ML20029A501

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Amend 133 to License DPR-72,revising Tech Spec 3.4.9.1 Including Figures 3.4-2,3.4-3 & 3.4-4 Re Pressure/Temp Curves
ML20029A501
Person / Time
Site: Crystal River 
Issue date: 02/07/1991
From: Berkow H
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20029A502 List:
References
NUDOCS 9102250110
Download: ML20029A501 (15)
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UNITED STATES NUCLEAR REGULATORY COMMISSION y

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FLORIDA POWER CORPORATION CITY OF ALACHUA CITY OF BU$ITREE[

CITY OF GATNE WTELE CITY OF KISSIMEE CITY OF LEESBURG CITY OF NEW SMYRNA BEACH AND DTILITIES COMMISSION, CITY _0F NEW SMYRNA BEACli CITY OF OCALA ORLANDO UTILITIES COMMISSIUR KRD CITY OF ORLANDO SfBYKOTTETTTES comISSION SEMIN0LE ELEClRIC COOPERATIVE, INC.

CITY OF TALLAHASSEE DOCKET NO. 50-302 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.133 License No. DPR-72 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Florida Power Corporation, et al.

(the licensees) dated October 31, 1989, as supplemented on March 30, 1990 and August 10, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the appilcation, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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Accordingly, the ' license. is-amended by changes to the Technical Specifications as indicated in the attachment to this. license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-72 is hereby amended to read as follows:

Technical Specifications-The Technical Specifications contained in Appendices A and B, as 1

revised through Amendment No.133, are here]y incorporated in the license. Florida-Power Corporation shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of. its date of issuance 'and shall be implemented within 30 days of issuance.

F0 THE NUCLEAR REGULATORY COMMISSION rbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II-Office of. Nuclear Reactor Regulat_ ion _

Attachment:

Changes to the Technical Specifications-i Date of Issuance: : February 7,1991 4

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ATTACWENT TO LICENSE AMENDMENT NO.133

-FACILITY OPERATING LICENSE NO. OPR-72 i

DOCKET NO. 50-302 l

- Replace the following aages of the Appendix "A" Technical Specifications with the attached pages. T1e revised pages are identified by amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.

Remove Insert 3/4 4-24 3/4 4-24 3/4 4-26 2,4 4-26 3/4 4 3/4 4-27 3/4 4-28 3/4 4-28 8 3/4 4-7 g 3f4 4,7

B 3/4 4-8 8 3/4 4-8 B 3/4 4-9 B 3/4 4-9 B.2/4 4-10 B 3/4 4-10 8 3/4.4-11 B 3/4 4-11 g

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i 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 1.0gCi/ gram Dose Equivalent 1131 CRYSTAL RIVER - UNIT 3 3/4 4-23

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a REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM-

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. LIMITING CONDITION FOR OPERATION 3.4.9.1

.Tha. 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, and inservice leak and hydrostatic testing with:

A' maximum heatup of 50*F in any one hour period, a.

b.

For the temperature ranges specified below, the cooldown rates should be as specified:

i.

-T > 280'F s 50*F in any 1/2 hour period.

ii.

150'F < T s 280*F s 25'F in any 1/2 hour period 111.

T s 150*F s 10*F in any 1/2 hour period and A maximum temperature change of less than or equal to 5'F in any

.c.

one hour peried during hydrostatic testing operations above system design pressure.

LApPLICABILITY:

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~ 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*F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

CRYSTAL RIVER UNIT 3 3/4 4-24 Amendment No. 82,133, y

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REECTOR COOLANT SYSTEM

. SURVEILLANCE REOUTREMENT3

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.4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to.be within the limits at least once per 30 minutes during system heatup, cooldown, and

-inservice leak and hydrostatic testing operations.

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4 CRYSTAL RIVF.R - UNIT 3 3/4 4-23 Amendment No. 30:

Figuro 3,4-2 l

REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR HEATUP FOR FIRST 15 EFPY l

2f.00 -

IS00' The regions of acceptable operation are below and to the right of the limit 2400-curve. Margins are included for the Pressure differential between point of 2300 sptern pressure measurement and the Pressure on the reactor sessel reEion 2200-controlling the limit curve. Margirts of 15 psig and 10*F are included for 7

I 2100-possible instrument error, 2000 i,' !

