ML19325E947
| ML19325E947 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/31/1989 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML19325E944 | List: |
| References | |
| NUDOCS 8911130084 | |
| Download: ML19325E947 (11) | |
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INDIK f DEFINITIONS SECTION PAE 1.0 DEFINITIONS (Continued)
WASTE GAS SYSTEM 1-6 VENTILATION EXHAUST TREATMENT SYSTEM l-7 PURGE PURGING 1-7 VENTING 1-7 INDEPENDENT VERIFICATION 1-7 LIQUID RADWASTE TREATMENT SYSTEM l-7 HEMBER(S) 0F THE PUBLIC 1-8 SITE BOVNDARY l-8 UNRESTRICTED AREA 1-8 PRESSVRE/ TEMPERATURE LIMITS REPORT l-8 OPERATIONAL NODES (TABLE 1.1) ,
1-9 FREQUENCY NOTATION (TABLE 1.2) 1-10 e
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! CRYSTAL RIVER - UNIT 3 la Amendment No.
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,, ..- l DEFINITIONS ,
i MEMBERf S) 0F THE PUBLIC j l
1.37 MEMBER (S) 0F THE PUBLIC shall include all individuals who by virtue of i their occupational status have no formal association with the plant. This i category shall include non-employees of the licensee who are permitted to use ;
portions of the site for recreational, occupational, or other purposes not ,
associated with plant functions. This category shall agi include non-employees '
l such as vending machine servicemen or postmen who, as part of their normal function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive ;
materials.
SITE BOUNDARY 1.38 The SITE B0VNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee.
UNRESTRICTED AREA 1.39 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY, access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or industrial, commercial, :
institutional, and/or recreational purposes.
PRESSURE / TEMPERATURE llMITS REPORT (PTLR) 1.40 The PRESSURE /TEHFERATURE LIMITS REPORT is the unit-specific document that I provides the reactor vessel pressure and temperature limits including heatup and cooldown rates for the current reactor vessel fluence period. These pressure -
and temperature limits shall be determined for each fluence period in accordance with Specification 6.9.1.7. Plant operation within these operating limits is addressed in Specification 3.4.9.1. ,
CRYSTAL RIVER - UNIT 3 1-8 Amendment No.
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, REACTOR COOLANT SYSIDi 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM ,
l LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressuriter) temperature and l pressure shall be within the acceptable limits provided in the j PRESSURE / TEMPERATURE LIMITS REPORT during heatup, cooldown, j criticality, and inservice leak and hydrostatic testing, i APPLICARILITY: At all times.
ACTION:
mth any of the acceptable limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out of-limit condition on the fracture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or 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 /> and reduce RCS T and pressure to less than 200'Fand500psig,respectively,withinthefo11Ning30 hours.
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l CRYSTAL RIVER UNIT 3 3/4 4-24 Amendment No.
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l9 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4,4.9.1.1 The Reactor Coolant System temperature and pressure shall be detemined to be within limits as specified in the PRESSURE / TEMPERA 1URE LIMITS REPORT at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
CRYSTAL RIVER - UNIT 3 3/4 4-24 Amendment No.
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CRYSTAL RIVEP,- UNIT 3 3/4 4-26 Amendment No.
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CRYSTAL RIVER - UNIT 3 3/4 4-27 Amendment No. '
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CRYSTAL RIVER - UN!T 3 3/4 4-28 Amendment No. !
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ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORT 6.9.1.6 Routine reports of operating statistics and shutdown experience shall i be submitted on a monthly basis to the Document Control Desk, U.S.
Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Office, submitted no later than the 15th of each month following the calendar month covered by the report.
PRESSURE / TEMPERATURE LIMITS REPORT 6.9.1.7 The PRESSURE / TEMPERATURE LIMITS REPORT is the unit-specific document that 3rovides the reactor vessel pressure and temperature limits i inclucing heatup and cooldown rates for the current reactor vessel i fluence period. Plant operation within these operating limits is addressed in specification 3.4.9.1 (Pressure / Temperature Limits).
Pressure and temperature limits shall be established and documented '
in the PRESSURE / TEMPERATURE LIMITS REPORT before each reactor vessel fluence period or any remaining part of a reactor vessel fluence period for specification 3.4.9.1, Pressure / Temperature Limits. The analytical i methods used to determine the pressure and temperature limits shall be those previously reviewed and approved by NRC in Topical Report B&W- ;
10046A, Rev. 2,
- Methods of Compliance with Fracture Toughness and Operational Requirements of 10CFR50, Apper. dix G " June 1986. The pressure / temperature limits shall be determined so that all applicable limits of the safety analysis are met. The PRESSURE / TEMPERATURE LIMITS REPORT, shall be provided upon issuance, for each reactor vessel fluence period to the NRC Document Control Desk with copies to the >
Region 61 Administrator and Resident Inspector.
