ML20082T375

Amend 124 to License NPF-6,revises pressure-temp Limits in Facility TS Section 3.4 & Associated Bases for Heatup, Cooldown,Leak Test & Criticality of RCS for 21 Effective Full Power Years

ML20082T375
Person / Time
Site:	Arkansas Nuclear
Issue date:	09/10/1991
From:	Quay T Office of Nuclear Reactor Regulation
To:
Shared Package
ML20082T382	List: ML20082T375
References
NPF-06-A-124 NUDOCS 9109180223
Download: ML20082T375 (14)

Text

. -

- _ _ _. --=

1. pm eg%,

UNITED STAT ES

1. ['3.c.,

NUCLE AR HEGULATORY COMMISSION eI nuamotos o. c. rosss

\\...../

ENTERGY OPERATIONS, INC.

DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING, LICENSE Amendment No.124 License No. NPF-6 1.

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

A.

The application for amendment by Entergy Operations, Inc.

(the licensee) dated June 18, 1991, as supplemented July 22, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regula-tions set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance:

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

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

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

l e

P 3

e

..c

y..

l l

i

-g.

2.

Accordingly, the license is amended by changes to the lechnical Specifications as indicated in the attachment to this license

{

amendment, and Paragraph 2.C.(2) of Facility Operating License No.

NPF-6 is hereby amended to read as follows:

li 2.

Technical,5pecifications t

The Technical Specifications contained in Appendix A, as l

revised through Amendment No.124, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

j 3.

The license amendment is effective as of November 18, 1991.

}

FOR THE NUCLEAR REGULATORY COMMISSION l

3 hwk

, h,

& Theodore R. Quay, Diiector r

Project Directorate IV-1 l

Division of Reactor Projects Ill, IV, and V Office of Nuclear Reactor Regulation i

Attachment-Changes to the Technical i

Specifications l

Date of Issuance:

September 10, 1991 I

f

[

?

t I

t

. -. ~., -. -

t I

[

ATTACHMENT (J l12iN$E AMENDMENT NO.i?4 l

[

fACIL41Y OPERATING LICENSE NO. NPF-6 t

i DOCKET NO. 50-368 I

Revise the following pages of the Appendix "A" Technical Specifications

(

with the attached pages.

The revised pages are identified by Amendment j

number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

l r

REMOVE PAGES INSERT PAGES 3/4 4-72 3/4 4-22 3/4 4-22a 3/4 4-22a i

3/4 4-23 3/4 4-23

{

3/4 4-23a 3/4 4-23a 3/4 4-23b 3/4 4-23b f

B 3/4 4-5 8 3/4 4-5 B 3/4 4-6 B 3/4 4-6 B 3/4 4 9 0 3/4 4-9 i

B-3/4 4-10 B 3/4 4-10 l

l y

?

?

t I

)

i t

v L-t 1

l t

t

-m f

_.m...

E

.t y.-.

.__.a_

n.

U 250

.3

. ; 1._,. _,..

' 9,_

_ _ p_

4 g

_ %q. _

7 5

_p 7.._

qp 7.,, ; r..

j

. _ _. 7_

j

, 4_,_._.

w 3

i,

.r i-1 1 1

u

_1 ' *._

'I 4 _. _

t-41 p j-

7. L%,.q.1J,.

.-- -H-goo D

__.ppp.Yr1 OPERAT, ION

'i l

J4 L _LLW UNACCEPTABLE

i.

i TrF iiw i.

.'r....+_

---v]-' P F

- -d, r...-

r 0

._ 7

_ u-.-

- t r._~., -. I._ _t : 7._

o

- -. ~, -

--- r g

u.

. 73

,__.r-

,j

,L g

H 150 se

___2_

q 7 _..2. p

. aa 4_.

_.y

.J l

--.g.

_ - -- [

8 2

1

..p.~_,

_ __.. _77

_..L!_

t, j

3 m

9.A- -

- 100 j.a_

A_

m

=

_.___t

_.g g

4

" ;,I '-

.r

- *-- - +

O 1-

_N. y.v _ _I 7 t.,I_t M i

. : 2_. - * - ~ ~ _

I

- r-i Ly'_ J_-h7 y._ _}....._ g g;,..Lg J.y

~

k

_ ACCEPTABL,v,._g7g.._._y4T--

~

~*l-'t '\\

l, s

4

~

'~

3.

