations ad RC i vice eak and ostat testin .

Verify: 30 minutes l a. RCS pressure and RCS temperature are dLL 6 6106.- l

, within the limits specified ,M g l l Figures 3.4.9-1 and 3.4.9 , and ,

CUM (, OldOU3A._

b. RCS heatup and cooldown rates are 0?erceriOrts C'MT s 100*F in any I hour peri j g pg,g g i D -

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, SURVEILLANCE QUIREMENTS (continued)

[ SURVEILLANCE I FREQUENCY SR 3.4.9.2 l Verify RCS pressure and RCS temperature are Once within l . within the criticality limits specified in 15 minutes

! Figure 3.4.9-3. prior toi d y control rod l ., ,.e,.Agq withdrawal for CUu (OCkO CDb MD\ (Cd W\ $ bY W S C $iUf' C""iC'

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l SR 3.4.9.3 --------------------NOTE-------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation j pump.

Verify the difference between the bottom head coolant temperature cnd the reactor kb D esr essel (RPV) coolant temperature NOFO 6tRhuc Cdv (d(c l

ydCi(CR{ddiOq'

, %mf l

SR 3.4.9.4 -------------------NOTE--------------------

Only required to be met in MODES 1, 2, 3, and 4 during startup of a recirculation pump.

Ver f the d ffere ce between he reactor coolant temperature in the recirculation dn -

, loop to be started and the RPV coolant y(M O h d h %

(dbf6&dkhbdh l temperature is s 50*F.

filk (continued) i 6

i HATCH UNIT 2 3.4-23 J,x ndaant 5 A 35-

. . )

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued) h SURVEILLANCE FREQUENCY l Yhet o- u N 4, SR 3.4.9.5 ------------------ -

A

NOTE-fpdwhen Only required to be $,f #o ' " g #,k b ino tensioning the reactor vessel head bolting f* 5,c.+/- A- J studs. . . < f o, va

_________ _____________________________ , w;,, ,n

_ [Cld6N5/O/)i(1G ~s 4 ,, y Verify reactor vessel flange 1fn'd head flange temperatures are 1 90*F.

30 minutes f4 Q A t j

SR 3.4.9.6 ------ ---------- /--------- ------ -

tr uired t be NOT[ er formed un gnb yd l l !

30 utes af er RCS temperatu s 0 in y

, QM M MO .

dgyd Md. ll Ver y rea or vessel fiang and i b[Nf 6[- flange tempe tures are 2 90' .

ad y

3 es s

b\'% --------

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

dt be perfo d until 12 ou af R temper ture s 120 in MO 4.

Ver fy r ctor vessel ange d d 12 urs f ange tem atures are _ F. (

r i

i 1

HATCH UNIT 2 3.4-24 fr aient S. 135

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

i' I

, Only required to be met when the reactor vessel head is tensioned.

I Verify reactor vessel flange and head flange temperatures are > 9CfF.

4 s

cLnb i Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after RCS temperature is s loo *F in MODE 4, hours thereafter.

AND Once within 30 minutes after RCS temperature is spfF in MODE 4, and 30 minutes thereafter.

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Pressure / Temperature Limits for Inservice Hydrostatic and Inservice Leakage Tests HATCH UNIT 2 3.4-25 Amendment No. 135

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Pressure / Temperature Limits for Inservice Hydrostatic and Inservice Leakage Tests l

HATCH UNIT 2 3.4-25 96 8/2/96

a =

RCS P/T Limits 3.4.9 b

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Pressure / Temperature Limits for Non-Nuclear Heatup, 4 Low Power Physics Tests, and Cooldown Following a Shutdown ,

(

HATCH UNIT 2 3.4-26 frendment No. 13 6

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Figure 3.4.9-2 (page 1 Of 1)

Pressure / Temperature Limits for Non-Nuclear Heatup, low Power Physics Tests, and C00ldown Following a Shutdown HATCH UNIT 2 3.4-26 96 8/2/96 j

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Pressure /Temptrature Limits for Criticality 4

HATCH UNIT 2 3.4-27 %cn &^nt Nc. 135-

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Figure 3.4.9-3 (page 1 Of 1)

Pressure / Temperature Limits for Criticality HATCH UNIT 2 3.4-27 96-23 8/2/96

, .a Attachment 2 to Enclosure 3 FOR INFORMATION Technical Specifications Bases Unit 1 and Unit 2 Revised Associated Bases Pages and Marked-Up Pages i

i i

i

._ _ J

, , 1 i

4 i .RCS P/T Limits B 3.4.9

!. B 3.4 REACTOR COOLANT SYSTEM (RCS) i j B 3.4.9 'RCS Pressure and Temperature (P/T) Limits BASES i

j BACKGROUND All components of the RCS are designed to withstand effects

of cyclic loads due to system pressure and temperature

! changes.. These loads are introduced by startup (heatup) and

shutdown (cooldown) operations, power transients, and

! reactor trips. This LCO limits the pressure and temperature 1

changes during RCS heatup and cooldown, within the design j assumptions and the stress limits for cyclic operation. ;

! This Specification contains P/T limit curves for non-nuclear - l l heatup and cooldown, and inservice leakage and hydrostatic ;

j testing, and also limits the maximum rate of change of. ]

reactor coolant temperature. The criticality curve provides 1 limits for. both nuclear heatup and criticality.

l_

i

, Each P/T limit curve defines an acceptable region of

! operation for a particular. operating condition. The usual l use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable

curve to determine that operation is within the allowable i region.

i

The LCO establishes operating limits that provide a margin l to brittle failure of the reactor vessel and piping of the ;

i reactor coolant pressure boundary (RCPB). The vessel is the !

j component most subject to brittle failure. Therefore, the l

} LC0 limits apply mainly to the vessel. '

10 CFR 50, Appendix G (Ref.1), requires the establishment

of P/T limits for material fracture toughness requirements

of the RCPB materials. Reference I requires an adequate j margin to brittle failure during normal operation, j anticipated operational occurrences, and system hydrostatic

. tests. It mandates the use of the ASME Code, Section III, j Appendix.G (Ref. 2). !

l The actual shift in the RT, of the vessel material is l !

. established periodically by removing and evaluating the i irradiated reactor vessel material specimens, in accordance l l with ASTM E 185 (Ref. 3) and Arpendix H of 10 CFR 50

(Ref. 4). The operating P/T limit curves are adjusted, l I

t (continued) i HATCH UNIT I B 3.4-46 96 8/2/96 u_ - - , , ,

RCS P/T Limits B 3.4.9 BASES BACKGROUND as necessary, based on the evaluation findings and the (continued) recommendations of Reference 5.

The P/T limit curve for inservice leak and hydrostatic testing, and the curve for non-nuclear heatup and cooldown include separate curves for the bottom head, beltline, and upper vessel and flange regions. These curves are derived from stress analysis of these vessel regions.

The criticality limits include the Reference 1 requirement that they be at least 40 F above the heatup curve or the 4 cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.

4 The consequence of violating the LC0 limits is that the RCS

! has been operated under conditions that can result in

. brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event

, these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. ASME Code, Section XI, Appendix E

- (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

APPLICABLE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (DBA) analyses. They are prescribed during normal operation 1 to avoid encountering pressure, temperature, and temperature rate of change conditions th'at might cause undetected flaws

, .to propagate and cause nonductile failure of the RCDB, a condition that is unanalyzed. References 8 and 12 approved l the curves and limits specified in this section. Since the P/T limits are not derived from any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.

(continued)

HATCH UNIT 1 B 3.4-47 96 8/2/96

, , 1 l

RCS P/T Limits B 3.4.9 BASES 4

APPLICABLE RCS P/T limits satisfy Criterion 2 of the NRC Policy SAFETY ANALYSES Statement (Ref. 9).

