ML021210033

Technical Specifications for Amendment No. 207, Deletes TS Figures 2-1A (Reactor Coolant System (RCS) - Temperature Limits for Heatup) & 2-1B (RCS Pressure Temperature Limits for Cooldown) & Replaces with Figure 2-1

ML021210033
Person / Time
Site:	Fort Calhoun
Issue date:	04/22/2002
From:	Wang A Office of Nuclear Reactor Regulation
To:	Ridenoure R Omaha Public Power District
References
Download: ML021210033 (14)

Text

TECHNICAL SPECIFICATIONS - FIGURES TABLE OF CONTENTS PAGE WHICH FIGURE DESCRIPTION FIGURE FOLLOWS 1-1 TMLP Safety Limits 4 Pump Operations ................................ 1-3 2-1 RCS Pressure-Temperature Limits for Heatup, Cooldown, and Inservice Test ........ 2-6c 2-lB Deleted 2-3 Deleted 2-12 Boric Acid Solubility in W ater ..................................... 2-19h 2-10 Spent Fuel Pool Region 2 Storage Criteria ............................. 2-39e 2-8 Flux Peaking Augmentation Factors .................................. 2-53 viii Amendment No. f 6,26,131 ,141,16 .,.7, 1772,188,192,97, 207

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

(5) DELETED (6) Both steam generators shall be filled above the low steam generator water level trip set point and available to remove decay heat whenever the average temperature of the reactor coolant is above 300'F. Each steam generator shall be demonstrated operable by performance of the inservice inspection program specified in Section 3.17 prior to exceeding a reactor coolant temperature of 300'F.

(7) Maximum reactor coolant system hydrostatic test pressure shall be 3125 psia. A maximum of 10 cycles of 3125 psia hydrostatic tests are allowed.

(8) Reactor coolant system leak and hydrostatic test shall be conducted within the limitations of Figure 2-1.

(9) Maximum secondary hydrostatic test pressure shall not exceed 1250 psia. A minimum measured temperature of 73 'F is required. Only 10 cycles are permitted.

(10) Maximum steam generator steam side leak test pressure shall not exceed 1000 psia.

A minimum measured temperature of 73 'F is required.

(11) If no reactor coolant pumps are operating, a non-operating reactor coolant pump shall not be started while TC is below 385 °F unless at least one of the following conditions is met:

2-2a Amendment No. 39,-56,66*71-,*,-36, 161,88,27,

2.0 LIMITING CONDITIONS FOR OPERATION 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.

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

Specification The reactor coolant pressure shall be limited during plant operation in accordance with Figure 2-1 and as follows:

(1) Allowable combinations of pressure and temperature (Tj)for a specific heatup rate shall be below and to the right of the applicable limit lines as shown on Figure 2-1.

(2) Allowable combinations of pressure and temperature (T,) for a specific cooldown rate shall be below and to the right of the applicable limit lines as shown on Figure 2-1.

(3) The heatup rate of the pressurizer shall not exceed 100'F in any one hour period.

(4) The cooldown rate of the pressurizer shall not exceed 200'F in any one hour period.

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

(a) Immediately initiate action to restore the temperature or pressure to within the limit.

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

(c) 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 />.

(6) Before the radiation exposure of the reactor vessel exceeds the exposure for which they apply, Figure 2-1 shall be updated in accordance with the following criteria and procedures:

2-3 Amendment No. 2,?4,-61,207

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

(a) Figure 2-1 is valid for a fast neutron (E Ž 1MeV) fluence of 2.15X10 19 n/cm 2 which corresponds to 40 EFPY.

(b) The limit line on the figure shall be updated for a new integrated power period as follows: the total integrated reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated fast neutron exposure (E Ž 1 MeV).

(c) The limit lines in Figure 2-1 shall be moved parallel to the temperature axis (horizontal) in the direction of increasing temperature a distance equivalent to the transition temperature shift during the period since the curves were last constructed. The boltup temperature limit line shall remain at 82°F as it is set by the RTNDT of the reactor vessel flange and not subject to fast neutron flux. The lowest service temperature shall remain at 164°F because components related to this temperature are also not subject to fast neutron flux.