Wj OPERATION i,

i c i i

50

,4

,;..,.p _. _.. - _.. _ _

., L -

@~. '

. _ -,J

.u.a._;._

m j y.

w o

9_

_(

_7.

w

.e., _.. :

8 4, _.rT' pL _27 L.,_ - _ r2_.,.., 4,.

m g

, ;,L,

. t__i T,..

i,..,

i i

i Ow r

, n-.. 3-..y

-_.--t-20 30 40 50 00 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 1.0pCl/ gram Dose Equivalent 1131 i

AR h sA5 - UNIT 2 3/4 4-21

+

~

REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LI_MITS REACTOR COOLANT SYSTEM e

LIMITING CONDITION FOR OPERATION 3.4.9.1.

The Reactor Coolant System (except the pressurizer) tempersa..e

}

and pressure shall be Ifnited in accordance with toe limit ifnes shosn on Figures 3.4-2A, 3.4-2B and 3.4-2C during heatup,.cooldown, criticality,-and l inservice leak and hydrostatic testing operations with:

i a.

A maximum heatup of 50'F, 60'F, 70'T or 80*F in any one hour period in accordance with curves A, B, C or D, respectively, in Figure 3.4-2A.

e b.

A maximum cooldown rate based on

- i RCS Temperature (Tc)

Maximum Cooldown Rate j

l T > 220'r 100'F per hour (constant) or 50*F in any half hour period (step) 140'F 5 T 5 220'F 60'F per hour (constant) or 30'r i

in any half hour period (atep) 7 < 140*F 25'F per hour (constant) or 12.5'F C

2a any half hour period (step)

A maximum temperature change of 5 10'F in any one hour period c.

during inservice hydrostatic and leak testing operations above f

the heatup and cooldown limit curves.

?

i APPLICABILITY: At all times.

3 ACIl03:

With any of the above limits exceeded, restore the temperature and/or 3

pressure to within the acceptable region of the applicable curve within 30 l

minutes;. perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness prope tfes of the Roactor

~

Coolant System; determine that the Reactor Coolant System remains

' ~

acceptable - for continued operatiu.;s or be-in at least l'OT STANDBY within __

-[

the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and~ reduce the RCS T and pressure to less than 200'F and i

less than 500 paia, respectively, witEin the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

I

' I i

i 1

ARKANSAS - UNIT 2 3/4 4-22 Amendment No. 124

?

e 4

---

tw

.+e,,

~,

,wr,

.--...r,s

.e-.n-

+a w --

v----,+-vm

REACTOR COOLANT SYSTEM EURVEILLANCE REOUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals shown in Table 4.4-5.

The results of these examinations shall be used to update Figures ~ J-2A, 3.4 2B and 3.4-2C.

ARKANSAS - UNIT 2 3/4 4-22a Ar-ndment No. 124

]

Figure 3.4 2A ARKANSAS NUCLEAR ONE UNIT 2 i

HEATUP CURVE 21 EFPY REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS r

2500 llllll l

~~

' CURVE A -

~

+

jjjjjjjj j

-~~

CURVE 8 :::: 5

- CURVE D

+-

4

,II

iiiiiii, 5

2000 CURVE C ---

n.

uI j!

j i-5

} lj f

f

?

t CURVE E mm l

l

(

1500 IIII I

li l

ACCEPTABLE l

I!

I/' '

-- I REGION gw Il l/ ~ }/

t:!

ll{

f(

i i

LOWEST

] /

SERVICE

/

- j ((

1000

-w TEMPERATURE m

170 'F L1

/

_g/

f I

I

\\ l

/

A

il e

-\\1

/

j-w tiill f

r(

.I,,

3 li A

500

_d. I I IIk I

I

,Ahl.

z 454 -"

l I

';II glll i!jj ;y MINIMUM 80LTUP TEMPERATURE (70*F)

~

.lli l

l

'ill l

l o

0 100 200 300 400 500 600 l-INDICATED REACTOR COOLANT TEMPERATURE T, F C

1 l

H'U LIMIT CURVE K13 RATE LIMIT A

50'F/HR B

60*F/HR C

70*F/HR l~

D 80*F/HR (NON-CRmCAL CORE)