, (continued)

LCO The elements of this LCO are:

_ a. RCS pressure and temperature are within the limits '

' specified in Figures 3.4.9-1 and 3.4.9-2 during RCS non-nuclear heatup and cooldown operations, and RCS inservice leak and hydrostatic testing. Additionally, heatup and cooldown rates are s 100 F during any RCS heatup or cooldown, and inservice leak and hydrostatic testing;

b. The temperature difference between the reactor vessel i bottom head coolant and the reactor pressure vessel (RPV) coolant is s 145*F during recirculation pump )

i startup;

)

c. The temperature difference between the reactor coolant !

in the respective recirculation loop and in the reactor vessel is s 50 F during recirculation pump startup;

\

d. RCS pressure and temperature are within the criticality limits specified in Figure 3.4.9-3, prior to achieving criticality; and

e. The reactor vessel flange and the head flange 4

temperatures are 1 76 F when tensioning or detensioning the reactor vessel head bolting studs.

f. The reactor vessel flange and head flange temperatures are 2 76*F when the reactor vessel head is tensioned.

i

g. For the case when the vessel head is either off or on but not tensioned and fuel is in the vessel, all three !

sections of the vessel (upper vessel, beltline, and ;

bottom head) may be lowered to a minimum of 68 F. )

When the head is being tensioned, or is already tensioned, the beltline and bottom head regions may be lowered to 68 F, as long as there is not any pressure or heatup/cooldown. The upper vessel, however, has a (continued)

HATCH UNIT 1 B 3.4-48 96 8/2/96

RCS P/T Limits B 3.4.9 I

BASES

-LC0 g. (continued) higher minimum temperature requirement with the head tensioned, as previously delineated.

The 68 F temperature is based on fuel shutdown margin considerations, since this is a more restrictive temperature than would be obtained from 10 CFR 50, Appendix G, considerations. With no fuel in the vessel, the temperature may drop to as low as 40 F, because this is the highest qualification temperature to meet toughness requirements for all reactor materials.

These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide ,

margin to nonductile failure.

l The rate of change of temperature limits controls the thermal gradient through the vessel wall and is used as ;

input for calculating the heatup, cooldown, and inservice 1 leakage and hydrostatic testing P/T limit curves. Thus, the i LC0 for the rate of change of temperature restricts stresses !

caused by thermal gradients and also ensures the validity of the P/T limit curves.

Violation of the limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCS components. The consequences depend on several factors, as follows:

a. The severity of the departure from the allowable operating pressure temperature regime or the severity of the rate of change of temperature;

b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick i vessel walls to become more pronounced); and l

c. The existences, sizes, and orientations of flaws in the vessel material.

(continued)

HATCH UNIT 1 B 3.4-49 96 8/2/96

RCS P/T Limits B 3.4.9 BASES (continued)

APPLICABILITY The potential for violating a P/T limit exists at all times.

For example, P/T limit violations could result from ambient temperature conditions that result in the reactor vessel metal temperature being less than the minimum allowed temperature for boltup. Therefore, this LC0 is applicable even when fuel is not loaded in the core.

~

ACTIONS A.1 and A.2 Operation outside the P/T limits while in MODES 1, 2, and 3 must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.

The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.

Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify the RCPB integrity remains acceptable ,

and must be completed if continued operation is desired. !

Several methods may be used, including comparison with '

pre-analyzed transients in the stress analyses, new ;

analyses, or inspection of the components. 1 ASME Code, Section XI, Appendix E (Ref. 6), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable to accomplish the evaluation of a mild violation. More severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed if continued operation is desired.

Condition A is modified by a Note requiring Required i Action A.2 be completed whenever the Condition is entered.

The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.

Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

(continued)

HATCH UNIT 1 B 3.4-50 96-23--8/2/96 l

RCS P/T Limits B 3.4.9 BASES ACTIONS 8.1 and B.2 (continued)

If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress, or a l sufficiently severe event caused entry into an unacceptable region. Either possibility indicates a need for more l careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. With the reduced ;

pressure and temperature conditions, the possibility of l propagation of undetected flaws is decreased.

j Pressure and temperature are reduced by placing the plant in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

i C.1 and C.2 l Operation outside the P/T limits in other than MODES 1, 2, I and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are 4

restored.

Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB i 4

integrity is acceptable and must be completed before approaching criticality or heating up to > 212 F. Several methods may be used, including comparison with pre-analyzed <

transients, new analyses, or inspection of the components.

ASME Code, Section XI, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline. !

J Condition C is modified by a Note requiring Required Action C.2 be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the

. effects of the excursion outside the allowable limits.

(continued)

HATCH UNIT 1 B 3.4-51 96 8/2/96 I

1 RCS P/T Limits l B 3.4.9 i

}

BASES i

ACTIONS C.1 and C.2 (continued)

Restoration alone per Required Action C.1 is insufficient because higher than. analyzed stresses may have occurred and may have affected the RCPB integrity.

SURVEILLANCE SR 3.4.9.1 ,

REQUIREMENTS Verification that operation is within limits.is required '

every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room >

indication available to monitor RCS status. Also, since temperature change limits are specified in hourly l increments, 30 minutes permits a reasonable time for ,

assessment and correction of minor deviations. !

Surveillance for heatup, cooldown, or inservice leakage and .

hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the -

activity are satisfied.

Verification of Figures 3.4.9-1 and 3.4.9-2 is required during non-nuclear heatups and cooldowns, and inservice leak i and hydrostatic testing. Verification of the s 100'F change !

in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period is required during any heatup or ,

cooldown.

SR 3.4.9.2 A separate figure is used when the reactor is critical. l {

Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.

Performing the Surveillance within 15 minutes prior to initial control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time criticality is achieved.

This SR, for clarity, is modified by a Note stating that it is only required to be met when the reactor is critical and immediately prior to control rod .:ithdrawal for the purpose of achieving criticality.

(continued)

HATCH UNIT 1 B 3.4-52 96 8/2/96 l i

RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.3 and SR 3.4.9.4 REQUIREMENTS (continued) Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. In addition, compliance with these limits

- ~ ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 7) are satisfied.

Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.

4 If the 145 F temperature differential specified in SR 3.4.9.3 cannot be determined by direct indication, an alternate method may be used as described below:

The bottom head coolant temperature and the RPV coolant can be assumed to be s;145 F if the following can be confirmed:

a

a. F or more loop drive flows were > 40 percent of i-.ted flow prior to the RPT,

b. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems have not injected since the RPT,

c. Feedwater temperature has remained > 300 F since the 1 RPT, and '

1

d. The time between the RPT and restart is < 30 minutes. l General Electric test data from BWR plants shows that stratification up to the 145 F differential does not occur

! any sooner than I hour following the RPT (Refs. 10 and 11).

Adding HPCI and RCIC injection, and feedwater temperature constraints provides assurance that the temperature differential will not be exceeded within 30 minutes of the l RPT.

i An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.4 is to compare the temperatures of the operating recirculation loop and the idle loop.

(continued)

HATCH UNIT 1 B 3.4-53 96-23--8/2/96 l

4 0 RCS P/T Limits B 3.4.9 i

BASES I

SURVEILLANCE SR 3.4.9.3 and SR 3.4.9.4 (continued)

REQUIREMENTS SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that requires the Surveillance to be performed only in MODES 1, i 2, 3, and 4. In MODE 5, the overall stress on limiting components is lower. Therefore, AT limits are not required.

. SR 3.4.9.5 :nd SR 3.4.9.6 l Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations

' approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values !

require assurance that these temperatures meet the-LCO limits.

