ML20129G009

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Proposed Tech Specs Re Page 134 Deleted Under Amend 236 & Remain Deleted
ML20129G009
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 10/23/1996
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20129F988 List:
References
NUDOCS 9610290277
Download: ML20129G009 (4)
v • d • e


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Attachment I  !

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Updated Pages for Proposed Change to the Technical l Specifications Regarding Power Uprate (JPTS-91-025)  ;

l Request for Amendment Submitted Under JPN-92-028 1 with Updates Submitted Under JAFP-96-0306  ;

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l 5 254c l 285

Please Note: Page 134 was deleted under Amendment 236 and should remain deleted.

, Changes to Page 134, submitted under the Power Uprate Submittal, are no j longer needed due to page deletion.

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9610290277 961023 3 ,

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JAFNPP , j 1.0 (cont'd) ,

opened to perform necessary operational activities. deficiency subject to regulatory review.

2. At least one door in each airlock is closed and S. Secondary Containment Intearity - Secondary containment sealed. integrity means that the reactor building is intact and the 1 following conditions are met:
3. All automatic containment isolation valves are operable or de-activated in the isolated position. 1. At least one door in each access opening is closed. .
4. All blind flanges and manways are closed. 2. The Standby Gas Treatment System is operable.

N. Rated Power - Rated power refers to operation at a reactor 3. All automatic ventilation system isolation valves are I power of 2,536 MWt. This is also termed 100 percent operable or secured in the isolated position. -

power and is the maximum power level authorized by the operating license. Rated steam flow, rated coolant flow, T. Surveillance Freauency Notations / Intervals rated nuclear system pressure, refer to the values of these '

parameters when the reactor is at rated power (Reference The surveillance frequency notations / intervals used in these ,

, 1). specifications are defined as follows:

Reactor Power Operation - Reactor power operation is any Notations intervals Freauency O.

operation with the Mode Switch in the Startup/ Hot t Standby or Run position with the reactor critical and above D Daily At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1 percent rated thermal power. W Weekly At least once per 7 days i M Monthly At least once per 31 days P. Reactor Vessel Pressure - Unless otherwise indicated, O Quarterly or At least once per 92 days t reactor vessel pressures listed in the Technical every 3 months Specifications are those measured by the reactor vessel SA Semiannually or At least once per 184 days steam space sensor, every 6 months A Annually or Yearly At least once per 366 days Q. Refuelina Outaae - Refueling outage is the period of time 18M 18 Months At least once per 18 months (550 between the shutdown of the unit prior to refueling and days) the startup of the Plant subsequent to that refueling. R Operating Cycle At least once per 24 months (731 days)

R. Safety Limits - The safety limits are limits within which S/U Prior to each reactor startup the reasonable maintenance of the fuel cladding integrity NA Not applicable and the reactor coolant system integrity are assured.

Violation of such a limit is cause for unit shutdown and j review by the Nuclear Regulatory Cenmission before resumption of unit operation. Or Wion beyond such a 1 Amendment No. 11,131,1SS,227,233.

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JAFNPP (A) ROUTINE REPORTS (Continued)

4. CORE OPERATING LIMITS REPORT
a. Core operating limits shall be established prior to startup from each reload cycle, or prior to any remaining portion of a reload cycle for the following:
  • The Average Planar Linear Heat Generation Rates (APLHGR) of Specification 3.5.H; i
  • The Minimum Critical Power Ratio (MCPR) and MCPR low flow adjustment factor, K,, of Specifications 3.1.B and 4.1.E;
  • The Linear Heat Generation Rate (LHGR) of Specification 3.5.l;
  • The flow biased APRM and Rod Block Monitor (RBM) rod block settings of Table 3.2-3; and
  • The Power / Flow Exclusion Region of Specification 3.5.J.

and shall be documented in the Core Operating Limits Report (COLR).

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b. The analytical methods used to determine the core operating limits shall be i those previously reviewed and approved by the NRC as described in:
1. " General Electric Standard Application for Reactor Fuel," NEDE-24011-P, latest approved version and amendments.
2. " James A. FitzPatrick Nuclear Power Plant SAFER /GESTR - LOCA Loss-of-Coolant Accident Analysis," NEDC-31317P, October,1986 including latest revision, errata and addenda. I
3. " Loss-of-Coo! ant Accident Analysis for James A. FitzPatrick Nuclear Power Plant," NEDO-21662-2, July,197'/ including latest errata and addenda.
4. "BWR Owners' Group Long-term Stability Solutions Licensing Methodology," NEDO-31960-A, June 1991.
5. "9WR Owners' Group Long-term Stability Solutions Licensing Methodology," NEDO-31960-A, Supplement 1, March 1992.

Amendment No.1S2, 23S, 254c

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7.0 REFERENCES

(1) E. Janssen, " Multi-Rod Burnout at Low Pressure," ASME Ps.per (11) Section 5.2 of the FSAR.

62-HT-26, August 1962.  ;

(12) TID 20583, " Leakage Characteristics of Steel Containment .

(2) K.M. Backer, "Bumout Conditions for Flow of Boiling Water in Vessel and the Analysis of Leakage Rate Determinations." .

Vertical Rod Clusters," AE-74 (Stockholm, Sweden), May 1962. (13) Regulatory Guide 1.163, " Performance-Based Containment - '

Leak-Test Program", dated September 1995.

(3) FSAR Section 11.2.2.

(14) Section 14.6 of the FSAR. .

(4) FSAR Section 4.4.3. ,

(15) ASME Boiler and Pressure Vessel Code, Nuclear Vessels, Section 111. Maximum allowable intemal pressure is 62 psig. *

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(5) 1.M. Jacobs, " Reliability of Engineered Safety Features as a Function of Testing Frequency," Nuclear Safety, Vol. 9, No. 4, ['

July-August 1968, pp 310-312. (16) 10 CFR Part 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B -

(6) Deleted Performance Based Requirements", Effective Date October 26, 1995 (7) 1.M. Jacobs and P.W. Mariott, APED Guidelines for .

I Determining Safe Test Intervals and Repair Times for (17) Deleted Engineered Safeguards - April 1969. '

(18) General Electric Report NEDC-32016P, " Power Uprate Safety (8) Bodega Bay Preliminary Hazards Report, Appendix 1, Docket Analysis for the James A. FitzPatrick Nuclear Power Plant," I 50-205, December 28,1962. December 1991 (proprietary).  !

(9) C.H. Robbins, " Tests of a Full Scale 1/48 Segment of the (19) James A. FitzPatrick Calculation JAF-CALC-RAD-OOOO8, Humbolt Bay Pressure Suppression Containment," " Radiological Consequences of Design Basis Accidents at ,

GEAP-3596, November 17,1960. James A. FitzPatrick," November 1991. ,

t (20) General Electric Report GE-NE-187-45-1191P, " Containment I (10) Ending

" Nuclear Safety December Program 31,1966, ORNL-4071." Annual Progress Report for Period Systems Evaluation," (proprietary).

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Amendment No. 190,227,234, 285 i

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