ML20094A577
| ML20094A577 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 07/30/1984 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20094A570 | List: |
| References | |
| NUDOCS 8408060008 | |
| Download: ML20094A577 (6) | |
Text
{{#Wiki_filter:F LIMITINGCdNDITIONFOROPERATION SURVEILLANCE REQUIREMENTS l 3.6 ' PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY Applicability: Applicability: l Applies to'the operating status of the Applies to the periodic examination and reactor coolant system. testing requirements for the reactor cooling system. Objective: Objective: To assure the integrity and safe oper- To determine the condition of the reactor ation of the reactor coolant system. coolant system and the operation of the safety devices related to it. Specification: Specification: 1 A. Thermal and Pressurization A. Thermal and Pressurization Limitations Limitations
- 1. The average rate of reactor coolant 1. During heatups and cooldowns, the temperature change during normal following temperatures shall be heatup or cooldown shall not exceed permanently logged at least every 100*F/hr when averaged over a one- 15 minutes until the difference hour period except when the vessel between any two readings taken over temperatures are above 450*F. a 45 minute period is less than 5*F.
The shell flange to shell tempera-ture differential shall not exceed a. Reactor vessel shell adjacent 145'F. to shell flange
- b. Reactor vessel shell flange
- c. Recirculation loops A and B
- 2. The reactor vessel shall not be 2. Reactor vessel shell temperature and pressurized for hydrostatic and/or reactor coolant pressure shall be
. leakage tests, and critical core permanently logged at least every operation shall not be conducted 15 minutes whenever the shell tem-unless the reactor vessel temperature perature is below 220*F and the is above that defined by the reactor vessel is not vented.
appropriate curves on Figures 3.6.1 and 3.6.2. In the event this Test specimens of the reactor vessel 4 requirement is not met, achieve base, weld and heat affected zone stable reactor conditions with reactor metal subjected to the highest flu-vessel temperature above that defined ence of greater than 1 Mev neutrons by the appropriate curve and obtain shall be installed in the reactor an engineering evaluation to determine vessel adjacent to the vessel wall the appropriate course of action at the core midplane level. The to take. specimens and sample program shall conform to the requirements of ASTM E 185-66. Selected h$h p o PM 3 Amendment No. 123
LIMITING CON ITION FOR OPERATION SURVEfLLANCE REQUIREMENTS 3.6.A Thermal and Pressurization 4.6.A Thermal and Pressurization Limitations (Cont'd) Limitations (Cont'd) neutron flux specimens shall be
. removed at the frequency required by Table 4.6.2 at.d tested to experimentally verify adjustments to Figures 3.6.1 and 3.6.2 for predicted NDTT irradiation shifts.
- 3. The reactor vessel head bolting 3. When t'Te reactor vessel head bolt-studs shall not be under tension ing studs are tensioned and the unless the temperature of the reactor is in a Cold Condition, vessel head flange and the head the reactor vessel shell l 1s greater than 55*F. temperature immediately below the head flange shall be permanently recorded.
- 4. The pump in an idle recirculation 4. Prior to and during startup of an loop shall not be started unless idle recirculation loop, the tem-the temperatures of the coolant perature of the reactor coolant within the idle and operating re- in the operating and idle loops circulation loops are within 50*F shall be permanently logged.
of each other.
- 5. The reactor recirculation pumps 5. Prior to starting a recirculation shall not be started unless the pump, the reactor coolant temper-coolant temperatures between the atures in the dome and in the dome and the bottom head drain bottom head drain shall be
- are within 145*F. compared and permanently logged.
- 6. Thermal-Hydraulic Stability Core thermal power shall not exceed a 257. of rated thermal power without forced recirculation.
- 8. Coolant Chemistry B. Coolant Chemistry
- 1. The reactor coolant system radio- 1. a. A reactor coolant sample shall activity concentration in water be taken at least every 96 shall not exceed 20 microcuries hours and analyzed for radio-of total lodine per ml of water. activity content.
- b. Isotopic analysis of a reactor coolant sample shall be made at least once per month.