(d) The Technical Specification 2.3(3) shall be revised each time the curves of Figure 2-1 are revised.

Basis All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor coolant system temperature and pressure changes."' These cyclic loads are introduced by normal unit load transients, reactor trips and startup and shutdown operation.

During unit startup and' shutdown, the rates of temperature and pressure changes are limited. The design number of cycles for heatup and cooldown is based upon allowable heatup/cooldown rates and cyclic operation.

2-4 Amendment No. 22,47,64,74,77,100,114, 16-i-,--97, 207

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

The maximum allowable reactor coolant system pressure at any temperature is based upon the stress limitations for brittle fracture considerations. These limitations are derived by using the rules contained in Section XI* 2) of the ASME Code including Appendix A and G, Westinghouse Electric Company/Combustion Engineering's P-T (W/CE's) limit curve methodology(") and the rules contained in 10 CFR 50, Appendix G, Fracture Toughness Requirements. This ASME Code assumes that a crack 10-11/16 inches long and 1-25/32 inches deep exists on the inner surface of the vessel. Furthermore, operating limits on pressure and temperature assure that the crack does not grow during heatups and cooldowns.

The reactor vessel beltline material consists of six plates. The nilductility transition temperature (TNDT) of each plate was established by drop weight tests. Charpy tests were then performed to determine at what temperature the plates exhibited 50 ft-lbs. absorbed energy and 35 mils lateral expansion for the longitudinal direction. NRC technical position MTEB-5-2 was used to establish a reference temperature for transverse direction (RTNDT) of -12'F.

The initial RTNDT value for the Fort Calhoun submerged arc vessel weldments was determined to be -56°F with a standard deviation of 17'F. By applying the shift prediction methodology of Regulatory Guide 1.99, Revision 2, a weld material adjusted reference temperature (RTNDT) was established at 10°F based on the mean value plus two standard deviations. The standard deviation was determined by using the root-mean-squares method to combine the margin of 28°F for uncertainty in the shift equation with the margin of 17'F for uncertainty in the initial RTNDT value.

Similar testing was not performed on all remaining material in the reactor coolant system.

However, sufficient impact testing was performed to meet appropriate design code requirements(3 ) and a conservative RTNDT of 50'F has been established.

As a result of fast neutron irradiation in the region of the core, there will be an increase in the TNDT with operation. The techniques used to predict the integrated fast neutron (E _> 1 MeV) fluxes of the reactor vessel are described in Reference 5 with the result that the integrated fast neutron flux (E Ž 1 MeV) is 1.73x10'9 n/cm 2 , including tolerance at the inside surface of the critical reactor vessel beltline weld material, over the 40 years design life of the vessel.

Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured transition shift for a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in calculated flux magnitude. The maximum exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calibrated azimuthal neutron flux variation. The maximum integrated fast neutron (E >1 MeV) exposure of the reactor vessel at the critical reactor vessel beltline location including tolerance is computed to be 1.73x10 19 n/cm 2 at the vessel inside surface for 40 years operation at 2-5 Amendment No. 22,4 ,,-,7,, 4,,

+97,207

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued) 1500 MWt and 85 % load factor. The predicted shift at this location at the 1/4t depth from the inner surface is projected to be 252°F, including margin, using the shift prediction equation of Regulatory Guide 1.99, Revision 2. The actual shift in TNDT will be re established periodically during the plant operation by testing of reactor vessel material samples which are irradiated cumulatively by securing them near the inside wall of the reactor vessel as described in Section 4.5.3 and Figure 4.5-1 of the USAR. To compensate for any increase in the TNDT caused by irradiation, limits on the pressure-temperature relationship are periodically changed to stay within the stress limits during heatup and cooldown. Analysis of the three removed irradiated reactor vessel surveillance specimens"8 9 and10),

" combined with weld chemical composition data and implementation of extreme low radial leakage core loading designs beginning in Cycle 14, indicate that the fluence at the end of 40 Effective Full Power Years (EFPY) at 1500 MWt will be 2.15x10 9 n/cm 2 on the inside surface of the reactor vessel. This results in a total shift of the RTNDT of 237.76°F, including margin, for the area of greatest sensitivity (weld metal) at the 1/4t location using Regulatory Guide 1.99, Revision 2, and a shift of 187.97°F at the 3/4t location. Operation through fuel Cycle 34 will result in less than 40 EFPY.

