Text

Amend 82 to License DPR-72,updating Reactor pressure-temp Limits to 8 EFPYs
	ML20137W646
Person / Time
Site:	Crystal River
Issue date:	09/23/1985
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Florida Power Corp, City of Alachua, FL, City of Bushnell, FL, City of Gainesville, FL, City of Kissimmee, FL, City of Leesburg, FL, City of New Smyrna Beach, FL, Utilities Commission, City of New Smyrna Beach, FL, City of Ocala, FL, City of Orlando, FL, Orlando Utilities Commission, Sebring Utilities Commission, Seminole Electric Cooperative, City of Tallahassee, FL
Shared Package
ML20137W651	List: ML20137W646; ML20137W667;
References
	DPR-72-A-082 NUDOCS 8510040356
	Download: ML20137W646 (14)
	v • d • e

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pn aeauq'o, UNITED STATES E

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NUCLEAR REGULATORY COMMISSION 5

,j WASHINGTON, D. C. 20555 4

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FLORIDA POWER CORPORATION CITY OF ALACHUA CITY OF BUSHNELL CITY OF GAINESVILLE CITY OF KISSIMMEE CITY OF LEESBURG CITY OF NEW SMYRNA BEACH AND UTILITIES COMMISSION, CITY OF NEW SMYRNA BEACH CITY OF OCALA s

b ORLANDO UTILITIES COMMISSION AND CITY OF ORLANDO 4

SEBRING UTILITIES COMMISSION SEMIN0LE ELECTRIC COOPERATIVE, INC.

CITY OF TALLAHASSEE DOCKET NO. 50-302 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT i

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 82 License No. OPR-72 1.

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

A.

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

(the licensees) dated February 14, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act)', and the Commission's rules and regulations set forth in 10 CFR Chapter I; i

B.

The facility will operate in conformity with the application, 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 i

safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common 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.

i 8510040356 850923 PDR ADOCK 05000302 P

PDR 1

l

2.

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 Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 82, are hereby 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.

FOR THE NUCLEAR REGULATORY COMMISSION i blc' ohn F. Stolz, Chief Operating Reactors Bra.ch #4 sion of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: September 23, 1935 l

1 l

L

l ATTACHMENT TO LICENSE AftENDPENT NO. 82 FACILITY OPERATING LICENSE NO. DPR-72 DOCKET NO. 50-302 i

I Replace the following pages of the Appendix "A" Technical Specifications with i

the enclosed pages. The revised pages are identified by Amendment number and i

contair. vertical lines indicating the area of change. The corresponding j

overleaf pages are also provided to maintain document completeness.

Pages 3/4 4-24 e

3/4 4-26 3/4 4-27 3/4 4-28 i

l B 3/4 4-9 t

j B 3/4 4-10 B 3/4 4-11 i

B'3/4 4-12 j

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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 > 1.0 Ci/ gram Dose Equivalent 1131 CRYSTAL RIVER - UNIT 3 3/4 4-23

REACTOR COOLANT SYSTEM 3/4.4.9 ' PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except tne 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:

a.

A maximum heatup of 1000F in any one hour period, b.

For the temperature ranges specified below, the cooldown rates should be as specified (in any one hour period):

1.

T >270 g

< 100 F/Hr, ii.

270 F > T> 170 F 7 30,F/Hr, iii.

170 F 1 T 310F/Hr, and c.

A maximum temperature change of less than or equal to 50F in any one hour period during hydrostatic testing operations above system design pressure.

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-cf-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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce RCS Tavg and pressure to less than 2000F and 300 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limiu at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4 4

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s CRYSTAL RIVER - UNIT 3 3/4 4-25 Acondment flo. 30

MeustE 3.C-2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS' FOR HEATUP FOR FIRST 8 EFPY 1800 i

t THE REGIONS OF ACCEPTABLE OPERA-l I

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l' 0F THE LIMIT CURVES. MARGINS ARE 2400 INCLUDED TOR THE PRESSURE DIFTER-ENTIAL BETWEEN POINT OF SYSTEM PRESSURE MEASUREMENT AND THE PRES-SURE ON THE REACTOR VESSEL REGION

pF 2200 CONTROLLING THE LIMIT CURVE.

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i MARGINS OF 25 PSIG AND 10 F ARE i,

INCLUDED FOR POSSIBLE INSTRUMENT IRROR.

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RCS TEMPERATURE, Tc(*F)

CRYSTAL RIVER-UNIT 3 3/4 4-26 Amendment No. 82

(

Pleunt 3. 4.-3 REhCTO'R COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR COOLDOWN FIRST 8 EFPY 7E tm THE REGIONS OF ACCEPTABLE j

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OPERATION ARE BELOW AND TO r,

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CURVES. MARGINS ARE IN-

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WITH NO RC PUMPS OPER-I

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TO THE REACTOR VESSEL 4

b SHALL BE USED.

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A MAXIMUM STEP TEMPER-I y goo 9

g ATURE CHANGE OF 75 F IS

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ALLOWABLE WHEN REMOVING f

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ALL RC PUMPS FROM OPER-iI h

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TEMPERATURE CHANGE IS

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(AFIER STOPPING ALP, I' RC PUMPS).

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CRYSTAL RIVER-UNIT 3 3/4 4-27 Amendment No. 82

FleultE 3 C-O REACTOR COOLANT SYSTEM PRESSURE -TEMPERATURE LIMITS FOR HEATUP & COOLDOWN LIMITS POR INSERVICE LEAK AND HYDROSTATIC TESTS FOR FIRST 8 EFPY seco HEATUP COOLDORi l'

l THE REGIONS OF ACCEPTABLE OPERA-i TION ARE BELOW AND TO THE RIGHT t '

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FIGURE 3.4-3 ARE APPLICABLE.