- 2. The reactor coolant water shall 2. During startups and at steaming not exceed the following limits rates less than 100,000 pounds per with steaming rates less than hour, a sample of reactor coo 1 ant 100,000 pounds per hour, except shall be taken every four hours l as specified in 3.6.8.3: and analyzed for chlorine content.
Conductivity .. 2 pmho/cm Chloride ion .. 0.1 ppm Amendment No. 124 1
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' TABLE 4.6.2 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM HITHDRAWAL SCHEDULE Effective Full Fluence 2
RT,, o r Capsule Power Years (n/cm ) (weld metal)
-Number (EFPY) (1/4 T) (*F)
L g 1 4.17- 1.8 x 10 55
- 2. 15 6.3 x 10 91
. (approx.) (approx.)
. - 3 End of. Life 1.4 x 10 136 (approx.)
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. Amendment No.. 124A J
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FIGURE 3. 6.2 PILGRIM REACTOR VESSEL PRESSURE- TEMPERATURE LIMITS SUBCRITICAL/ CRITICAL HEAT UP S CbOL DOWN I i l
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AMENDMENT N O. 128A i
h .
. Bases:
3.6.A and 4.6.A Thermal and Pressurization Limitations (Cont'd) The reactor coolant system is a primary barrier against the release of fission
- products-to the environs. In order to provide assurance that this barrier is maintained at a high degree of integrity, restrictions have been placed on the operating conditions to which it can be subjected.
Appendix G to 10CFR50 defines the temperature-pressurization etstrictions for hydrostatic and leak tests, pressurization, and critical operation. These e limits have been calculated for Pilgrim and are contained in Figures 3.6.1 and 3.6.2.
-For Pilgrim pressure-temperaturo restrictions, two locations in the reactor vessel are limiting. The-closure region controls at lower pressures and the beltilne controls at higher pressures.
The nil-ductility transition (NDT) temperature is defined as the temperature below which ferritic steel breaks in a brittle rather than ductile manner. s Radle.tlon exposure from fast neutrons (>1 mev) above about 10 nyt may L shift the NOT temperature of the vessel metal above the initial value. Impact tests from the first material surveillance capsule removed from the reactor vessel have established the magnitude of the RTuor shift for the beltilne. The shift, which is. greatest for the weld metal, is tabulated below for various
' . fluence levels and EFPY of operation:
RPV Wall EFPY Fluence (1/4T) RTuor 1, 6.68* :2.8 x 10 n/cm' 61*F
'8.0 3.4 x 10 n/cm2 68'F d
10.0 4.4 x 10 n/cm 2 76*F
'12.0 5.1 x 10 n/cm2 82*F 14.3 6.1 x 10 n/cm" 90*F
*PNPS Effective Full Power Years (EFPY) as of end of Cycle 6 (12/83) i Neutron flux wires and samples of vessel material are installed in the reactor vessel adjacent to the vessel wall at the core midplane level. The wires and samples will be periodically removed and tested to experimentally verify the values used for Figures 3.6.1 and 3.6.2. The withdrawal schedule of Table
-4.6.2 has been established as required by ICCFR50, Appendix H.
The pressure-temperature limitations of Figures 3.6.1 and 3.6.2 applicable to the beltline reflect an. Initial RTuor of O'F. This initial value is based Amendment No. 139
Bases: 3.6.A and 4.6.A Thermal and Pressurization Limitations (Cont'd) on unirradiated test data adjusted for specimen orientation in accordance with USNRC Branch Technical Position MTEB 5-2.
' The pressure-temperature limitations of Figures 3.6.1 and 3.6.2 applicable to the closure region reflect an RTuor of -5'F, also based on test data adjusted for specimen orientation. The curves apply to 100% bolt preload condition, but are conservative for lesser bolt preload conditions.
, For critical core operation when the water level is within the normal range for power operation and the pressure is less than 20% of the preservice system hydrostatic test pressure (313 psi), the minimum permissible temperature of
< the highly stressed regions of the closure flange is RTuor + 60 - 55'F.
The closure region is more limiting than the feedwater nozzle with regards to a both stress intensity and RTuor. Therefore the pressure-temperature limits of the closure are controlling. s i i-f i 139A Amendment,,No. L}}