The flange temperatures must be verified to be above the limits 30 minutes before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. Verification of flange i

' temperatures is also required while detensioning is in progress until all reactor vessel head bolts are completely

, detensioned. (The head is considered tensioned if one or more bolts are partly or completely tensioned.) When in MODE 4 with RCS temperature s 86*F, 30 minute checks of the <

flange temperatures are required because of the reduced l margin to the limits. When in MODE 4 with RCS temperature '

s 106 F, monitoring of the flange temperature is required every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to ensure the temperature is within the i limits specified.

1 The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable bated on the rate of temperature change possible at these temperatures.

SR 3.4.9.5 is modified by a Note that requires the Surveillance to be met only when tensioning /detensioning the l reactor vessel head bolting studs SR 3.4.9.6 is modified by a Note that requires the Surveillance to be met when the

, head is tensioned.

(continued)

HATCH UNIT 1 B 3.4-54 96 8/2/96

e RCS P/T Limits B 3.4.9 BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix G, January 1996. l

2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.

3. ASTM E 185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," July 1982.

4. 10 CFR 50, Appendix H.

5. Regulatory Guide 1.99, Revision 2, May 1988.

6. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E.

7. FSAR, Section 14.3.6.2.

. 8. George W. Rivenbark (NRC) letter to J. T. Beckham, Jr.

(GPC), Amendment 126 to the Operating License, dated June 20, 1986.

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

10. GE-NE-668-13-0393, " Recirculation Pump Restart Without Vessel Temperature Indication for E.I. Hatch Nuclear Plant," December 28, 1993.

11. DRF A00-05834/6, " Safety & 10 CFR 50.92 Significant Hazards Consideration Assessment for RPV ;

Stratification Prevention Improvements at Edwin I. ,

Hatch Nuclear Plant Units 1 and 2," April 1994. '

12. (To be added when amendment is received.) l l

l HATCH UNIT 1 B 3.4-55 96 8/2/96

l l

l Reactor Steam Dome Pressure B 3.4.10

! B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.10 Reactor Steam Dome Pressure l

BASES l

BACKGROUND The reactor steam dome pressure is an assumed value in the determination of compliance with reactor pressure vessel l overpressure protection criteria and is also an assumed l initial condition of design basis accidents and transients.

l l APPLICABLE The reactor steam dome pressure of s 1058 psig is an l SAFETY ANALYSES initial condition of the vessel overpressure protection l

analysis of Reference 1. This analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system, primarily the l safety / relief valves, during the limiting pressurization l transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor i steam dome pressure; therefore, the limit on this pressure l- ensures that the assumptions of the overpressure protection j analysis are conserved. Reference 2 also assumes an initial i reactor steam dome pressure for the analysis of design basis accidents and transients used to determine the limits for fuel cladding integrity (see Bases for LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain (see Bases for LC0 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)").

l l Reactor steam dome pressure satisfies the requirements of Criterion 2 of the NRC Policy Statement (Ref. 3).

l l

l l LC0 The specified reactor steam dome pressure limit of l s 1058 psig ensures the plant is operated within the j assumptions of the overpressure protection analysis.

Operation above the limit may result in a response more severe than analyzed.

l APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these (continued) l HATCH UNIT 1 B 3.4-56 96 8/2/96 l i

l 7eactor Steam Dome Pressure B 3.4.10

! BASES J

APPLICABILITY MODES, the reactor may be generating significant steam and (continued) events which may challenge the overpressure limits are i possible. l l

In MODES 3, 4, and 5, the limit is not applicable because I the reactor is shut down. In these MODES, the reactor l pressure is well below the required limit, and no anticipated events will challenge the overpressure limits.

l ACTIONS M l With the reactor steam dome pressure greater than the limit, prompt action should be taken to reduce pressure to below the limit and return the reactor to operation within the bounds of the ar,alyses. The 15 minute Completion Time is ,

reasonable considering the importance of maintaining the l pressure within limits. This Completion Time also ensures that the probability of an accident occurring while pressure !

is greater than the limit is minimized. 1 M

! If the reactor steam dome pressure cannot be restored to I within the limit within the associated Completion Time, the i plant must be brought to a MODE in which the LC0 does not l apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

l l SURVEILLANCE SR 3.4.10.1 l REQUIREMENTS l Verification that reactor steam dome pressure is s 1058 psig ensures that the initial conditions of the vessel overpressure protection analysis is met. Operating experience has shown the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency to be sufficient for identifying trends and verifying eperation within safety analyses assumptions.

[ (continued)

HATCH UNIT 1 B 3.4-57 96 8/2/96 l l

i i

l Reactor Steam Dome Pressure B 3.4.10 BASES (continued)

REFERENCES 1. FSAR, Appendix M.

2. FSAR, Section 14.3.

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

l

! l l

l l

i I

l .

4 b

HATCH UNIT 1 B 3.4-58 96 8/2/96 l

RCS P/T Limits B 3.4.9 8 3.4 REACTOR COOLANT SYSTEM (RCS) i B 3.4.9 RCS Pressure and Temperature (P/T) Limits :

I l

i BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

l This Specification contains P/T limit curves for non-nuclear j heatup and cooldown, and inservice leakage and hydrostatic testing, and also limits the maximum rate of change of reactor coolant temperature. The criticality curve provides limits for both nuclear heatup and criticality. l Each P/T limit curve defines an acceptable region of operation for a particular operating condition. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure. Therefore, the l

LC0 limits apply mainly to the vessel.

I 10 CFR 50, Appendix G (Ref. 1), requires the establishment j of P/T limits for material fracture toughness requirements '

of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code, Section III, !

Appendix G (Ref. 2).

The actual shift in the RT, of the vessel material is l established periodically by removing and evaluating the i-irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Appendix H of 10 CFR 50 ,

(Ref. 4). The operating P/T limit curves are adjusted, l l

(continued) 1 HATCH UNIT 2 B 3.4-46 96 8/2/96 I

1 RCS P/T Limits l B 3.4.9 i

i l BASES

BACKGROUND as necessary, based on the evaluation findings and the

! (continued) recommendations of Reference 5.

l The P/T limit curve for inservice leak and hydrostatic testing, and the curve for non-nuclear heatup and cooldown l include separate curves for the bottom head, beltline, and upper vessel and flange regions. These curves are derived i from stress analysis of these vessel regions.

l

' l l

l The criticality limits include the Reference I requirement that they be at least 40 F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.

The consequence of violating the LC0 limits is that the RCS l has been operated under conditions that can result in l brittle failure of the RCPB, possibly leading to a l nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed l to determine the effect on the structural integrity of the RCPB components. ASME Code, Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

APPLICABLE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (DBA) analyses. They are prescribed du 'ng normal operation to avoid encountering pressure, tempera: m a, and temperature !

rate of change conditions th'at might cause undetected flaws l to propagate and cause nonductile failure of the RCPB, a condition that is vaanalyzed. References 8 and 12 approved l the curves and limits specified in this section. Since the P/T limits are not derived from any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits ace acceptance limits themselves since they preclude operation in an unanalyzed condition.

l (continued) l l HATCH UNIT 2 B 3.4-47 96 8/2/96 l

i RCS P/T Limits

! B 3.4.9 i

i BASES APPLICABLE RCS P/T limits satisfy Criterion 2 of the NRC Policy l SAFETY ANALYSES Statement (Ref. 8).

