ML20114F394

Amend 57 License NPF-42 Revising TS 4.4 by Deleting Reactor Vessel Matl Specimen Withdrawal Schedule in Table 4.4.5

ML20114F394
Person / Time
Site:	Wolf Creek
Issue date:	10/05/1992
From:	Black S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20114F396	List: ML20114F394 ML20114F401
References
NUDOCS 9210130087
Download: ML20114F394 (11)

Text

_ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _

gue os

'o UNITED $TATES

'g NUCLEAR REGULATORY COMMISSION n

{

y W ASHING TON, 0. C. 20555

%...../

WQ1F CREEK NI) CLEAR OPERATING CORPORAT101{

WOI.F CREEK GENERATING STAT 10M DOCKET NO. 50-482 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 57 License No. NPF-42 1.

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

A.

The application for amendment to the Wolf Creek Generating Station (the facility) facility Operating License No. NPF-42 filed by the Wolf Crtek Nuclear Operating Corporation (the Corporation), dated June 11, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance:

(i) that the activities authorizeo 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 i,ccordance with 10 CFR Part 51 of the Commission's regulati'ns and all applicable requirements have-been satisfied.

i p

AO f

P

2-2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of facility Operating License No. NPF-42 is hereby amended to read as follows:

2.

Technical Specifications The Technical Specifications contain1d in Appendix A, as revised through Amendment No. 57, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hew by incorporated in the license. The Corporation shall operate the facility in accordt.nce with the Technical Specifications and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of,

FOR THE NUCLEAR REGULATORY COMMISSION

,. ( ;

!..

• l

'[

Suzanne'C. Black, Director Project Directorate IV-2 Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation

Attachment:

Ct.anges to the Technical Spec'fications Date of I suance:

October 5, 1992

.n ca.,

a

4 ATTACHMENI TC LICEN"3E AMENDMENT NO. 57

[AClllTY OPERATING LICENSE NO. NPF-42 DOCKET NO. 50-482 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

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

REM 0VE iliiBI Vlli Vill 3/4 4-29 3/4 4-29 3/4 4-32 B 3/4 4-8 B 3/4 4-8 B 3/4 4-15 B 3/4 4-15

_ =. -

I LIMI, TIN 3. CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INST RUME NT AT 10N - (Contin Jed)

TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS DELETED Radioattive Gaseous-Effluent Monitoring Instrumentation DELETED Explosive Gat Monitoring Instrumentation.................

3/4 3-58 TABLE 3.3-13 EXPLO51VE GAS MONITORING INSTRUMENTATION.............

3/4 3-59 TABLE 4.3-9 EXPLO51VE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........................

3/4 3-61 3/4.3.4 TURBINE OVER5 PEED PROTECT 10N..............................

3/4 3-63 3/4.4 REACTOR COOLANT SYSTEM 3/4,4.1 REACTOR CODLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation...............................

3/4 4-1 H t StandL..

3/4 4-2 Hot Snutdo -

3/4 4-3 Loops filled............................

3/4 4-5 Cold Shutcor,:

Col d Shutdown - Loops Not F i11 ed.........................., 3/4 4-6 3/4.4.2 SAFETY VALVES Shutdow.r....

3/4 4-7 Operating.

3/4 4-8 3/4.4.3 PRL55URIZER.

3/4 4-9 3/4.4.4 RELIEF VALVE 5..

..3 3/4 4-10 3/4.4.5 STEAM GENERATOR 5..........................................

3/4 4-11 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATOR 5 TO BE INSPECTED DURING INSERVICE INSPECTION......................., 3/4 4-16 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION......................

3/4 4-17 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection 5ystems...........................

3/4 4-18 Operational Leakage.......................................

3/4 4-19 l

WOLF CREEK - UNIT 1 VII Amendment No.15, 42

LIMITING CONDlT10N3 FOR OPERATION AND SURVflLtANCE RE0VIREMENTS l

SECTION EAGE TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSilRE ISOLATION i

VA!VES..........................................

3/4 4-21 3/4.4.7 CHEMISTRY...........................................

3/4 4-22 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............. 3/4 4-23 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS....................................

3/4 4-24 3/4.4.8 SPECIFIC ACTIVITY...................................

3/4 4-25 FIGURE 3.4-1 DOSE EQUIVALENT l-131 REACTOR COOLANT SPECIFIC ACTIVITY LlHIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY

> 1 pCi/ GRAM DOSE EQUIVALENT I-131..............

3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM................................

3/4 4-28 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System..........................

3/4 4-29 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 7 EFPY.........................

3/4 4-30 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP T0 7 EFPY......................... 3/4 4-31 TABLE 4.4-S DELETED Pressurizer...

3/4 4-33 Overpressure Protection Systems................. 3/4 4-34 FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE COLD OVERPRESSURE HITIGATION SYSTEM.................. 3/4 4-36 3/4.4.10 STRUCTURAL INTEGRITY................................ 3/4 4-37 3/4.4.11 REACTOR COOLANT SYSTEM VENTS........................

3/4 4-38 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS........................................

