ML19340A475
| ML19340A475 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/22/1978 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML19340A473 | List: |
| References | |
| TAC-08884, TAC-8884, NUDOCS 8004070486 | |
| Download: ML19340A475 (8) | |
Text
{{#Wiki_filter:. b 05-1397-00 REACTOR C00LAfiT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS
' REACTOR COOLANT SYSTEM
- LIMITING CONDITION FOR OPERATION 3.4.'9.1_ The Reactor Coolant System (except the pressurizer) temperature
-and pressure shall be limited in accordance with- the limit lines shown on Figures 3.4-2, 3.4-3, and 3.4-4 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with maximum heatup and cooldown
,, rates as indicated on the applicable figure.
l APPLICABILITY: At all times. " ACTION: With any of the above 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 i fracture toughness properties of the Reactor Coolant System; determine l l that the Reactor Coolant System remains acceptable for contined operation or be in at least. HOT STANDBY within'the next 6 hours and reduce RCS I T and pressure to less than 200 F and 145 psig, respectively, within t$E9following 30 hours. I l l - 1 - l l . 1 i l l l - i .
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~4.4.9.1.1,The Raactor Coolant System temperature and pressure shall be l
determined ~to be within the limits at least once per 30 minutes during ! system heatup. .cooldown, and inservice leak and hydrostatic testing operations . 4.4.'9.1.2 The reactor vessel raterial irradiation surveiliance specimens
- shall be removed and examined, to determine changes in material properties, at 'the intervals shown in Table,4.4-5 The results of these examinations shall be used to update ' Figures 3.4-2, 3.4-3 and 3.4-4.
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3 . HAX HEATUP RATE < 50*F/HR
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b - _ POINT _ PRESS TEMP I ~ A 253 80*F h . B B 1625 , 327 F ,
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C 2250 377*F D 0 420*F E 2250 420*F a a E [c 1000 -
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CR-3 3, // -3 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR COOLDOWN FOR THE FIRST 5 EFPY 2500 - J
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gy ~K_ , ' L ' l, , o POINT PRESS VEMP S A 145 80*
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B 250 160* g C 525 195* g E D 525 , 215* - E 720 215* 500 - C D F 1850 330 0 0 2250 377 .
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Indicated Reactor Coo,lant Temperature, Tc , F -
! CRYSTAL RIVER 3 .
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'2500 I
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3 1 2000 - TE!iP C.D. MAX RATES 550-280 100 F/HR T i 280-150 50 F/HR
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ct A 220 75 1000 - - . 8 420 165 C 525 213 E D 525 255 E 675 256 500 - C D F 2250 410 B . G 2500 420 . A MARGlHS OF 25 PSIG AND id'F ARE I i t i INCLUDED FOR'POSSIBLE INSTRUMENTATION 100 200 300 400 500 ERROR i Temperature, 'F i l
05- n97-00~d _ REACTOR C00LAfiT SYSTEM- _ BASES The ACTION statement permitting POWER OPERATION to continue for limited time periods with; the primari coolant's . specific activity >1.0 uCi/ gram DOSE EQUIVALEitT- I-131, but within the allowable limit shown on Figure 3.4-1, acco-
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nrnodates .possible iodine spiking phenomenon which may. occur .following changes in THERMAL POWER. Operation with specific activity levels exceeding.1.0
-uC1/ gram DUSE EQUIVALENT 1-131 but within the limits shown on Figure 3.4-1 must:be restricted to no more than 10-percent.of the unit's yearly. operating l time since the activity.1evels allowed by Figure 3.4-1 increase the 2. hour thyroid dose at the site boundary by a factor of up to 20 following a postu-lated steam generator tube rupture.
Reducing T to 5000F pre' vents the release of activity should a steam generator. tube, Sture.since the saturation pressure ~ of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive. specific activity levels .in.the primary coolant will be detected in sufficient time to take corrective action; Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduc-tion in frequency'of isotopic analyses following power changes may be per-missible if justified by the data obtained. _3/4 4.9 : PRESSURE / TEMPERATURE LIMITS-
- All components in the Reactor' Coolant System are designed to withstad the effects -of cyclic loads due to system temperature and pressure changes.
. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operdions. The varicus categories of load cycles used -for design purposes are provided in Section 4.1.2.4 of the FSAR. During . ;
.heatup and cooldown, the rates:of. temperature and pressure changes are limited '
so that the. maximum specified heatup and cooldown rates are consistent with the design assumptions and ' satisfy the stress-limits for cyclic operation.