E S0*F/HR (CRmCAL CORE) r ARKANSAS - UNIT 2 3/4 4-23 Amendment No.124

Figure 3.4 2B ARKANSAS NUCLEAR ONE UNR 2 COOLDOWN CURVE 21 EFPY-REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS 2500 f

i I

I g

05 2000 l'

b I

i g

- -e ;

m D

ACCEPTABLE l

f REGION 1500 I

/l ll l

E I

l l/

E i

I

/

g

iii I n

/

3 6

l-t f4 II l-l l

l

.N

$),

-1003 LOWEST [

l

)

I

' ' l lj i

i a

SERVICEi 4

i

/

c TEMPERATURE-__.#

W 170*F Nmtfl

/I H4 l!!

xN I

III II I

I 500

~

l z

454 -

II,

l!*l i l

l i

,;l l

L MINIMUM BOLTUP TEMPERATURE (70*F) g

iI l

I 0

O 100 200 X *t 400 500 600 INDICATED REACTOR COOLANT TEMPERATURE T, F c

RCS TEMP (Tc)

C/D R ATE STEP

• T > 220*F 100'F/HR s 50*F IN ANY 1/2 HR PERIOD 140F s T s 220'F 60'F/HR s 30*F IN ANY 1/2 HR PERIOD T = 140*F 25'F/HR s 12.5'F IN ANY 1/2 HH PERIOD

• Not to esosed the sp+cifMHf instantaneous decrease in temperature with e subsequent thirty minute hold sRKANSAS - UNIT 2 3/4 4-23a Amendment No.124

~..

Figure 3.4-2C ARKANSAS NUCLEAR ONE UNR 2 INSERVICE HYDROSTATIC TEST CURVE - 21 EFPY REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS 2500 I

i J

{

2000

/

l tu ti i

g I

ACCEPTABLE D

1 1,

l REGION 1500

, LOWEST b

I IIII

'II IIIII Q-E TEMPERATURE-~~

[i SERVICE iiii 170 'F -_'- J f

g'.

E

'I!!!

l N' INSERVICE h /l HYDROSTATIC -

D

!))

l1) i II I

iib I

III II' 1000

illli l

l 0-

!!i l!

O

.i N

!'I, lll I!I!

l 500 z

4s4 -

lll' l

l

!!I

,.i - HINIMUM BOLTUP TEMPERATURE (70'F) p l

!!Ii 0

O 100 200 300 400 500-

--600 INDICATED REACTOR COOLANT TEMPERATURE T, F c

A maximum temperature change of s 10'F in any one hour period during Inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.. Otherwise,the heatup arn!

cooldown limit curves apply.

ARKANSAS - UNIT 2 3/4 4-23b Amendment No.124

TABLE 41425 2?.g'

- REACTOR VESSEL MATERIAL ~ IRRADIATION SURVEILLANCE SCHEDULE g

. z. -

SPECIMEN

' REMOVAL INTERVAL

^

E -

'1.69 EFPY'.

l A

ro 2.

19 EFPY 3.

30 EFPY

.i

- 4. -

Standby j

. S '.

Standby

'i 6.

Standby i

r 2

'!q a

R a

n g

E t

l I

w I

I, l

.}'I j

'89 a

m-.

w,

.c

~ ~

REACTOR COOLANT SYSTEM BASES steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 gpm and a concurrent loss of offsite electrical power.- The values for the limits on specifle. activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters of the Arkansas Nuclear One site, such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site. This reevaluation may result in higher limits.

Th.

'I

)N statement permitting POWER OPERATION to continue for limited ods with the primary coolant's specific activity > 1.0 pC1/ gram 3

4VALEMT l-131, but within the allowable limit shown on. Figure 1, accommodates possible iodine spiking phenomenon which.may occur following changes in THERMAL POWER.

Reducing T to ( S00'T prevents the release of activity should a steam generator fu6e rupture since the saturation pressure of the primary' coolant is.below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity. levels in the primary coolant will be detected in sufficient time to take corrective action.