3/4 5-1 WOLF EREEK - UNIT 1 Vill Amendment No. 40,57

. -, +

,.-m,,

m-

l 3/4.4,S PRESSURE / TEMPERATURE LIMITS BEACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the i+mit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

A naximum heatup of 60*F in any 1-hour period for indicated 1,,,

a.

less than or equal to 200'F, b.

A maximum heatup of 100'F in any 1-hour period for indicated T,y, greater than 200*F, c.

A maximum cooldown of 100*F in any 1-hour period, and d.

A maximum temperature change of less than or equal to 10*F in any l-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Cuolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT 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 than200*Fand500psig,respectively,withinthefoWowing30 hours.

SURVElllANCE REQ _ULREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressura 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, as required by 10 CFR Part 50, Appendix H.

The results of these examinations l

shall be used to update Figures 3.4-2, 3.4-3, and 3.4-4 WOLF CREEK - UNIT 1 3/4 4-29 Amendment No. 40,57

MATERAL PROPERTY BASIS Controlling Material:

RV Lcwer Shell Copper Content:

0.07 Wel ht s Nickel Content:

0.62 We ht s initial RT 40 Deg NDT RT After 7 EFPY:

1/4T: 108.0 Deg F NOT 3/4T: 99.47 Dog F

.3000 TTi Ill;Illli 11 ll 1

l l _

lll111111!

!*I ii

!!illl. ::i Il 1

1 I

L i

catildtikuutt l.3 iit; i,

lorr/m HuTue cuevt i'

5 mto on ser/m L l'

Ll!6 1.,1

'l i

Mi i

I

' ' Q'

,ntAtuP euart I

2500-I d-fI 1: /; 1/

'I

li i

i t

+

ll.

i'i tii ti i ;!!!!,

il lil t/'

l l'i :/

i i

titli.il i

lit litii illiTi' I

(LI_

l/L / fi - /;

j it II

,0

,! ! i I

ttu itst i ~ g[,

fi f,pl,/ }

i jj l

i l

LI"If

'l (5 2000

- /:

fif

(

l si g

111

__L t

i

/-

)

/J J

l 11 1:

l

/

) Ji J

/

/

I ti g

_,n e

li I

11 11]

/

i 111LI

'iiiiii til l

t

/

/j f

/l calTICAWTTMMIT,j,,

1 D

':i

IIIIII'l A

f 1500 -

m IWum I t

x i i se r/ m MtATur cunyt

,' -dfI

/;/

/

j NtATUP cytyt j.i

-glill I

l l ~! / dl

/

_i l 9

~

3

,ri i

1 / zy-

/

}

j i

/

/Wy/

}

j i

~~

g1000 ll

'j I

I

[la /, D,

,t i

Ij

/

7 M

l l1 J

1 i

If l l

1

'l catticAttiv uutt aut0 Ou

~~

500

! I letttylet m DtOSTAfic TES1 - --

I TDiPitATURI (283*f) FOR Tit -

$ER41CE PfR100 UP TO ? ffPY ---

8 O

i l

's 1

m 0

100 200 300 400 500 INDICATED TEMPERATURE (DEG. F)

FIGURE 3.4-2 REACTOR COOLANT SYSTEd HEATUP LIMITATIONS l

APPLICABLE UP TO 7 EFPY t

WLF CREEK - UNIT 1 3/4 4-30 Amendment No. 40 l

M.ATER!AL PROPERTY BASIS

/

Controlling Material:

RV Lower Shell Copper Content:

0.07 Wei ht s Nickel Content:

0.62 Welhtz initial RT 40 Deg NDT RT After 7 EFPY:

1 4T: 108.0 Dog F NDT 3 4T: 99.47 Dog F 3000 3;;,7);;;;igji);;]ij;jjgq;;i jjgi ij;;

;jj; g

i'#1 l'i ii

' ~-

4

'!;ii.iti:i ' iii i.

la lli,ii

{i.

• 2

.+i i l Itij_i lliI i

e

I 1 _ ji tila;!is'ti

• !jll t,i l!i4li ~ tI !,

i.

I I

! i d' ' i id l il i

Ill'I'I!'

'II'

' ! 'illil 1 v ill i l ' l i t t i l 8 ' tu'i!

2500 i!!'l ih i ';!I i;iji in

.t i;il; /4 ilitijifej '

t-s i'**:

i i '*

i

{'ii' I l

4i J l} * * !.i I [ lie l

I{* f 1'liliitil il 1l1!illtl^riti ill ilil ( ll l liff J 4 'I I'I I 'I'III' i ! ! 'l 'l M 'I i! i'! m $2000 I l iil' >lli llill! ii i .I' ll /il !!ili !a @i, 'll g i ,i' iil ii'l _ili' }illi il,ii/ i ti>; I, i a l l[I ll l !{ !ll 14llill l I 'I!' I i ls' Il l, l i/ ! F1 l lilil 1;_lj .i i II i i ' i ' ! 'Id - I l-liI! $ 1500 ' ! I I -- I I !i r i i 'If-1;ttil; a:

iit -

f P Tljili j i t i/ li ili i i ~ c; O \\\\liil>,- j di !Ii i;il ilill i Si jj i i ~-- ij , 7; g1000 coom um cum r, J i I ij i 1 1 ja op 3 i l i t i te ?f)W \\ \\ \\ \\ \\ 500 - 4e - - 9 se j - noe. 1 1 0 i 0 100 200 300 400 500 INDICATED TEMPERATURE (DEG. F) FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 7 EFPY 3/4 4-31 Am nt NO. 0 WOLF CREEK - UNIT 1

