Text

Amend 103 to License DPR-72,changing Tech Specs to Support Operations for Fuel Cycle 7
	ML20237E421
Person / Time
Site:	Crystal River
Issue date:	12/14/1987
From:	Berkow H; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20237E423	List: ML20237E421; ML20237E430;
References
	NUDOCS 8712280331
	Download: ML20237E421 (23)
	v • d • e

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UNITED STATES y

g NUCLEAR REGULATORY COMMISSION

j WASHINGTON, D. C. 20555 l

%.....w o

FLORIDA POWER CORPORATION

~ ~ ~ ~ ~

CITY OF ALACHUA l

CITY OF BU5hhELL CITY OF _AINE5VILLE G

CITY OF KI55IfffEE l

CITY OF LEE 3hbRt-CITY OF NEW S.M.YRNA BEACH AND.U.TIETT.IES COMMI.S._S_I_O_N, CITY OF NEW SMYRNA BEACH

-- g gg ORLANDO UTILITIES EDMil55 ION AND CITY OF ORLANDO 5EBhlhC l'TILITIE5 EDMQ331DY"~~~

SEF0hblE ELEtTRIC COOPERATIVE,~~INC.

CITY OF TALEXKA35EE-~

~-

DOCKET NO. 50-30,2 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.103 License No. DPR-72 j

1.

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

A.

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

(the licer. sees) dated April 15, 1987, as supplemented July 17, 1987, l

~

September 16, 1987 and October 27, 1987, 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; B.

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

There is reasonable assurance (1) 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 Commission'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.

8712280331 871214' hDR ADOCK 05000302 PDR

I l 2.

Accordingly, the liccrse is anended by changes to the Technical Specifications as indicated in the attachrnent to this license amendment.,

{

and paragraph F.C.(2) of Facility Operating License No. DPR-72 is herety amended to read as follows:

T_echnical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.103, 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 C0ffilSSION

.erbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects-I/II-Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical I

Specifications Date of Issuance: December 14, 1987 l

O L____.____ _. _ _ _ _ _ _ _.

-a_.- - _ _ _ -. - - - - - - - - - - -

ATTACHMENT TO LICEh5E,,ApENppEE N0.J03 FACJ,LI,1), OF,E, RATING L1CEN,S,E, N,0. DPR-72 DOCKET N0_._50-302 Replace the following pages of the Appendix "A" Technical Specifications with the attecbed pages. The revised pages are ider,tified by Anendment nuriber ar.d contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document corpleteress.

Remove Insert 2-3 2-3 2-7 2-7 B2-1 B2-1 B2-2 B2-2 j

B2-3 B2-3 l

3/4 1-25 3/4 1-25 I

3/4 1-27 3/4 1-27 3/4 1-27a 3/4 1-27a 3/4 1-28 3/4 1-28 1

3/4 1-29 3/4 1-29 l

3/4 1-29a 3/4 1-29a i

3/4 1-30 3/4 1-30 3/4 1-34 3/4 1-34 3/4 1-37 3/4 1-37 l

3/4 2-1 3/4 2-1 i

3/4 2-2 3/4 2-2 3/4 2-2a 3/4 2-2a i

3/4 2-3 3/4 2-3 3/4 2-3a 3/4 2-3a 3/4 2-11 3/4 2-11 3/4 3-8 3/4 3-8 B3/4 1-2 B3/4 1-2 B3/4 2-1 B3/4 2-1 B3/4 2-2 B3/4 2-2 B3/4 2-3 B3/4 2-3 5-4 5-4

'J a--

n

____m-m____ _ _ _ _

Figure 2.1-2 Reactor Core Safety Limits

. 120

~

(-35.59,112)

(32.85,112)

-110 Acceptable

- 100

(-48.81,99.67)

Oper tion

(-35.59,89.6)

(32.85,89.6) 90

^

i(48.43,84.37)

(-48.81,77.27)

W Acceptable o

3 & 4 Pump

[-- 70 Operation l

,(48.43,61,97)

- 60 y-3-- 50 E

"-- 40 E

E

- 30 g-e t-- 20 E

a

- 10

- 50 30 10 0

10 20 30 40 50 60 Axial Power Imbalance, %

l I

CRYSTAL RIVER - UNIT 3 2-3 Amendment Nos. J$,Jg,3l,fy,fg, E%,65,77,103

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS l

REACTOR PROTECTION SYSTEM SETPOINTS 2.2.1 The Reactor Protection System Instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

I APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

With a Reactor Protection System instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

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1 CRYSTAI RIVER - UNIT 3 2-4

,..%.~.e.*

,. - ~ ~ -,.

u.

