ML091170368

Updated Tech Spec Pages, from Licensee, License Amendment Request Lic 08-0078, Administrative Revisions to the Technical Specifications to Correct Typographical Errors and Provide Clarification

ML091170368
Person / Time
Site:	Fort Calhoun
Issue date:	02/18/2009
From:	Edwards M Omaha Public Power District
To:	Lynnea Wilkins Division of Operating Reactor Licensing
Shared Package
ML091170346	List: ML091170332 ML091170368
References
TAC MD9186
Download: ML091170368 (10)

Text

LIC-08-0078 Page 1 Proposed Technical Specification Pages (Clean)

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System 2.1.1 Operable Components Applicability Applies to the operable status of the reactor coolant system components.

Objective To specify certain conditions of the reactor coolant system components.

Specifications Limiting conditions for operation are as follows:

(1) Reactor Critical All four (4) reactor coolant pumps shall be in operation.

Exceptions The limitations of this specification may be suspended during the performance of physics tests provided the power level is < 10-1% of rated power and the flow requirements of Table 2-11 No. 2 are met.

(2) Hot Shutdown or 350°F < Tcold < 515°F (a) The reactor coolant loops listed below shall be operable:

(i) Reactor coolant loop 1 and at least one associated reactor coolant pump.

(ii) Reactor coolant loop 2 and at least one associated reactor coolant pump.

(b) At least one of the above reactor coolant loops shall be in operation.

Exceptions All reactor coolant pumps may be de-energized for up to one hour provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 10°F below saturation temperature.

2.1 - Page 1 Amendment No. 46,47,56, 256

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Operable Components (Continued)

(12) Reactor Coolant System Pressure Isolation Valves (a) The integrity of all pressure isolation valves listed in Table 2-9 shall be demonstrated, except as specified in (b). Valve leakage shall not exceed the amounts indicated.

(b) In the event that the integrity of any pressure isolation valve specified in Table 2-9 cannot be demonstrated, reactor operation may continue, provided that at least two valves in each high pressure line having a nonfunctional valve are in and remain in the mode corresponding to the isolated condition. Manual valves shall be locked in the closed position; motor operated valves shall be placed in the closed position and power supply deenergized.

(c) If Specifications (a) and (b) above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Basis The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation and maintain DNBR above the minimum DNBR limit during all normal operations and anticipated transients.

When Specification 2.1.1(2) is applicable, the reactor coolant pumps (RCPs) are used to provide forced circulation heat removal during heatup and cooldown. Under these conditions, decay heat removal requirements are low enough that a single reactor coolant system (RCS) loop with one RCP is sufficient to remove core decay heat. However, two RCS loops are required to be OPERABLE to provide redundant paths for decay heat removal. Only one RCP needs to be OPERABLE to declare the associated RCS loop OPERABLE. Reactor coolant natural circulation is not normally used but is sufficient for core cooling. However, natural circulation does not provide turbulent flow conditions. Therefore, boron reduction in natural circulation is prohibited because mixing to obtain a homogeneous concentration in all portions of the RCS cannot be assured.

2.1 - Page 4 Amendment No. 56,70,77,92,161,188, 221 TSBC-05-002

TECHNICAL SPECIFICATIONS 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate Applicability Applies to the temperature change rates and pressure of the Reactor Coolant System (RCS).

Objective To specify limiting conditions of the reactor coolant system heatup and cooldown rates.

Specification The combination of RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR and as designated below:

a. Allowable combinations of pressure and temperature (Tc) for a specific heatup rate shall be below and to the right of the applicable limit lines as shown on the pressure and temperature (P-T) limit Figure(s) in the PTLR.

b. Allowable combinations of pressure and temperature (Tc) for a specific cooldown rate shall be below and to the right of the applicable limit lines as shown on the P-T limit Figure(s) in the PTLR.

c. The heatup rate of the pressurizer shall not exceed 100°F in any one hour period.

d. The cooldown rate of the pressurizer shall not exceed 200°F in any one hour period.

Required Actions (1) When any of the above limits are exceeded, the following corrective actions shall be taken:

1. Immediately initiate action to restore the temperature or pressure to within the limit.

2. Perform an analysis to determine the effects of the out of limit condition on the fracture toughness properties of the reactor coolant system.