Information.obtained on lodine spiking will be used to assess the parameters associated with spiking phenomena.

A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS

-All components in-the Reactor Coolant System are designed to withstand the effects _of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 5.2.1.5 of the FSAR.

During startup and' shutdown, the rates of tesperature and pressure changes-are limited so that the maximum specified heat'ap and cooldown retes do not-exceed the design assumptions and satisfy the stress limits for' cyclic operation.

Operation within-the limits of the appropriate heatup and cooldown curves-assure the integrity of-the reactor vessel against fracture induced by combined thermal and pressure stresses. As the vessel is: subjected to =

increasing fluence, the toughness of the limiting material continues to declino, and even more restrictive' pressure / temperature limits must be observed.- The current limits. Figures 3.4-2A, 3.4-2B and 3.42-C are for up to anc. including 21 Effective Full Power Years (EFPY) of operation.

l ARKANSAS - UNIT 2 B 3/4 4-5 Amendment No. 72,124 l

J REACTOR COOLANT SYSTEM BASES The reactor vessel materials have been tested to determine their initial RT

the results of these test are shown in Table B 3/4.4-1.

ReactoropkNt.fonandresultantfastneutron(E>1Mev)frradiationwill cause an increase in the RT on Figure 3.4-2A, 3.4-2B an P I The heatup and cooldown limit curves shown 4-2C inciude predicted adjustments for this shift in RT at the end of the applicable service period, as well as adjustments Nr the location and for possible errors in the pressure and N

temperature sensing instruments.

It should be noted that the location adjustment considered the operation of three RCPs from a RCS temperature of 70'F and above.

The heatup, cooldown, and hydrostatic test limits are presented in tabular form in Table B 3/4.4-2.

The shift in the material fracture toughness, as represented by RT is calculated usir.y Regulatory Guide 1.99, Revision 2.

For21EFPY,ab[he 1/41 position, the adjusted reference temperature (ART) value is 111'F.

At the 3/4t position the ART value is 96'F.

These values are conservatively based on a reactor vessel inner surface fluence of 3.74 x 10nyt.

The fluence at the 1/4t point is 2.33 x 10nyt and the fluence of the 3/4t point is 9.06 x 10'*nyt.

These values are used with procedures developed in the ASME Boiler and Pressure V3ssel Code,Section III, Appendix G to calculate heatup and cooldown limits in accordance with the requirements of 10 CFR Part 50, Appendix G.

To develop composite pressure / temperature limits for the heatup transient, the isothermal, 1/4t heatup, and 3/4t heatup pressure / temperature limits are compared for a given thermal rate.

Then the most restrictive pressure / temperature limits are combined over the complete temperature interval resulting in a composite limit curve for the reactor vessel beltline for the heatup event.

To develop composite pressure / temperature Ilmit for the cooldown event,the isothermal pressure / temperature limits must be calculated.

The isothermal pressure / temperature limit is then compared to the pressure / temperature lia.it associated with both the constant cooldown rate and the corresponding step change rate (an instantaneous drop in temperature followed by a hold period). The more restrictive allowable pressure / temperature limit is chosen resulting in a composite limit curve for the reactor vessel beltline.

Both 10CFR Part 50, Appendix G and ASME Code Section III Appendix d, require the development of pressure / temperature limits which are applicable to inservice hydrostatic tests.

The minimum temperature for the inservice hydrostatic test pressure can be determined by entering the curve at the test pressure (1.1 times normal operating pressure) and locating the corresponding temperature. This curve is shown for 21 EFPY on Figure 3.4-2C.

Similarly, 10CFR Part 50 specifies that core critical limits be established based on material considerations.

This limit is shown on the heatup curve, Figure 3.4-2B.

