LR-N06-0061, Revised Technical Specification Bases for Arts/Mellla Implementation

Revised Technical Specification Bases for Arts/Mellla Implementation

ML060450521
Person / Time
Site:	Hope Creek
Issue date:	02/06/2006
From:	Benyak D Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LCR H04-01, LR-N06-0061
Download: ML060450521 (5)

Text

I PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 FEB 0 6 2006 0 PSEG NuclearLLC LR-N06-0061 LCR H04-01 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 REVISED TECHNICAL SPECIFICATION BASES ARTS/RIELLLA IMPLEMENTATION HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354

References:

1. LR-N04-0062, "Request: for License Amendment: ARTS/MELLLA Implementation," dated June 7, 2004

2. LR-N05-0032, "Supplement to Request for License Amendment:

ARTS/MELLLA Implementation," dated February 18, 2005

3. LR-N05-0214, "Response to Request for Additional Information:

Request for Change to Technical Specifications: ARTS/MELLLA Implementation," dated August 3, 2005 PSEG Nuclear LLC (PSEG) is incorporating clarifying information into the Technical Specification Bases in support of the license amendment request in Reference 1 to revise the Technical Specifications (TS) for the Hope Creek Generating Station to reflect an expanded operating domain resulting from implementation of Average Power Range Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA). The amendment request also includes changes in the methods used to evaluate annulus pressurization (AP) and jet loads resulting from the postulated recirculation suction line break (RSLB).

Information regarding the average power range monitor (APRM) flow biased scram, provided in References 1, 2 and 3, is being incorporated into TS Bases 2.2.1. The clarifying information being added to TS Bases 2.2.1 is consistent with the Hope Creek Updated Final Safety Analysis Report (UFSAR).

95-2168 REV. 7/99

Document Control Desk FEB 06 2006 LR-N06-0061 The revised marked up TS Bases page provided in Attachment 1 to this letter replaces the marked up page provided previously in Reference 1.

PSEG has determined that the information contained in this letter and attachment does not alter the conclusions reached in the 10CFR50.92 no significant hazards analysis previously submitted.

If you have any questions or require additional information, please contact Mr. Paul Duke at (856) 339-1466.

Sincerely, Darin M. Benyak Director - Regulatory Assurance Attachment C: Mr. S. Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. S. Bailey, Project Manager - Salem & Hope Creek U. S. Nuclear Regulatory Commission Mail Stop 08B1 Washington, DC 20555 USNRC Senior Resident Inspector - Hope Creek (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, New Jersey 08625

Attachment I LR-N06-0061 LCR H04-01 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATION BASES PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. NPF-57 are affected by this change request:

Technical Specification Page Bases 2.2.1 B 2-7

1%

LIMITING SAFETY SYSTEM SETTINGS I

BAtSES REACTOR PROTECTION SYSTEM4 INSTRMI~ENTATION SETPOINTS (Continued)

Averwae Power Range Monitor (Continued)

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to chanje power by a significant amount, the rate of power rise is very slow. Generally the heatt flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than ES of RATED THERMAL POWER per minute and the APR system would be more thain adequate to assure shutdown lbefore the power could exceed the Safety Limit.

The' 5L neutron flux trip remains active until the mode switch is placed in the Run position.

The APRI trip system is calibrated using heat balance data taken during steady state conditions. Fission chaubers provide the basic Input to the system and therefor the monitors respond directly and quickly to changes due to transient operation for the cas;e of the Fixed Neutron Flux-Upscale set-point; .e, for a power increase. the THERNAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel. For the Flow Biased Simulated Thermal

- Power-Upscale setpoint, a time constant of 6 t 0.6 seconds is introduced Into the flow biased APAt in order to simulate the fuel thermal transient characteristics. A more conservative maxiau. value is used for the flow bias;ed setpoint as shown in Table 2.2.1-2.1.<

The APRN setpoints were selected to provide adequate margin for the Safety Limits and yet a& p crating ma r a b educes the sbi of unn ces-sarv shutdo thiftl rofercnctd trp so l stb d/" y tne d formula/1n Spoeylieat10, 3.2.2 t order/to msa in the e argi clfic V 7n FF I5S s g&atr F ;nor ;ul to 64tP. T

3. Reactor Vessel Stem Dom Pressure-Hh High pressure in the nuclear system could cause a rupture to the nuclear system process. barrier resulting in the release of fission products. A pressure incnease while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flUx, counteracting the pnissure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximua allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power/flow conditions when the turbine control valve fast closure and turbina stop va1vt closure trip are bypassed. For a load rejection or turbine trip under these conditions, the transient .nalysis indicated an adequate margin to the thermal hydraulic limit.

HOPE CREEK 9 2-7

Attachment I LR-N06-0061 LCR H04-01 Insert TS Bases 2.2.1l Although it is part of the Hope Creek design configuration and Technical Specifications, the APRM flow-biased simulated thermal power scram is not credited in any Hope Creek safety licensing analyses.