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i o 1000 5 i'i Applicable for heatup rates of s 50'F

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in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period W

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1 INDICATED ACS INLET TEMPERATUAE (deg F)

CRYSTAL RIVER UNIT 3 3/4 4-26 AMENDMENT NO. E2,133,

Figure 3.4-3 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR COOLDOWN FOR FIRST 15 EFPY 2600 The regions of acceptable operation are 2500 - below and to the right of the limit curst. Margins are included for the 2400 pressure differential between point of rptem pressure measurement and the 2300 pressure on the reactor sessel region controlling the limit curve. Margins 2200 - of 23 psig and 10*F are included for l

possible instrument error.

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50 100 150 200 250 300 350 400 450 500 550 600 IN0ZCATE0 ACC INLET TEMPEAATUAE (deg F)

CRYSTAL RIVER UNIT 3 3/4 4-27 AMENDMENT NO. U,133,

Figure 3.4-4 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR INSERVICE LEAK & HYOROSTATIC TESTS FOR FIRST 15 EFPY 2600 The regions of acceptable operation are 2500 below and to the right of the limit

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2100 Pouible instrument error.

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For cooldowm, Notes 112 on Figure 3.4 3 are app!! cable

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50 100 150 200 250 300 350 400 450 500 550 600 INDICATED AOS INLET TEMPEAATURE (ceg F)

CRYSTAL RIVER UNIT 3 3/4 4-28 AMENWENT NO. F2.133

DELETED CRYSTAL RIVER UNIT 3 8 3/4 4-7 Amendment No.133,

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' BASES TABLE 4-1

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hk REACTOR VESSEL TOUGHNESS

.h RT MATERIAL

'CU NI RT TRANS UPPER SHELF ADJUSTED NOT FOR i

EOMPONENT TYPE W/0 W/0 NDT/F FT-L8 9 1/4 t. 'F 9 3/4 T. *F 21 FULL POWER YEARS I

Nozzle Belt SA-508 CL 2

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+10 183 117 101-

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  • Upper Shell SA-5338

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+20 88 162 126

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+20 90 162 126 Lower Shell SA-5338

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+45 88 142 121

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+43 63 Upper Long Weld

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199 153

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199 153 Upper Circum Weld

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NA 133 4

(60%)

Upper Circum Weld

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187 NA (40%)

0" Middle Circus Weld

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66****

222 168 E!

(100%)

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tower Long Weld

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199 153 j'

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Lower Circum Weld

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66****

69 65 i

(100%)

Dut 1st Nozzle Weld

.19

(+20)****

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Middle Circus A typical weld.41

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+90 (100%)

211 180

  • Surveillance Base Metal A
    • Surveillance Base Metal 8
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REACTOR COOLANT SYSTEM BASES I

The heatup analysis also covers the determination of pressure temperature i

limitations for the case in which the outer wall of the vessel becomes the controlling location.

The thermal gradients established during hettup produce l

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 bound curve similar to that described for the heatup of the inner wall cannot be defined.

Consequently, 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 Figure 3.4 2, 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 50'F per hour. During cooldown, similar l

types of thermal stress occur. Thus, the cooldown limit curve Figure 3.4-3, is also a composite curve which was prepared based upon the same, type analysis as the heatup curve with the exception 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.

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.

is caused by the high local stresses at the inside corner of the nozzle whichNozzle s 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.

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 caterial-in the bcitline 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 8 3/4-4-10 Amendment No. P2,133,

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1 Irradiation damage to the beltline region can be quantifiW y dttura m, t%

decrease in the temperature at which the metal changes from Wge G bi'bre fracture (ART region have bb) determined for those materials for which sufficient am

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materials were available and are listed on Table 4 1.

The adjusted refeNnce temperatures on Table 41 are calculated by adding the predicted radiation.

induced change in the reference temperature (6RT

) and the unirradiated reference temperature.

(TheassumedunirradiateNT region and of the outlet nozzle steel forgings was 6N.of the closure head i

) The adjusted RT of the beltline region materials at the end of the twenty first full powerndtS year are listed on Table 4-1 for the one-quarter and three-quarter wall thickness l

of the vessel wall.

Tt.e actual shif t in RTuor 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 measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.

The limit curves must be recalculated i

when the RTwot determined from the surveillance capsule is different from the calculated RTuoi or the equivalent capsule radiation exposure. The pressure and f

temperature limits shown on Figure 3.4 4 for inservic6 leak and hydrostatic l

testing, have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

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 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 co,mponents will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program fct 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 steaming 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 assumed in the sa)fety analysis. demonstrate that the valves are fully open at the CRYSTAL RIVER - UNIT 3 B 3/4 4 11 Amendment No.133',$

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INTENTIONALLY BLANK l

l CRYSTAL RIVER - UNIT 3 B 3/4 4-12 Amendment No. 15, 82

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