ML020930466
| ML020930466 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/28/2002 |
| From: | Donohew J NRC/NRR/DLPM/LPD4 |
| To: | |
| References | |
| TAC MB1638 | |
| Download: ML020930466 (11) | |
Text
SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:
2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained
_>1.23 for the WRB-2 DNB correlation, and > 1.30 for the W-3 DNB correlation.
2.1.1.2 The peak centerline temperature shall be maintained
- 5080 OF, decreasing by 58 OF per 10,000 MWD/MTU of burnup.
2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained < 2735 psig.
2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.2.2 If SL 2.1.2 is violated:
2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.
Wolf Creek - Unit 1 2.0-1 Amendment No. 2, 144
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 6)
Reactor Trip System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 1.3% of AT span.
AT (1 (1+ 'cs)
+ 1- S) +IT3sS (s+ T~SS) (] -K-4S[) -TI+K3(P-P)-f,(J)}
(1+T6S Where: AT is measured RCS AT, OF.
AT0 is the indicated AT at RTP, OF.
s is the Laplace transform operator, sec 1 .
T is the measured RCS average temperature, OF.
T' is the nominal Tavg at RTP, < *.
P is the measured pressurizer pressure, psig.
P' is the nominal RCS operating pressure _>* psig.
K1 - . K2 = */OF K3 = /psig t1 =
- sec T2 =
- sec 'U3 = sec
-r=* sec T5 =
- sec T6 =- *sec f,(AI) = * {* % + (qt- qb)} whenqt-qb< *%RTP 0% of RTP when * % RTP < qt - qb < * %RTP
- {(qt-qb)-*%}
when qt-qb>* % RTP where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.
The values denoted with
- are specified in the COLR.
Wolf Creek - Unit 1 3.3-19 Amendment No. 4-2-, 144
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 6)
Reactor Trip System Instrumentation Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 2.6% of AT span.
(1+/- r0s)1 < ATJ KT-K(
A(1I+T2 S)
+/-S
+T 3 S7TOK45 1Is
+T 7 S) (1+/- T6S)
K7FS)
T (1+
Ij' s-)T"
-S f J(A Where: AT is measured RCS AT, OF.
ATo is the indicated AT at RTP, OF.
s is the Laplace transform operator, sec 1 .
T is the measured RCS average temperature, °F.
T" is the indicated Tavg at RTP (Calibration temperature for AT instrumentation),
- O
°F.
K4=
- K5 = */OF for increasing Tavg K<6 = /F when T > T"
- /°F for decreasing Tavg */°F when T < T" 1: =* sec 2* sec 3=
- sec r6 =*sec T7 *sec f2 (Al) =
- The values denoted with
- are specified in the COLR.
Wolf Creek - Unit 1 3.3-20 Amendment No. 42-3, 144
RCS Pressure, Temperature and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:
- a. Pressurizer pressure is greater than or equal to the limit specified in the COLR;
- c. RCS total flow rate > 37.1 x 104 gpm and greater than or equal to the limit specified in the COLR.
APPLICABILITY: MODE 1.
-- -- N O T E -------------------------. .................
Pressurizer pressure limit does not apply during :
- a. THERMAL POWER ramp > 5% RTP per minute; or
- b. THERMAL POWER step> 10% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE ------.--- A.1 Restore RCS DNB 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Not applicable to RCS total parameter(s) to within flow rate. limit.
One or more RCS DNB parameters not within limits.
(continued)
Wolf Creek - Unit 1 3.4-1 Amendment No. 1-2-3, 144
RCS Pressure, Temperature and Flow DNB Limits 3.4.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. --------- NOTE----------- B.1.1 Restore RCS flow rate to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Required Action B.2 must within limits.
be completed whenever Condition B is entered. OR B.1.2.1 Reduce THERMAL 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> RCS flow rate not within POWER to < 50% RTP.
limits.
AND B.1.2.2 Reduce Power Range 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Neutron Flux - High trip setpoints to < 55% RTP.
AND B.1.2.3 Reduce THERMAL 74 hours8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br /> POWER to < 5% RTP.
AND (continued)
Wolf Creek - Unit 1 3.4-2 Amendment No. 123
RCS Pressure, Temperature and Flow DNB Limits 3.4.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 NOTE THERMAL POWER does not have to be reduced to comply with this Required Action.