Changes to this report shall become effective after review and :
acceptance by the PRC, and the approval of the Director, Nuclear Plant i Operations. .
CRYSTAL klVER - UNIT 3 6 15 Amendment No.
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BASES ._
The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower buund curve similar to that described for the heatup of the inner wall cannot be defined. Cunsequently, for the cases in which the outer wall of the vessel ;
becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.
The heatup limit curve, as shown in the PRESSVRE/ TEMPERATURE LIMITS REPORT (PTLR) is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 100'F per hour. During cooldown, similar types of thermal stress occur.
Thus, the cooldown limit curve in the PTLR is also a composite curve which was prepared based upon the same type analysis as the heatup curve with the exception l that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing comprehensive stresses at the outside wall. Additionally, during cooldown and heatup at the higher temperatures, the most conservative limits are imposed by thermal and loading cycles on the steam generator tubes. These limits are the vertical segments of the limit lines shown in the PTLR. (These limits will not require adjustments due to the neutron fluences.)
During the first several years of service life, the most limiting Reactor Coolant System regions are the closure head region (due to mechanical loads resulting from bolt pre load) and the reactor vessel outlet nozzles. Nozzle sensitivity is caused by the high local stresses at the inside corner of the nozzle which i can be two to three times the membrane stresses of the shell. After the first several years of neutron radiation exposure, the beltline region of the reactor vessel becomes the most limiting region due to material irradiation. 1 For the service period for which the limit curves are established, the pressure / temperature limits were obtained through a point by point comparison of the limits imposed by the closure head region, outlet nozzles, and the most sensitive material in the beltline region. The lowest pressure calculated for these three regions becomes the maximum allowable pressure for the fluid temperature used in the calculation. The calculated pressure / temperature curves are adjusted by 25 PSI and 10*F for possible instrument errors. The pressure limit is also adjusted for the pressure differential between the point of pressure measurement and the limiting component for all combinations of reactor coolant pump operations.
CRYSTAL RIVER - UNIT 3 B 3/4 4 10 Amendment No.
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. . i Irradiation damage to the beltline region can be quantified by determining the l decrease in the temperature at which the metal changes from ductile to brittle !
fracture (ART,,37). The unirradiated transverse impact properties of the beltline region have been determined for those materials for which sufficient amourits of materials were available and are listed on Table 4-1. The adjusted reference temperature (ARTuo7) and the unirradiated reference temperature. (The assumed (
unirradiated RT j forgings was 6007The
)of adjusted the closure RT uo7 shead of theregion beltlineand of the region outlet nozzle materials at the steel end of the eighth full power year are listed on Table 4-1 for the one quarter and three-quarter wall thickness of the vessel wall.
Bases Figure 4-1 111ustrates the calculated peak neutron fluence, for several locations through the reactor vessel beltline region wall and at the center of ,
the surveillance capsules, as a function of exposure time. Bases Figure 4-2 illustrates the design curves for predicting the radiation-induced ARTm7 as a function of the material's copper and phosphorus content and neutron fluence, i Thus, using these two figures and information on Table 4-1, shifts in the RTwo ;
can be predicted over the full service life of the vessel.
The actual shift to RTu37 of the beltline region material will be established periodically during operation by removing and evaluating the reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the ;
irradiation samples and vessel inside the radius are essentially identical, the i measured transition shift for a sample can be applied with confidence to the c adjacent section of the reactor vessel. The limit curves must be recniculated '
when the RTn7 determined from the surveillance capsule is different frcm the calculated RT for the equivalent capsule radiation exposure. The pressure and temperature 1Eits shown in the PTLR for reactor criticality, and for inservice l .
leak and hydrostatic testing, have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. ;
t The limitations imposed on pressurizer heatup and cooldown and spray water !
temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in l
accordance with the ASME Code requirements.
3/4 4.10 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components, except steam '
generator tubes, ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.
The internals vent valves are provided to relieve the pressure generated by i l steaming in the core following a LOCA so that the core remains sufficiently I covered, inspection and manual actuation of the internals vent valves 1) ensure OPERABILITY, 2) ensure that the valves are not stuck open during normal operation, and 3) demonstrate that the valves are fully open at the forces assumed in the safety analysis.
CRYSTAL RIVER - UNIT 3 8 3/4 4-11 Amendment No.
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