Text

Amend 77 to License DPR-72,revising Tech Specs to Support Cycle 6 Operation
	ML20132E261
Person / Time
Site:	Crystal River
Issue date:	07/16/1985
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
ML20132E265	List: ML20132E261; ML20132E290;
References
	DPR-72-A-077 NUDOCS 8508010756
	Download: ML20132E261 (73)
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UNITED STATES yie g

NUCLEAR REGULATORY COMMISSION s ;

j WASHINGTON, D. C. 20555

/

FLORIDA POWER CORPORATION CITY OF ALACHUA CITY OF BUSHNELL CITY OF GAINE5VILLE CITY OF KISSIMMEE CITY OF LEESBURG CITY OF NEW SMYRNA BEACH AND UTILITIES COMMISSION, CITY OF NEW SMYRNA BEACH CITY OF OCALA ORLANDO UTILITIES COMMISSION AND CITY OF ORLANDO SEBRING UTILITIES COPNISSION SEMINOLE ELECTRIC COOPERATIVE, INC.

CITY OF TALLAHASSEE DOCKET NO. 50-302 CRYSTAL-RIVER UNIT 3 NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 77 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 April 25, 1985, 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 application, the provisions of the Act, and the rules and regulations of the Comission; 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 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.

A 302 p

PM

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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. OPR-72 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 77, 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 Joh f' F. Stolz, Thief Op rating Reactors Br ch #4 vision of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: July 16, 1985

_______o

e I

l ATTACHMENT TO LICENSE AMENDMENT NO. 77 FACILITY OPERATING LICENSE NO. DPR-72 DOCKET NO. 50-302 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The 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, faSe 2-3 3/4 2-6 2-7 3/4 2-11 B 2-1 3/4 3-6 8 2-2 3/4 3-7 8 2-5 3/4 3-8 B 2-6 3/4 3-19 8 2-8 3/4 3-36 3/4 1-14 3/4 3-39 3/4 1-16 3/4 4-4a 3/4 1-19 3/4 4-9 3/4 1-25 3/4 5-4 3/4 1-27 3/4 5-5 3/4 1-27a 3/4 7-2 3/4 1-28 3/4 7-25 3/4 1-28a 3/4 7-35 3/4 1-29 8 3/4 1-2 3/4 1-29a B 3/4 2-1 3/4 1-30 B 3/4 2-2 3/4 1-31 8 3/4 2-3 3/4 1-34 8 3/4 7-1 3/4 1-37 5-4 3/4 1-38 i

3/4 1-38a 3/4 1-39 3/4 1-40 3/4 2-1 3/4 2-2 3/4 2-2a 3/4 2-3 3/4 2-3a (New page) 3/4 2-4 j

o Figure 2.1-2 REACTOR CORE SAFETY LIMIT

- - 120

(-33.8,112)

(31.112)

- 110 Acceptable 4 Pump

(-48.5,99.6)

Operation

- 100

(-33.8,89.6)

__ 90 (31,89.63 (48.2,85.2) 0

(-48.5,77.2) 3 & 4 Pump

- 70 Operation g-2 (48.2.62.8)

- 60

--i 5-- 50 3

E-- 40 g.

30 e

$-- 20 W

Z

- 10 a

a a

a i

f I

I I

I

-60

-50

-40

-30

-20

-10 0

10 20 30 40 50 60 Axial Power Imbalance, %

i Amendments Nos. Jg, 7), M, /1, CRYSTAL RIVER UNIT 3 2-3 pg, pg, $;. 77

l FIGURE 2.2-1 TRIP SETPOINT FOR NUCLEAR OVERPOWER BASED ON RCS FLOW AND AXIAL POWER IMBALANCE

(-17,108)

,~.t 10 (17,108)

My = 1.0 100 Acceptable M2 = -1.815 4 Pump

(-34.7,90.3 Operation

- 90

(-17,80.67)

(17,80.67) ev (34.7,75.86)

Acceptable

- 70 3 & 4 Pump y

(-34.7,62.97)

Operation o

60

~~

(34.7.48.53)

T

,% -- 40 D4 s-- 30 m

20

8. - -

3

%E- - 10 m

a i

i i

1 1

g g

-60

-50

-40

-30

-20

-10 0

10 20 30 40 50 60 Axial Power Imbalance, %

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I I' N' N' g," 7y CRYSTAL RIVER UNIT 3 2-7

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 and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented 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 and, therefore, THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the BAW-2 DNB correlation.

The DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. 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 minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is !!mited to 1.30. This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curve presented in Figure 2.1-1 represents the conditions at which a DNBR of 1.30 or greater is predicted for the maximum possible thermal power 112% when the reactor coolant flow is 139.7 x 106 lbs/hr, which is 106.5% of the design flow rate for four operating reactor coolant pumps. This curve is based on the following nuclear power peaking factors with potential fuel densification effects:

F' = 2.82 F

= 1.71 F = 1.65 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 Amendments Nos. J$, JL R,77 l

~-- - - -

A

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SAFETY LIMITS SA5E5 The reactor trip envelope appears to approach the safety limit more closely than 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 margin to the safety limit.

2 The curves of. Figure 2.1-2 are based on the more restrictive of two thermal limits and account for the effects of potential fuel densification and potential fuel rod bow:

N 1.

The 1.30 DNBR limit produced by a nuclear power peaking factor of F

= 2.82 l

or the combination of the radial peak, axial peak and position of the axkl peak that yields no less than a 1.30 DNBR.

i 2.

The combination of radial and axial peak that causes central fuel melting at i

the hot spot. The limit is 20.5 kw/f t.

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 minimum 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-maximum thermal power combinations shown in BASES Figure 2.1. The curves of BASES Figure 2.1 represent the conditions at which a minimum DNBR of 1.30 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation.

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

The DNBR as calculated by the BAW-2 DNB correlation continually increases from point of minimum DNBR, so that the exit DNBR is always higher.

Extrapolation of the correlation beyond its p~ublished quality range of 22% is j stified on the basis of u

experimental data.

1 1

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Amendments Nos. J$, J/, /f, 77 CRYSTAL RIVER - UNIT 3 B 2-2 n

e s

LIMITING SAFETY SYSTEM SETTINGS BASES RCS Outlet Temperature - High The RCS Outlet Temperature High trip less than or equal to 613 prevents the reactor outlet, temperature from exceeding the design limits and acts as a backup trip for a!! power excursion transients.

I Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE The power level trip setpoint produced by the reactor coolant system flow is based on a flux-to-flow ratio which has been established to accommodate flow decreasing transients from high power.

1 The power level trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level setpoint produced by the power-to-flow ratio provides overpower DNB protection for all modes of pump

)

operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.2-1 are as follows:

1.

Trip would occur when four reactor coolant pumps are operating if power is greater than or equal to 108% and reactor flow rate is 100%, or flow rate is less than or equal to 95.88% and power level is 100%

l 2.

Trip would occur when three reactor coolant pumps are operating if power. is greater than or equal to 80.67% and reactor flow rate is 74.7%, or flow rate is less than or equal to 69.44% and power is 75%

For safety calculations the maximum calibration and instrumentation errors for the power level were used.

i l.

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Amendments Nos. Jf, 17, J/, JJ, )),

CRYSTAL RIVER - UNIT 3 B 2-3 ff. If,77 i

4 1

s LIMITING SAFETY SYSTEM SETTINGS BASES The AXIAL POWER IMBALANCE boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kw/f t limits or DNBR limits. The AXIAL POWER IMBALANCE reduces the power level trip the flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced.

produced by, flow ratio reduces the power level trip and associated reactor power-reacto The flux-to-power-imbalance boundaries by 1.08% for a 1% flow reduction.

l RCS Pressure - Low, High, and Variable Low The High and Low trips are provided to limit the pressure range in which reactor operation is permitted.

j During a slow reactivity insertion startup accident from low power or a slow reactivity insertion from high power, the RCS Pressure-High setpoint is reached before the Nuclear Overpower Trip Setpoint. The trip setpoint for RCS Pressure-High, 2300 psig, has been established to maintain the system pressure below the safety limit, 2750 psig, for any design transient. The RCS Pressure-High trip is backed up by the pressurizer code safety valves for RCS over pressure protection is therefore, set lower than the set pressure for these valves,2500 psig. The RCS Pressure-High trip also backs up the Nuclear Overpower trip.

