ML041210256

From kanterella
Jump to navigation Jump to search

Technical Specifications for Amd 162 and 153, Respectively
ML041210256
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 04/28/2004
From:
Office of Nuclear Reactor Regulation
To:
Billings, S A, NRR/DLPM, 301-415-1485
Shared Package
ML041210016 List:
References
TAC MB8128, TAC MB8129
Download: ML041210256 (16)
v • d • e


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 > 1.17 for WRB-1 DNB correlation for OFA fuel.

2.1.1.2 The peak fuel centerline temperature shall be maintained

< 47000F.

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 I 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.

Prairie Island Units land 2 Unit 1 -Amendment No. 162 Unit 2 - Amendment No. 153 2.0-1

PHYSICS TESTS Exceptions-MODE 2 3.1.8 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 PHYSICS TESTS Exceptions-MODE 2 LCO 3.1.8 During the performance of PHYSICS TESTS, the requirements of LCO 3.1.3, "Isothermal Temperature Coefficient (ITC)";

LCO 3.1.4, "Rod Group Alignment Limits";

LCO 3.1.5, "Shutdown Bank Insertion Limits";

LCO 3.1.6, "Control Bank Insertion Limits"; and LCO 3.4.2, "RCS Minimum Temperature for Criticality" may be suspended and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6, and 16.e may be reduced to "3" required channels, provided:

I

a. RCS lowest loop average temperature is > 5351F;
b. SDM is within the limits provided in the COLR; and
c. THERMAL POWER is < 5% RTP.

APPLICABL-flY:

During PHYSICS TESTS initiated in MODE 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit.

A. 1 Initiate boration to restore 15 minutes SDM to within limit.

AND A.2 Suspend PHYSICS TESTS 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> exceptions.

Prairie Island Units 1 and 2 Unit I -Amendment No. 162 Unit 2 - Amendment No. 153 3.1.8-1

FQ(Z) 3.2.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued)

A.4 Perform SR 3.2.1.1 and Prior to increasing SR 3.2.1.2.

THERMAL POWER above the limit of Required Action A.1 B. ------

NOTE-----------

B.1 Reduce AFD limits > 1% for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after Required Action B.4 each 1% F ' (Z) exceeds each F I (Z) shall be completed t

d whenever this Condition is entered.

AND B.2 Reduce Power Range 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after FQ (Z) not within limits.

Neutron Flux-High trip each F I' (Z) setpoints > 1% for each 1%

determination that the maximum allowable power of the AFD limit is reduced.

AND B.3 Reduce Overpower AT trip 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after setpoints > 1% for each 1%

each FI (Z) that the maximum allowable Q

power of the AFD limit is determination reduced.

AND Prairie Island Units I and 2 Unit I - Amendment No. 162 Unit 2-Amendment No. 153 3.2.1-2

F Q (Z) 3.2.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued)

B.4 Perform SR 3.2.1.1 and Prior to SR 3.2.1.2.

increasing THERMAL POWER above the maximum allowable power of the AFD limits 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 not met.

Prairie Island Units 1 and 2 Unit I - Amendment No. 162 Unit 2 - Amendment No. 153 3.2.1-3

AFD 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCO 3.2.3 The AFD in % flux difference units shall be maintained within the limits specified in the COLR.

NOTE--------------------------------------

The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

APPLICABILlY:

MODE 1 with THERMAL POWER > 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits.

A.1 Reduce THERMAL 30 minutes POWER to < 50% RTP.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD within limits for each OPERABLE 7 days excore channel.

Prairie Island Units I and 2 Unit 1 - Amendment No. 162 Unit 2 - Amendment No. 153 3.2.3-1

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE

5.

Source Range Neutron Flux 2 (d) 2 H, I SR 3.3. 1.1 SR 3.3.1.8 SR3.3.1.11 SR 3.3.1.16

< 1.0E6 cps 3 (a), 4(a), 5 (a) 2 1,J SR3.3.1.1 SR 3.3.1.7 SR 3.3. 1.11 SR 3.3.1.16

< 1.OE6 cps

6. Overtemperature AT 1,2 4

E SR 3.3.1.1 SR 3.3.1.3 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.12 SR 3.3.1.16 Refer to Note I (Page 3.3.1-23)

7. Overpower AT 1,2 4

E SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.12 SR 3.3.1.16 Refer to Note 2 (Page 3.3.1-24)

I

8. Pressurizer Pressure
a.

Low l(e) 4 3

K SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 E

SR 3.3.1.1 SR 3.3.1.7 SR 3.3. 1.10

> 1845 psig

< 2400 psig

b.