.e.u-a.w-%-ee.-

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Figure 2.2-1 Trip Setpoint for Nuclear Overpower Based on RCS Flow and Axial Power Imbalance

(-17,108)_

_ (17,108)

~

M = 1.0 Acceptable M = -1.833 2

1 100 4 pump Operation

(-34.7,90.3)<

90

(-17,80.67)

(17,80.67)

^

80 (35,75) u 70 I

(-34.7,62.97),

Acce$aup g

o Operation

=

.E l

h-- 50 (35,47.67)

.ee 3-- 40 u

30 L

I E

g-20 1

%e E. 10 i L,

i 40

-30

-20

-10 0

10 20 30 40 50 Axial Power Imbalance, %

CRYSTAL RIVER - UNIT 3 2-7 Amendment Nos. 77,M,JJ,J$,

EE,55,5f 77, 103 l

e

= = _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _. -

2.1 SAFETY LIMITS BASES 2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding l

and possible cladding perforation which would result in the release of fission i

products to the reactor coolant.

Overheating of the fuel cladding is prevented l

by restricting fuel operation to within the nucleate boiling regime. where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime would result in excessive cladding temperature because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter -during operation but THERMAL POWER and Reactor Coolant Temperature and pressure can be related to DNB using a Critical Heat Flux (CHF) correlation.

The' local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux is indicative of the margin to DNB.

The B&W-2 and BWC CHF correlations have been developed to predict DNB for axially uniform and non-uniform heat flux distributions.

The B&W-2 correlation applies to Mark-B fuel and the BWC correlation applies to Mark-BZ fuel.

The ninimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30 (B&W-2) and 1.18 (BWC).

A DNBR of 1.30 (B&W-2) or 1.18 (BWC) corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur.

The curve presented in Figure 2.1-1 represents the conditions ' at which the cinimum allowable DNBR or greater is predicted for the thermal power and number of operating reactor coolant pumps.

This curve is based on the following nuclear power peaking factors with potential fuel densification effects:

I N

N N

F

= 2.82 F = 1.71 F = 1.65 0

AH Z

The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum allowable control rod withdrawal, and form the core DNBR design basis.

CRYSTAL RIVER - UNIT 3 B 2-1 Amendment Nos. N,0.M,77,103

i

)

l EATETY LIMITS BASES The reactor trip envelope appears to approach the safety limit more closely than l

it actually does because the reactor trip pressures are measured at a location where the indicated pressure is about 30 psi less than core outlet pressure, providing a more conservative u rgin to the safety limit.

1 1

The curves of Figure 2.1-2 are based on the more restrictive of two thermal l

limits and consider the ef fects of potential fuel densification and potential I

fuel rod bow:

1 1.

The DNBR limit produced by a nuclear power peaking factor of F

= 2.82 I

or the combination of the radial peak, axial peak and posit $on of the axial peak that yields no less than the DNBR limit.

8 2.

The combination of radial and axial peak that causes central fuel melting at the hot spot.

I Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.

The specified flow rates for curves 1 and 2 of Figure 2.1-2 correspond to the expected minisua flow rates with four pumps and three pumps respectively.

The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-saxinua thermal power combinations shown in BASES Figure 2.1.

The curves of BASES Figure 2.1 represent the conditions at which the DNBR limit l

predicted at the maximum possible thermal power for the number of reactor l

coolant pumps in operation.

These curves include the potential effects of ' fuel rod bow and fuel densification.

l 1

l l

l CRYSTAL RIVER - UNIT 3 B 2-2 Amendment Nos. 75,77,2(,77. 103

'SATETY LIMITS BASES For each curve of BASES Figure 2.1, a pressure-temperature point above and to the lef t of the curve would result in a DNBR greater than 1.30 (B&W-2) or 1.18 (BWC) or a local quality at the point of minimum DNBR less than 22% (B&W-2) or 26% (BWC) for that particular reactor coolant pump situation.

The curve for three pump operation is more restrictive than any other reactor coolant pump cituation because any pressure / temperature point above and to the lef t of the three pump curve will be above and to the left of the other curves.

2.1.3 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of l

radionuclides contained in the reactor coolant from reaching the containment atmosphere.

1 I

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Boiler and pressure Vessel Code which permits a maximum transient pressure of 110%, 2750 psig, of design pressure.