TABLE 4.4-5 (Deleted) WOLF CREEK - UNIT 1 3/4 4-32 Amendment No. 57 m -e ,r p --,r,---- g,e

b REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, oth e inherent plant characteristics, e.g., pump heat addition and preraurizer heater capacity, may 1 mit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges. 2. These limit lines shall be calculated periodically using methods provided below. 3. The secondary side of the steam generator must not be prescurized above 200 psig if the temperature of the steam generator 12 below 70*F. 4. The pressurizer heatup and cooldown rates shall not exceed 100'F/h and 200'F/h, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 583*F. 5. System preservice hydrotests and in-service leak and hy(cotests shall be performed at pressures in accordante with the requirements of ASME Boiler and Pressure Vessel Codt, Section XI. The fracture toughness properties of the ferritic materials in the ree.ctor vessel are determined in accordance with the 1972 Winter Addenda to Sectiun III cf the ASME Boiler and Pressure Vessel Code. Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNOT, at the end of 7 effective full power years (EFPY) of service life. The 7 EFPY service life per'od is chosen such that the limiting RT at the 1/4T location in the core region is NDT greater than the RT of the limiting unitradiated material.- The selection of NDT such a limiting RT assures that all components in the Reactor Coolant System NDT will be operated conservatively in accordance with applicable Cade requirements. The reactor vessel materials have been tested to determine their initial RTNDT; the rssults of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron.(E greater than 1 MeV) irradiation can cause an increase in the RT Therefore, an adjusted reference temperature. NDT. -based upon the fluence and copper content and nickel: content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ARTNDT computed by either Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this. shift in RTNDT at the end of 7 EFPY as well as adjustments for possible errors in the pressure and temperature sensing instruments. i WOLF CREEK - UNIT 1 B 3/4 4-7 Amendment No. 40

l l ELAN @ COOL A_NT SYSTEM BAsfs ILRESSURE/ TEMPERATURE LIMlU (Continued) Values of ART determined in this manner may be used until the results of the next schedu"Nd capsule from the material curveillance program, evaluated according to ASTM E185, are available. Captules will be removed in accordance with the requirements of ASTM fl85-73 and 10 CFR Part 50, Appendix H. The lead factor represents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict the future radiation damage to the reactor vessel material by using the lead far. tor and the withdrawal time of the capsule. The heatup determined from the and cooldown curves must be recalculated when the ART,3,he equivalent capsule surveillance capsule exceeds the calculated ART,33 for t radiation exposure. Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section 111 of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50. The general method for calculating heatup and cooldown limit curves is basea upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section 111 as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margin's for protection against nonductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RT,33, is used and this includes the radiation-induced shift, ART,3,, corresponding to the end of the period for which heatup and cooldown curves are generated. The ASME approach for calculating the allowable limit curves for various heatup and cocidewn rates specifies that the total stress intensity factor, K, for the combined thermal and pressure stresses at any time during heatup ior cooldown cannv. be greater than the reference stress intensity factor, Ka, for the metal temperature at that time. K is cbtained from the reference a fracture toughness curve, defined in Appendix G to the ASME Code. The Ka carve is given by the equation: Ka - 26.78 + 1.223 exp (0.0145(T-RT,3, + 160)] (1) WOLF CREEK - UNIT 1 B 3/4 4-8 Amendment No. 40, 57

j l BASIS Q B OVERPRESSVH (Continued) RCP eliminates the possibility of a 50*f differerce existing between indicated and actual RCS temperature as e result of heat transport effects. Considering instrument uncertainties only, an indicated RCS temperature of 350*F is suf ficiently high to allow full RCS pressurization in accordance with Appendix G limitations. Should an overpressure event occur in these conditions, the pressurizer safety valves provide acceptable and redundant overpressure protection. The Maximum Allowed PORV Setpoint for the Cold Overpressure Hitigation System will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required 'oj 10 CFR Part 50, Append'y H. }]L4.10 STRUCTVRAL INTEGRIJJ 1he inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these componsits will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i). Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler ind Pressure Yessel Code,1974 Eaition a',d Audenda through Summer 1975. 3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY of a reactor vessel head vent path ensures the capability .ts to perform this function. The valve redundancy of the Reactor Coolant System vent paths secves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure vent valve power supply or conti01 system does not prevent isolation of the vent path. The function, capabilities, and testing requirements of the Reactor Coolant System vants are consistent with the requirements of item II.B.1 of NUREG-0737, " Clarification of THI Action Plan Requirements," Nosember 1980. WOLF CREEK - UNIT 1 B 3/4 4-15 Amendment No. 40, 57 ) -}}