. (continued) '

l LCO The elements of this LCO are:

a. RCS pressure and temperature are within the limits i specified in Figures 3.4.9-1 and 3.4.9-2 during RCS l non-nuclear heatup and cooldown operations, and RCS i inservice leak and hydrostatic testing. Additionally, heatup and cooldown rates are 5 100*F during any RCS l heatup or cooldown, and inservice leak and hydrostatic i

testing;

b. The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel

! (RPV) coolant is s 145"F during recirculation pump startup; i

c. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is s 50*F during recirculation pump startup; l d. RCS pressure and temperature are within the ;

l criticality limits specified in Figure 3.4.9-3, prior !

! to achieving criticality; and i

e. The reactor vessel flange and the head flange l temperatures are 1 90 F when tensioning or detensioning the reactor vessel head bolting studs.

f. The reactor vessel flange and head flange temperatures are 1 90*F when the r.eactor vessel head is tensioned.

! . g. For the case when the vessel head is either off or on but not tensioned and fuel is in the vessel, all three sections of the vessel (upper vessel, beltline, and bottom head) may be lowered to a minimum of 68 F.

When the head is being tensioned, or is already tensioned, the beltline and bottom head regions may be lowered to 68*F, as long as there is not any pressure

, or heatup/cooldown. The upper vessel, however, has a i

(continued)

HATCH UNIT 2 B 3.4-48 96 8/2/96

s o l

l l RCS P/T Limits B 3.4.9 r BASES LC0 g. (continued) higher minimum temperature requirement with the head tensioned, as previously delineated.

The 68 F temperature is based on fuel shutdown margin considerations, since this is a more restrictive temperature than would be obtained from 10 CFR 50, Appendix G, considerations. With no fuel in the ,

vessel, the temperature may drop to as low as 50 F, because this is the highest qualification temperature to meet toughness requirements for all reactor materials.

These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.

The rate of change of temperature limits controls the '

thermal gradient through the vessel wall and is used as input for calculating the heatup, cooldown, and inservice leakage and hydrostatic testing P/T limit curves. Thus, the LC0 for the rate of change of temperature restricts stresses ,

caused by thermal gradients and also ensures the validity of the P/T limit curves. -

Violation of the limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCS components. The consequences depend on several factors, as follows-I

a. The severity of the departure from the allowable operating pressure temperature regime or the severity of the rate of change of temperature;

b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounce!); and

c. The existences, sizes, and orientations of flaws in the vessel material.

l (continued)

HATCH UNIT 2 B 3.4-49 96 8/2/96 l

f 1 RCS P/T Limits l B 3.4.9 BASES (continued) l APPLICABILITY The potential for violating a P/T limit exists at all times.

For example, P/T limit violations could result from ambient l temperature conditions that result in the reactor vessel i metal temperature being less than the minimum allowed I

temperature for boltup. Therefore, this LC0 is applicable even when fuel is not loaded in the core.

~

ACTIONS A.1 and A.2 i

i Operation outside the P/T limits while in MODES 1, 2, and 3 must be corrected so that the RCPB is returned to a

condition that has been verified by stress analyses. 3 1

The 30 minute Completion Time reflects the urgency of !

restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be l accomplished in this time in a controlled manner.

Besides restoring operation within limits, an evaluation is i required to determine if RCS operation can continue. The )

l evaluation must verify the RCPB integrity remains acceptable l l and must be completed if continued operation is desired.

l Several methods may be used, including comparison with l pre-analyzed transients in the stress analyses, new l analyses, or inspection of the components.

l ASME Code, Section XI, Appendix E (Ref. 6), may be used to '

support the evaluation. However, its use is restricted to evaluation of the vessel beltline.

i The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable to accomplish the evaluation of a mild violation. Moie severe violations may l

require special, event specific strass analyses or l inspections. A favorable evaluation must be completed if continued operation is desired.

Condition A is modified by a Note requiring Required Action A.2 be completed whenever the Condition is entered.

The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.

l Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

I (continued)

HATCH UNIT 2 B 3.4-50 96 8/2/96 l

RCS P/T Limits B 3.4.9 BASES ACTIONS B.1 and 8.2 (continued)

If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because either the RCS remained in an unacceptable P/T l region for an extended period of increased stress, or a sufficiently severe event caused entry into an unacceptable region. Either possibility indicates a need for more

_ careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. With the reduced pressure and temperature conditions, the possibility of propagation of undetected flaws is decreased.

Pressure and temperature are reduced by placing the plant in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 and C.2 Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored.

Sesides restoring the P/T limit parameters to within limits, an cu luation is required to determine if RCS operation is al l owed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 212 F. Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components.

ASME Code, Section XI, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.

Condition C is modified by a Note requiring Required Action l C.2 be completed whenever the Condition is entered. The l Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.

(continued)

HATCH UNIT 2 B 3.4-51 96 8/2/96 l l

RCS P/T Limits B 3.4.9 BASES ACTIONS C.1 and C.2 (continued)

Restoration alone per Required Action C.1 is insufficient i because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

SURVEILLANCE SR 3.4.9.1 REQUIREMENTS i Verification that operation is within limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature change limits are specified in hourly l l

increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations.

l Surveillance for heatup, cooldown, or inservice leakage and

hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.

Verification of Figures 3.4.9-1 and 3.4.9-2 is required <

during non-nuclear heatups and cooldowns, and inservice leak l and hydrostatic testing. Verification of the s 100 F change in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period is required during any heatup or cooldown.

l SR 3.4.9.2 j A separate figure is used when the reactor is critical. l Consequently, the RCS pressure and temperature must be verified within the appro)riate limits before withdrawing control rods that will mace the reactor critical.

Performing the Surveillance within 15 minutes prior to initial control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time criticality is achieved.

This SR, for clarity, is modified by a Note stating that it is only required to be met when the reactor is critical and immediately prior to control rod withdrawal for the purpose of achieving criticality.

! (continued)

HATCH UNIT 2 8 3.4-52 96 8/2/96

. o RCS P/T Limits B 3.4.9 4

BASES i

SURVEILLANCE SR 3.4.9.3 and SR 3.4.9.4

! REQUIRMENTS (continued) Differential temperatures within the applicable limits l

t ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design

, allowances. In addition, compliance with these limits l ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 7) are satisfied.

Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump l

start.

If the 145*F temperature differential specified in '

SR 3.4.9.3 cannot be determined by direct indication, an

! alternate method may be used as described below:

, The bottom head coolant temperature and the RPV coolant can be assumed to be s 145*F if the following can be confirmed:

a. One or' more loop drive flows were > 40 percent of rated flow prior to the RPT,

b. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems have not l injected since the RPT,

c. Feedwater temperature has remained > 300 F since the RPT, and

d. The time between the RPT and restart is < 30 minutes.

General Electric test data from BWR plants shows that stratification up to the 145*F differential does not occur any sooner than I hour following the RPT (Refs.10 and 11).

Adding HPCI and RCIC injection, and feedwater temperature constraints provides assurance that the temperature differential will not be exceeded within 30 minutes of the RPT.

l An acceptable means of demonstrating compliance with the

' temperature differential requirnment in SR 3.4.9.4 is to compare the temperatures of the operating recirculation loop and the idle loop.

(continued)

HATCH UNIT 2. B '.4-53 96 8/2/96 l

s <

RCS P/T Lir.iits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.3 and SR 3. 4. 9 d (continuad)

RE0UIREMENTS SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 4.

In MODE 5, the overall stress on limiting components is lower. Therefore, AT limits are not required.

SR 3.4.9.5 and SR 3.4.9.6 l Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LC0 limits.

The flange temperatures must be verified to be above the limits 30 minutes before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. Verification of flange temperatures is also required while detensioning is in progress until all reactor vessel head bolts are completely detensioned. (The head is considered tensioned if one or more bolts are partly or completely tensioned.) When in MODE 4 with RCS temperature s 100 F, 30 minute checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature s 120 F, monitoring of the flange temperature is required every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to ensure the temperature is within the limits specified.