The Reactor Coolant System piping, valves and fittings, are designed to USAS.B 31.7, February, 1968 Draft Edition, which permits a maximum transient pressure of 110%, 2750 psig, of component design pressure. The Safety Limit of 2750 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant system is hydrotested at 3125 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

l CRYSTAL RIVER - UNIT 3 B 2-3 Amendment No. 16.f7. 103 1

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS i

nl The Reactor Protection System Instrumentation Trip Setpoint specified in Table 2.2-1 The Trip are the values at which the Reactor trips are set for each parameter.

l Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety ilmits. Operation with a trip setpoint less but within its specified Allowable Value is conservative than its Trip Setpoint acceptable on the basis that the difference between each Trip Serpoint and the Ailowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

The Shutdown Bypass provides for bypassing certain functions of the Reactor Protection System in order to permit control rod drive tests, zero power PHYSICS TESTS and The purpose of the Shutdown Bypass RCS certain startup and shutdown procedures.

Pressure-High trip is to prevent normai operation with Shutdown Bypass activated. This high pressure trip setooint is lower than the normal low pressure trip setpoint so that the reactor must be tripped before the bypass is initiated. The Nuclear Overpower Trip Setpoint of less than or equal to 5.0% prevents any significant reactor power from being produced. Sufficient natural circulation would be available to remove 5.0% of RATED THERMAL POW'ER if none of the reactor coolant pumps were operating.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic Reactor Protection System instrumentation channeis and provides manual reactor trip capability.

Nuclear Overpower A Nuclear Overpower trip at high power level (neutron flux) provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.

During normal station operation, reactor trip is initiated when the reactor power level reaches 104.9% of rated power. Due to calibration and instrument errors, the maximum actual power at which a trip would be actuued could be 1i2% which was used in the safety analysis.

CRYSTAL RIVER - UNIT 3 B 2-4 Amendment flo. 75, $7, EE, JE

REACTIVITT CONTROL SYSTDIS REGULATING R0D INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3 6 The regulating rod groups shall be limited in physical insertion to the acceptable operation region as shown on Figures 3.1-1, 3.1-2, 3.1-3, and 3.1-4, with a rod group overlap of 25 ! 5\\ between sequential withdrawn groups 5 and 6, and 6 and "/.

APPLICABILITY:

MODES 1* and 288 ACTION:

a.

With the regulating rod groups inserted in the unacceptable operation region, immediately initiate and continue boration at greater than or equal to 10 GpM of 11,600 ppm boric acid solution or its equivalent, until out of the unacceptable operation region.

Additionally, either:

1. Restore the regulating groups to within the acceptable region limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of exceeding the acceptable operation region, or

2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figures within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of exceeding the acceptable operation region, or

3. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of exceeding the acceptable operation region.

b.

With the regulating rod groups inserted in the restricted operation region or with any group sequence or overlap outside the specified

limits, except for surveillance testing pursuant to Specification 4.1.2.1.2, either:

1. Restore the regulating groups to within the acceptable region limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of exceeding the acceptable operation region, or

2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figures within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of exceeding the acceptable operation region, or

3. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of exceeding the acceptable operation region.

'See Special Test Exceptions 3.10.1 and 3.10.2.

With K gg greater than or equal to 1.0.

e i

CRYSTAL RIVER - UNIT 3 3/4 1-25 Amendment Nos. f$,64,77, 103

1 REAC*IVITY CONTROL SYu d5 REGULATING ROD INSERTION LIMITS SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each regulating group shall be determined to be within the insertion, sequence and overlap limits at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except when:

The regulating rod insertion limit alarm is inoperable, then a.

verify the groups to be within the insertion limits at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; b.

The control red drive sequence alann is inoperacle, then verify the groups to be within the sequence and overlap limits at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

CRYSTAL RIVER - UNIT 3 3/4lj5

Figure 3.1-1 Regulating Rod Group Insertion Limits for Four-Pump Operation From 0 to 500_+10 EFPD 110

-(200,102)

(300,102) 100 (270,102)

(270,92) gg UNACCEPTABLE.

RESTRICTED

'g 80 (250,80)

OPERATION OPERATION I

g 70 SHUTD0tlN y

MARGIN LIMIT 60

?>

3 50 (120,50)

(175,50) e

~

5-l 40 ACCEPTABLE c.

