ML17265A466

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Proposed Tech Specs 4.2.1,revising Description of Fuel Cladding Matl & Updating List of References Provided in TS 5.6.5 for COLR
ML17265A466
Person / Time
Site: Ginna Constellation icon.png
Issue date: 11/24/1998
From:
ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17265A463 List:
References
NUDOCS 9812070083
Download: ML17265A466 (16)
v • d • e


Text

Attachment IV R.E. Ginna Nuclear Power Plant Mark-up of Existing Ginna Station Technical Specifications Included pages:

4.0-1 5.0-20 5.0-21 98i2070083 98ii24 PDR ADQCK 05000244 P PGR

J Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The site for the R.E. Ginna Nuclear Power Plant is located on the south shore of Lake Ontario, approximately 16 miles east of Rochester, New York.

The exclusion area boundary distances from the plant shall be as follows:

Direction Distance m N (including offshore) 8000 NNE 8000 NE 8000, ENE 8000 E 747 ESE 640 SE 503 SSE 450 S 450 SSW 450 SW 503 WSW 915 W 945 WNW 701

'W 8000 NNW 8000 4.2 Reactor Core cH o<

4.2.1 Fuel Assemblies The reactor shall contain 121 fuel asse blies. Each assembly shall consist of a matrix of zircalloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO,) as fuel material. Limited substitutions of zircon~urn a o or stainless steel filler rods for fuel rods, in z.i rc~loyj Z %<~~> accor ance with approved applications of fuel rod configurations, may e use . Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved

~c" ~AN codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limite number of lead test assemblies that have not completed repr sentative testing may be placed in nonlimiting core. regions.

eye e sp~

(continued)

R.E. Ginna Nuclear Power Plant 4.0-1 Amendment No. 61

11 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC,. specifically those described in the following documents:"
1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

(Methodology for LCO 3. 1. 1, LCO 3. 1.3, LCO 3. 1.5, LCO 3. 1.6, LCO 3.2. 1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)

2. WCAP-9220-P-A, "Westinghouse ECCS Evaluation Model-1981 Version," Revision 1, February 1982.

(Hethodolog for LCO 3.2. 1.)

3. WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report," 'September 1974.

(Methodology for LCO 3.2.3.)

4. WCAP-8567-P-A, "Improved Thermal Design Procedure;"

February 1989.

(Hethodology for LCO 3.4. 1 whe usin P.

5. WCAP 11397-P-A, "Revised Thermal Design Procedure,"

April 1989.

(Methodology for LCD 3.4. 1'hen using RTDP.)

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6. 'CAP-'10054-P-A and WCAP-1008+ "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code,"

August 1985.

(Methodology for LCO 3.2. 1) p

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~c.i >>. WCAP-10924-P-A, Volume 1, . 1, and Adden "Westinghouse Large-Break LOCA Best-Estimate 1,2,3, Methodology, Vo 1: Model Description and a > a so ," December 1988.

(Hethodology for LCO 3.2; 1)

C. t.l 8. WCAP-10924-P-A, Volume 2, . 2, and nda "Westinghouse Large-Break LOCA Best- stimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped with Upper Plenum Injection," December 8.

(Methodology for LCO 3.2. 1)

(continued)

R.E. Ginna Nuclear Power Plant 5.0-20 Amendment No. 61

eporting Requirements 5.6 5.6 Reporting, Requirements.

5:6.5 COLR (continue

9. WCAP-10924-P-A, Rev. 2 and WCAP-12071, "Westinghouse Large-Break LOCA Best Estimate Methodology, Volume 2:

Application to Two-Loop PWRs Equipped With Upper Plenum Injection, Addendum 1: Responses to NRC guestions,"

December 1988.

( Hethodolo g y for LCO 3.2.1)

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Revs'Sia ~

WCAP-10924-P olume 1, 1, Adden um 4, "We t hous OCA Best- stimate Methodology Model ascription and Validation Model Revisions," ~ugus 990 Pldrdk l1% I

( ethodology for LCO 3.2. 1)

C. The core operating limits shall be determined such that limits (e.g., fuel thermal mechanical limits, all'pplicable core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDH, transient analysis limits, and accident analysis limits) of the safety analysis are met.

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

5.6.6 Reactor Coolant S stem RCS PRESSURE AND TEMPERATURE LIMITS

~RT PL a ~ RCS pressure and temperature limits for heatup, cooldown, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits"

b. The power operated relief valve lift settings required to Overpressure Protection (LTOP) support the Low Temperature System, and the LTOP enable temperature shall be established arid documented in the PTLR for the following:

LCO 3.4.6, "RCS Loops - MODE 4";

LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled";

LCO 3.4. 10, "Pressurizer Safety Valves"; and LCO 3.4. 12, "LTOP System."

