ML20141G772

Proposed Tech Specs Section 5.3.1 Incorporating Addl Info Into TS 6.9.1.9, Colr

ML20141G772
Person / Time
Site:	Wolf Creek
Issue date:	07/03/1997
From:	WOLF CREEK NUCLEAR OPERATING CORP.
To:
Shared Package
ML20141G767	List: ML20141G764 ML20141G772
References
NUDOCS 9707100332
Download: ML20141G772 (6)

Text

e

,, Attachment IV to ET 97-0067

~

,, Page 1 of 3 ,

a k

t ,

l  !

i

! i l C r

r i

l ATTACIDENT IV -i r

!. PROPOSED CURRENT TECHNICAL SPECIFICATION CHANGES i i

j i

i i

I -

i i

l i n

-1 l

l 1

l l

i l

i 4

)

i i' 9707100332 970703 ~~

PDR ADOCK 05000482

P PDR,

. _ _ _ ,. m _ , e r -- _, . ._.

F

. Attachment IV to ET 97-0067 l 'g * , Page 2 of 3 1

( , DESIGN FEATURES

• 5.3 REACTOR CORE l FUEL ASSEMBLIES e-b'y

'".^s',-"n

5.3.1 g,_ e n7 7"e cem.ce',e

m: r".,;2" gga smgc^^* 9^ #ma.s

-g, g._a 9? 'u9 ~~ e-b',:^'

3, ,uya ege7 ""'" ^ --"

. _amy 7.j s ..am.:mm ga gn g g 5 cug. ,g cem,i..:m; m, y, 7 myi a e. g.g:mim 7 7

C l ?de 9u!' ed by ? CyC'e !??'"#- '^'^?d I*ed 3mg7g ^'.by c -' ",-*".fe",g~2y s .g ,s h^ 3 mg_:,,g 7-.:,e s.,gi,_,m.s e,4ea ns.,.g i  : r'er T"e !-52' -^'e ' ecd: ; "9! 52ee e ?" u- ^^~M9 -"r!ch e^' ^' .

?M9 -9;"? pe ce-!" ??S of^rd '"9159! 5e d-" ta phydc9 d^d;- te

'"e ! " r' c^ e '^~d:^; The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or ZlRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in l l taccordance with approved applicatior.3 of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC approved codes and methods and i shown by test or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies -

that have not completed representative testing may be F. laced in non-limiting core regions.

CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. All control rod assemblies shall be hafnium, silver-indium-cadmium, or a mixture of both types. All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

i a. In accordance with the Code requirements specified in Section 5.2 of ,

the FSAR, with allowance for normal degradation pursuant to the j applicable Surveillance Requirements, 1

b. For a pressure of 2485 psig, and l

c. For a temperature of 650 F, except for the pressurizer which is  !

680 F.  !

VOLUME i 5.4.2 The total volume of the Reactor Coolant System, including pressurizer and surge lino, is 12,135 + 100 cubic feet at a nominal T, of 557*F.

5.5 METEOROLOGICAL TOWER LOCATION l

l

! 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

l l

WOLF CREEK - UNIT 1 5-6 Amendment No. ?, 49 49 '9.61 i

)

I

! ,., Attachment IV to ET 97-0067

y. Page 3 of 3 ADMINISTRATIVE CONTROLS

• CORE OPERATING LIMITS REPORT (COLR) (Continued) l Nuclear Enthalpy Rise Hot Channel Factor - F[f : Specification 3.9.1.b - Refueling Boron Concentration).

j) NRC Safety Evaluation Reports dated July 1,1991," Acceptance for Referencing of Topical Report WCAP 12610 ' VANTAGE + Fuel Assembly Reference Core Report' (TAC NO. 77258)," and September 15,1994, " Acceptance for Referencing of Topical Report WCAP-12610, Appendix B, Addendum 1, l ' Extended Burnup Fuel Design Methodology and ZlRLO Fuel Performance l Models' (TAC No. M86416)" (WCAP-12610-P-A) l (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor-F a(Z) i The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-hydraulic limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or i supplements thereto, shall be provided upon issuance, for each reload cyc;e, I to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.

S_PECIAL REPORTS '

G.9.2 Special reports shall be submitted to the Regional Administrator of the 'l NRC Regional Office within the tima period specified for each report. j i

6.10 RECORD RETENTION l in addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least 5 years:

a. Records and logs of unit operation acveg time interval at each power level;

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety; j

c. All REPORTABLE EVENTS; j

d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications-

e. Records of changes made to the procedures requireo by Specification

! 6.8.1; i

~

f. Records of radioactive shipments;

g. Records of sealed source and fission detector leak tests and results; and

h. Records of annual physicalinventory of all sealed source material of record.

WOLF CREEK- UNIT 1 6-21b Amendment No. 64, 92

• , Attachment V to ET 97-0067

• • Page 1 of 3 l

r 4

4 1

l ATTACID4ENT V PROPOSED IMPROVED TECHNICAL SPECIFICATION CHANGES l

i r

1 1

i l

I i

i l

J 1

I l

l

(

l 1

i

Y Attachment V to ET 97-0067 Y, Page 2 of 3 R porting Requirements 5.6  !

. l 5.6 Reporting Requi;ements l 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

7. NRC Safety Evaluation Report dated November 13, 1986, for j "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code" (WCAP-10266 P A, Rev. 2).

8. NRC Safety Evaluation Report dated May 17, 1988,

" Acceptance for Referencing of Westinghouse Topical Report WCAP 11596 - Qualificatian of the Phoenix P/ANC Nuclear Design System for Pressurized Water Reactor Cores."

9. NRC Safety Evaluation Report dated June 23, 1986, l

" Acceptance for Referencing of Topical Report WCAP 10%5 P I and WCAP 10966-NP-ANC: A Westinghouse Advanced Nodal Computer Code." i

c. The core operating limits shall be determined such that all l applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits Emergency Core Cooling Systems (ECCS) i limits, nuclear limits such as SDM, transient analysis limits. l and accident analysis limits) of the safety analysis are met, j

d. The COLR, including any midcycle revisions or supplements, shall i be provided upon issuance for each reload cycle to the NRC.  !

5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Specification 3.4.3, "RCS Pressure and Temperature (P/T)

Limits", and

2. Specification 3.4.12. " Low Temperature Overpressure l Protection System".

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and (continued) i WCGS ITS 5.0 30 5/15/97

e

.\ Attachment V to ET 97-0067 l'. Page 3 of 3 INSERT A

10. NRC Safety Evaluation Reports dated July 1, 1991, " Acceptance for Referencing of Topical Report WCAP-12610 ' VANTAGE + Fuel Assembly Reference Core Report' (TAC NO. 77258)," and September 15, 1994, " Acceptance for Referencing of Topical Report WCAP-12610, Appendix B, Addendum 1, ' Extended Burnup Fuel Design Methodology and ZIRLO Fuel Performance Models' (TAC No.

M86416)" (WCAP-12610-P-A).

i l

l J