The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limitt could be exceeded. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable based on the rate of temperature change possible at these temperatures.

SR 3.4.9.5 is modified by a Note that requires the Surveillance to be met only when tensioning /detensioning the l reactor vessel head bolting studs. SR 3.4.9.6 is modified by a Note that requires the Surveillance to be met when the head is tensioned.

(continued) l' HATCH UNIT 2 B 3.4-54 96 8/2/96

l s

j l

l RCS P/T Limits B 3.4.9 BASES (continued) l l

l REFERENCES 1. 10 CFR 50, Appendix G, January 1996. l l l

2. ASME, Boiler and Pressure Vessel Code, Section III, l Appendix G. l

3. ASTM E 185-82, " Standard Practice for Conducting i Surveillance Tests for Light-Water Cooled Nuclear l

Power Reactor Vessels," July 1982.

4. 10 CFR 50, Appendix H.

L 5. Regulatory Guide 1.99, Revision 2, May 1988. ,

I

6. ASME, Boiler and Pressure Vessel Code, Section XI, l

Appendix E.

7. FSAR, Section 15.1.26.

8. Kahtan N. Jabbour (NRC) letter to W. G. Hairston, III l (GPC), Amendment 118 to the Operating License, dated i Jr.nuary 10, 1992.

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

10. GE-NE-668-13-0393, " Recirculation Pump Restart Without Vessel Temperature Indication for E.I. Hatch Nuclear P1 ant," December 28, 1993.

11. DRF A00-05834/6, " Safety & 10 CFR 50.92 Significant Hazards Consideration Assessment for RPV

' Stratification Prevention Improvements at Edwin I.

Hatch Nuclear Plant Units 1 and 2," April 1994.

12. (To be added when amendment is received.) l l

l I

t HATCH UNIT 2 B 3.4-55 96 8/2/96 i

a .

Reactor Steam Dome Pressure B 3.4.10 B 3.4 REACTOR COOLANT SYSTEM (RCS) l B 3.4.10 Reactor Steam Dome Pressure BASES BACKGROUND The reactor steam dome pressure is an assumed value in the determination of compliance with reactor pressure vessel l overpressure protection criteria and is also an assumed initial condition of design basis accidents and transients.

APPLICABLE The reactor steam dome pressure of s 1058 psig is an SAFETY ANALYSES initial condition of the vessel overpressure protection analysis of Reference 1. This analysis assunes an initial maximum reactor steam dome pressure and evaltates the response of the pressure relief system, primarily the safety / relief valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumotions of the overpressure protection analysis are conserved. Reference 2 also assumes an initial reactor steam dome pressure for the analysis of design basis accidents and transients used to determine the limits for fuel cladding integrity (see Bases for LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain (see Bases for LC0 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)").

Reactor steam dome pressure satisfies the requirements of Criterion 2 of the NRC Policy Statement (Ref. 3).

LCO The specified reactor steam dome pressure limit of

s 1058 psig ensures the plant is operated within the assumptions of the overpressure protection analysis.

Operation above the limit may result in a response more severe than analyzed.

APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these (continued)

HATCH UNIT 2 B 3.4-56 96 8/2/96 l

1 i .

Reactor Steam Dome Pressure B 3.4.10 BASES APPLICABILITY MODES, the reactor may be generating significant steam and (continued) events which may challenge the overpressure limits are possible.

In MODES 3, 4, and 5, the limit is not applicable because the reactor is shut down. In these MODES, the reactor pressure is well below the required limit, and no anticipated events will challenge the overpressure limits.

ACTIONS A_d With the reactor steam dome pressure greater than the limit, l prompt action should be taken to reduce pressure to below l the limit and return the reactor to operation within the l bounds of the analyses. The 15 minute Completion Time ',s reasonable considering the importance of maintaining tre pressure within limits. This Completion Time also ensures i that the probability of an accident occurring while pressure j is greater than the limit is minimized. 1 IL1 If the reactor steam dome pressure cannot be restored to within the limit within the associated Completion Time, the plant must be brought to a MODE in which the LCO does net apply. To achieve this status, the plant must be brought to ;

at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completiol !

Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.10.1 REQUIREMENTS Verification that reactor steam dome pressure is :s 1058 psig ensures that the initial conditions of the vessel l overpressure protection analysis is met. Operating experience has shown the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency to be sufficient for identifying trends and verifying operation within safety analyses assumptions.

(continued)

HATCH UNIT 2 B 3.4-57 96 8/2/96 l

1 1 s Reactor Steam Dome Pressure B 3.4.10 BASES (continued)

REFERENCES 1. FSAR, Supplement SA.

2. FSAR, Section 15.

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

J v

I I l 4

i

. 1 e

i 4

4 f

HATCH UNIT 2 B 3.4-58 96 8/2/96 l

-.p s .

RCS P/T Lisits B 3.4.9

/

8 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.g RCS Pressure and Temperature (P/T) Limits

! BASES i

BACKGROUND All components of the RCS are designed to withstand effects i

i of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and l shutdown (cooldown) operations, power transients, and j

' reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design i

assumptions and the stress limits for cyclic operation.

This Specification contains P/T limit curve hha pa cooldown, and inservice leakage and hydrostatic testingkand l also limits the maximum rate of change of reactor coolant i temperature. The criticality curve provides limits for both l Odheatup and criticality. g d (

E f

} Kd)(YD :;ach: P/T nth:.limit Thecurve usual defines ancurves use of the acceptable regionY is operational or-::mi i

i gg 1 guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and . ;

compared to the applicable curve to determine that operation

is within the allowable region.

1 i The LCO establishes operating limits that provide a margin ;

j to brittle failure of the reactor vessel and piping of the i reactor coolant pressure boundary (RCPB). The vessel is the 4

component most subject to brittle failure. Therefore, the l LCO limits apply mainly to the vessel.

4 10 CFR 50, Appendix G (Ref.1), requires the establishment

of P/T limits for material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate 5

margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code, Section III, Appendix G (Ref. 2). .-

t6 The actual shift established in the RT,,,

periodically of the vessel by removing materialt "^ bey 7A/

and evaluating irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Appendix H ofy CFR 50 (Ref. 4). The operating P/T limit curves . 7,ttadjusted, (continued)

HATCH UNIT 1 B 3.4-46 REVISION 1

_. _ _ _ _ __.___ ____ _ __ _ _m _ _

i I

RCS P/T Lizits B 3.4.9 BASES BACKGROUND as necessary, based on the evaluation findings and the I (continued) recommendations of Reference 5. '

I The P/T limi c vesarec9sitecurvese a lished by O k cg g h arimposi lim der ed a stress alys of tho 4

L po ions o the reac r essel a d head at are e t I

%t4h%{ rest ict tempe

e. At any re rate of ific pre ure temperatur ang% one loc n within the d

actor l Nt. vesse ill dictat the nos restri limit. ros the

! spa of e P/T mit curves, iff rent catio are no

{ re rictiv a , thus, the cu are e i es of the j ( st restri e regions.i f The h' ' tup curve re esentsadif[rentsetMestric[s 7 1 i than h cooldown u e because e di ect ons of th th al adient; thro h the ssel wa are rev ed. Jh i 5

t al gr t' reversa alt s the lo a on o he spile l l

y ress betwe ien,the outer a inner wa) s g ,

i l The criticality limits include the Reference I requirement

that they be at least 40'F above the heatup curve or the 1 4

cooldown curve and not lower than the minimum pemissible

temperature for the inservice leakage and hydrostatic ;

] testing. l 1

! The consequence of violating the LC0 limits is that the RCS

! has been operated under conditions that can result in

brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. ASME Code, Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that c.auses an excursion outside the limits.