The RCS Pressure-Low,1800 psig, and RCS Pressure-Variable low,(11.59 Tout F -5037.8) psig, Trip Setpoints have been established to maintain the DNB ratio greater than or equal to 1.30 for those design accidents that result in a pressure reduction. It also prevents reactor operation at pressures below the valid range of DNS correlation limits, protecting against DNB.

Due to the calibration and instrumentation errgrs, the safety analysis used a RCS Pressure-Variable Low Trip Setpoint of (11.59 Tout F -5077.8) psig, i

3 4

Amendments Nos. J$, J), A/, JJ, $$,

II' N' CRYSTAL RIVER - UNIT 3 B 2-6

t a

LIMITING SAFETY SYSTEM SETTINGS BASES 1

Reactor Containment Vessel Pressure - High The Reactor Containment Vessei Pressure-High Trip Set oint 6 4 psig, provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the containment vessel or a loss-of-coolant accident, even in the absence of a RCS Pressure -

Low trip.

Reactor Coolant Pumo Power Monitors In conjunction with the power / imbalance / flow trips, the Reactor Coolant Pump Power Monitors trip prevents the minimum core DNBR from decreasing below 1.30 by tripping the reactor due to more than one reactor coolant pump not operating.

A reactor coolant pump is considered to he not operating when the power required By the pump is greater than or equal to 2G2% (14,400 kw] or is less than or equal to 20.9% (1152 kw) of the operating power (.5500 kw). In order to avoid spurious trips during.nomal operation, the trip setpoints have been selected to maximize the operating band while assuring that a reactor trip will occur upon loss of power to the pump. The 20.9% trip setpoint and response time are based on the maximum time within which an RCPPN-RPS trip must occur to provide CNBR protection for the four pump coastdown. Florida Power fias agreed to take credit for the pump i

overpower trip in order to assure that certain potential faults (such as a seismically induced fault high signall will not prevent this instrumentation from providing the protective action (i.e. a trip signal). Thus, the maximum setting, approximately 262% (14,400 kwl, was selected.

J CRYSTAL RIVER - UNIT 3 8 2-7 Amendment No. 16, 17, H, 35, 64 l

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t n.--..

BASES FIGURE 2.1 PRESSURE / TEMPERATURE LIMITS AT MAXIMUM ALLOWABLE POWER FOR MINIMUM DNBR 2400 CURVE 2 2200 3 PUMP S

E b

E E

y 2000 CURVE 1 5

4 PUMP u

8 1800 f

f 580 600 620 640 REACTOR OUTLET TEMPERATURE, F REACTOR COOLANT FLOW PUMPS OPERATING CURVE FLOW (% OESIGN)

POWER (RTP)

(TYPE OF LIMIT) 1 139.7 x 106(106.5%)

112%

4 PUMPS (DNBR) 2 104.4 x 10 ( 79.6%)

89.6%

3 PUMPS (DNBR) l CRYSTAL RIVER UNIT 3 B 2-8 Amendments Nos.17, 32, A7, 77

7 REACTIVITY CONTROL SYSTEMS l

BORIC ACID PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.7 At least one boric acid pump in the boron injection flow path required by

^

Specification. 3.1.2.2a shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow path through the boric acid pump in Specification 3.1.2.2a is OPERABLE.

APPLICABILITY:

MODES 1,2,3 and 4 ACTION:

With no boric acid pump OPERABl.E, restore at least one boric acid pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 1% delta k/k at 2000F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least one boric acid pump to OPERABLE status with the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I SURypn i ANCE REQUIREMENTS 4.1.2.7 No additional Surveillance requireme'nts other than those required by Specification 4.0.3.

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CRYSTAL RIVER - UNIT 3 3/4 1-13 Amendment !!o. 32. H, 64 i

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 As a minimum, one of the following borated water sources shall be OPERABLE:

a.

A concentrated boric acid storage system and associated heat tracing with:

1.

A minimum contained borated water volume of 600 gallons, 2.

Between 11,600 and 14,000 ppm of boron, and 3.

A minimum solution temperature of 103 F.

b.

The borated water storage tank (BWST) with:

1.

A minimum contained borated water volume of 13,300 gallons, 2.

A minimum boron concentration of 2,270 ppm, and 3.

A minimum solution temperature of 40'F.

APPLICABILITY:

MODES 3 and 6.

ACTION:

With no borated water sources OPERABLE, suspend all operations involving CORE ALTERATION or positive reactivity changes until at least one borated water source is restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.8 The above required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1.

Verifying the boron concentratic.n of the water, 2.

Verifying the contained borated water volume of the tank, and CRYSTAL RIVER - UNIT 3 3/4 1 14

[

.l

,s REACTIVITY CONTROL SYSTEMS

~

e SURVEILLANCE' REQUIREMENTS (Continued) 3.

Verifying the concentrated beric acid storage system

/

selution temperature when it is the source.of borated water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the BWST temocrature when'it is the source of borated water and the outside air l

temperature is < 40*F.

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1 CRYSIAL RIVER. UNIT 3 3/4 1 15 8

O O do 0 9 p 9 A... __ _ _

f. a a t...f a D- @t.1 e

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1 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.9 Each of the following borated water sources shall be OPERABLE a.

A concentrated boric acid storage system and associated heat tracing with:

1.

A minimum contained borated water volume of 6,000 gallons, I

2.

Between !!,600 and 14,000 ppm of boron, and 3.

A minimum solution temperature of 103 F.

b.

The borated water storage tank (BWST) with:

I 1.

A minimum contained borated water volume of 413,200

gallons, 2.

Between 2,270 and 2,430 ppm of boron, and 3.

A minimum solution temperature of 40 F.

APPLICABILITY:

MODES 1,2,3 and 4 ACTION:

a.

With the concentrated boric acid storage system inoperable, restore the storage system to OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY i

and borated to a SHUTDOWN MARGIN equivalent to 1% delta k/k at 2000F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the concentrated boric acid storage system to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With the borated water storage tank inoperable, restore the tank to OPERABLE status within one hour 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 in HOT SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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Amend ants Nos. 8, 15, 19, J/,

CRYSTAL RIVER - UNIT 3 3/4 1-16 N ' N 'II 4

REACTIVITY CONTROL SYSTEMS i

ACTION: (Continued) c)

A power distribution map is obtained from the incore detectors and q and Fp are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and F

d)

The THERMAL POWER level is reduced to d60% of the THERMAL POWER allowable for the reactor coolant pump combination within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the Nuclear Overpower Trip Setpoint is reduced to s 70% of the THER!tR PouER allowabic for the reactor coolant pump combination, or e)

The remainder of the rods in the group with the inoperable rod are aligned to within 3 5% of the inoperable rod within one hour while 6

maintaining the rod sequence, insertion and overlap limits of Figures 3.1-1, 3.1-2, 3.1-3, 3.1-4, 3.1-3 and 3.1-6; the THERMAL POWER level shall be resricted pursuant to Specification 3.1.3.6 during subsequent operation.

l SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each control rod shall be determined to be within the group i

l average height limits by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Asymmetric Rod Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each control rod not fully inserted shall be determined to be OPERABLE by movement of at least 3% in any one direction at least once every 31 days.

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CRYSTAL RIVER - UNIT 3 3/41-19 Amendment No. f),77 i

4 I

REACTIVITY CONTROL SYSTEMS GROUP HEIGHT - AXIAL POWER SHAPING R00 GROUP LIMITING CONDITION FOR OPERATION 2

3.J.3.2 All axial power shaping rods (APSR) shall be OPERABLE, unless fully withdrawn, and shall be positioned within + 6.5% (indlcated position) of their group average height.