3. Determine that the reactor coolant system remains acceptable for continued operation or be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

(2) Before the radiation exposure of the reactor vessel exceeds the exposure for which they apply, the P-T limit Figure(s) shown in the PTLR shall be updated in accordance with the following criteria and procedures:

2.1 - Page 8 Amendment No. 22,74,161, 207,221

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.4 Power Distribution Limits (Continued)

(c) When the linear heat rate is continuously monitored by the excore detectors, withdraw the full length CEA's beyond the long term insertion limits of Specification 2.10.2(7) and maintain the Axial Shape Index, YI within the limits of Limiting Condition for Operations for the Excore Monitoring of LHR Figure provided in the COLR. If the linear heat rate is exceeding its limits as determined by the Axial Shape Index, YI, being outside the limits of the Limiting Condition for Operation for Excore Monitoring of LHR Figure provided in the COLR:

(i) Restore the reactor power and Axial Shape Index, YI, to within the limits of the Limiting Condition for Operations for Excore Monitoring of LHR Figure provided in the COLR within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or (ii) Be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(d) When calibration of the ex-core detectors has not been accomplished within the previous 30 equivalent full power days, then:

(i) reduce the axial power distribution monitoring trip setpoints as shown in the Axial Power Distribution LSSS for 4 Pump Operation Figure provided in the COLR by 0.03 ASI units; and (ii) reduce the axial power distribution monitoring Limiting Condition for Operations (LCO for Excore Monitoring of LHR and LCO for DNB Monitoring Figures provided in the COLR) by 0.03 ASI units.

When calibration of the ex-core detectors has not been accomplished within the previous 200 equivalent full power days, the power shall be limited to less than that corresponding to 75% of the peak linear heat rate permitted by Specification 2.10.4(1).

(2) Total Integrated Radial Peaking Factor The calculated value of FRT defined by FRT = FR (1+Tq) shall be within the limit provided in the COLR. FR is determined from a power distribution map with no non-trippable CEA's inserted and with all full length CEA's at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump combination. The azimuthal tilt, Tq, is the measured value of Tq at the time FR is determined.

2.10 - Page 15 Amendment No. 5,20,32,43, 47,117,126,141, 167, 249

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.13 Limiting Safety System Settings, Reactor Protective System (continued)

TABLE 2-11 RPS LIMITING SAFETY SYSTEM SETTINGS No. Reactor Trip Trip Setpoints 1 High Power Level (A) 109.0% of Rated Power 4-Pump Operation 2 Low Reactor Coolant Flow (B)(F) 4-Pump Operation 95% of 4 Pump Flow 3 Low Steam Generator Water Level 31.2% of Scale 4 Low Steam Generator Pressure (C) 500 psia 5 High Pressurizer Pressure 2400 psia 6 Thermal Margin/Low Pressure (B)(F) 1750 psia to 2400 psia (depending on the reactor coolant temperature as shown in the Thermal Margin/Low Pressure 4 Pump Operation Figure provided in the COLR) 7 High Containment Pressure (D) 5 psig 8 Axial Power Distribution (E) (as shown in the Axial Power Distribution LSSS for 4 Pump Operation Figure provided in the COLR) 9 Steam Generator Differential Pressure 135 psid A Setpoint cannot be set greater than 10% above measured power whenever reactor power is greater than 10% of rated power.

-4 B May be bypassed below 10 % power.

C May be bypassed below 600 psia.

D Bypass allowed for containment leak test.

E Inhibited below 15% power.

-1 F For physics testing at power levels less than 10 % of rated power, the low reactor coolant flow and thermal margin/low pressure trips may be bypassed until their reset points are exceeded if automatic bypass removal

-1 of 10 % of rated power is operable.