OPERATION 30 20 (60,15) 10 (0,5) i 0

0 50 100 150 200 250 300 Rod Index, % Withdrawn

. i 0

25 50 75 100 0

25 50 75 100 Group 5 Group 7 0

25 50 7,5 100 j

Group 6 j

Note: This Figure shall be used up to complete APSR i

withdrawal per Specification 3.1.3.9 CRYSTAL RIVER - UNIT 3 3/4 1-27 Amendment Nos. 7,2,7%,79, 77,74,ff,6F,77,103

9

-DELETED-s i

i il CRYSTAL RIVER UNIT 3 3/4 1-27a Amendment Nos. H,M,77,103 i

~

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Figure 3.1-2 Regulating Rod Group Insertion Limits for Four-Pump Operation After 500_+10 EFPD 110 (300,102)

(208.102) 65.102) 1.00 (260,92) 90 UNACCEPTABLE OPERATION RESTRICTED (250,80) u 80 y

OPERATION O

SHUTDOWN 70 MARGIN g

E LIMIT 60

=

(127,50)

(175,50) 50 s

ACCEPTABLE 40 b

OPERATION E

30 20 (64,15) 10 0

O 50 100 150 200 250 300 Rod Index, % Withdrawn 0

25 50 75 100 0

25 50 75 100 e

e i

t I

i i

f I

I Group 5 Group 7 0

25 50 75 100 a

i 1

i I

Group 6 Note: This Figure shall be used after complete APSR withdrawal per Specification 3.1.3.9 CRYSTAL RIVER - UNIT 3 3/4 1-28 Amendment Nos. 76,79,33,35,

)

E,EF,77,103 i

Q

4 1

I

-DELETED-e l

CRYSTAL RIVER UNIT 3 3/41 28a l

Figure 3.1-3 Regulating Rod Group Insertion Limits for Three-Pump p

Operation From 0 to 500.+JO EFP0 110 4

100 90, (300,77) 80 b

UNACCEPTABLE (200,77)

(270'77)

OPERATION (270,69.5) i 70 E

RESTRICTED (250,60.5) y 60 OPERATION y

SHUTDOWN VARGIN g

50 LIMIT o

40 J

(120,38)

(175,'33) y E

30 ACCEPTABLE 20 OPERATION 10 (60,11.75)

(0,4.25}

1 0

50 100 150 200 250 300 Rod Index, ", Withdrawn 0

25 50 75 100 0

25 50 75 100

)

i i

i e

i i

Group 5 Group 7 0

25 50 75 100 I

i i

I Group 6 Mote: This Figure shall be used up to complete APSR withdrawal per Specification 3.1.3.9 CRYSTAL RIVER - UNIT 3 3/4 1 29 Amendment Nos. JE,79,32,

E 54.77.103

S G

l

-DELETED-l 1

l l

l CRYSTAL RIVER UNIT 3 3/4 1-29a Amendment Nos. SE,EF,77,103

Figure 3.1-4 Regulating Rod Group Insertion Limits for Three-Pump Operation After 500+10 EFPD 110 l

100 l

l l

90 UNACCEPTABLE OPERATION (300,77) 80 b

(208,77)

(265,77) -

1 E

70 (260,69.5)

RESTRICTED (250.60.5) 60 OPERATION 5

SHUTDOWN y

MARGIN LIMIT q

50 cc 40 J

(127,38)

~

(175,38) y 30 ACCEPTABLE OPERATION 20 (64,11,75) 10 0

0 50 100 150 200 260 300 Rod Index, % Withdrawn O

25 50 75 100 0

25 50 75 100 g

i i

l 1

i I

I I

I Group 5 Group 7 0

25 50 75 100 l

t i

i i

I Group 6 l

Note: This Figure shall be used after complete APSR withdrawal per Specification 3.1.3.9 CRYSTAL RIVER - UNIT 3 3/4 1-30 Amendment Nos. 76,77,32, 4,

H,77,103 s -

- O

3 l

l

-DELETED-3 i

i e

l 1

CRYSTAL RIVER UNIT 3 3/4 1-31 Amendments Nos. )p, $$,,,

1 l

1

____-_______w

REACTIVITY CONTRot SYSTEMS R00 PROGRAM LIMITING CON 0! TION FOR OPERATION 3.1.3.7 Each control rod (safety, regulating and APSR) shall be pro-gramed to operate in the core position and rod group specified in Figure 3.1-7.

APPt!CABILITY: MODES 1* anc 2*.

ACTION:

e With any control rod not programed to coerate as specified above, be in HOT STAND 8Y within I hour.

SURVEILLANCE REQUIREMENTS 4.1.3.7 Each control rod shall be demonstrated to be programed to a.

operate in the specified core position and red group by:

1.