APPLICABILITY: MODES 1* and 2*.

ACTION:

With a maximum of one APSR inoperable or misaligned from its group average height by more than + 6.5% (indicated position), operation may continue provided that within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

a.

The APSR group is positioned such that the misaligned rod is restored to within limits for the group average height, or b.

It is determined that the imbalance limits of Specification i

3.2.1 are satisfied and movement of the APSR group is pre-vented while the rod remains inoperable or misaligned.

1 SURVEILLANCE REQUIREMENTS' I

i I

4.1.3.2.1 The position of each APSR rod shall be determined to be within the group average height limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Asynnetric Control Rod Monitor is inoperable, then verify the indi-vidual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.2.2 Unless all APSR are fully withdrawn, each APSR shall be determined to be OPERA 8LE by..oving the individual rod at least 3%

at least once every 31 days.

i

See Special Test Exceptions 3.10.1 and 3.10.2.

i CRYSTAL RIVER - UNIT 3 3/4 1-20 f

a

REACTIVITY CONTROL SYSTEMS REGULATING ROD INSERTION LIMITS 4

LIMITING CONDITION FOR OPERATION 3.1.3.6 The regulating rod groups sha!! be limited in physical insertion as shown on Figures 3.1-1,.3.1-la, 3.1-2, 3.1-3, 3.1-3a, and 3.1-4, with a rod group overlap of 25 1 5%

i between sequential withdrawn groups 5 and 6, and 6 and 7.

APPLICABILITY:

MODES l' and 2*#

ACTION:

With the regulating rod groups inserted beyond the above insertion limits, or with any group sequence or overlap outside the specified limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:

a.

Restore the regulating groups 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 figures 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 .

See Special Test Exceptions 3.10.1 and 3.10.2.

With Keff greater than or equal to 1.0.

CRYSTAL RIVER - UNIT 3 3/4 1-25 Amendments Nos. /%, $#,77

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REACTIVITY CONTACL S b dS REGULATING ROD INSERTICN LIMITT SURVEILLANCE REQUIREMENTS 4.1.3.5 The position of each regulating grcup shall be detennined to te 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 red insertion limit alann is incoerable then a.

verify the groups to be within the insertien 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 incperable, then verify the groups to be within the sequence and overlap limits at least once per 4 hcurs.

l CRYSTAL RI'IER - UNkT 3 3/2 ijs W

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FIGURE 3.1-1 REGULATING R00 GROUP INSERTION LIMITS FOR FOUR PUMP OPERATION FROM 0 EFPD TO 200 + 10 EFP0 l

)

110 (273, 100 (300,102) 90 (267,92 6

3 80 2

(250,80) 70

.E UNACCEPTABLE i

60 OPERATION T; 50 (175,50)

" 40

+

u" i

l 30 ACCEPTABLE OPERATION 20 (32,15) 10 ll (0,5) 0 i

i i

i 1

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

25 50 75 100 0

25' 50 75 100 t

i i

1 1

i i

i Group 5 Group 7 0

25 50 75 100 I

i 1

1 1

1 Group 6 4

i CRYSTAL RIVER UNIT 3 3/4 1-27 Amendments Nos. J, 7 Jp,17, )/,

34,#7,$#,77

o FIGURE 3.1-la REGULATING R00 GROUP INSERTION LIMITS FOR FOUR PUMP OPERATION FROM 200 110 TO 400 110 EFPD 110 (73,102) 100 (300,102)

(267,92) 90 i.,

f 80 2

(250,80)

L'NACCEPTABLE

[ 70 OPERATION

$ 60 ~

? E 175,50) 40 -

L f30 ACCEPTABLE 20 -

OPERATION 58,15) 10 -

0,5)

O O

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

25 50 75 100 0

25 50 75 100 t

i i

Group 5 Group 7 0

25 50 75 100 t

i i

Group 6 CRYSTAL RIVER UNIT 3 3/4 1-27a Amendments Nos. A$, $$,77

FIGURE 3.1-2 REGULATING R00 GROUP INSERTION LIMITS FOR FOUR PUMP OPERATION AFTER 400 +10 EFPD 110-(265,102)

(300,102) 100 -

(260,92) 90 250,80) 2 UNACCEPTABLE 70 OPERATION E 60 -

T 50 -

j (175,50) w 40 -

i.

g 30 20 ACCEPTABLE

~

OPERATION 10 58,15) 0.5) 0 O

50 100 150 200 250 300 Rod Index, 7 Withdrawn 0

25 50 75 100 0

25 50 75 100 t

i 1

i i

t i

i i

i Group 5 Group 7 0

25 50 75 100 t

i i

Group 6 CRYSTAL RIVER UNIT 3 3/4 1-28 Amendments flos. If,17, y, il,PC,if,77

I a

1 i

)

-DELETED-CRYSTAL RIVER UNIT 3 3/41 28a Amendment No, (f, 77

s

.\\

1, j

/

FIGURE 3.1-3 REGULATING R00 GROUP INSERTION LIMITS FOR THREE PUr OPERATION FROM 0 TO 200 +10 EFPD -

l s

,,i

??

s

,y

'1 110 100 < -

s 90 80,

i 2

'(275.77)

(30017) t 2 70 c

UNACCEPTALLE (267,69) s k 60 OPEWION

,.\\

T (250,60) j 50 -

w a

4 g 40 s

g 175,37.5)

= 30 (s.,

20

.ACCJFTABLE 10 (32,11.75)

OPERATION 0,4.25)

N O

1 i

i i

1 0

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

2,5 5,0 75 100 g

35 j0 7,5 100 Group 5 Group 7 s

J j

Group 6

\\.

J l

CRYSTAL RIVER UNIT 3 3/4 1-29 Amendment 1 Nos. If,17. Ji,3$. H., y 77 I'

.+

/

V F

FIGURE 3.1-3a REGULATING R00 GROUP INSERTION LIMITS FOR c

THREE PUMP OPERATION FROM 200 110 to 400 110 EFPD

'\\

,/c I

110 100 -

90 80 (275,77)

(300,77)

' 70 UNACCEPTABLE (267,69)

OPERATION g

(250,60) g 50 40

('. '

175,37.5) g30 t

o

" 20 ACCEPTABLE OPERATION 58,11.75) 10 0

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

25 50 75 100 0

25 50 75 100 t

i i

t 1

i i

1 Group 5 Group 7 st 6

0

{5 j0 7,5 1,00 Group 6 Amendments Hos. $$, $#, 77 t

B YSTAL RIVER UNIT 3 3/4 1-29a i

~

FIGURE 3.1-4 REGULATING ROD GROUP INSERTION LIMITS FOR THREE PUMP OPERATION AFTER 400 +10 EFPD 110 100 90 g 80 (267,77) g

- (300,77) 70 (260,69)

E y 60 (250,60) 3 50 UNACCEPTABLE OPERATION a

w 40 s'

y 30 (175,37.5) e 20 ACCEPTABLE OPERATION 10 (58,11.75)

( 0,4.2,5) 0 50 100 150 200 250 300 Rod Index ". Withdrawn 0

25 50 75 100 0

25 50 75 100 1

I I

t I

f I

I Group 5 Group 7 1

0 25 50 75 100 t

I i

i i

Group 6 CRYSTAL RIVER UNIT 3 3/4 1-30 Amendments Nos. 75, 77, 37, Af,

$$, 77

-DELETED-

)

\\

l l

CRYSTAL RIVER UNIT 3 3/4 1-31 Amendments Nos, Jp, pp,77

a 1

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

APPt.!CABILITY: MODES 1* and 2*.

ACTION:

With any control red not progranned to operate as specified above, be in HOT STAN08Y within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

SURVEILLANCE REQUIRE"ENTS 4.1.3.7 E.tch control rod shall be demonstrated to be programmed to 4.

operate in the specified core position and red group by:

1.