- The heatup limit curve, Figure-3~.4-2 is a composite' curve which' was .
prepared by determining the most conservative case, with either the'1/4T or 3/4T wall;1ocation controlling, for any heatup rate up to the indicated maxima:per hour. The:cooldown? limit curve, Figure 3.4-3, is a composite curve which was-prepared based upon-the same type analysis. The -reactor? vessel matdrials have been tested to determine their initialLRT ; . Reactor opbtion and resultant. fast neutron (E>1LMev) irradiation will l-cause an increase in the. RT , iture, based upon' the fluenckT*o"Therefore,,an c pper, and phosphorous adjusted reference content of.the tempera-material l- ;in question Vbeea vislM. ' The heatup and cooldown limit curves, L L ~ of: Figures 3.4-2 and 3.4-3 account for; predicted adjustments for possible
'errorsfin. thespressure. and: temperature sensing instruments.
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. D5-1397-0014 REACTORC00LAtlTSYSTEM(continued) 9ASES The actual shift in RT of the vessel material will be established periodically during operatikby removing and evaluating reactor vessel material irradiation surveillance specimens installed near the inside wall of the reac-tor vessel-in the core area. The normal heatup, normal cooldown, and inservice leak and hydrostatic test curves must be recalculated when the ART deter-minedfromthesurveillancecapsuleisdifferentfromthecalculatEART for the equivalent capsule radiation exposure. riDT e
The cl'osure. head regicn is significantly stressed at relatively low temperatures (due to mechanical loads resulting from bolt pre-load). The outlet nozzles of the reactor vessel affect the presssure-temperature limit . curves of the first several service periods. For the service period for which the limit curves are established, the maximum allowable pressure as a function of fluid temperature is obtained through a point-by-point comparison of the limits imposed by the closure head region, outlet nozzles, and beltline region. The maximum allowable pressure is taken to be the lower pressure of the three calculated pressures. The calculated pressure temperature limit curves are then adjusted by 25 psi and 10oF for possible errors in the pressure and temperature sensing instruments. The pressure limit is also adjusted for the pressure differential between the point of system oressure measurement and the limiting component for all operating reactor coolant pump combinations. The limit curves were prepared based upon the most limiting adjusted ntference temperature of all-the beltline region materials at the end of the fifth effec-tive full power year. The unirradiated transverse impact properties of the beltline region materials, required by Appendices G and H to 10 CFR 50, were determined for those materials for which sufficient amounts of material were available. The adjusted reference temperature is calculated by adding the predicted radiation-induced ART are calculated in accordab,. andThe the unirradiated. predicted ARTwith Regulatory Guide 1.99 using the resktive neutron fluence and copper and phosphorus contents. ; The assumed unirradiated ' reference temperature of the beltline region controlling material is 120 F. At the end-of five effective full power years this increases.to 252 F at ~1/4T and 135 F at 3/4T. The RT at the outlet juncture is taken to be 120*E. The RT intheclosurehNSbisassumedto be'60F.: Pressure-temperatureconsthSIntsarecomputedinaccordancewith BAW 10046A with postulated flaws taken to be fully contained within the material region of. interest. Figure 3.4-2 presents the ~pnassure-temperature limit curve for normal he,atup. This figure also presents the core criticality limits as required
.by Appendix G to 10 CFR 50. Figure 3.4-3 presents the pressure temperature
- limit curve for normal cooldown. Figure 3.4-4 presents the pressure-tempera-
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CRYSTAL RIVER -_ UNIT.3- B 3/4 4-7
Qf W ') ' i REACTOR C00LAi1T SYSTEM (continued)
. _B ASES ture limit curves for heatup and cooldown for inservice leak and hydrostatic testing.
All pressure-temperature limit curves are applicable up to the fif th
- effective full _ power year. The protection against non-ductile failure is .
assured by _ maintaining the coolant pressure below the upper limits of Figures 3.4-2, and 3.4-3, and 3.4-4 The pressure andLtemperature limits shown on Figures 3.4-2 and 3.4-3
# for reactor criticality and few inservice leak and hydrostatic testing have been provided to_ assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.
The number of reactor vessel frradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-3 to assure compliance with the requirements of Appendix H .to 10 CFR Part.50. The limitations imposed on pressuriter 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 per-
~ formed in accordance with the ASME Code requirements.
3/4.4.10 STRUCTURAL ItiTEGRITY 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 Boilcr and Pressure Vessel Code. _The internals vent. valves are provided to relieve..the pressure generated by-steam - ing in the core -following a LOCA so that the core remains sufficiently 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.
e CRYSTAL. RIVER - UNIT 3- . . B 3/4 4-8
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