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C 191 545 l

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407 2500 2

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RCS TEMPERATURE, Tc('F)

/cenMnt No. M CRYSTAL RIVER-UNIT 3 3/4 4-28

O N

BASES TABLE 4-1 b

REACTOR VESSEL TOUCilNESS m

E RT ADJUSTED NDT FOR MATERIAL CU P

S RT TRANS UPPER SilELF 8 FULL POWER YEARS C

COMPONENT TYPE NDT F FT-LB (d 1/4 T,OF Cd3/4T,OF 3

[

Nozzle Belt SA-508 CL 2

.054

.008

.006

+10 183 26 17

Upper Shell SA-533B

.20

.008

.016

+20 88 90 54

Upper Shell SA-533B

.20

.003

.016

+20 90 90 54 Lower Simil SA-533B

.12

.013

.015

-20 119 26 2

Lower Shell SA-5338

.12

.013

.015 445 88 91 67

*

Surveillance Weld

.30

.020

.005

+43 63 Upper Long Weld

.20

.009

(+ 20) * * *

66****

130 73 3

Upper Long Weld

.105

.091

.004

(+ 20) * * *

66****

I30 73 y

Upper Circum Weld

.106

.014

.013

(+ 20) * * *

66****

NA 53 (60%)

Upper Circum Weld

.19

.021

.016

(+20)* * *

66****

128 NA (40%)

Middle Circum Weld

.27

.014

.011

(+20)* "

66****

177 96 (100%)

Lower Long Weld

.22

.015

.013

(+20)* m 66****

134 75 (100%)

Lower Circum Weld

.20

.015

.021

(+20)" *

66****

30 20 (100%)

Out ist Nozzle Weld

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~ 90 136 112 Middle Circum Atypical weld

+

k Surveillance Base Metal A Surveillance Base Metal B R

R

- - Surveillance Weld E

Estimated Value E

S l

REACTOR COOLANT SYSTEM BASES q

The heatu) analysis also covers the determination of pressure-temperature limitations for j

the case mn 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 a

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; j

therefore, a lower bound curve similar to that described for the heatup of the inner wall j

cannot be defined. Consequently, for the cases in which the outer will of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on i

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 1000F per hour. During cooldown, similar types of thermal stress occur. Thus, the cooldown limit curve, Figure 3.4-3, is also a composite curve i

which was prepared based upon the same type analysis as the heatup curve with the i

exception that the controlling location is always the inside wall where the cooldown j

t.5ermal gradients tend to produce tenslie stresses while producing compressive 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 f

i generator tubes. These limits are the vertical segments of the limit lines on Figures 3.4-3 and 3.4-4, respectively. (These limits will not require adjustmer.:s due to the neutron l

fluences.)

t i

During the first several years of service life, the most limiting Reactor Coolant System I

regions are the closure head region (due to mechanical loads resulting from bolt pre-load) and the rea: tor vessel outlet nozzles. Nozzle sensitivity is caused by the high local stresses at the inside corner of the nozzle which can be two to three times the membrane i

stresses of the she!!. Af ter the first several years of neutron radiation exposure, the i

beltline region of the reactor vessel becomes the most limiting region due to material irradiation.

l For the service period for which the limit curves are established, the l

pressure / temperature limits were obtained through a point-by-point comparison of the j

limits imposed by the closure head region, outlet nozzles, and the most sensitive material in the belt!!ne region. The lowest pressure calculated for these three regions becomes the maximum a!!owable pressure for the fluid temperature used in the calculation.

The calculated pressure / temperature curves are adjusted by 25 PS! and 100F for possible instrument errors.

The pressure limit is also adjusted for the pressure differential I

between the point of pressure measurement and the limiting component for all combinations of reactor coolant pump operations.

i Irradiation damage to the beltline region can be quantified by determining the decrease in the temperature at which the metal changes from ductile to brittle fracture ( ART NDT).

The unirradiated transverse impact properties of the beltline reglon have been determined

{

for those materials for which sufficient amounts of materials were available and are CRYSTAL RIVER - UNIT 3 B 3/4 4-10 knendment No. 82 l

l listed on Table 4-1. The adjusted reference temperatures on Table 4-1 are calculated by adding the predicted radiation-induced change in the reference temperature ( ART NDT) and the unirradiated reference temperature. (The assumed unirradiated RTNQT of the closure head region and of the outlet nozzle steel forgings was 600F.) The adjusted RT NDTs of the beltline region materials at the 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 i!!ustrates the calculated peak neutron fluence, for several locations i

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 ART NDT as a function of the material's copper and phosphorus content and neutron fluence. Thus, using these two figures and information on Table 4-1, shif ts in the RT NDT can be predicted over the full service life of the vessel.

The actual shif t in RT NDT 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 shif t for a sample can be applied with 1

confidence to the adjacent section of the reactor vessel.

The limit curves must be i

recalculated when the RT NDT determined from the surveillance capsule is different from the calculated RT f

NDT or the equivalent capsule radiation exposure. The pressure and temperature limits shown on Figures 3.4-2 and 3.4-4 for reactor criticality, and for inservice leak and hydrostatic testing, have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

1 i

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.

4 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 Boller and Pressure Vessel Code.

l The internals vent valves are provided to relieve the pressure generated by steaming in the core fo!!owing 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.

)

CRYSTAL RIVER - UNIT 3 B 3/4 4-11 Amendment No. 82

I INTENTIONALLY BLANK CRYSTAL. RIVER - UNIT 3 03/44-12 Amendment fio. JJ, 82

INTENTIONALI.Y BLANK CRYSTAL RIVER - UNIT 3 B 3/4 4-13 Amendments flos. JA. 13. 82

ML20137W646

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