Selection and actuation front the control room and verifi-cation of movement of the proper rod as indicated by both the absoluta and relative position indicators:

a)

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

b)

For specifically affected individual rods, following maintenanca, test, reconnection or modification of power er instrumentation cables from the control rod drive control system to the control roa drive.

1 2.

Verifying that esca cable that has been disconnected has been properly matched and reconnected to the specified control rod drive.

b.

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

See Soecial Test Excacticas 3.10.1 and 3.10.2.

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

+

FIGURE 3.1-7 CONTROL R0D LOCATIONS AND GROUP DESIGNATIONS FOR CRYSTAL RIVER 3 CYCLE 6 I

I Fuel Transfer

[

Canal A

B 1

6 1

C 2

5 5

2 0

7 8

7 E

2 5

4 4

5 2

7 1

8 6

3 6

8 1

G S

4 3

3 4

5

-Y H

w-6 7

3 3

7 6

K 5

4 3

3 4

5 L

1 8

6 3

6 8

1 M

2 5

4 4

5 2

i N

l 7

8 7

0 l

l 2,

5 5

2 P

l l

l 1

6 1

R l

l l

8 i

l i I

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15

)

1 Group Nureer l

Grouc No. of Rods Function 1

8 Safety j

2 8

Safety 3

8 Safety 4

8 Safety 5

12 Control 6

8 Control 7

8 Control 8

8 APSRs Total 66 i

CRYSTAL RIVER UNIT 3 3/4 1-34 Amendments Nos. 19. J7, pf, 77

REACTIVITY CONTROL SYSTEMS AXIAL 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. Up to 390 EFPD, the APSR's may be positioned as necessary. The APSR's shall be completely withdrawn (100%) by 410 EFPD. Between 390 and 410 EFPD, the APSR's may be withdrawn. However, once withdrawn during this period, the APSR's shall not be reinserted.

APPLICABILITY: MODES I and 2*.

ACTION:

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

Restore the axial power shaping rod group to within the limits within 2 a.

hours, 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 REOUIREMENTS 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 21.0.

CRYSTAL RIVER - UNIT 3 3/4 1-37 Amendments Nos. J$, 77, f$, $f,

.,.-3 y--.--

....i..

,m w

w-

-,.y

4 9

a O

P

-DELETED-CRYSTAL RIVER UNIT 3 3/4 1-38 gendmentsNos.JE,17,77,M,H.

-DELETED-CRYSTAL RIVER UNIT 3 3/41-38a Amendirents Nos. JS, $9,77

-DELETED-

'afSTAL RIVER UNIT 3 3/4 1-39 Amendments Nos.17, 37, pg, g;, 77

o

-DELETED-1 CRYSTAL RIVER UNIT 3 3/4140 Amendment No. $f, 77 i

i

i 3/4.2 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, 3.2-la, 3.2-2, and 3.2-2a.

(

APPLICABILITY: MODE I 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 REOUIREMENTS 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 40% of RATED THERMAL POWER except when an AXIAL POWER IMBALANCE monitor is inoperable, then calcula,te the AXIAL POWER IMBALANCE in each core quadrant with an inoperable monitor at least once per hour.

See Special Test Exception 3.10.1.

i l

CRYSTAL RIVER - UNIT 3 3/42-1 Amendment No. #,77

._=.

FIGURE 3.2-1 AXIAL POWER I W LANCE ENVELOPE FOR FOUR PUMP OPERATIt)N FROM 0 TO 400 +10 EFPD

"" 110

(-23.102)

(15.102)

,,,y

(-25,92)

(15,92)

. 90

(-26,80)

- 80 (20,80)

- 70 g.- 60 2

(-28,50) j- - 50 (20,50) 2

. 40 7.

ACCEPTABLE E

UNACCEPTABLE OPERATION

".- 30 OPERATION

u I2 20

- 10 1

L t

t t

t l

-40

-30

-20

-10 0

10 20 30 40 Axial Power Imbalance, ;

i l

l CRYSTAL RIVER UNIT 3 3/4 2-2 Amendments iios. 7, E, 7%, 17, JE,

/$,3/,77

FIGURE 3.2-la AXIAL POWER IMBALANCE ENVELOPE FOR FOUR-PUMP OPERATION AFTER 400 +10 EFPD

)

l 1

110

(-21.102)

(15,102)

.. vu

(-25.92 (15,92) 90

(-28,80)

- 80 (20,80) 70 5-60 a.

(-28,50)

j. - 50 (20,50)

.2 T,.- 40 ACCEPTABLE 3

OPERATION UNACCEPTABLE OPERATION 30 L

f 2.- 20 10 f

I

-40

-30

-20

-10 0

10 20 30 43 Axial Power Imbalance, %

CRYSTAL RIVER UNIT 3 3/4 2-2a Amendments Nos. f$, $$,

77

FIGURE 3.2-2 AXIAL POWER IMBALANCE ENVELOPE FOR THREE PUMP OPERATION FROM 0 TO 400 +10 EFPD

,,100

.90

.80

(-20.67)

(15,77)

~

(-25,69)

(15,69) 40 (20*60)

(-26.60) io

.50

(-28,37.S'

[- 40 (20.37.5) 2 3

.30 w

20 ACCEPTABLE f

UNACCEPTABLE OPERATION OPERATION 10 i

i e

i i

-40

-30 20

-10 0

10 20 30 40 Axial Power Imbalance, 7, CRYSTAL RIVER UNIT 3 3/4 2-3 Amendments flos Jp, Jp, 32,

$1, 99, $$, 77

_.i

FIGURE 3.2-2a AXIAL POWER IMBALANCE FOR THREE PUMP OPERATION AFTER 400 +10 EFPD

- 100

. 90

- 80

(-20.67)

(15,77)

~

(-25,69)

(15,69)

(-28,60) 60 (20,60) h 2-- 50

(-28,37.5) g-- 40 (20.37 5)

Yu

,% -- 30 d.. 20 ACCEPTABLE OPERATION !

UNACCEPTABLE a.

OPERATION

- 10 a

i f

I e

l a

-40

-30

-20

-10 0

10 20 30 40 Axial Power Imbalance, 1 GYSTAL RIVER UNIT 3 3/4 2-3a Amendment No. 77

POWER DISTRIBUTION LIMITS NUCLEAR HEAT FLUX HOT CHANNEL FACTOR - FQ LIMITING CONDITION FOR OPER ATION 3.2.2 FQ shall be limited by the following relationships:

i FQ f 3.13 i

T I

where P =

THERM AL POWER and P & 1.0 RATED THERMAL POWER APPLICABILITY: MODE 1 ACHON:

j With FQ exceeding its limit:

Reduce THERMAL POWER at least 1% for each 1% F a.

Q exceeds the limit within 15 minutes and similarly reduce the Nuclear Overpower Trip Setpoint and Nuclear Overpower based on RCS Flow and AXIAL POWER IMBALANCE Trip Setpoint within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

Demonstrate through in-core mapping that Fo is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5%

of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Identify and correct the cause of the out of limit condition prior to increasing c.

THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that FQ is demonstrated through in-core mapping to be within its limit at a nominal 50%

of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THEP. MAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater i

RATED THERMAL POWER.

SURVEILLANCE REOUIREMENTS 4.2.2.1 FQ shall be determined to be within its limit by using the incore detectors to obtain a power distribution map:

i i

CRYSTAL RIVER - UNIT 3 3/42-4 Amendment No. 7,77 4

-,n.

0 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) a.

Prior to initial operation above 75 percent of RATED THERMAL POWER after each fuel loading, and b.

At least once per 31 Effective Full Power Days, The provisions of Specification 4.0.4 are not applicable.

c.

