ML17054C221

License Amendment Request for the Transition to Westinghouse Core Design and Safety Analyses - Revised Technical Specification Pages

ML17054C221
Person / Time
Site:	Wolf Creek
Issue date:	01/17/2017
From:	Wolf Creek
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17054C103	List: ET 17-0001, License Amendment Request for the Transition to Westinghouse Core Design and Safety Analyses, Enclosure 10 - CAW-16-4500 ML17053B393 ML17054C219 ML17054C220 ML17054C221 ML17054C222 ML17054C223 ML17054C224 ML17054C226 ML17054C227 ML17054C229 ML17054C230 ML17054C232 ML17054C233 ML17054C235 ML17054C236
References
ET 17-0001
Download: ML17054C221 (13)

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 the WRB-2 DNB correlation, and 1.13 for the ABB-NV DNB correlation, and 1.18 for the WLOP DNB.

2.1.1.2 The peak centerline temperature shall be maintained 5080 °F, decreasing by 58 °F 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. 123, 144,

RCS Boron Limitations < 500°F 3.1.9 3.1 REACTIVITY CONTROL SYSTEMS 3.1.9 RCS Boron Limitations < 500°F LCO 3.1.9 The boron concentration of the Reactor Coolant System (RCS) shall be greater than the all rods out (ARO) critical boron concentration.

APPLICABILITY: MODE 2 with keff < 1.0 with any RCS cold leg temperature < 500°F and with Rod Control System capable of rod withdrawal, MODE 3 with any RCS cold leg temperature < 500°F and with Rod Control System capable of rod withdrawal, MODES 4 and 5 with Rod Control System capable of rod withdrawal.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS boron concentration A.1 Initiate boration to restore Immediately not within limit. RCS boron concentration to within limit.

OR A.2 Initiate action to place the Immediately Rod Control System in a condition incapable of rod withdrawal.

OR A.3 --------------NOTE-------------

Not applicable in MODES 4 and 5.

Initiate action to increase Immediately all RCS cold leg temperatures to 500°F.

Wolf Creek - Unit 1 3.1-21 Amendment No.

RCS Boron Limitations < 500°F 3.1.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.9.1 Verify RCS boron concentration is greater than the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ARO critical boron concentration.

Wolf Creek - Unit 1 3.1-22 Amendment No.

RTS Instrumentation 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME T. One or more required T.1 Verify interlock is in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> channel(s) or train required state for existing inoperable. unit conditions.

OR T.2 Be in MODE 2. 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> U. One trip mechanism U.1 Restore inoperable trip 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperable for one RTB. mechanism to OPERABLE status.

OR U.2 Be in MODE 3. 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> (continued)

Wolf Creek - Unit 1 3.3-9 Amendment No. 123,

RTS Instrumentation 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME V. One channel inoperable. ---------------------NOTE-------------------

The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels.

V.1 Place channel in trip. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR V.2.1 Be in MODE 2 with 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> keff < 1.0.

AND V.2.2.1 Initiate action to fully insert 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> all rods.

AND V.2.2.2 Initiate action to place the 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> Rod Control System in a condition incapable of rod withdrawal.

OR V.2.3 Initiate action to borate the 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> RCS to greater than all rods out (ARO) critical boron concentration.

(continued)

Wolf Creek - Unit 1 3.3-10 Amendment No.

RTS Instrumentation 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME W. One channel inoperable. ---------------------NOTE-------------------

The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels.

W.1 Place channel in trip. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> X. Required Action and X.1.1 Initiate action to fully insert Immediately associated Completion all rods.

Time of Condition W not met. AND OR X.1.2 Initiate action to place the Immediately Rod Control System in a Two or more channels condition incapable of rod inoperable. withdrawal.

OR X.2 Initiate action to borate the Immediately RCS to greater than all rods out (ARO) critical boron concentration.

SURVEILLANCE REQUIREMENTS

---

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

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (continued)

Wolf Creek - Unit 1 3.3-11 Amendment No. 123,

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 6)

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

1. Manual Reactor Trip 1,2 2 B SR 3.3.1.14 NA 3(b), 4(b), 5(b) 2 C SR 3.3.1.14 NA

2. Power Range Neutron Flux

a. High 1,2 4 D SR 3.3.1.1 112.3% RTP SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16

b. Low 1(c), 2(f) 4 V SR 3.3.1.1 28.3% RTP SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16 4 W, X SR 3.3.1.1 2(h), 3(i) 28.3% RTP SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16

3. Power Range Neutron Flux Rate

a. High Positive Rate 1,2 4 E SR 3.3.1.7 6.3% RTP SR 3.3.1.11 with time SR 3.3.1.16 constant 2 sec

b. High Negative 1,2 4 E SR 3.3.1.7 6.3% RTP with Rate SR 3.3.1.11 time constant SR 3.3.1.16 2 sec

4. Intermediate Range 1(c), 2(d) 2 F,G SR 3.3.1.1 35.3% RTP Neutron Flux SR 3.3.1.8 SR 3.3.1.11 (continued)

(a) The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

(b) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(c) Below the P-10 (Power Range Neutron Flux) interlock.