APPLICABLE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (DBA) analyses. They are prescribed during nomal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed. Referenq@8approvedthe curves and limits specified in this sectio . Since the P/T limits are not derived from any DBA, ther are no asd l9 -

(continued)

HATCH UNIT 1 B 3.4-47 REVISION 1

i =

l l

INSERT FOR PAGE B 3.4-47 The P/T limit curve for inservice leak and hydrostatic testing, and the curve for non-nuclear i heatup and cooldown include separate curves for the bottom head, beltline, and upper vessel and l flange regions. These curves are derived from stress analysis of these vessel regions.

l l

[

l 1

I

(

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l UNIT I U 2 1

i a RCS P/T Linits

B 3.4.9 i

BASES l

l APPLICABLE acceptance limits related to the P/T limits. Rather, the '

, SAFETY ANALYSES P/T limits are acceptance limits themselves since they (continued) preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 9).

f LCO The elements of this LCO are: I

a. FR re6dre -tem rature are Yrithin-the-limits 4

/2 am a if n Fi s 3. and 3.4 -2, an atu 2 or oldo tes < 100'F ng eatup

, viS /hP / pooldo , and i ervice 1e, and h ostatic te 6,gl

/#'h b. The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is s 145'F during recirculation pump l startup;

c. The temperature difference between the reactor coolant i

in the respective recirculation loop and in the (

reactor vessel is s 50'F during recirculation pump startup;

d. RCS pressure and temperature are within the criticality limits specified in Figure 3.4.9-3, prior to achieving criticality; and

e. The reactor vessel flange and the head flange temperatures are 2 76'F when tensionin he reactor g vessel head bolting studs. ggjg G_fyd

These limits define allowable operating regions and permit a l 1arge number of operating cycles while also providing a wide margin to nonductile failure. j j

The rate of change of temperature limits controls the thermal gradient through the vessel wall and is used as input for calculating the heatup, cooldown, and inservice leakage and hydrostatic testing P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of

the P/T limit curves.

(continued)

HATCH UNIT 1 B 3.4-48 REVISION 1

s i

INSERT 1, page B 3.4-48 RCS pressure and temperature are within the limits specified in Figures 3.4.9-1 and 3.4.9-2 during RCS non-nuclear heatup and cooldown operations, and RCS inservice leak and )'

hydrostatic testing. Additionally, heatup and cooldown rates are < 100 F during any RCS heatup or cooldown, and inservice leak and hydrostatic testing.

I i

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l W2- .

INSER PAGE B 3.4-48 i

i l 1 l

i l f. The reactor vessel flange and head flange temperatures are > 76 F when the reactor vessel is tensioned. '

g. For the case when the vessel head is either off or on but noteetensione/, sec all t c55e(tio the vessel (upper vessel, beltline, and bottom head) may be lowered to a minimum of 68 F.

When the head is being tensioned, or is already tensioned, the beltline and bottom head

, regions may be lowered to 68 F, as long as there is not any pressure or heatup/cooldown.

l The upper vessel, however, has a higher minimum temperature requirement with the head l tensioned, as previously delineated.

l i The 68 F temperature is based on fuel shutdown margin considerations, since this a more j restrictive temperature than would be obtained from 10 CFR 50, Appendix G, L considerapt

& no bi in 4beMessel,&4em shre. mau dropio i 15 bd 65 MF becctase%is is % h hest g 4 meeed hk, ness regsremeds & ca rea sterials.

i i

i L

UNIT 1 U 2

RCS P/T Liaits B 3.4.9 4

BASES i

! ACTIONS C.1 and C.2 (continued) 4 that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored.

! Besides restoring the P/T limit parameters to within limits, i an evaluation is required to determine if RCS operation is

! allowed. This evaluation must verify that the RCPB

! integrity is acceptable and must be completed before j approaching criticality or heating up to > 212*F. Several

methods may be used, including comparison with pre-analyzed 1

transients, new analyses, or inspection of the components.

ASME Code, Section XI, Appendix E (Ref. 6), may be used to i support the evaluation; however, its use is restricted to i evaluation of the beltline.

l Condition C is modified by a Note requiring Required Action j C.2 be completed whenever the Condition is entered. The

, Note emphasizes the need to perform the evaluation of the l effects of the excursion outside the allowable limitr,.

Restoration alone per Required Action C.1 is insufficient

because higher than analyzed stresses may have occurred ard j may have affected the RCPB integrity.

I i SURVEILLANCE SR 3.4.9.1 REQUIREMENTS i Verification that operation is within limits is required every 30 minutes when RCS pressure and temperature i conditions are undergoing planned changes. This Frequency 4 is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature nh ;f change limits are specified in hourly increments, 30 minutes permits a reasonable time for i assessment and correction of minor deviations.

li Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria

given in the relevant plant procedure for ending the

! activity are satisfied.

[ e r u_

. Cektched. (continued) l i

HATCH UNIT 1 B 3.4-51 REVISION 1

l:

i l

l INSERT, PAGE B 3.4-51 )

Verification of Figures 3.4.9-1 and 3.4.9-2 is required during non-nuclear heatups and cooldowns, and inservice leak and hydrostatic testing. Verification of the 100 F change in any one hour period is required during any heatup or cooldown, c_

J 4

4 l

s

RCS P/T Liaits B 3.4.9 I BASES

)

i SURVEILLANCE SR 3.4.9.1 (continued)

}

j REQUIREMENTS p-7 f ' f y f 'R

/ cooldown operations'a .RCST,iaservice leakagedk'

~

! ph'ydrostatic testing -

/ L =

1

) SR 3.4.9.2 g,

! A separate is used when the reactor is approaching i criticality. Consequently, the RCS pressure and temperature i

must be verified within the appropriate limits before withdrawing control rods that will make the reactor i

critical. , , , - f. n. tat 4

i Performing the Surveillance within 15 minutes hf;ce control rod withdrawal for the purpose of achieving criticality i provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time

! sf tu ce;; trol-red-withdreweh < >$ tou y a m t,; - t J

i

! I SR 3.4.9.3 and SR 3.4.9.4 e /pJfs*

l "4 L> Differential temperatures within the applicable limits j

ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 7) are satisfied.

J l Performing the Surveillance within 15 minutes before

starting the idle recirculation pump provides adequate i

j assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump ;

. start. i j w~n If the 145'F temperature differential specified in

! SR 3.4.9.3 cannot be determined by direct indication, an i

alternate method may be used as described below:

1 The bottom head coolant temperature and the RPV coolant can q be assumed to be s 145'F if the following can be confirmed: .

j a. One or more loop drive flows were > 40 percent of

, rated flow prior to the RPT, i

t 4

(continued) !

! HATCH UNIT 1 B 3.4-52 REVISION 3 1

~

f S I

I INSERT, PAGE B 3.4-52 i

This SR, for clarity, is modified by a Note stating that it is only required to be met when !

the reactor is critical, and immediately prior to control rod withdrawal for the purpose of achieving criticality.

I i

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l

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5 s l RCS P/T Limits B 3.4.9 BASES

/ 024 SURVEILLANCE SR 3.4.9.5 SR 3.4.9.6v dd/ S$/MMM/

REQUIREMENTS

, (continued) Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations i

approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LC0 limits.

l The flange temperatures must be verified to be above the limits 30 minutes before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned 6 Un66(k Je_ limits are satisfied.4When in MODE 4 with RCS temperature s 86'F, 30 minute checks of the flange

. b( @{ - temperatures are required because of the renuced margin to the limits. When in MODE 4 with RCS temperature s 106*F, monitoring of the flange temperature is rer,uired every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to ensure the temperature is within the limits ,

specified.