Selection a$d actuation from the control. room and verifi-cation of movement of the proper red as indicated by both the absolute and relative position indicators:

a)

For all control rods, after the control rod dMve patchs are locked subsecuent to test. reprogramming or maintenance within the panels.

b)

For specifically affected individual rocs, follaring maintenance, test, reconnection or modification of power or instrumentation cables from the control red dHve control system to the control rod drive.

2.

VeMfying that each catie that has been disconnected has been properly matched and reconnected to the specifiec control rod dH ve.

b.

At least once each 7 days, veMfy that the control red dMye patch panels are locked.

see special Test i. ace:t1ons 3.10.1 and 3.10.2.

CRYSTAL RIVER - UNIT 3 3/4 1 33 Amendment No.1 S e.

Ficure 3.1-7 Control Rod Locations and Group Designations for Crystal River 3 Cycle 7 l

X Fuel Transfer Canal 1

A B

1 6

1 C

3 5

5 3

+

D 7

8 7

E 3

5 4

4 5

3 F

1 8

6 2

6 8

1 G

S 4

2 2

4 5

H W"

6 7

2 2

7 6

Y K

5 4

2 2

4 5

t, 1

8 6

2 6

8 1

11 3

5 4

4 5

3 N

l 7

8 7

l8 7

0 l

3 5

5 3

l P

l l

1 6

1 R

a Z

~

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 X

Group Number

~

Grouc No. of Rods

_ Function 1-8 Safety 2

8 Safety 3

8 Safety 4

8 Safety 5

12 Control 6

8 Control 7

8 Control 8

8 APSRs Total 68 CRYSTAL RIVER - UNIT 3 3/4 1-34 Amendment Nos.17,77,H,77,103

~

c

REACTIVITY CONTROL SYSTEMS AYTAL POWER SHAPING ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.9 Except as required for surveillance testing per Technical Specification 3.1.3.3, the following limits apply to axial power shaping rod (ApSR) insertion.

Op to 490 ETPD, the APSR's may be l

positioned as necessary.

The APSR's shall be completely withdrawn (100%) by 510 ETPD.

Between 490 and 510 ETpD, the APSR's say be l

withdrawn.

However, once withdrawn during this period, the APSR's shall not be reinserted.

APPLICABILITY: MODES 1 and 2'.

ACTION:

With the axial power shaping rod group outside the above insertion limits, either;

a. Restore the axial power shaping rod group to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figure within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

c. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.9 The position of the axial power shaping rod group shall be determined to be within the insertion limits at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With Kegg 2 1.0.

CRYSTAL RIVER - UNIT 3 3/4 1-37 Amendment flos. 76,77,75,67,77, 103 l

J

-DELETED-i CRYSTAL RIVER UXIT 3 3/4 1-38 gendmentsNos.JE,JP,H,$$,pg,

J /4.2 -

POWER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE LIMITING CONDITION FOR OPERATION 3.2.1 AXIAL POWER IMBALANCE shall be maintained within the limits shown on Figures 3.2-1, and 3.2-2.

l APPLICABILIH:

MODE 1 above 40% of RATED THERMAL POWER *.

ACTION:

With AXIAL POWER IMBALANCE exceeding the limits specified above, either:

a. Restore the AXIAL POWER IMBALANCE to within its limits within 15 minutes, or

b. Be in at least HOT STANDBY within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

SURVEILLANCE REQUIREMENTS l

4.2.1 The AXIAL POWER IMBALANCE shall be determined to be within limits in each core quadrant at least once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when above 4 0*$ of RATED THERMAL POWER except when an AXIAL POWER IMBALANCE monitor is inoperable, then calculate the AXIAL POWER IMBALANCE in each core quadrant with an inoperable monitor at least once per hour.

' See Special Test Exception 3.10.1.

l CRYSTAL RIVER - UNIT 3 3/4 2-1 Amendment !!os. H,77,103 W_---_._-.---__-_._.-._

Figure 3.2-1 Axial Power Imbalance Envelope for Four-Pump Operation From 0 EFPD to EOC l

1 110--

(-20_,102 )

(15,102) 100-(-25,92)

(15,92)

,(20,80)

(-28,80)<

80 70 - -

\\

l l

b

$ 60 -

n.

l

(-28,50)o l 50 -

" (20,50) o 4

ACCEPTABLE UNACCEPTABLE 13 40

. OPERATION OPERATION

%a:

30 - -

E t

20 -

g 10 - -

0

-50

-40

-30

-20

-10 0

10 20 30 40 50 Axial Power Imbalance, %

C0YSTAL RIVEP.- UNIT 3 3/4 2-2

-Amendment Nos. 7,7,76,79,32,

6,f A,77,103

i i

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4

-DELETED-

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CRYSTAL RIVER UNIT 3 3/4 2-2a Amendment Nos. #6,57,77, 103 1