(d) Above the P-6 (Intermediate Range Neutron Flux) interlock.

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

(f) With Keff 1.0.

(h) With Keff < 1.0, and all RCS cold leg temperatures 500°F, and RCS boron concentration the all rods out (ARO) critical boron concentration, and Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(i) With all RCS cold leg temperatures 500°F, and RCS boron concentration the ARO critical boron concentration, and Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

Wolf Creek - Unit 1 3.3-17 Amendment No. 123, 131, 132, 165,

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 6)

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

5. Source Range Neutron 2(e) 2 I,J SR 3.3.1.1 1.6 E5 cps Flux SR 3.3.1.8 SR 3.3.1.11 3(b), 4(b), 5(b) 2 J,K SR 3.3.1.1 1.6 E5 cps SR 3.3.1.7 SR 3.3.1.11

6. Overtemperature T 1,2 4 E SR 3.3.1.1 Refer to Note 1 SR 3.3.1.3 (Page 3.3-19)

SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

7. Overpower T 1,2 4 E SR 3.3.1.1 Refer to SR 3.3.1.7 Note 2 SR 3.3.1.10 (Page SR 3.3.1.16 3.3-20)

8. Pressurizer Pressure

a. Low 1(g) 4 M SR 3.3.1.1 1930 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

b. High 1,2 4 E SR 3.3.1.1 2395 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

9. Pressurizer Water 1(g) 3 M SR 3.3.1.1 93.9% of Level - High SR 3.3.1.7 instrument span SR 3.3.1.10

10. Reactor Coolant Flow - 1(g) 3 per loop M SR 3.3.1.1 88.9% of Low SR 3.3.1.7 normalized flow SR 3.3.1.10 SR 3.3.1.16 (continued)

(a) The Allowable Value defines the Limiting Safety System Setting. See the Bases for the Trip Setpoints.

(b) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

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

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

Wolf Creek - Unit 1 3.3-18 Amendment No. 123, 140,

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;

b. RCS average temperature is less than or equal to the limit specified in the COLR; and

c. RCS total flow rate 361,200 gpm and greater than or equal to the limit specified in the COLR.

APPLICABILITY: MODE 1.

---

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

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. 123, 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.

SR 3.4.1.3 Verify RCS total flow rate is 361,200 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 ----------------------------NOTE-----------------------------------

Not required to be performed until 7 days after 95% RTP.

Verify by precision heat balance that RCS total flow 18 months rate is 361,200 gpm and greater than or equal to the limit specified in the COLR.

Wolf Creek - Unit 1 3.4-4 Amendment No. 123, 144,

MSSVs 3.7.1 Table 3.7.1-1 (page 1 of 1)

OPERABLE Main Steam Safety Valves versus Maximum Allowable Power NUMBER OF OPERABLE MSSVs PER MAXIMUM ALLOWABLE POWER STEAM GENERATOR (% RTP) 4 70 3 51 2 31 Wolf Creek - Unit 1 3.7-3 Amendment No. 123,

Reporting Requirements 5.6 5.6 Reporting Requirements 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 (FNH),

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 T and Overpower T Trip Setpoints,

10. Specification 3.4.1: Reactor Coolant System pressure, temperature, and flow DNB limits, and

11. Specification 2.1.1: 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. WCAP-11397-P-A, Revised Thermal Design Procedure.

2. WCAP-10216-P-A, Relaxation of Constant Axial Offset Control -

FQ Surveillance Technical Specification.

3. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology.

(continued)

Wolf Creek - Unit 1 5.0-25 Amendment No. 123, 142, 144, 158, 159, 164, 179,

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

4. WCAP-16009-P-A, Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM).

5. WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON.

6. WCAP-16045-P-A, Addendum 1-A, Qualification of the NEXUS Nuclear Data Methodology.

7. WCAP 10965-P-A, ANC: A Westinghouse Advanced Nodal Computer Code.

8. WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core Report.

9. WCAP-8745-P-A, Design Bases for the Thermal Power T and Thermal Overtemperature T 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.

(continued)

Wolf Creek - Unit 1 5.0-26 Amendment No. 123, 142, 144, 158, 164, 179, 209, 213,