High 1,2 (a)

WVith Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(d)

Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(e)

Above the P-7 (Low Power Reactor Trips Block) interlock.

Prairie Island Units 1 and 2 Unit I - Amendment No. 162 Unit 2 - Amendment No. 153 3.3.1-18

RTS Instrimentation 3.3.1 Table 3.3.1-1 (page 7 of 8)

Reactor Trip System Instrumentation Note I: Overtemperature AT The Overtemperature AT Function Allowable Value is defined by the following Trip Setpoint.

AT < ATo{K, -K 2(T -T'F)[ (1+

K3(P -P) - fl(A&I)}

Where: AT is measured Reactor Coolant System (RCS) AT, IF.

ATo is the indicated AT at RTP, IF.

s is the Laplace transform operator, sec '.

T is the measured RCS average temperature, IF.

T' is the nominal TaVg at RTP, = *'F.

P is the measured pressurizer pressure, psig P'is the nominal RCS operating pressure, =

  • psig K2 = */'F K3 = */psig T I =
  • sec T 2 =
  • sec f1(AI) = {* + (qt - qb)}

when q, - qb < *% RTP

  • { (qt - qb) -

when q, - qb > *% RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and q, + qb is the total THERMAL POWER in percent RTP.

  • As specified in the COLR.

Prairie Island Unit 1 - Amendment No. 162 Units 1 and 2 3.3.1-23 Unit 2-AmendmentNo. 153

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 8 of 8)

Reactor Trip System Instrumentation Note 2: Overmower AT The Overpower AT Function Allowable Value is defined by the following Trip Setpoint.

AT < ATo{K4 -K 5 1+3T _K6(T-T')

Where:

AT is measured RCS AT, IF.

ATo is the indicated AT at RTP, 'F.

s is the Laplace transform operator, sec'.

T is the measured RCS average temperature, "F.

T' is the nominal Tavg at RTP, = *'F.

I I

I K5 = */'F for increasing Tavg

= */'F for decreasing Tavg K6 = */'F when T > T'

= */'F when T < T' l.

3 =

  • sec I

I

  • As specified in the COLR.

Prairie Island Units 1 and 2 Unit 1 - Amendment No. 162 Unit 2-Amendment No. 153 3.3.1-24

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.3 Radioactive Effluent Report E-----------

NOT -----------------------------------------

A single submittal may be made for the plant. The submittal shall combine sections common to both units.

The Radioactive Effluent Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the objectives outlined in the ODCM 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:

TS 2.1.1, "Reactor Core SLs";

LCO 3.1.1, "SHUTDOWN MARGIN (SDM)";

LCO 3.1.3, "Isothermal Temperature Coefficient (ITC)";

LCO 3.1.5, "Shutdown Bank Insertion Limits";

LCO 3.1.6, "Control Bank Insertion Limits";

LCO 3.1.8, "PHYSICS TESTS Exceptions - MODE 2";

Prairie Island Unit I - Amendment No. 162 Units l and 2 5.0-34 Unit 2 - Amendment No. 153

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))";

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FH)";

LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";

LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation" Overtemperature AT and Overpower AT Parameter Values for Table 3.3.1-1; LCO 3.4.1, "RCS Pressure, Temperature, and Flow - Departure from Nucleate Boiling (DNB) Limits"; and 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, specifically those described in the following documents:

1. NSPNAD-8101-PA, "Qualification of Reactor Physics Methods for Application to PI Units" (latest approved version);
2. NSPNAD-8102-PA, "Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units" (latest approved version);
3.

NSPNAD-97002-PA, "Northern States Power Company's "Steam Line Break Methodology", (latest approved version);

4. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology";
5.

WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code";

6.

Deleted;

7.

WCAP-1 0924-P-A, "Westinghouse Large Break LOCA Best Estimate Methodology";

Prairie Island Unit 1 -Amendment No. 162 Units land 2 5.0-35 Unit 2-Amendment No. 153

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

8. XN-NF-77-57 (A), XN-NF-77-57, Supplement I (A), "Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II";
9.

WCAP-13677, "10 CFR 50.46 Evaluation Model Report:

W-COBRA/TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLOTm Cladding Options";

10. NSPNAD-93003-A, "Transient Power Distribution Methodology",

(latest approved version);