Note that this limit does not consider the l

core reactivity safety analyses that actually control the temperature at l

which the core can be brought critical, i

ARKANSAS - UNIT 2 B 3/4 4-6 Amendment No.124

TABLE B 3/4.4-2 ARKANSAS NUCLEAR ONE UNIT 2 21 EFPY - TECHNICAL SPECIFICATION PRESSURE-TEMPERATURE LIMITS COOLDOWN HEATUP HYDROSTATIC BELTLINE BELTLINE BELTLINE COMPOSITE CURVE COMPOSITE CURVE COMPOSITE CURVE RCS P-ALLOWABLE P-ALLOWABLE (PSIA)

P-ALLOWABLE TEMPERA 1URE (PSIA)

ISO 50 F/

60 F/

70 F/

80 F/

(PSIA)

DEG. F THERMAL HOUR HOUR HOUR HOUR 70 358.6 448.4 433.8 419.7 406.5 393.8 655.0 l

72.5 368.6 464.0 433.8 419.7 406.5 393.8 675.8 l

80 l

82.5 378.6 85 388.6

.93.8 699.8 482.0 433.8 419.7 406.5 90 95 408.6 97.5 418.6 502.8 433.8 419.7 406.5 393.8 727.5 100 102.5 428.6 107.5 448.6-110 526.8 439.2 420.2 406.5 393.8 759.5 120 488.6 554.6 452.6 427.9 409.3 393.8 796.6 130 586.7 472.1 442.1 418.4 399.1 839.4 132.5 528.6 140 498.6 623.9 497.8 462.6 434.1 409.7 889.0 150 558.6 666.8 529.5 489.2 455.7 427.0 946.2 160 628.6 716.4 567.7 522.4 483.5 449.7 1012.4 170 698.6 773.8 613.9 562.4 518.1 479.5 1088.9 180 788.6 840.2 668.0 610.1 560.1 515.4 1177.4 190 888.6 916.9 731.0 666.3 609.8 559.6 1279.6 200 998.6 1005.5 803.9 732.1 668.2 611.3 1397,8 210 1108.0 1108.0 888.2 808.7 737.4 673.4 1528.2 220 1226.5 1226.5 987.6 897.8 817.9 744.9 1678.0 230 1363.4 1363.4 1102.3 1001.2 911.1 830.0 1851.3 240 1521.8 1521.8 1234.4 1121.1 1019.7 927.3 2051.5 250 1704.8 1704.8 1386.7 1259.9 1146.1 1042.4 2283.0 260 1916.5 1916.5 1562.1 1420.5 1291.9 1173.7 2550.7 270 2161.1 2161.1 1768.1 1606.4 1460.0 1328.4 280 2443.9 2443.9 2005.6 1821.3 1656.6 1504.7 j

290 2770.8 2770.8 2279.1 2070.0 1883.2 1712.2 300 2594.0 2357.5 2144.3 19=8.5 2690.0 2447.0 2226.3 310 320 2798.3 2542.5 ARKANSAS - UNIT 2 B3/4 4-9 AMENIPfENT NO.124

REACTOR COOLANT SYSTEM BASES The Lowest Service Temperature is the minimum allowable temperature at pressures above 20% of the pre-operational system hydrostatic test pressure (624 psia). This temperature is defined as equal to the most limiting RT for the balance of the Reactor Coolant System component (c$0Iervativelyestimatedso50*F)plus100'F,perArticleNB2332of Section III of the ASME Boiler and Pressure Vessel Code. Temperature instrument uncertainty is conservatively estimated as 20'F.

The horizontal line between the minimum _ boltup temperature and the Lowest Service Temperature is defined by the ASME Boiler and Pressure

- Vessel Code as -20% of the pre-operational hydrostatic test pressure.

The minimum boltup temperature is the minimum allowable temperature at pressures below 20% of the pre-operational system hydrostatic test The minimum is defined as the initial RT for the material of pressure.

thehigherstressedregionofthereactorvesselpiggTany effects for irradiation per Article G-2222 of Section III of the ASME Boiler and Pressure Vessel lCode. The initial reference temperature of the reactor vessel and closure head flanges.was determined using the certified material-test reports and Branch Technical Position MTEB 5-2.

The maximum initial RT ass ciated with the stressed region of-the vessel flange is 30'F.

NDT The minimum boltup temperatu7e including temperature instrument uncertainty is 30'F + 20'T = 50'F.

However, for additfonal conservatism, a minimum boltup temperature of 70'T is utilized.

-The number of reactor vessel: irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

.The limitations imposed on the pressurizer heatup and cooldown rates are provided to assure.that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

ARKANSAS - UNIT 2 B 3/4 4-10 Amendment No. 47,124 4

a