4.2.2.2 The measured F of 4.2.2.1 above, shall be increased by 1.4%

n to account for manufactaring tolerances and further increased by 7.5%

to account for measurement uncertainty, i

r 2

CRYSTAL RIVER - UNIT 3 3/4 2-5

POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F[g LIMITING CONDITION FOR OPERATION 3.2.3 F

shall be limited by the following relationship:

I H

(1.71 h + 0.3 (1-Pf F

Thermal Power p

Rated Thermal Power and Pf1.0 APPLICABILITY:

MODE 1.

ACTION:

With F[H exceeding its limit:

a.

Reduce THERMAL POWER at least 1% for each 1% that F exceeds the limit within 15 minutes and similarly reduce the Nuclear verpower Trip j

Setpoint and Nuclear Overpower based on RCS Flow and the AXIAL POWER IMBALANCE Trip Setpoint within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Demonstrate through in-core mapping that F$ g is within its limit within 24

,i b.

l hours after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

L 4

c.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that F[g is demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter attaining 95% or greater RATED THERMAL POWER.

1 k

CRYSTAL RIVER - UNIT 3 3/42-6 Amendment No. J$,77

TABLE 3.2-2 OUADRANT POWER TILT LIMITS i

STEADY STATE TRANS!ENT MAXIMUM l

LIMIT LIMIT LIMIT QUADRANT POWER TILT as Measured by:

Symmetrical Incore Detector System 3.20 9.08 20.0 1

i Power Range Channels 1.61 6.96 20.0 l

Minimum Incore Detector System 1.73 4.40 20.0 I

i j

1 1

h l

CRYSTAL RIVER - UNIT 3 3/4 2-11 Amendments Nos. J$, Jp, y,77

~

o e

POWER O!STRIBUTION LIMITS CNB PARAMETERS LIMITING CCN0! TION 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 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 enan 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 REQUIREMENTS 4.2.5.1 Each of the parameters of Table 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 flew rate shall be datermined to be within its limit by measurement at least once per 18 months.

1 f

CRYSTAL RIVER - UNIT 3 3/4 2-12 l

l l

t

a

. With tha numb r,cf channels OPE?aBLE.cne less than recuired.

ACTION 5 by the Minimum Channels OPE?ABLE requirement and with the THERFAL POWER level:

a. " (10'N ames on the Inter nediate Range (IR) in-I rumentation, restore the inocerable channel to

~

OPERABLE gatus prior to inct, easing THERMAL PCWER above 10 amps on the IR instrumentation.

> 10-10 amps en the IR instrumentation, opera:icn b.

may centinue.

With the number of channels CPERABLE cne less than re-ACTION 6 quired by the Minimum Channels OPERABLE requirer.ent, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 within one hour and at leas:

ence per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

With the number of OPERASLE channals one less than the-ACTION 7 Tc:al Number of Channels STARTUP and/or POWER OPERATION r ay preceed provided all of the felicwing conditions are '

satisfied:

a.

Within 1 hour:

flace the incperable char.nel in the :rinced i

1.

ccndition, or 2.

Remove power supplied.c the central rec trip davice associated with the inocerative enannel.

l b.

One additional channel may be bypassed for uc to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for sur'veillance testing per Specification 4.3.1.1, and the incperable channel above may be bypassed for uq to 30 minutes in anv 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period wnen necessary to test the tric breaker associatec with the icgic of the cnannel being tested per Specification 4.3.1.1.

The incoerable channel above may not be bypassed to test the logic of a channel 1

J of the trip system associated with the inoperacle channel.

.With the nu=ter of channels OPERABLE less than recuired ACTION 8 by the Minimum Channels OPEPJELE requi-sment, be in at least HOT STAND 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 25 With the nuntier of channels OPERABLE one less than the required itinimum Channels OPERABLE requirement, plant cperation may continue until the next required Channel Functional Tes: pro-vided the inoperable channei is placed in the tripped corscition within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

CRYSTAL RIVER-UNIT 3 3/4 3-5 Amendment fic. 55

TABLE 3.3-2

~

Q REACTOR PROTECTION SYSTEM INSTRUMENTATION RESPONSE TIMES d

9 r-Functional Unit Response Times

&<rn M

1.

Manual Reactor Trip Not Applicable C{

2.

Nuclear Overpower

60.266 seconds 3.

RCS Outlet Temperature - liigh Not Applicable 4.

Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE

61.842 seconds 5.

RCS Pressure - Low d 0.44 seconds 6.

RCS Pressure - High 5 0.44 seconds 3

7.

Variable Low RCS Pressure Not Applicable y

w 8.

Pump Status Based on RCPPMs*

51.44 seconds 9.

Reactor Containment Pressure - liigh Not Applicable

. E,b*

Oh Neutren detectors are exempt from response time testing. Response time of the neutron flux signal a

y portion of the channel shall be measured from detector output or input of first electronic component 0

in channel.

2:

8

Time response testing of the RCPPMs may exclude testing of the current and voltage sensors and the watt transducer.

D e

TABLE 4.3-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS g

~

w d

-i>

CHANNEL MODES IN WHICII r-CilANNEL CilANNEL FUNCTIONAL SURVEILLANCE 3

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

$w c

1.

Manual Reactor Trip N.A.

N.A.

S/U(!)

N.A.

f 2.

Nuclear Overpower S

D(2) and Q(7)

M 1, 2 u

3.

RCS Outlet Temperature--High S

R M

1, 2 4

Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE 5(4)

M(3) and Q(7, 8)

M 1, 2 5.

RCS Pressure--Low 5

R M

1, 2 6.

RCS Pressure--High 5

R M

1, 2 7.

Variable Low RCS Pressure S

R M

1, 2, 8.

Reactor Containment Pressure-High 5

R M

1, 2 Y

9.

Intermediate Range, Neutron Flux and Rate S

R(7)

S/U(IX5) 1, 2 and

10.

Source Range, Neutron Flux and Rate S

R(7)

S/U(IX5) 2, 3, 4 and 5 11.

Control Rod Drive Trip Breaker N.A.

N.A.

M and S/U(1) 1, 2 and

12.

Reactor Trip Module N.A.

N.A.

M 1, 2, and

13.

Shutdown Bypass RCS S

R M

2**,3**,4**,3**

Pressure--High F

I4.

Reactor Coolant Pump Power Monitors S

R(9)

M 1, 2 l

2 "a

$n O.

i TABLE 4.3-1 (Continued)

NOTATION With any control rod drive trip breaker closed.

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 ITP) is above 30% of RATED THERMAL POWER (RTP),

compare out-of-core measured AXIAL POWER IMBALANCE (APl ) to incore measured AXIAL POWER IMBALANCE o

(APl ) as follows:

i (APlo - api )

= Imbalance Error t

TP Recalibrate if the absolute value of the Imbalance Error is equal to or greater than 3.5%

(4)

AXIAL POWER IMBALANCE a,nd 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 CHANNEL CALIBRATION.

(8)

- Flow rate measurement sensors may be excluded from CHANNEL CALIBRATION.

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

(9)

Current and voltage sensors may be excluded from CHANNEL CALIBRATION.

i i

9 CRYSTAL RIVER - UNIT 3 3/43-8 Amendment No. H,77 i

, ~ _

4 i

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM. S._ INSTRUMENTATION SURVEILLANCE REQUIREMENTS n

m CHANNEL MODES IN WillCil y

CilANNEL CilANNEL FUNCTIONAL SURVEILLANCE r-FUNCTIONAL UNIT CilECK CALIBRATION TEST REQUIRED E<

3.

REACTOR BUILDING SPRAY G

a.

Reactor Building Pressure y

High-liigh coincident with HPI Signal S

R M(4) 1,2,3 h.

Automatic Actuation Logic N/A N/A M(1) (3) (5) 1,2,3 4.

OTHER SAFETY SYSTEMS a.

Reactor 13uilding Purge Exhaust g

Duct Isciation on liigh Radioactivity y

1.

Gaseous S

Q M

All Modes 2

b.

Steam Line Rupture Matrix 1.

Low SG Pressure N/A R

N/A 1,2,3 l

2.