The limit lines in Figure 2-1 are based on Reference 2, Appendix G, W/CE's methodology for P-T limit curve generation", and ASME Code Case N-640 as discussed below.

Reference Stress Intensity Factor The reference stress intensity factor (KIR) used in the development of the limit lines in Figure 2-1 is based on ASME Code Case N-640. This Code case allows the use of K,,

(lower bound of static initiation critical stress intensity factor) and is an approved exemption by the NRC in accordance with 10 CFR 50.60(b). K1c is obtained from a reference fracture toughness curve for reactor pressure vessel low alloy steels as defined in Appendix A to Section XI of the ASME Code and is approximated by the following equation:

K = 33.20 + 20.734e[°° 2°cr-RTNDTl where, K = Crack initiation reference stress intensity factor, Ksi fin T = temperature at the postulated crack tip, 'F RTNDT = adjusted reference nil ductility temperature (also called ART) at the postulated crack tip, 'F For any instant during the postulated heatup or cooldown, K,, is calculated using the metal temperature at the tip of the flaw, as well as the value of ART at that flaw location.

2-6 Amendment No. 22,47,64,74-,:R 100, 114,121,161,19,7, 207

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

Regulatory Requirements The Reference 2, Appendix G equation relating KIM, KIT, and KIR is re-arranged as shown below to solve for the allowable pressure stress intensity factor, KlM, as a function of time with the calculated KIR and KIT values to determine the allowable pressure stress intesity factor and consequently the allowable pressure:

(1) For Service Level A and B operation:

KIM= KIR -KIT 2

and (2) For Hydrostatic and Test Conditions when the core is not critical and tests are performed at isothermal conditions (i.e., thermal stress intensity factor, KIT, = 0)

KIR KIM =

1.5 where, KiM = Allowable pressure stress intensity factor based on coolant temperature, Ksi Vrin KIR = Reference stress intensity factor based on coolant temperature, Ksi /im KIT = Thermal stress intensity factor based on coolant temperature, Ksi 1v Tn Calculational of P-Allowable To develop P-T limits, the reactor vessel (RV) beltline region is the only location that receives sufficient neutron fluence to substantially alter the fracture toughness of the RV material. Hence, the beltline region is the most limiting with respect to allowable pressure at any specific temperature. This reduction in fracture toughness is determined using an adjusted reference temperature (ART), which is calculated in accordance with Regulatory Guide 1.99 Revision 2. The allowable pressure is based on the highest ART. The RV beltline region is analyzed assuming a semi-elliptical surface flaw oriented in the axial direction with a depth of one quarter of the RV beltline thickness and an aspect ratio of one to six.

2-6a Amendment No. 207

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

This postulated flaw is analyzed at both the inside diameter location (referred to as the 1/4t location) and the outside diameter location (referred to as the 3/4t location) to assure the most limiting condition is achieved. KIT is determined in accordance with Reference 2, Appendix G-2214.3. The thermal stress intensity factors are determined from the calculated temperature profilh through the beltline wall using thermal influence coefficients specifically generated for this purpose. The method employed uses a polynomial fit of the temperature profile and superposition using influence coefficients to calculate KIT. The influence coefficients are dependent upon the geometrical parameters associated with the postulated defect, the geometry of the reactor vessel beltline region, and the assumed unit loading. These influence coefficients were calculated using a 2-dimensional finite element model of the reactor vessel. The influence coefficients were corrected for three dimensional effects using Reference 2, Appendix A procedures. The KIT and KiM are calculated at any time point in a transient using influence coefficients generated by applying unit loads on a finite element model of the reactor vessel beltline region. The influence coefficients are calculations of stress intensity factors at the 1/4t and 3/4t crack depth location under the following unit loads:

a. for KIM.P, pressure load of 1 ksi, ksi-Vin/ksi

b. for KIT.L, linear through-wall gradient with peak temperature of 1 'F, ksi- u/n/ -F

c. for KIT-Q, quadratic through-wall gradient with peak temperature of 1VF, ksi-V Tn/ `F

d. for KITC, cubic through-wall gradient with peak temperature of 1°F, ksi- Vi-n/ °F Each stress intensity factor is calculated using a standard quarter point element formulation at the respective crack tips. Since all calculations performed are linear, superposition is then used to scale and combine these influence coefficients as necessary to determine the stress intensity factor for a given temperature profile.

2-6b Amendment No. 207

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

In the case of KIT, the through-wall temperature profiles are then fit to the third order polynomial below:

T (x) = Co + CL (l-X/h) + CQ (1-X/h)2 + CC (1-X/h)3 where, T(x) = Temperature at radial location x from inside wall surface, OF Co,CL,CQ,Cc = Coefficients in polynomial fit, °F x = Distance through beltline wall,. in h = Beltline wall thickness, in The coefficients of this polynomial are then combined through the following equation to calculate KIT at the 1/4t and 3/4t locations.

KIT - Total = CL *KIT-L + CQ *KIT. + Cc *KIT-C To calculate the allowable pressure, P-Allowable, the resultant K]T.Total from above in conjunction with Equation (1) from Reference 2, Appendix G-2215, is described as follows.

for Normal Level A and B loads P-Allowable = KIR - Klr.Total 2*Klm-p for Hydrostatic and Test Conditions, where Isothermal conditions result in KIT.Total -- 0 P-Allowable = KIR 1.5*Klm-p The P-T limits developed using the method described above account for the temperature differential between the RV base metal and the reactor coolant bulk fluid temperature only. However, uncertainties for instrumentation error, elevation, and flow induced differential pressure differences between the RV beltline and pressurizer are accounted for as follows:

1) Temperature instrumentation uncertainty of 14'F is applied to the entire temperature range of Figure 2-1.

2-6c Amendment No. 207

TECHNICAL SPECIFICATIONS Figure 2-1 FORT CALHOUN STATION UNIT I COMPOSITE P/T LIMITS, 40 EFPY 3000 2800 2600 2400 2200

0. 2000 ul 1800 CL, 1600 W

UJ N

1400 1200 LI 9L 1000 800 600 400 200 0

0 100 200 300 400 500 600 Te INDICATED REACTOR COOLANT SYSTEM TEMPERATURE *F RCS Pressure-Temperature Omaha Public Power District Limits for Heatup, Cooldown, and Inservice Test Fort Calhoun Station-Unit No. I Ai...........N.  ?, 77,1 0, 114,,,64, 19,;,*,

, ,207

FIGURE 2-1B This Figure has been deleted.

Amendment No. "4 ,"

• 0,114,-1-5-,161,-1-9' 207

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Curves (Continued)

2) Pressure correction factors that account for the difference in pressure between the reactor vessel beltline and pressurizer pressure instrumentation due to elevation and RCP flow are as follows:

RCS Temperature < 210'F = 61 psi RCS Temperature > 210'F = 67 psi

3) Below 350'F, pressure instrumentation uncertainty is accounted for in the LTOP system setpoints. Above 350'F, a pressure instrumentation uncertainty of 50 psi is applied to Figure 2-1.

Lowest Service Temperature = 50'F + 100'F + 14°F = 164°F. As indicated previously, an RTNDT for all material with the exception of the reactor vessel beltline was established at 50'F. Reference 2,Section III, NB-2332 requires a lowest service temperature of RTNDT + 100°F for piping, pumps and valves.

Below this temperature a pressure of 20 percent of the system hydrostatic test pressure cannot be exceeded. Taking into account pressure correction factors for elevation and flow, this pressure is (.20)(3125) - 61 = 564 psia, where 61 psi is the pressure correction factor.