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Figure 3.2-2 Axial Power Imbalance $nvelope for Three-Pump Operation From 0 EFPD to EOC l

110--

100- -

90-

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(15,77)

(-20,77) -

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(-25,69.5) 70-(15,69.5) 4

(-28,60.5)i 60 i(20,60.5) u I

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(-28,38) o ACCEPTABLE UNACCEPTABLE

)!30-OPERATION OPERATION e

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-50

-40

-30

-20

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CRYSTAL RIVER - UNIT 3 3/4 2-3 Amendment Nos. 75,79,32,77, FE.EA,77,103

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_ _a

-DELETED-CRYSTAL RIVER UNIT 3 3/4 2-3a Amendment flos. 77,103

TABLE 3.2-2 OUADRANT POWER TILT LIMITS STEADY STATE TRANSIENT MAXIMUM LIMIT LIMIT LIMIT QUADRANT POWER TILT as Measured by:

Synsetrical Incore Detector Systes 4.12 10.03' 20.0 Power Range Chanriels 1.96 6.96 20.0

~Minimus Incore Detector System 1.9 4.40-20.0 l

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'r 4

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CRYSTAL RIVER - UNIT 3 3/4 2-11 Amendment Nos. 15,79,M,77, 103

)

POWER DISTRIBUTION LIMITS, DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNS related parameters shall be maintained within the limits shown on Table 3.2-1:

a.

Reactor Coolant Hot. Leg Temperature b.

Reactor Coolant Pressure c.

Reactor Coolant Flow Rate i

APPLICABILITY: MODE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the param-eter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than St of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQU!REMENTS i

4.2.5.1 Each of the parameters of Taele 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months.

CRYSTAL RIVER - UNIT 3 3/4 2-12

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TABLE 4.3-1 (Continued) l NOTATION l

- With any control rod drive trip-breaker closed..

1

- - When Shutdown Bypass is actuated.

(1) - If not performed in previous 7 days'.

(2) - Heat balance only, above 15% of RATED THERMAL POWER.

(3) - When THERMAL POWER (TP) is above 30% of RATED THERMAL POWER (RTP) compare out-of-core seasured AXIAL POWER IMBALANCE (APIo) to incore measured AXIAL POWER IMBALANCE (APIj) as follows:

RTP (APIo - APIj) = Imbalance Error TP Recalibrates if the absolute value of the Imbalance Error is equal to or greater than 2.5%.

l (4) - AXIAL POWER IMBALANCE and loop flow indications only.

(5) - Verify at least one decade overlap if not verified in previous 7 days.

(6) - Each train tested every other month.

(7) - Neutron detectors may be excluded from CRANNEL CALIBRATION.

(8) - Plow rate measurement sensors may be excluded from CRANNEL CALIBRATION.

However, each flow acasurement sensor shall be calibrated at least once per 18 months.

(9) - Current and voltage sensors may be excluded from CRANNEL CALIBRATION.

l l

CRYSTAL RIVER - UNIT 3 3/4 3-8 Amendment Hos. #7,77,103 1

~

a

3/4.1 REACTIVITY CONTROL SYSTEMS BASES i

o i

n i i 3/4.1.1 BORATION CONTROL 3/4.1.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions,2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor wiu be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition. During Modes 1 and 2 the SHUTDOWN MARGIN is known to be within limits if all control rods are OPERABLE and withdrawn to or beyond the insertion ilmits.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration and RCS Tav The most restrictive condition for Modes 1, 2, and 3 occurs at EOL, with Tav, [t no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident a minimum SHUTDOWN MARGIN of 0.60% delta k/k is initially required to control the reactivity transient.

Accordingly, the SHUTDOWN MARGIN required is based upon this limiting condition and is consistent with FSAR safety analysis assumptions.

l 3/4.1.1.2 BORON DILUTION A minimum flow rate of atieast2700 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual through the Reactor Coolant System in the core during boron concentration reductions in the

~~

Reactor Coolant System. A flow rate of at least 2700 GPM wi!! circulate an equivalent Reactci Coolant System volume of 12,000 cubic feet in approximately 30.

minutes. The reactivity change rate associated with boron concentration reduction will be within the rapahif t y for operator recognition and control.

t 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance requiremers for measurement of the MTC each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. The confirmation that the measured MTC value is within its limit provides assurance that the coefficient will be maintained within acceptable values throughout each fuel cycle.