The 30 minute Frequency reflects the urgency of maintaining

the temperatures within limits, and also limits the timo that the temperature limits could be exceeded. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable based on the rate of temperature change possible at these temperatures. gt gjh6dk M #d'NM. (h

~g "l" SR 3.4.9.5 is modified by ote that requir s e Surveillance to be f'%,,0,/ only when tensionin reactor vessel head U ting studs. SR 3.4.9.6 modified by a Note that recuires the Surveillance to be},: nit"td

. O minut ter F; temperatu <8 in Moife 4.

l 3.4. 7 is d fie by a te tr uires t S e lance t e init' ate 12 ho aft RCS tempera e

$1 F in Mod The s co ai d in e SRs ar ne s ry to eci when er ctor ssel/ nge a h d f} nge mpe tures e eq r d to be v ified b.e with

{,thelimit pecified.

REFERENCES 1. 10 CFR 50, Appendix G (J. grub'L \C k

2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.

1 1

(continued) 2 HATCH UNIT 1 B 3.4-53 REVISION 1 i

s INSERT, PAGE B 3.4-53 Verification of flange temperatures is also required while detensioning is in progress until all reactor vessel head bolts are completely detensioned. (The head is considered tensioned if one or more bolts are partly or completely tensioned).

N l

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j d

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I f

i P

i -

}

i 1

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RCS P/T Limits B 3.4.9 BASES l

l REFERENCES 3. ASTM E 185-82, " Standard Practice for Conducting (cor.tinued) Surveillance Tests for light-Water Cooled Nuclear Power Reactor Vessels," July 1982.

4. 10 CFR 50, Appendix H.

l 5. Regulatory Guide 1.99, Revision 2, May 1988.

6. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E.

7. FSAR, Section 14.3.6.2.

8. George W. Rivenbark (NRC) letter to J. T. Beckham, Jr.

l (GPC), Amendment 126 to the Operating License, dated

! June 20, 1986.

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

10. GE-NE-668-13-0393, " Recirculation Pump Restart Without Vessel Temperature Indication for E.I. Hatch Nuclear P1 ant," December 28, 1993.

11. DRF A00-05834/6, " Safety & 10 CFR 50.92 Significant

, Hazards Consideration Assessment for RPV Stratification Prevention Improvements at Edwin I.

Hatch Nuclear Plant Units 1 and 2," April 1994, i

l L. ( A k <DDre u% m n e ,a f h n < < <:~S,)

I l

l l

l i

HATCH UNIT 1 B 3.4-54 ., REVISION 3

t 5 l RCS P/T Liaits B 3.4.g i

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.9 RCS Pressure and Temperature (P/T) Limits l

BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and i

reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation LOCO-(kCICd l

This Specification cooldown, contains and inservice leakageP/T andlimit curvestesting hydrostatic fortheatupk,q and also limits the maximum rate of change of reactor coolant temperature. The criticality curve provides limits for both ORCle y p heatup and criticality. -

Each P/T limit curve defines an acceptable reg or p The usual use of the curves is operational (s$$F gfgIgg .

guidance during heatup or cooldown maneuvering, when -

Opc(cdarc Odvd% pressure and temperature indications are monitored and (

" compared to the applicable curve to determine that operation '

~

is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the l component most subject to brittle failure. Therefore, the i

LCO limits apply mainly to the vessel.

10 CFR 50, Appendix G (Ref.1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code, Section III, Appendix G (Ref. 2). .

(69 The actual shift in the RT of the vessel materialWE ,k established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4). The operating P/T limit curves "/fy adjusted, CLCc' (continued) (

FMTCH UNIT 2 B 3.4-46 REVISION 1

RCS P/T '.isits B 3.4.9 BASES I

BACKGROUND as necessary, based on the evaluation findings and the (continued) recommendations of Reference 5.

The P/T li t urves are co s cur s tablished by

) s erimpos ng its derive from r ss ana ses of t'ios (JL ftCL db por ons f the actor y sel and d that a the was restr ve. At a spe fic pres re temperat e, a b temper re rate of ia e, one 1 cation ithin th actor g g g- vesse w 1 dictate t span of th P/Tlimi c ves, most re rictive it. Ac fferent loc ions re the re rictive, ind t s, t rves are compos of the ;

, trestrict%,rygions e heat rve r ents iffe nt et r tr et th t coo cury ause t direc ns of e the 1 grad s thro the v el wal a re rs . The' t a r lent rev sal t st 1 ation the silej e ress en th uter and ner w s. J .

l The criticality limits include the Reference I requirement that they be at least 40*F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic

! testing.

The consequence of violating the LC0 limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a

'nonisolable leak or loss of coolant accident. In the event l these limits are exceeded, an evaluation must be performed

to determine the effect on the structural integrity of the RCPB components. ASME Code, Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

APPLICABLE The P/T limits are not derived from Design Basis Accident SAFETY ANALYSES (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws ,

to propagate and cause nonductile fail of the RCPB, a condition that is unanalyzed. Referen approved the curves and limits specified in this see o. Since the P/T limits are not derived from any DBA, there are no l

CLRcf @

l t

(continued)

HATCH UNIT 2 B 3.4-47 REVISION 1

9 3 l

INSERT FOR PAGE B 3.4-47 The P/T limit curve for inservice leak and hydrostatic testing, and the curve for non-nuclear heatup and cooldown include separate curves for the bottom head, beltline, and upper vessel and flange regions. These curves are derived from stress analysis of these vessel regions.

l i

i l

l 4

l i

d U 1 UNIT 2

RCS P/T Linits B 3.4.g BASES APPLICABLE acceptance limits related to the P/T limits. Rather, the SAFETY ANALYSES P/T limits are acceptance limits themselves since they (continued) preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 8).

LCO The elements of this LC0 are:

a. [RS' ure tempe ure are ithin the limits ~~I ifi .- 3 , Etup

, # 'or down es ar 00*F ng R eatup 4 / coold d ins ice 1e and h ostatic sting;

b. The temperature difference between the reactor vessel udp / y bottom head coolant and the reactor pressure vessel P (RPV) coolant is $ 145'F during recirculation pump startup;

c. The temperature difference between the reactor coolant -

in the respective recirculation loop and in the (

reactor vessel is s 50*F during recirculation pump startup;

d. RCS pressure and temperature are within the criticality limits specified in Figure 3.4.9-3, prior to achieving criticality; and

e. The reactor vessel flange and the head flange th h Lhscr temperaturesare290*FwhentensioningfereactorO(

vessel head bolting studs. de NCLChfd. These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.

The rate of change of temperature limits controls the thermal gradient through the vessel wall and is used as input for calculating the heatup, cooldown, and inservice leakage and hydrostatic testing P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves.

(continued) (

HATCH UNIT 2 B 3.4-48 REVISION 1

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

s % ,

INSERT 1, page B 3.4-48 RCS pressure and temperature are within the limits specified in Figures 3.4.9-1 and 3.4.9-2 during RCS non-nuclear heatup and cooldown operations, and RCS inservice leak and hydrostatic testing. Additionally, heatup and cooldown rates are $ 100 F during any RCS heatup or cooldown, and inservice leak and hydrostatic testing. ,

4 e

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l

+ - - ,. ,,

s l t

l l

)

L l INSER'I PAGE B 3.4-48 ;

1 l

f. The reactor vessel flange and head flange temperatures are > 90 F when the reactor vessel is l tensioned.

g-4he, Ve

g. For the case when the vessel head is either off or on but not tensionedctions (, all of three se%e) the vessel (upper vessel, beltline, and bottom head) may be lowered to a minimum of 68 F.

l When the head is being tensioned, or is already tensioned, the beltline and bottom head l regions may be lowered to 68 F, as long as there is not any pressure or heatup/cooldown.

l The upper vessel, however, has a higher minimum temperature requirement with the head tensioned, as previously delineated.