2.13 - Page 5 Amendment No. 252

TECHNICAL SPECIFICATIONS TABLE 2-1 (continued)

ENGINEERED SAFETY FEATURES SYSTEM INITIATION INSTRUMENT SETTING LIMITS Functional Unit Channel Setting Limit

6. 4.16 KV Emergency Bus Low a. Loss of Voltage (2995.2 + 104, -20.8) volts } Trip Voltage 5.9 seconds(4) }

b. Degraded Voltage i) Bus 1A3 Side 3988.8 volts } Trip (4.8 +/- .5) seconds }

ii) Bus 1A4 Side 3990.6 volts } Trip (4.8 +/- .5) seconds }

7. Low Steam Generator Water Auxiliary Feedwater Actuation 28.2% of wide range tap span Level

8. High Steam Generator Delta Auxiliary Feedwater Actuation 119.7 psid Pressure (1) May be bypassed below 1700 psia and is automatically reinstated prior to exceeding 1700 psia.

(2) May be bypassed below 600 psia and is automatically reinstated prior to exceeding 600 psia.

(3) Simultaneous containment high pressure, pressurizer low/low pressure, and steam generator low pressure.

(4) Applicable for bus voltage 2995.2 - 20.8 volts only. (For voltage (2995.2 - 20.8) volts, time delay shall be > 5.9 seconds).

2.14 - Page 5 Amendment No. 41,65,86,154,159, 255

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.7 Not used.

5.8 Procedures 5.8.1 Written procedures and administrative policies shall be established, implemented and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, 1978;

b. The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33;

c. Fire Protection Program implementation; and

d. All programs specified in Specification 5.11 through 5.24.

5.8.2 Temporary changes to procedures of 5.8.1 above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant supervisory staff, at least one of whom holds a Senior Reactor Operator's License.

5.0 - Page 5 Amendment No. 9,19,38,84,99, 115,149,157,160,184,202,216, 228, 252

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.18 Process Control Program (PCP) (Continued)

a. Shall be documented and records of reviews performed shall be retained as required by the Quality Assurance Program. This documentation shall contain:

1. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

2. A determination that the change will maintain the overall conformance of the solidified waste program to existing requirements of federal, state, or other applicable regulations.

b. Shall become effective after the review and acceptance by the Plant Review Committee and the approval of the plant manager.

c. Temporary changes to the PCP may be made in accordance with Technical Specification 5.8.2.

d. Shall be submitted to the Nuclear Regulatory Commission in the form of a complete, legible copy of the entire PCP as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the PCP was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed and shall indicate the date (e.g., month/year) the change was implemented.

5.19 Containment Leakage Rate Testing Program

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program, dated September 1995," as modified by the following exceptions:

(1) If the Personnel Air Lock (PAL) is opened during periods when containment integrity is not required, the PAL door seals shall be tested at the end of such periods and the entire PAL shall be tested within 14 days after RCS temperature Tcold > 210°F.

(2) Type A tests may be deferred for penetrations of the steel pressure retaining boundary where the nominal diameter does not exceed one inch.

(3) Elapsed time between consecutive Type A tests used to determine performance shall be at least 24 months or refueling interval.

b. The containment design accident pressure (Pa) is 60 psig.

5.0 - Page 16 Amendment No. 152,185, 202, 220, 237

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.19 Containment Leakage Rate Testing Program (Continued)

c. The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.1% of containment air weight per day.

d. Leakage Rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is 1.0 La. During unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La Maximum Pathway Leakage Rate (MXPLR) for Type B and C tests and 0.75 La for Type A tests.

2. Personnel Air Lock testing acceptance criteria are:

(i) Overall Personnel Air Lock leakage is 0.1 La when tested at Pa.

(ii) For each PAL door, seal leakage rate is 0.01 La when pressurized to 5.0 psig.

e. Containment Purge Valve (PCV-742A/B/C/D) testing acceptance criterion is:

For each Containment Purge Valve, leakage rate is < 18,000 SCCM when tested at Pa.

f. If at any time when containment integrity is required and the total Type B and C measured leakage rate exceeds 0.60 La Minimum Pathway Leakage Rate (MNLPR), repairs shall be initiated immediately. If repairs and retesting fail to demonstrate conformance to this acceptance criteria within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, then containment shall be declared inoperable.

g. The provisions of Specification 3.0.1 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

h. The provisions of Specifications 3.0.4 and 3.0.5 are applicable to the Containment Leakage Rate Testing Program.

5.20 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license or

2. A change to the USAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

5.0 - Page 17 Amendment No. 185,202,215, 237