CRYSTAL RNER - UNIT 3 53/411 Asandment flo. At, 64

_________L_-____

REACTIVITY CONTROL SYSTEMS B&SES 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 525'F.

This limitation is required to ensure that (1) the moderator temperature coef ficient is within its analyzed temperature range, (2) the protective instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor pressure vessel is above its minimum RTNDT temperature.

3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is

&vailable during each mode of f acility operation.

The components required - to perform this function include (1) borated water sources, (2) makeup or DHR pumps, (3) separate flow paths, (4) boric acid pumps, (5) associated heat I

tracing systems, and (6) an emergency power supply from OPERABLE emergency busses.

With the RCS average temperature above 200*F, a minimum 6f two separate and redundant boron injection systems are provided to ensure single functional l

capability in the event an assumed failure renders one of the systems inoperable.

Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall l

facility safety from injection system failures during the repair period.

The boration capability of either system is suf ficient to provide a SHUTDOWN l

MARGIN from all operating conditions of 1.0t ok/k after xenon decay and cooldown to 200*F.

The maximum boration capability requirement occurs from full power equilibrium xenon conditions and requires either 5,400 gallons of 11,600 ppm boric acid solution from the boric acid storage tanks or 45,000 gallons of 2,270 ppm borated water from the borated water storage tank.

The requirements for a minimum contained volume of 415,200 gallons of borated water in the borated water storage tank ensures the capability for borating the RCS to the desired level.

The specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.4.

Therefore, the larger volume of borated water is specified.

Also, the 6,000 gallons minimum BAST requirement per Specification 3.1.2.9 is conservative for this cycle.

With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1.0t Ak/k af ter xenon decay and cooldown from 200*F to 140'F.

This condition requires either 490 gallons of 11,600 ppa boron from the boric acid storage system or 2,502 gallons of 2,270 ppm boron from the borated water storage tank.

To envelope future cycle BWST and BAST contained borated water volume requirements, a minimum volume of 13,500 gallons and 600 gallons, respectively, are specified.

CRYSTAL P.IVER - UNIT 3 B 3/4 1-2 Amendment Nos. U,20,32,4,

SF.77,103

]

1 POWER DISTRI'UTION LIMITS B

3/4.2 BASES l

The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Trequency) events by:

(a) maintaining the minimus DNBR in the core within the limit during l

normal operation and during short ters transients, (b) maintaining the peak linear power density 1 18.0 kW/ft during normal operation, and (c) maintaining the peak power density 120.5 kW/f t during short term transients.

In addition, the above criteria aust be set in order' to meet the assumptions used for the loss-of-coolant accidents.

The power-iabalance envelope defined in Figures 3.2-1, and 3.2-2 and the insertion limit curves, Figures 3.1-1, 3.1-2, 3.1-3, and 3.1-4, are based on

.LOCA analyses which-have defined the maxinua linear heat rate such that the maximum clad temperature will not exceed the Final Acceptance Criteria of 2200'T following a LOCA.

Operation outside of the power-labalance envelope alone does not constitute a situation that would cause the Final Acceptance Criteria to be exceeded should a LOCA occur.

The power-imbalance envelope represents the boundary of operation limited by the Final Acceptance Criteria only if the control rods are at the insertion limits, as defined by Figures 3.1-1, 3.1-2, 3.1-3, and 3.1-4, and if the steady state limit QUADRANT POWER TILT exists.

Additional conservatism is introduced by application of:

a.

Nuclear uncertainty factors.

b.

Thermal calibration uncertainty.

c.

Fuel densification effects.

l d.

Hot rod manufacturing tolerance factors.

The conservative application of the above peaking augmentation factors considers I

the potential peaking penalty due to fuel rod blow.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met.

The definitions of the design limit nuclear power peaking factors as used. in these specifications are as follows:

Fg Nuclear Heat Flux Bot Channel Factor, is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and rod dimensions.

CRYSTAL RIVER - UNIT 3 B 3/4 2-1 Amendment Hos. M,77, 103

_-__-_____-_a

i POWER DISTRIBUTION LIMITS BASES FN Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of aH the integral of linear power along the rod on which minimus DNBR occurs to the average rod power.

It has been determined by extensive analysis of possible operating power shapes that the design limits on nuclear power peaking and on minimus DNBR at full power are met, provided:

Fq 1 3.13; FN g 1,71 AH Power Peaking is not a directly observable quantity and therefore limits have

-been established on the bases of the AXIAL POWER IMBALANCE produced by the power peaking.