Automatic Actuation Logic N/A N/A M(3) 1,2,3 3"a

.f c.

Emergency Feedwater ga 1.

MFW Pump Turbines A and B Control Oil Low S

R N/A 1,2,3 o*

3.

OTSG A or B Level w

Low-Low S

R N/A 1,2,3,4

r

==

e.

wg

=,es e

N 4.

l e

c

,.t e m e

m

~.

m m

=:

. -l w

N "n

- n

~

N

>.~. l m

n o

,.,.n 4

.e w

2l e

e

.m"'

'e"'

.. : u

- O p,

.r..

.m

==

4

<,j ;-

.s:

4 l

(,,,:,

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I l

e 4,

l 4

e.d

.-.u

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",.,I Z<

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i w

Z Z

1

< =.

n

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.c..

b.

z

.'t w

/,

m-

m. l.

=,

e.s c.

o Z==* 3 14

=.

% A

==*

a t.

~

s

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.C. fI

.. =

m.

e# g

..k!

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= -

=,

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Z

=i u -:. : u ":

g

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.= 5 w-

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e.

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g.:: =

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vt g = u = : :.,. _.

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3 o

CRYST.2d. RIV*_A - UNIT 3 3/4 MG Menc=ent ::o. 33, 37, 66 D

9 d.

m

^

~

o O;c 5

y n

TABLE 3.3-9 g

ItEMOTE SilU1DOWN MONITORING INSTRUMENTATION

o C

MINIMllM h

READOUT MEASUREMEN T CilANNELS LOCATION.

RANGE OPERABLE w

INS TitUMENT l.

Reactor Trip Breaker CRO switchgear room open-close i per trip breaker and 12's foot elevation I per secondary trip breaker Indica tion Reinote shutdown panel 320.6200F I per loop 2.

Reactor Coolant Temperature - Th Remote shutdown panel 0-2300 psig I

3.

Iteactor Coolant Pressure lleinote shutdown panel 0-320" 11 0 I

2

's.

I'vessurizer Level lle: note shutdown panel 0-1200 psig i per steam generator Y

w 3.

Steam Generator Pressure 6.

Steam Generator Level 4160 ES B switchgear 0-230"11 0 I per steam generator 2

room 108 foot elevation lle: note shutdown panel 0-3000 F l per cooler

{

7.

Decay lleat itemoval l empera tur e i!r Inscrmediale Building 0-2000 psig i per pump S

3.

Motor Driven Einergency 95 foot elevation

[

l'eedwater Pressure 9.

Nnclear Services Closed Auxiliary Bnilding 0-300 psig i

s" 95 foot elevation E

CyricCoollag Pumps Discliarge Pressure

10. Nuclear Services Closed Auxiliary Bulleling 0-2500F l per cooler 95 foot elevation w

Cycle Cooling Cooler Oestlet Tennperature

TABLE 4.3.6 n

oj REMOTE SHUTDOWN MQTiG iNG INSTRUMENTATION SURVEILLANCE REQUIREMENTS

-t>r-so E

CIIANNEL CHANNEL rn INSTRUMENT CitECK CALIBRATION

o e

C2 1.

Reactor Trip Breaker Indication

'M N.A.

~e

' " 2.

Reactor Coolant Temper: :ure Th -

M R

2'

~

3.

Reactor Coolant Pressure M

R 4.

Pressurizer Level M

R w3 5.

Steam Generator Level M

R

6. ~ Steam Generator Pressure M

R l

7.

Decay Heat Removal M

R Temperature 8.

Motor, Driven Eme[gency M

R Feedwater Pressure 9.

Nuclear Services Closed M

R Cycle Cooling Pumps Discharge Pressure E

10.

Nuclear Services Closed M

p Cycle Cooling Cooler E

Outlet Temperature

/

n

~.

~

s

'?

/

M e

p-1 e

s

~~

, TABLE 4.3-7 POST-ACCIDEN1 MONITORING INS't"RUMENTATION SURVEILLANCE REQUIREMENTS

' ~ O

o d

INSTRUMENT CHANNEL CHANNEL

-i CHECK CALIBRATION

>r-W l.

Power Range Nuclear Flux M

Q*

2.

Reactor Building Pressure M

R w

e 3.

Source Range Nuclear Flux C

M R*

2 4.

Reactor Coolant Outlet Temperature M

R

[

5.

Reactor Coolant Total Flow Rate M

R 6.

RC Loop Pressure M

R 7.

Pressurizer Level M

R 8.

Steam Generator Outlet Pressure M

R g

9.

Steam Generator Level M

R (Primary EFW Flow Detector) 10.

Borated Water Storage Tank Level M

R 11.

Startup Feedwater Flow Rate M

R 12.

Reactor Coolant System Subcooling Margin Monitor M

R 13.

PORY Position Indicator (Primary Detector)

M R

14.

PORY Position Indicator (Backup Detector)

M R

g 3

15.

PORY Block Valve Position Indicator a.

M R

16.

Safety Valve Position Indicator (Primary Detector)

M R

s#

17.

Safety Valve Position Indicator (Backup Detector)

M R

E I S.

Emergency Feedwater Ultrasonic Flow Indicator M

R

{

(Backup EFW FI ow Detector)

.N

Neutron detectors may be excluded from CilANNEL CALIBRATION

.D s

e REACTOR COOLANT SYSTEM POWER OPERATED RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.3.2 The power operated relief valve (PORV) and its associated block valve shall be OPERABLE.

APPLICABILITY:

MODES 1,2, and 3.

ACTION:

a.

With the PORV inoperable, within I hour either restore the PORY to OPERABLE status or close the associated block valve and remove power from the block valve; otherwise, 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With the block valve inoperable, within I hour either restore the block valve to OPERABLE status or close the block valve and remove power from the block valve or close the PORY and remove power from the associated solenoid valve; otherwise, 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The provisions of Specification 3.0.4 are not applicable.

c.

SURVEILLANCE REQUIREMENTS 4.4.3.2.1 In addition to the requirements of Specifications 4.0.5, the PORV shall be demonstrated OPERABLE at least once per 18 months by performance of a CHANNEL CALIBRATION.

l 4.4.3.2.2 The block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel.

J i

l l

CRYSTAL RIVER - UNIT 3 3/4 4-4a Amendments Nos. $$, H, 7), 77

)

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) or specifications.

Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

2.

Degradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either taside or outside of a tube.

3.

Degraded Tube means a tube containing imperfections 220% of the nominal wall thickness caused by degradation.

4.

% Degradation means the percentage of the tube wall thickness affected or removed by degradation.

5.

Defect meansan imperfection of such severity that it exceeds the plugging limit.

A tube containing a defect is defective. Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube.

6.

Pluggina Limit means the imperfection depth at or beyond which the tube shall be removed from service because it may become unserviceable prior to the next inspection and is equal to 40% of the nominal tube wall thickness.

7.

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

8.

Tube Inspection means an inspection of the entire steam generator tube as far as possible.

l b.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2 (and Table 4.4-6, if the provisions of Specification 4.4.5.2.d are utilized).

4.4.5.5 Reports Following each inservice inspection of steam generator tubes, the number of a.

l tubes plugged in each steam generator shall be reported to the Commission within 15 days.

CRYSTAL RIVER - UNIT 3 3/44-9 Amendment No, 3), 77

e REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b.

The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the

. inspection. This Special Report shall include:

1. _ Number and extent of tubes inspected.

2.

Location and percent of wall-thickness penetration for each l

indication of an imperfection.

3.

Identification of tubes plugged.

c.

Results of steam generator tube inspections wnich fall into Cate-gory C-3 and require prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

4.4.5.6 The steam generator shall be demonstrated OPERABLE by verifying steam generator level to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

CRYSTAL RIVER - UNIT 3 3/4 4-10 Amendment No. 3 3

a EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T

> 280*F y

LIMITING CONDITION FOR OPERATION 3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a.

One OPERABLE high pressure injection (HPI) pump, b.