Perform SR 3.4.1.3. Prior to THERMAL POWER exceeding 50% RTP AND Prior to THERMAL POWER exceeding 75% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER reaching
> 95% RTP C. Required Action and C.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is greater than or equal to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> the limit specified in the COLR.
(continued)
Wolf Creek - Unit 1 3.4-3 Amendment No. 1-2-3, 144
RCS Pressure, Temperature and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.1.2 Verify RCS average temperature is less than or equal 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to the limit specified in the COLR.
I SR 3.4.1.3 Verify RCS total flow rate is > 3.71 x 104 gpm and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> greater than or equal to the limit specified in the COLR.
SR 3.4.1.4 Not required to be performed until 7 days after
_>95% RTP.
Verify by precision heat balance that RCS total flow 18 months rate is > 37.1 x 104 gpm and greater than or equal to the limit specified in the COLR.
Wolf Creek - Unit 1 3.4-4 Amendment No. 41-2, 144
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.
The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.
5.6.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.
5.6.5 CORE OPERATING LIMITS REPORT (COLR)
- a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
- 1. Specification 3.1.3: Moderator Temperature Coefficient (MTC),
- 2. Specification 3.1.5: Shutdown Bank Insertion Limits,
- 3. Specification 3.1.6: Control Bank Insertion Limits,
- 4. Specification 3.2.3: Axial Flux Difference,
- 5. Specification 3.2.1: Heat Flux Hot Channel Factor, FQ(Z),
- 6. Specification 3.2.2: Nuclear Enthalpy Rise Hot Channel Factor (FXH),
- 7. Specification 3.9.1: Boron Concentration,
- 8. SHUTDOWN MARGIN for Specification 3.1.1 and 3.1.4, 3.1.5, 3.1.6, and 3.1.8,
- 9. Specification 3.3.1: Overtemperature AT and Overpower AT Trip Setpoints, (continued)
Wolf Creek - Unit 1 5.0-27 Amendment No. 4 23,442, 144
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 10. Specification 3.4.1: Reactor Coolant System pressure, temperature, and flow DNB limits, and
- 11. Specification 2.1.11: Reactor Core Safety Limits.
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1. WCNOC Topical Report TR 90-0025 W01, "Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating Station."
- 2. WCAP-1 1397-P-A, "Revised Thermal Design Procedure."
- 3. WCNOC Topical Report NSAG-006, "Transient Analysis Methodology for the Wolf Creek Generating Station."
- 4. WCAP-1 0216-P-A, "Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification."
- 5. WCNOC Topical Report NSAG-007, "Reload Safety Evaluation Methodology for the Wolf Creek Generating Station."
- 6. NRC Safety Evaluation Report dated March 30, 1993, for the "Revision to Technical Specification for Cycle 7."
- 7. WCAP-10266-P-A, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code."
- 8. WCAP-11596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores."
(continued)
Wolf Creek - Unit 1 5.0-28 Amendment No. 1-23,-4---2, 144
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 9. WCAP 10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code."
- 10. WCAP-1261 0-P-A, "VANTAGE+ Fuel Assembly Reference Core Report."
- 11. WCAP-8745-P-A, "Design Bases for the Thermal Power AT and Thermal Overtemperature AT Trip Functions."
- 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.
5.6.6 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, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
- 1. Specification 3.4.3, "RCS Pressure and Temperature (P/T) Limits,"
and
- 2. Specification 3.4.12, "Low Temperature Overpressure Protection System ."
- b. The analytical methods used to determine the RCS pressure and temperature and Cold Overpressure Mitigation System limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
(continued)
Wolf Creek - Unit 1 5.0-29 Amendment No. 123, 130,412,144
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)
- 1. NRC letter dated December 2, 1999, "Wolf Creek Generating Station, Acceptance for Referencing of Pressure Temperature Limits Report (TAC No. MA4572)," and
- 2. WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January, 1996.
- c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6.7 Not Used.
5.6.8 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.
5.6.9 Not Used.
5.6.10 Steam Generator Tube Inspection Report
- a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission.
- b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a report within 12 months following completion of the inspection. This Special Report shall include:
- 1) Number and extent of tubes inspected, (continued)
Wolf Creek - Unit 1 5.0-30 Amendment No. 23,,1*30, 142 1