It has been determined that the above hot channel factor limits will be met provided the following conditions are maintained.

1.

Control rods in a single group move together with no individual rod insertion differing by more than i 6.5% (indicated position) from the group average height.

2.

Regulating rod groups are sequenced with overlapping groups as required in Specification 3.1.3.6.

3.

The regulating rod insertion limits of Specification 3.1.3.6 and the axial power shaping rod insertion limits of Specification 3.1.3.9 are maintained.

4.

AXIAL POWER IMBALANCE limits are maintained.

The AXIAL POWER IMBALANCE is a seasure of the difference in power between the top and botton halves of the core. Calculations of core average axial peaking factors for many plants and measurements from operating plants under a variety of operating conditions have been correlated with AXIAL POWER IMBALANCE.

~

The correlation shows that the design pwer shr;,e is not exceeded if the AXIAL POWER IMBALANCE is maintained witein M e limits of Figures 3.2-1, g

and 3.2-2.

The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to mininua allowable control rod insertion and are the core DNBR design basis.

Therefore, for operation at a fraction of RATED THERMAL POWER, the design limits are net.

When using incore detectors to make power distribution maps to determine Fg andFfH' a.

The measurement of total peaking factor, FgMeas, shall be increased by 1.4 percent to account for manufacturing tolerances and further increased by 7.5 percent to account for measurement error.

l CRYSTAL RIVER - UNIT 3 B 3/4 2-2 Amendment ilos. 75,79,77, 103

1 POWER DISTRIBUTION LIMITS BASES b.

The measurement of enthalpy rise hot channel factor, T{H, shall be increased by 5 percent to account for measurement error.

For Condition II events, the core is protected from exceeding allowable fuel melt limit locally, and from going below miniaua allowable DNBR by automatic protection on power, AXIAL POWER IMBALANCE, pressure and temperature.

Only conditions 1 through 3, above, are mandatory since the AXIAL POWER IMBALANCE is an explicit input to the Reactor Protection System.

The QUADRANT POWER TILT limit assures that the radial power distribution satisfies the design values used in the power capability analysis.

Radial power

. distribution acasurements are made during startup testing and periodically during power operation.

For QUADRANT POWER TILT, the safety (measurement independent) limit for' Steady State is 4.92, for Transient State is 11.07, and l

for the Maximum Limit is 20.0.

The QUADRANT POWER TILT limit at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts.

The limit was selected to provide an allowance for the uncertainty associated with the power tilt.

In the event the tilt is not corrected, the margin for uncertainty on Fg is reinstated by reducing the power by 2 percent for each percent of tilt in excess of the limit.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.

The limits are-consistent with the FSAR initial assumptions and have been analytically demonstrated adequate to maintain a DNBR within the limit throughout each analyzed transient.

l The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with acasured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis.

)

1 CRYSTAL RIVER - UNIT 3 8 3/4 2-3 Amendment Nos. 19,77. 103 l

DESIGN FEATURES DESIGN PRESSURE AND TDIPERATURE 5.2.2 The Reactor Containment building is designed and shall be maintained for l

a maximum internal pressure of 55 psig and a temperature of 281'F.

l 5.3 REACTOR CORE 1

FUEL ASSDIBLIES 5.3.1 The reactor core shall contain 177 fuel assemblies with each fuel assembly containing 208 fuel rods clad with Zircaloy - 4.

Each fuel rod shall have a nominal active fuel length of 144 inches and contain a maximum total weight of 2253 grams uranium.

The initial core loading shall have a maximum enrichment of 2.83 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.0 (nominal) weight percent U-235.

CONTROL RODS 5.3.2 The reactor core shall contain 60 safety and regulating (including extended life control rods) and 8 axial power shaping (APSR) control rods.

Except for the extended life control rods, the safety and regulating control rods shall contain a nominal 134 inches of absorber material.

The extended life control rods shall contain a nominal 139 inches of absorber material.

The nominal values of absorber material l

shall be 80 percent silver, 15 percent indium, and 5 percent cadmium.

I Except for the extended life control rods, all ecntrol rods shall be clad with stainless steel tubing.

The extended lifs control rods shall be clad with Inconel.

The APSRs shall contain a weni. f 3 inches of absorber material at their lower ends.

The absorber material for the l

APSRs shall be 100\\ Inconel.

1 l

l j

l CRYSTAL RIVER - UNIT 3 5-4

.91endment !Ios. 76,25,72,77,92, 103 l

- _ _ - _ - -.

ML20237E421

Text

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