(continued)

R.E. Ginna Nuclear Power Plant 5.0-21 Amendment No. 61

I k

INSERT 1 WCAP-13677-P-A, "10 CFR 50.46 Evaluation Model Report:

WCOBRA/TRAC Two-Loop Upper Plenum Injection Model Updates to Support ZIRLO Cladding Option," February 1994 .

(Methodology for LCO 3.2.1).

INSERT 2 WCAP-12610-P-A, "VANTAGE + Fuel Assembly Reference Core Report,"

April 1995.

(Methodology for LCO 3.2. 1).

'1 f

Attachment V R.E. Ginna Nuclear Power Plant i

Proposed Ginna Station Technical Specifications

Design Features 4.0 4.0 DESIGN FEATURES

4. 1 Site Location The site for the R.E. Ginna Nuclear Power Plant is located on the south shore of Lake Ontario, approximately 16 miles east of Rochester, New York.

The exclusion area boundary distances from the plant shall be as follows:

Direction Distance m N (including offshore) 8000 NNE 8000 NE 8000 ENE 8000 E 747 ESE 640 SE 503 SSE 450 S 450 SSW 450 SW 503 WSW 915 W 945 WNW 701 NW 8000 NNW 8000 4.2 Reactor Core 4.2. 1 Fuel Assemblies The reactor shall contain 121 fuel assemblies. Each assembly shall consist of a matrix of zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO,) as fuel material. Limited substitutions of zircaloy, ZIRLO, or stainless steel filler rods for fuel rods, in accordance with NRC approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or cycle specific analyses to comply with all fuel safety design bases.

A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

I (continued)

R.E. Ginna Nuclear Power Plant 4.0-1 Amendment No. Q

4 eporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

(Methodology for LCO 3. 1. 1, LCO 3. 1.3, LCO 3. 1.5, LCO 3. 1.6, LCO 3.2. 1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)

2. WCAP-13677-P-A, "10 CFR 50.46 Evaluation Model Report:

WCOBRA/TRAC Two-Loop Upper Plenum Injection Model Updates to Support ZIRLO' Cladding Option,"

February 1994.

(Methodology for LCO 3.2. 1.)

3. WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report," September 1974.

(Hethodology for LCO 3.2.3.)

4. WCAP-12610-P-A, "VANTAGE + Fuel Assembly Reference Core Report," April 1995.

(Methodology for LCO 3.2. 1).

5. WCAP 11397-P-A, "Revised Thermal Design Procedure,"

April 1989.

(Methodology for LCO 3.4. 1 when using RTDP.)

6. WCAP-10054-P-A and WCAP-10081-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code,"

August 1985.

(Methodology for LCO 3.2. 1)

7. WCAP-10924-P-A, Volume 1, Revision 1, "Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1:

Model Description and Validation Responses to NRC guestions," and Addenda 1,2,3, December 1988.

(Hethodology for LCO 3.2. 1)

(continued)

R.E. Ginna Nuclear Power Plant 5.0-20 Amendment No. g

li r

I

eporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

8. WCAP-10924-P-A, Volume 2, Revision 2, "Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 2:

Application to Two-Loop PWRs Equipped with Upper Plenum Injection," and Addendum 1, December 1988.

(Methodology for LCO 3.2. 1)

9. WCAP-10924-P-A, Volume 1, Revision 1, Addendum 4, "Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1: Model Description and Validation, Addendum 4: Model Revisions," March 1991.

(Methodology for LCO 3.2. 1) c ~ The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDH, transient analysis limits, and accident analysis limits) of the safety analysis are met.

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

5.6.6 Reactor Coolant S stem RCS PRESSURE AND TEMPERATURE LIMITS REPORT PTLR a 0 RCS pressure and temperature limits for heatup, cooldown, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits"

b. The power operated relief valve lift settings required to support the Low Temperature Overpressure Protection (LTOP)

System, and the LTOP enable temperature shall be established and documented in the PTLR for the following:

LCO 3.4.6, "RCS Loops - MODE 4";

LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled";

LCO 3.4. 10, "Pressurizer Safety Valves"; and LCO 3.4. 12, "LTOP System."

(continued)

R.E. Ginna Nuclear Power Plant 5.0-21 Amendment No. g

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