ML061020354

Tech Spec Pages for Correction of Amendment No. 107, Regarding Removal of Requirement to Perform End-of-Life Moderator Temperature Coefficient Measurement (Tac No. MC6566)

ML061020354
Person / Time
Site:	Seabrook
Issue date:	04/13/2006
From:	Plant Licensing Branch III-2
To:
References
TAC MC6566
Download: ML061020354 (7)

Text

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REACTIVITY CONTROL SYSTEMS BORATION CONTROL MODERATOR TEMPERATURE COEFFICIENT SURVEILLANCE PEQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a. The MTC shall be measured and compared to the BOL limit specified In the COLR, prior to Initial operation above 5% of RATED THERMAL POWER, after each fuel loading; and

b. The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified Inthe COLR (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm*. In the event this comparison Indicates the MTC Is more I negative than the 300 ppm surveillance limit specified in the COLR, the MTC shall be remeasured, and compared to the EOL MTC limit specified In the COLR, at least once per 14 EFPD during the remainder of the fuel cycle.

• Measurement of the MTC in accordance with Surveillance Requirement 4.1L3.b may be suspended provided that the benchmark eriteiia in WCAP-I 3749-P-A and theRevised Piediction specified in the COLR are satisfied.

SEABROOK- UNIT I 3/4 1-5 Amendment No. X, 107

ADMINISTRATIVE CONTROLS 6.8.1.6.b (Continued)

'12. NYN-95048, Letter from T. C. Feigenbaum (NAESCo) to NRC, "License Amendment Request 95-05: Positive Moderator Temperature Coefficient",

May 30, 1995. -

Methodology for Specification:

3.1.1.3 - Moderator Temperature Coefficient

13. WCAP-1 261 0-P-A, "VANTAGE + Fuel Assembly Reference Core Report".

April, 1995, (Westinghouse Proprietary).

Methodology for Specification:

3.2.2 - Heat Flux Hot Channel Factor

14. WCAP-1 0216-P-A, Revision IA (Proprietary), "Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification", February, 1994.

Methodology for Specification:

3.2.1 - AXIAL FLUX DIFFERENCE 3.2.2 - Heat Flux Hot Channel Factor

15. WCAP-9272-P-A, (Proprietary), "Westinghouse Reload Safety Evaluation Methodology", July, 1985.

Methodology for Specifications:

2.1 - Safety Limits 3.1.1.1 - SHUTDOWN MARGIN for MODES 1,2,3, and 4 3.1.1.2 - SHUTDOWN MARGIN for MODE 5 3.1.1.3 - Moderator Temperature Coefficient 3.1.2.7 - Isolation of Unborated Water Sources - Shutdown 3.1.3.5 - Shutdown Rod Insertion Limit 3.1.3.6 - Control Rod Insertion Limits 3.2.1 - AXIAL FLUX DIFFERENCE 3.2.2 - Heat Flux Hot Channel Factor 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor 3.2.5 - DNB Parameters 3.5.1.1 - Accumulators for MODES 1, 2, and 3 3.5.4 - Refueling Water Storage Tank for MODES 1, 2, 3, and 4 3.9.1 - Boron Concentration

16. WCAP-1 3749-P-A, (Proprietary) "Safety Evaluation Supporting the Conditional Exemption of the Most Negative Moderator Temperature Coefficient Measurement," March, 1997.

Methodology for Specifications:

3.1.1.3 - Moderator Temperature Coefficient SEABROOK - UNIT 1 6-18 Amendment No. 0, 22, 34, 66, 96,44, 107

ADMINi!STRATIVE CONTROLS 6.8.1.6.c The core operating limits shall be determined so that all applicable limits (e.g.,

fuel thermal-mechanical 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 for each reload cycle, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector.

SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator within the time period specified for each report.

6.9 (THIS SPECIFICATION NUMBER IS NOT USED) 6.10 RADIATION PROTECTION PROGRAM 6.10.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.

6.11 HIGH RADIATION AREA 6.11.1 Pursuant to paragraph 20.1601(c) of 10 CFR Part 20, in lieu of the "control device" or "alarm signal" required by paragraph 20.1601 (a) and (b), each high radiation area, as defined in 10 CFR Part 20, in which the intensity of radiation is equal to or less-than 1000 mR/h at 30 cm (12 in.) from the radiation source orfrom any surface that the radiation penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel continuously escorted by such individuals may be exempt from the RWF' issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 mR/h, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area; or

b. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or SEABROOK - UNIT I 6-19 Amendment No. 88&4G4, 107

ADMINISTRATIVE CONTROLS HIGH FADIATION AREA 6.11.1 (Continued)

C. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the Radiation Work Permit.

6.11.2 In addition to the requirements of Specification 6.1 1.1, areas accessible to personnel with radiation levels greater than 1000 mR/h at 30 cm (12 in.) from the radiation source or from any surface that the radiation penetrates shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Manager on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

For individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mR~h that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device.

6.12 PROCESS CONTROL PROGRAM (PCP)

Changes to the PCP:

a. Shall be documented and records of reviews performed shall be retained as required by the Operational Quality Assurance Program (OQAP). This documentation shall contain:

1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

2) A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

b. Shall become effective after review and acceptance by the SORC and approval of the Station Director.

SEABROOK - UNIT 1 6-20 Amendment No. 66-8y44, 107

ADMINISTRATIVE CONTROLS 6.13 OFFSITE DOSE CALCULATION MANUAL (ODCM)

Changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained as required by the Operational Quality Control Program (OQAP). This documentation shall contain:

1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I.to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

1). Shall become effective after review and acceptance by the SORC and the approval of the Station Director.

c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and each affected page shall indicate the revision number the change was implemented.

6.14 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEMS*

6.14.1 Licensee-initiated major changes to the Radwaste Treatment Systems (liquid, gaseous, and solid):

a. Shall be reported to the Commission in the Annual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by tile SORC. The discussion of each change shall contain:

1) A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;

2) Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

• Licensees may choose to submit the information called for in this Specification as pair of the FSAR update, pursuant to 10 CFR 50.71.

SEABRO)OK - UNIT 1 6-21 Amendment No. 22, 66, 8",404,107

ADMINISTRATIVE CONTROLS 6.14.1 (Continued)

3) A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems;

4) An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the License application and amendments thereto;

5) An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the License application and amendments thereto;

6) A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the change is to be made;

7) An estimate of the exposure to plant operating personnel as a result of the change; and

8) Documentation of the fact that the change was reviewed and found acceptable by the SORC.

b Shall become effective upon review and acceptance by the SORC.

6.15 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak Test Program, dated September 1995," as modified by the following exception:

3. NEI 94-01-1995, Section 9.2-3: The first ILRT performed after October 30, 1992 shall be performed no later than October 29, 2007.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 49.6 psig.

rhe maximum allowable containment leakage rate, La, at Pa, shall be 0.15% of primary containment air weight per day.

SEABROtOK - UNIT 1 6-22 Amendment No. 22, 66, 8&,404,107

ADMINISTRATIVE CONTROLS CONTAINMENT LEAKAGE RATE TESTING PROGRAM 6.15 (Continued)

The provisions of SR 4.0.2 do not apply to the test frequencies specified inthe Containment Leakage Rate Testing Program.

The provisions of SR 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

Containment leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests and < 0.75 La for Type A tests.

Overall air lock leakage rate acceptance criterion is < 0.05 L. when tested at 2 Pa.

Each containment 8-inch purge supply and exhaust isolation valve leakage rate acceptance criterion is < 0.01 La when tested at Pa.

SEABRO)OK - UNIT 1 6-23 Amendment No. o-h,64494,107