11. NAD-PI-003, "Prairie Island Nuclear Power Plant Required Shutdown Margin During Physics Tests";
12. NAD-PI-004, "Prairie Island Nuclear Power Plant F I (Z) Penalty With Increasing [Fc (Z) / K(Z)J Trend";
13. WCAP-10216-P-A, Revision IA, "Relaxation of Constant Axial Offset Control/ FQ Surveillance Technical Specification";
14. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions";
15. WCAP-1 1397-P-A, "Revised Thermal Design Procedure";
16. WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report";
17. WCAP-7588 Rev. 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods";

Prairie Island Unit 1 -Amendment No. 162 Units l and 2 5.0-36 Unit 2 - Amendment No. 153

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

18. WCAP-7908-A, "FACTRAN - A FORTRAN IV Code for Thermal Transients in a U02 Fuel Rod";
19. WCAP-7907-P-A,"LOFTRAN Code Description";
20. WCAP-7979-P-A, "TWINKLE - A Multidimensional Neutron Kinetics Computer Code";
21. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code";
22. WCAP-1 1394-P-A, "Methodology for the Analysis of the Dropped Rod Event";
23. WCAP-1 1596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores";
24. WCAP-12910 Rev. 1-A, "Pressurizer Safety Valve Set Pressure Shift";
25. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis"; and
26. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses".

Prairie Island Units 1 and 2 Unit I - Amendment No.162 Unit 2 - Amendment No. 153 5.0-37

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

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, and hydrostatic testing, OPPS arming, PORV lift settings and Safety Injection Pump Disable Temperature as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits";

LCO 3.4.6, "RCS Loops - MODE 4";

LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled";

LCO 3.4.10, "Pressurizer Safety Valves";

LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) -

Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature";

LCO 3.4.13, "Low Temperature Overpressure Protection (LTOP) -

Reactor Coolant System Cold Leg Temperature (RCSCLT) < Safety Injection (SI) Pump Disable Temperature"; and LCO 3.5.3, "ECCS - Shutdown".

Prairie Island Unit I -Amendment No. 162 Units I and 2 5.0-38 Unit 2-Amendment No. 153

Reporting Requirements 5.6 5.6 Reporting Requirenients 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

b.

The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (includes any exemption granted by NRC to ASME Code Case N-5 14).

c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Changes to the curves, setpoints, or parameters -in the PTLR resulting from new or additional analysis of beltline material properties shall be submitted to the NRC prior to issuance of an updated PTLR.

5.6.7 Steam Generator Tube Inspection Report

1. Following each in-service inspection of steam generator tubes, if there are any tubes requiring plugging or sleeving, the number of tubes plugged or sleeved in each steam generator shall be reported to the Commission within 15 days.
2.

The results of steam generator tube in-service inspections shall be included with the summary reports of ASME Code Section XI inspections submitted within 90 days of the end of each refueling outage. Results of steam generator tube in-service inspections not associated with a refueling outage shall be submitted within 90 days of the completion of the inspection. These reports shall include: (1) number and extent of tubes inspected, (2) location and percent of wall-thickness penetration for each indication of an imperfection, and (3) identification of tubes plugged or sleeved.

Prairie Island Unit 1 7Amendment No. 162 Units l and 2 5.0-39 Unit 2-Amendment No.153

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Steam Generator Tube Inspection Report (continued)

3. Results of steam generator tube inspections which fall into Category C-3 require notification to the Commission prior to resumption of plant operation, and reporting as a special report to the Commission within 30 days. This special report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
4.

The results of inspections performed under Specification 5.5.8.b for all tubes that have defects below the F* or EF* distance, and were not plugged, shall be reported to the Commission within 15 days following the inspection. The report shall include:

a.

Identification of F* and EF* tubes, and

b.

Location and extent of degradation.

5. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC staff prior to returning the steam generators to service should any of the following conditions arise:
a. If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steamline break) for the next operating cycle.
b. If circumferential crack-like indications are detected at the tube support plate intersections.
c.

If indications are identified that extend beyond the confines of the tube support plate.

d. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.

Prairie Island Unit I - Amendment No. 162 Units I and 2 5.0-40 Unit 2 - Amendment No. 153

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Steam Generator Tube Inspection Report (continued)

e.

If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds IE-02, notify the NRC and provide an assessment of the safety significance of the occurrence.

EM Report When a report is required by Condition C or J of LCO 3.3.3, "Event Monitoring (EM) 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.

5.6.8 Prairie Island Units I and 2 Unit 1 - Amendment No. 162 Unit 2 - Amendment No. 153 5.0-40a

Retrieved from "https://kanterella.com/w/index.php?title=ML041210256&oldid=5402274"