The 68 F temperature is based on fuel shutdown margin considerations, since this a more restrictive temperature than would be obtained from 10 CFR 50, Appendix G, considerg l

[(0% w fuel in % vessel, h 4emperedure, maa drop b as W as 60"F beconsedhts iSh kiqhest cLtd4iGcBhton temperchre 4o ,niee+ hjness YdfuW6hed5 Wof d l

yencdce nefencds. 3 l

I i )

l U 1 UNIT 2k l

i %. ,

RCS P/T Limits B 3.4.9 BASES l

l ACTIONS C.1 and C.2 (continued) 1 i that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are

! restored.

Besides restoring the P/T limit parameters to within limits,

~

an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before .

approaching criticality or heating up to > 212*F.. Several l l methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components.

ASME Code, Section XI, Appendix E (Ref. 6), may be used to ,

support the evaluation; however, its use is restricted to ;

evaluation of the beltline. l l Condition C is modified by a Note requiring Required Action I l C.2 be completed whenever the Condition is entered. The 1 Note emphasizes the need to perform the evaluation of the i

! effects of the excursion outside the allowable limits. i Restoration alone per Required Action C.1 is insufficient l because higher than analyzed stresses may have occurred and

! may have affected the RCPB integrity.

I SURVEILLANCE SR 3.4.9.1 l REQUIREMENTS Verification that operation is within limits is required every 30 minutes when RCS pressure and temperature l

conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room l indication available to monitor RCS status. Also, since l temperature ;;t; ;f change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of. minor deviations.

Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.

Q ECL O tt(L. 5 s been toAe per le with d

ot ha requir y durin i

ste 4eatup nd 4 { veil l

CthchcL(. (continued) 1 HATCH UNIT 2 B 3.4-51 REVISION 1 r

6 t

INSERT, PAGE B 3.4-51 Verification of Figures 3.4.9-1 and 3.4.9-2 is required during non-nuclear heatups and cooldowns, and inservice leak and hydrostatic testing. Verification of the $ 100 F change in any one hour period is required during any heatup or cooldown.

I i

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RCS P/T Liaits 8 3.4.9 BASES 1

SURVEILLANCE SR 3.4.9.1 (continued)

REQUIREMENTS c6o]down~ operations,and RCS-inservi[e]1iakigeind /

, hydrostat'ic ',testing.f- ' _

.n

.)

SR 3.4.9.2

, -S;y r e-l A separate Haft is used when the reactor is ;;;r::f.;r.g critical,Wy. Consequently, the RCS pressure and temperature l must be verified d'hin the appropriate limits before l withdrawing control rods that will make the reactor critical.

g%%

Performing the Surveillance within 15 minutes before J control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be a

exceeded between the time of the Surveillance and the time m, m<, o - = > . _ _ _ j ,

c ro r.a rily h uh, cad !

Jul-C, + i/a hw r /t l l SR 3.4.9.3 and SR 3.4.9.4 '

'd Differential temperatures within the applicable limits e

ensure that thermal stresses resulting from the startup of

>w an idle recirculation pump will not exceed design l w ;A allowances. In. addition, compliance with these limits w Mg*pa - ensures that the, assumptions of the. analysis for the startup of an idle recirculation loop (Ref. 7) are satisfied.

l D Performing the Surveillance within 15 minutes before

starting the idle recirculation pump provides adequate

, '*,: g# assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump

..u start.-

) ' m "]}$v ' '

If the 145'F temperature differential specified in

SR 3.4.9.3 cannot be detemined by direct indication, an alternate method may be used as described below:

.& , . . m

- l y

The bottom head coolant temperature and the RPV coolant can.

be assumed to be s 145'F if the following can be confimed:

a. One or more loop drive flows were > 40 percent of l g> rated flow prior to the RPT, 1

(continued)

J L " HATCH UNIT 2 B 3.4-52 REVISION 3

) yL;3

. 6 %

e i

i INSERT 2, PAGE B 3.4-52 This SR, for clarity, is modified by a Note stating that it is only required to be met when the reactor is critical, and immediately prior to control rod withdrawal for the purpose of achieving criticality.

0 k

i k

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I 1

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h 1

l a

4.

RCS P/T Liaits i B 3.4.9 BASES-g CbVI SURVEILLANCE SR 3.4.9.5

R 3.4.9.6w*"'*'N'** '

REQUIREMENTS (continued) Limits on the reactor vessel flange and head flange temperatures are ger.arally bounded by the other P/T limits ,

during system heatup and cooldown. However, operations i approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the ;

LCO limits. 1 The flange temperatures must be verified to be above the

limits 30 minutes before and while tensioning the. vessel i

i QOt'6 (ONN/ the limits are satisfied.fWhen in MODE 4 with RCSa ) head bolting stud obched. r te=perature 5 loo *F, 30 niinute checks of the flange l \ temperatures are required because of the reduced margin to 3 the limits. When in MODE 4 with RCS temperature s 120*F, l monitoring of the flange temperature is required every 1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to ensure the temperature is within the limits j specified.

4 The 30 minute Frequency reflects the urgency of maintaining

! the temperatures within limits, and also limits the time

that the temperature limits could be exceeded. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i Frequency is reasonable based on the rate of temperature change possible at these temp t res. gg i SR 3.4.9.5 is modified by Note that requires 3th )

i Surveillance to be "-t't,, d only when tensionincj>the

! reactor vessel head bolting studs. SR 3.4.9.6 is modified j by_ a Note that rec!uires the Surveillance to be h4S'"4 O minu s after F temp ature 1 0"F Mode 4. 1 S 3.4 .7 is no fi b a Note at re ires the

} Sur lance to ini ated 12 urs ter RCS pe ture j $1 F in Mo 4. Te tes ntain thes SRs r l i ne ss ry to pecify when e eacto vess 1 f ange nd ad f ange ratures are req ed to e veri d +a -_ with *)

(helimi specif

]

acoA benoned.

w ,

f REFERENCES 1. 10 CFR 50, Appendix G d(ututt(L y R%

. a 4 2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.

]

4 i (continued)

HATCH UNIT 2 B 3.4-53 REVISION 1

g l

l i

INSERT, PAGE B 3.4-53 Verification of flange temperatures is also required ;.ldte detensioning is in progress until '

all reactor vessel head bolts are completely detensioned. (The head is considered tensioned if one or more bolts are partly or completely tensioned).

4 %

RCS P/T Limits B 3.4.9 i

BASES l REFERENCES 3. ASTM E 185-82, " Standard Practir.e for Conducting (continued) Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," July 1982.

4. 10 CFR 50, Appendix H.

5. Regulatory Guide 1.99, Revision 2, May 1988.

6. ASME, Boiler and Pressure Vessel Code, Section XI, !

Appendix E.

7. FSAR, Section 15.1.26.

8. Kahtan N. Jabbour (NRC) letter to W. G. Hairston, III (GPC), Amendment 118 to the Operating License, dated January 10, 1992.

' i i 9. . NRC No.93-102, " Final Policy Statement on Technical !

ci; Specification Improvements," July 23, 1993.

- 10. GE-NE-668-13-0393, " Recirculation Pump Restart Without Vessel Temperature Indication for E.1. Hatch Nuclear m.naQA4 Plant," December 28, 1993.-

g

$ Y- 11. DRF A00-05834/6, " Safety & 10 CFR 50.92 Significant

t. , o

Hazards Consideration Assessment for RPV

. . N' Stratification Prevention. Improvements at Edwin I.

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