Boltup Temperature = 10°F + 14°F = 24°F. A conservative value of 82°F will be used and maintained. At pressure below 564 psia, a minimum vessel temperature must be maintained to comply with the manufacturer's specifications for tensioning the vessel head. This temperature is based on RTNDT methods. This temperature corresponds to the measured 10°F RTNDT of the reactor vessel flange, which is not subject to radiation damage, plus 14°F instrument error.

The temperature at which the heatup and cooldown rates change in Figure 2-1 reflects the point at which the most limiting heatup and cooldown rates with respect to the inlet temperature (Tc) change.

2-7 Amendment No. 22,47,64,74,100,

+6+,207

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)

References:

(1) USAR, Section 4.2.2 (2) ASME Boiler and Pressure Vessel Code (3) USAR, Section 4.2.4 (4) USAR, Section 3.4.6 (5) WCAP-15443, Revision 0, Fast Neutron Fluence Evaluation for the Fort Calhoun Unit 1 Reactor Pressure Vessel, July 2000.

(6) Technical Specification 2.3(3)

(7) Article IWB-5000, ASME Boiler and Pressure Vessel Code,Section XI (8) TR-O-MCM-001, Revision 1, Omaha Public Power District, Fort Calhoun Station Unit No. 1, Evaluation of Irradiated Capsule W-225, August 1980.

(9) TR-O-MCM-002, Omaha Public Power District, Fort Calhoun Station Unit No. 1, Evaluation of Irradiated Capsule W-265, March 1984.

(10) BAW-2226, Omaha Public Power District, Fort Calhoun Station Unit No. 1, Evaluation of Irradiated Capsule W-275, November 1994.

(11) Safety Evaluation of Topical Report CE NPSD-683, Revision 6, "Development of a RCS Pressure and Temperature Limits Report (PTLR) for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications" (TAC No.

MA9561).

2-7a Amendment No. 22,47,64,74, 100,

+6+i9?, 207

_.U LL.N1 I. LU.tl'h1U lNu., tUiX UIr'tl(,IIu.. N 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves Applicability Applies to the status of the pressurizer and main steam safety valves.

Objective To specify minimum requirements pertaining to the pressurizer and main steam safety valves.

Specifications To provide adequate overpressure protection for the reactor coolant system and steam system, the following safety valve requirements shall be met:

(1) The reactor shall not be made critical unless the two pressurizer safety valves are operable with their lift settings adjusted to ensure valve opening at 2485 psig +/-1 % and 2530 psig (2) Whenever there is fuel in the reactor, and the reactor vessel head is installed, a minimum of one operable safety valve shall be installed on the pressurizer. However, when in at least the cold shutdown condition, safety valve nozzles may be open to containment atmosphere during performance of safety valve tests or maintenance to satisfy this specification.

(3) At least four of the five Main Steam Safety Valves (MSSVs) associated with each steam generator shall be OPERABLE in MODES 1 and 2. Lift settings shall be at 985 psig

+3/-2%, 1000 psig +3/-2%, 1010 psig +3/-2%, 1025 psig +3/-2%, and 1035 psig

+3/-2%.("

a. With less than four of the five MSSVs associated with each steam generator OPERABLE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT SHUTDOWN within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(4) Two power-operated relief valves (PORVs) shall be operable during heatups and cooldowns when the RCS temperature is less than 515'F, and in Modes 4 and 5 whenever the head is on the reactor vessel and the RCS is not vented through a 0.94 square inch or larger vent, to prevent violation of the pressure-temperature limits designated by Figure 2-1.

a. With one PORV inoperable during heatups and cooldowns when the RCS temperature is less than 515'F, restore the inoperable PORV to operable within 7 days or be in cold shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

b. With both PORVs inoperable during heatups and cooldowns when the RCS temperature is less than 515'F, be in cold shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

c. With one PORV inoperable in Modes 4 or 5, within one hour ensure the pressurizer steam space is greater than 53 % volume (50.6% or less actual level) and restore the inoperable PORV to operable within 7 days. If adequate steam space cannot be established within one hour, then restore the inoperable PORV to operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the PORV cannot be restored in the required time, depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

2-15 Amendment No. 39,47,54i-, 46-*, i-,

i+-9,207