One OPERABLE low pressure injection (LPI) pump, c.

One OPERABLE decay heat cooler, and d.

An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) on a safety injection signal and manually transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a.

With one ECCS subsystem inoperable, restore the inoperable subsystem to OPCRABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

CRYSTAL RIVER - UNIT 3 3/4 5-3

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power opdrated or automatic)in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

b.

By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suction during LOCA conditions. This visual inspection shall be performed:

1.

For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and 2.

Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.

By verifying the correct position of each mechanical position stop for the c.

following HPI stop check valves prior to restoring the HP! system to OPERABLE status following periodic valve stroking or maintenance on the valves.

1.

MUV-2 2.

MUV-6 3.

M UV-10 d.

By verifying that the flow controllers for the following LPI throttle valves l

operate properly prior to restoring the LPI system to OPERABLE status following periodic valve stroking or maintenance on the valves.

1.

DHV-Il0 2.

DHV-Ill e.

At least once per 18 months by:

1.

Verifying automatic isolation and interlock action of the DHR system from the Reactor Coolant System when the Reactor Coolant System pressure is greater than or equal to 284 psig.

l CRYSTAL RIVER - UNIT 3 3/4'5-4 Amendment No. 77, 77

.~

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i 2.

Verifying the correct position of each mechanical position,stop for each of the stop check valves listed in Specification 4.5.2.c.

3.

Verifying that the flow controllers for the throttle valves listed in l

Specification 4.5.2.d operate properly.

4.

A visual inspection of the containment emergency sump which verifies that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of l

structural distress or corrosion.

}

5.

Verifying a total leak rate less than or equal to 6 gallons per hour for the j

LPI system at:

a)

Normal operating pressure or a hydrostatic test pressure of greater than or equal to 150 psig for those parts of the system downstream of the pump suction isolation valve, and 4

b)

Greater than or equal to 55 psig for the piping from the containment emergency sump isolation valve to the pump suction isolation valve.

f.

At least once per 18 months during shutdown by 1.

Verifying that each automatic valve in the flow path actuates to its correct position on a high pressure or low pressure safety injection test q

signal, as appropriate.

i 2.

l Verifying that each HPI and LPI pump starts automatically upon receipt l

of a high pressure or low pressure safety injection test signal, as appropriate.

1 g.

Following completion of HPI or LPI system modifications that could have altered system flow characteristicsl, by performance of a flow balance test during shutdown to confirm the following injection flow rates into the Reactor

).

Coolant System:

l HPI System - Single Pumo LPI System - Single Pump Single pump flow rate greater than -

1. Injection Leg A - 2800 to 3100 or equal to 500 gpm at 600 psig.

gpm.

While injecting through 4 Injection Legs,

2. Injection Leg B - 2800 to 3100 i

the flow rate for all combinations of 3 gpm.

injection Legs greater than or equal to 350 gpm at 600 psig.

j I

Flow balance tests performed prior to complete installation of modifications are valid if performed with the system change that could alter flow characteristics in i

effect.

