ML023370460
| ML023370460 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/29/2002 |
| From: | Spaulding D NRC/NRR/DLPM/LPD3 |
| To: | Cayia F Nuclear Management Co |
| Spaulding, D W, NRR/DLPM, 415-2928 | |
| Shared Package | |
| ML023370142 | List: |
| References | |
| TAC MB4956, TAC MB4957 | |
| Download: ML023370460 (6) | |
Text
Wisconsin Electric Power Company A.
Maximum Power Levels NMC is authorized to operate the facility at reactor core power levels not in excess of 1540 megawatts thermal.
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 207, are hereby incorporated in the license. NMC shall operate the facility in accordance with Technical Specifications.
C.
Deleted D.
Deleted E.
Spent Fuel Pool Modification The licensee* is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.
F.
NMC shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FFR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Point Beach Nuclear Plant Modified Amended Security Plan," with revisions submitted through March 23, 1988; "Point Beach Nuclear Plant Modified Amended Security Force Training and Qualification Plan," with revisions submitted through August 6, 1982; and "Point Beach Nuclear Plant Modified Amended Security Contingency Plan," with revisions submitted through March 6, 1981. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
Reference to the licensee in License Conditions 3.E, 3.G and 3.J refers to Wisconsin Electric Power Company and is maintained for historical purposes.
Amendment No. +97, 20-1-, 207 Point Beach Unit 1
Definitions 1.1 1.1 Definitions RATED THERMAL POWER (RTP)
SLAVE RELAY TEST STAGGERED TEST BASIS THERMAL POWER RTP shall be a total reactor core heat transfer rate to the reactor coolant of 1540 MWt.
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
- a.
All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation;
- b.
With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and
- c.
In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for OPERABILITY and verifying the OPERABILITY of each required -'"ve relay.
The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total channel steps.
A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
Point Beach 1.1-5 Unit 1 - Amendment No.-20-t, 207 Unit 2 - Amendment No.-26, 21 2 I
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 CORE OPERATING LIMITS REPORT (COLR). (continued)
(8) LCO 3.2.3, "Axial Flux Difference (AFD)"
(9) LCO 3.3.1, "Reactor Protection System (RPS) Instrumentation Overtemperature AT" (10) LCO 3.3.1, "Reactor Protection System (RPS) Instrumentation Overpower AT" (11) LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits" (12) LCO 3.9.1, "Boron Concentration"
- b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. When an initial assumed power level of 102 percent of the original rated thermal power is specified in a previously approved method, 100.6 percent of uprated rated thermal power may be used only when the main feedwater flow measurement (used as the input for reactor thermal output) is provided by the Caldon leading edge flowmeter (LEFM) as described in reports 11 and 12 listed below. When main feedwater flow measurements from the LEFM are unavailable, a power measurement uncertainty consistent with the instruments used shall be applied.
Future revisions of approved analytical methods listed in this Technical Specification that currently reference the original Appendix K uncertainty of 102 percent of the original rated thermal power should include the condition given above allowing use of 100.6 percent of uprated rated thermal power in the safety analysis methodology when the LEFM is used for main feedwater flow measurement.
The approved analytical methods are described in the following documents:
(1)
WCAP-14449-P-A, "Application of Best Estimate Large Break LOCA Methodology to Westinghouse PWR's with Upper Plenum Injection," Revision 1, October 1999. (cores containing 422V+ fuel)
(2)
WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.
(3)
WCAP-1 1397-P-A, "Revised Thermal Design Procedure," April 1989.
Point Beach 5.6-3 Unit 1 - Amendment No 20+,
207 Unit2-AmendmentNo
-206, 212
Reporting Requirements 56 5.6 Reporting Requirements 5.6.4 CORE OPERATING LIMITS REPORT (COLR)
(continued)
(4)
WCAP-14787-P, Rev. 1, "Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wisconsin Electric Power Company Point Beach Units 1 & 2 (Fuel Upgrade &
Uprate to 1656 MWt-NSSS Power with Feedwater Venturis, or 1679 MWt-NSSS Power with LEFM on Feedwater Header),
February, 2002.
(5)
WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using The NOTRUMP Code," August 1985.
(6)
WCAP-10054-P-A, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code:
Safety Injection into the Broken Loop and COS] Condensation Model," Addendum 2, Revision 1, July 1997.
(7)
WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions,"
September 1986.
(8)
WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control," Revision 1A, February 1994.
(9)
WCAP-10924-P-A, "Large Break LOCA Best Estimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped with Upper Plenum Injection," and Addenda, December 1988. (cores not containing 422 V+ fuel)
(10)
WCAP-10924-P-A, "LBLOCA Best Estimate Methodology:
Model Description and Validation: Model Revisions," Volume 1, Addendum 4, August 1990. (cores not containing 422 V+ fuel)
(11)
Caldon, Inc., Engineering Report-80P, "TOPICAL REPORT:
Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM-/TM System,"
Revision 0, March 1997.
(12)
Caldon, Inc., Engineering Report-1 60P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate With the LEFM /
TM System," Revision 0, May 2000.
- 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 SDM, 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.
Point Beach 5.6-4 Unit 1 - Amendment No. 20+/-, 207 Unit 2-Amendment No. q06.
212
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
- a.
RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, LTOP enabling, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
(1) LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits" (2) LCO 3.4.6, "RCS Loops-MODE 4" (3) LCO 3 4.7, "RCS Loops-MODE 5, Loops Filled" (4) LCO 3.4.10, "Pressurizer Safety Valves" (5) LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)"
b The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the NRC Letters dated October 6, 2000 and July 23, 2001.
- c.
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto 5 6.6 PAM Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.7 Tendon Surveillance Report Abnormal conditions observed during testing will be evaluated to determine the effect of such conditions on containment structural integrity. This evaluation should be completed within 30 days of the identification of the condition. Any condition which is determined in this evaluation to have a significant adverse effect on containment structural integrity will be considered an abnormal degradation of the containment structure.
Any abnormal degradation of the containment structure identified during the engineering evaluation of abnormal conditions shall be reported to the Point Beach 5.6-5 Unit 1 - Amendment No. -2E-207 Unit 2-Amendment No.2E6 212
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Tendon Surveillance Report (continued)
Nuclear Regulatory Commission pursuant to the requirements of 10 CFR 50.4 within thirty days of that determination. Other conditions that indicate possible effects on the integrity of two or more tendons shall be reportable in the same manner. Such reports shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedure and the corrective action taken.
5.6.8 Steam Generator Tube Inspection Report (a)
After each inservice examination, the number of tubes plugged or repaired in each steam generator shall be reported to the commission as soon as practicable.
(b)
The complete results of the steam generator tube inservice inspection shall be included in a report for the period in which the inspection was completed.
Reports shall include.
- 1.
Number and extent of tubes inspected.
- 2.
Location and percent of all thickness penetration for each indication
- 3.
Identification of tubes plugged or repaired.
(c)
Reports required by Table 5.5.8-1, "Steam Generator Tube Inspection," shall provide the information required by Specification 5.6.8 (b) and a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence. The report shall be submitted to the Commission prior to resumption of plant operation.
Point Beach 56-6 Unit 1 -Amendment No. 207 Unit 2 - Amendment No. 212