ML20129G009
| ML20129G009 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| 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) | |
Text
-
l 4
I Attachment I i
i i
Updated Pages for Proposed Change to the Technical Specifications Regarding Power Uprate (JPTS-91-025)
Request for Amendment Submitted Under JPN-92-028 1
with Updates Submitted Under JAFP-96-0306
)
i t
i l
J Panes l
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.
4 9610290277 961023 ppR ADOCK O 3
P
, 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 3
operable or de-activated in the isolated position.
1.
At least one door in each access opening is closed.
i 4.
All blind flanges and manways are closed.
2.
The Standby Gas Treatment System is operable.
j t
N.
Rated Power - Rated power refers to operation at a reactor 3.
All automatic ventilation system isolation valves are i
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
[
L 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:
O.
Reactor Power Operation - Reactor power operation is any Notations intervals Freauency t
operation with the Mode Switch in the Startup/ Hot 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.
5
~
m.
m.
m
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 Reactor Protection System (RPS) APRM flow biased trip settings of Table 3.1-1; 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).
- b. The analytical methods used to determine the core operating limits shall be 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
. - - -.. -. ~.. - -... - - _...
. ' l JAFNPP t
7.0 REFERENCES
(1)
E. Janssen, " Multi-Rod Burnout at Low Pressure," ASME Ps.per (11) Section 5.2 of the FSAR.
l 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.
l (14) Section 14.6 of the FSAR.
(4) FSAR Section 4.4.3.
(15) ASME Boiler and Pressure Vessel Code, Nuclear Vessels, (5) 1.M. Jacobs, " Reliability of Engineered Safety Features as a Section 111. Maximum allowable intemal pressure is 62 psig. *
~
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) " Nuclear Safety Program Annual Progress Report for Period Ending December 31,1966, ORNL-4071."
Systems Evaluation," (proprietary).
I i
l t
Amendment No. 190,227,234, 285 i
i
.