ML19326C231
| ML19326C231 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/28/1975 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML19326C229 | List: |
| References | |
| NUDOCS 8004220804 | |
| Download: ML19326C231 (2) | |
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loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations. The number of thermal and loading cycles used for design purposes are shown in Table 4-8.of the FSAR. The maximum unit heatup and cooldown rate of 100 F per hour satisfies stress limits for cyclic opera-tion. (2) The 200 psig pressure limit for the secondary side of the steam generator at a temperature less than 100 F satisfies stress levels for tem-peratures below the DU. (3) The plate material and welds in the core region of the reactor. vessel have been tested to verify conformity to specified re-quirements and a maximum NDTT value of 10 F has been detemined based on Charpy V-notch tests. The maximum NDTT value obtained for the steam generator shell material and welds was 40 F.
Figures 3.1.2-1 and 3.1.2-2 contain the limiting reactor coolant system pressure-temperature relationship for operation at DU(4) and below to assure that stress levels are low enough to preclude brittle fracture. These stress levels and their bases are defined in Section 4.3.3 of the FSAR.
As a result of fast neutron irradiation in the region of the core, there will be an increase in the NDTT with accumulated nuclear operation. The predicted i
maximum NDTT increase for the 40-year exposure is shown on Figure 4-10. (4)
The actual shift in NDTT will be determined periodically during plant operation by testing of irradiated vessel material samples located in this reactor vessel. (5) The results of the irradiated sample testing will be evaluated and compared to the design curve (Figure 4-11 of FSAR) being used to predict the increase in transition temperature.
.i The design glue for fast neutron (E > 1 Mev) exposure of the reactor vessel is 3.0 x 10 n/cm2sec at 2568 MWt rated power and an integrated exposure of 3.0 x 1019
~
n/cm2 for 40 years operation. (6) The calculated maximum values are 2.2 x 1010 n/cm2sec and 2.2 x 1019 n/cm2 integrated exposure for 40 years operation at 80 percent load. (4)
Figure 3.1.2-1 is based on the design value which is considerably higher than the calculated value. The DTT value for Figure 3.1.2-1 is br. sed on the projected NDU at the end of the first two years of operation.
During these two years, the energy output has been con-servatively estimated to be 1.7 x 10f thermal megawatt days which is equiva-lent to 655 days at 2568 MWt core power. The projected fast neutron exposure of the reactor vessel for the two years is 1.7 x 1018 n/cm2 which is based on the 1.7 x 106 thermal megawatt days and the design value for fast neutron exposure.
The actual shift in ND U will be established periodically during plant opera-tion by testing vessel material samples which are irradiated cumulatively by securing them near the inside wall of the vessel in the core area. To com-pensate for the increases in the ND U caused by irradiation, the limits on the pressure-temperature relationship are periodically changed to stay within the established stress limits during heatup and cooldown.
The NDTT shift and the magnitude of the thermal and pressure stresses are sen-sitive to integrated reactor power and'not to instantaneous power level.
Figures 3.1.2-1 and 3.1.2-2 are applicable to reactor core thermal ratings up to 2568 MWt.
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The pressure limit line on Figure 3.1.2-1 has been selected such that the reactor vessel stress resulting from internal pressure will not exceed 15 percent yield strengch considering the following:
A.
A 25 psi error in measured pressure.
B.
System pressure is measured in either loop.
C.
Maximum differential pressure between the point of system pressure measurement and reactor vessel inlet for all operating pump combinations.
For adequate conservatism, in lieu of portions of the Fracture Thoughness Testing Requirements of the proposed Appendix G to 10 CFR 50, a maximum pressure of 550 psig and a maximum heatup rate of 50 F/hr has been imposed below 275 F as shown on Figure 3.1.2-1.
The spray temperature difference restriction based on a stress analysis of the spray line nozzle is imposed to maintain the thermal stresses at the pressuri-zer spray line nozzle below the design limit. Temperature requirements for the steam generator correspond with the measured NDTT for the shell.
The heatup and cooldown rates stated in this specification are intended as the maximum changes in temperature in one direction in a one hour period. The actual temperature linear ramp rate may exceed the stated limits for a time period provided that the maximum total temperature difference does not exceed the limit and that a temperature hold is observed to prevent the total temperature difference from exceeding the limit for the one hour period.
REFERENCES (1) PSAR, Section 4.1.2.4 (2) ASME Boiler and Pressure Code,Section III, N-415 (3)
FSAR, Section 4.3.10.5 (4) FSAR, Section 4.3.3 (5)
PSAR, Section 4.4.5 (6) FSAR, Sections 4.1.2.8 and 4.3.3 d
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