(

CRYSTAL RIVER - UNIT 3 3/45-3 Amendment No. 77, 77

.,~

- - --- - ~ ' -

~~~ " ~ - ~ ~ ~ ~ ^ ~

i e l 1

l 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves shall be OPERABLE.

I APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or the Nuclear Overpower Trip Setpoint is reduced per Table 3.7-1; otherwise, 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 in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The provisions of Specification 3.0.4 are not applicable, i

l SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5, are applicable for the main steam line code safety valves of Table 4.7-1.

i CRYSTAL RIVER - UNIT 3 3/4 7 1

s.

TABLE 3.7-1 MAXIMUM ALLOWABLE NUCLEAR OVERPOWER TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES l

Maximum Allowable Nuclear Maximum Number of Inoperable Safety Overpower Trip Setpoint Valves on Any Steam Generator (Percent of RATED THERMAL POWER) 1 96.35 2

31.95 3

67.5 1

l CRYSTAL RIVER - UNIT 3 3/47-2 Amendment No. SA, 77

PLANT SYSTEMS 3/4.7.9 HYDRAULIC SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.9.1 All hydraulic snubbers listed in Table 3.7-3 shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

With one or more hydraulic snubbers Inoperable, replace or restore the inoperable snubber (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 1

4.7.9.1 Hydraulic snubbers will be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

a.

Each hydraulic snubber with seat material fabricated from ethylene i

propylene or other materials demonstrated compatible with the operating environment and approved as such by the NRC, shall be determined OPERABLE at least once af ter not less than 4 months but within 6 months of initial criticality and in accordance with the inspection schedule of Table 4.7-4 thereafter, by a visual inspection of the snubber. Visual inspections of the snubber shall inclu'de, but are not necessarily limited to, inspection of the hydraulle fluid reservoirs, fluid connections, and linkage connections to the piping and anchors. Initiation of the Table 4.7 4 inspection schedule shall be made assuming the unit was prey!ously at the 6 month inspection

interval, b.

Each hydraulic snubber with seal material not fabricated from ethylene propylene or other materials demonstrated compatible with the operating environment shall be determined OPERABLE at least once per 31 days by a visual inspection of the snubber. Visual inspection of the snubbers shall include but are not necessarily limited to, inspection of the hydraulle fluid reservoirs, fluid connections, and linkage connections to the piping and anchors.

i l

i i

CRYSTAL RIVER - UNIT 3 3/4 7-25 Amendment No. M,77 1

I f

4

__._---..,-r----

e PLANT SYSTEMS HYDRAULIC SNU88ERS (Continued)

SURVEILLANCE REQUIREMENTS (Continued) c;--'At least once per 18 months during shutdown a representative sample of at least 10 hydraulic snubbers or at least 10% of all snubbers listed in Table 3.7-3, whichever is less, shall be selected and functionally tested to verify correct piston movement, lock up and bleed. Snubbers greater than 50,000 lbs capacity may be excluded from functional testing requirements.

Snubbers selected for functional testing shall be selected on a rotating basis. Snubbers identified in Table 3.7-3 as either "Especially Difficult to Remove" or in "High Radiation Zones" may be exempted from functional testing provided these snubbers were demonstrated OPERABLE during previous functional tests. Snubbers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming opera-tion.

For each snubber found inoperable during these func-tional tests, an additional minimum of 10% of all snubbers or 10 snubbers, whichever is less, shall also be functionally tested until no more failures are found or all snubbers have been functionally tested.

1 i

CRYSTAL RIVER - UNIT 3 3/4 7-26

n W

d

-i>

F TABLE 4.7-4 5

HYDRAULIC SNUBBER INSPECTION SCHEDULE N

e E

NUMBER OF SNUBBERS FOUND INOPERABLE NEXT REQUIRED q

DURING INSPECTION OR DURING INSPECTION INTERVAL (*)

INSPECTION INTERVAL *

w 0

18 months + 25%

1 12 monthsI 25%

~

2 6 months 25 %

~

3 or 4 124 days 1 25 %

5, 6, or 7 62 days + 25%

Greater than or equal to 8 31 days _i25%

w7 70 Snubbers may be categorized into two groups, " accessible" and " inaccessible". This categorization shall be based upon the snubber's accessibility for inspection during reactor operation. These two groups may be inspected independently according to the above schedule.

.. The required inspection interval shall not be lengthened more than one step at a time.

N i

3 O

M e

a PLANT SYSTEMS 3/4.7.10 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION

~

Each sealed source containing radioactive material either in 3.7.10.1 excess of 100 microcuries of beta and/or ganna emitting material or 5 microcuries of alpha emitting material shall be free of > 0.005 micro-curies of removable contamination.

APPLICABILITY: At all times.

ACTION:

Each sealed source with removable contamination in excess of a.

the above limit shall be immediately withdrawn from use and:

1.

Either decortaminated and repaired, or 2.

Disposed of in accordance with Cannission Regulations.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable, o

' SURVEILLANCE REQUIREMENTS 4.7.10.1.1 Test Recuirenents - Each sealed source shall be tested for leakage and/or contamination by:

a.

The licensce, or b.

Othdr persons specifically authori:ed by the Ccanission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.

4.7.10.1.2 Test Frecuencies - Each category of sealed sources shall be tested at the frequency described below, Sources in use (excluding startup sources and fission detectors a.

previously suojected to core flux) - At least once per six months for all sealed sources containing radioactive material:

CRYSTAL RIVER - UNIT 3 3/4 7-36 i

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 SHUTDOWN MARGIN l

A sufficient - SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions,2) the reactivity transients associated with 2

postulated acddent conditions are controllable within acceptable limits, and 3) the reactor will be maintained suffidently subcritical to preclude inadvertent criticality in the shutdown condition. During Modes 1 and 2 the SHUTDOWN MARGIN is known i

to be within ilmits if all control rods are OPERABLE and withdrawn to or beyond the insertion limits.

4 SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration and RCS TavE. The most restrictive condition for Modes 1,2, and 3 occurs at EOL, with Tavt t no load operating temperature, and is associated with a postulated steam Ene 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.

d Accordingly, the SHUTDOWN MARGIN required is based upon this limiting condition

]

and is consistent with FSAR safety analysis assumptions, g

3 3/4.1.1.2 BORON DILUTION A minimum flow rate of atleast2700 GPM provides adequate mixing, prevents 4

stratification and ensures that reactivity changes will be gradual through the i

Reactor Coolant System in the core during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 2700 GPM will circulate an equivalent Reactor 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 capability for operator recognition and control.

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure i

that the assumptions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance requiremera 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 ilmit provides assurance that the coefficient will be maintained within acceptable values throughout each fuel cycle.

i 1

i CRYSTAL RIVER - UNIT 3 53/411 Amendment flo. 32, 64 e Wee eep

---.m,_--.-

,,g..

.,m_._

w,.,..

_,m

,.y.,,,

_y__

y_ _,

REACTIVITY CONTROL SYSTEMS BASES 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 325 0F. This limitation is required to ensure that (1) the moderator temperature coefficient 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 j

The boron injection system ensures that negative reactivity control is available during_each mode of facility operation. The components required to perform this 4

function include (1) borated water sources, (2) makeup or DHR pumps, (3) separate flow paths, (4) boric acid pumps, (3) associated heat tracing systems, and (6) an emergency power supply from OPERABLE emergency busses.

With the RCS average temperature above 2000F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability l

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 facility safety from injection system failures during the repair period.

The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from all operating conditions of 1.0% A k/k after xenon decay and cooldown to 2000F. The maximum boration capability requirement occurs from full power equilibrium xenon conditions and requires either 4,980 gallons of 11,600 ppm boric acid solution from the boric acid storage tanks or 35,681 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 i

to the desired level. The specified quantity of borated water is consistent with the l

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 l

Specification 3.1.2.9 is conservative for this cycle.

I With the RCS temperature below 2000F, 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 2000F is sufficient to provide a SHUTDOWN i

MARGIN of 1.0%A k/k after xenon decay and cooldown from 2000F to 1400F. This conditio-requires either 390 gallons of !!,600 ppm boron from the boric acid I

storage system or 1,990 gallons of 2,270 ppm boron from the borated water storage tank. To envelop future cycle BWST and BAST contained borated water volume requirements, a minimum volume of 13,500 gallons and 600 gallons, respectively are specified.

CRYSTAL RIVER - UNIT 3 B3/41-2 Amendments flos. 7$,19, J7, AS,

$4, 77

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining 1

the minimum DNBR in the core ;p_ l.30 during normal operation and during short term transients, (b) maintaining the peak linear power density f 18.0 kW/ft during normal operation, and (c) maintaining the peak power density g 20.3 kW/ft during short term transients. In addition, the above criteria must be met in order to meet the assumptions used for the loss-of-coolant accidents.

The power-imbalance envelope defined in Figures 3.2-1, 3.2-la, 3.2-2, and 3.2-2a and the insertion limit curves, Figures 3.1 1, 3.1-la, 3.1-2, 3.1-3, 3.1-3a, 3.1-4, 3.1-9, and 3.1-10 are based on LOCA analyses which have defined the maximum !!near heat rate such that the maximum clad temperature will not exceed the Final Acceptance Criteria of 22000F following a LOCA. Operation outside of the power-imbalance 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 !!mits, as defined by Figures 3.1-1, 3.1-la, 3.1 2, 3.1-3, 3.1-3a, 3.1 4, 3.1-9, and 3.1 10, 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, d.

Hot rod manufacturing tolerance factors.

The conservative app!! cation of the above peaking augmer.tation factors compensates for 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 th"se specifications are as follows:

F9 Nuclear Heat Flux Hot 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 Araendment No, J$, 77

l POWER DISTRIBUTION' LIMITS BASES Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the N

F integral of linear power along the rod on which minimum DNBR occurs to the g

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 minimum DNBR at full power are met, -

provided:

qf 3.13; F f 1.71 F

Power Peaking is not a directly observable quantity and therefore limits have been estab!Ished 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 3 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 insettion 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 measure of the difference in power between the top and bottom 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 power shape is not exceeded if the AXIAL POWER IMBALANCE is maintained within the limits of Figures 3.2 1, 3.2-la, 3.2-2, and 3.2.2a.

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 insertion and are the core DNBR design basis. Therefore, for operation at a fraction of RATED THERMAL POWER, the design limits are met. When using incore detectors to make power distribution maps to determine FQ and F NH 3

The measurement of total peaking factor, FQMeas, shall be increased by 1.4 a.

percent to account for manufacturing tolerances and further increased by 7.5 percent to account for measurement error.

CRYSTAL RIVER - UNIT 3 83/42-2 Amendments Nos. Jp,19, 77

l POWER DISTRIBUTION LIMITS BASES b.

The measurement of enthalpy rise hot channel factor, FNH, shall be increased by 3 percent to account for measurement error.

For Condition !! events, the core is protected from exceeding 20.5 kW/f t locally, and from l

going below a minimum DNBR of 1.30 by automatic protection on power, AXIAL POWER IMBALANCE, pressure and temperature.

Only conditions I 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 measurements are made during startup testing and periodically during power operation.

For QUADRANT POWER TILT, the safety (measurement independent) limit for Steady State is 4.49,for Transient State is 11.07, and for the Maximum Limit is 20.0.

l 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 Fq is reinstated by l

reducing the power by 2 percent for each percent of tilt in excess of the limit.

3/fs.2.5 DNB PAR AMETERS 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 i

have been analytically demonstrated adequate to maintain a DNBR of 1.30 or greater throughout each analyzed transient.

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 t

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 i

the flow Indication channels with measured 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 B 3/4 2 3 Amendment No. D, 77 i

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o 3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CY,CLE 3/4.7.1.1 SAFETY VALVES The OPERABil.!TY of the main steam line code safety valves ensures that the secondary system pressure w!!! be limited to within its design pressure of 1050 psig during the most severe anticipated system operational transient.

The maximum relieving capacity is' associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section !!! of the ASME Boller and Pressure Vessel Code,1971 Edition. The total re!!eving capacity for all valves on all of the steam lines is 13,007,774 lbs/hr which is 118.3 percent of the total secondary steam flow of !!.0 x 106 lbs/hr at 100% RATED THERMAL POWER.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Nuclear Overpower channels. The reactor trip setpoint reductions are derived on the fo!!owing bases:

SP =

X-AY x NOTS XI where:

SP = reduced Nucelar Overpower Trip Setpoint in percent of Rated Thermal Power.

X = total actual re!!eving capacity of each steam generator in Ibs/hr (6,503,887 lbs/hr).

A = maximum number of inoperable safety valves per steam generator.

Y = maximum relieving capacity of each of the larger capacity safety valves in Ibs/hr (843,739 lbs/hr).

x 1 = total required relieving capacity (of each steam generator for 112%

Rated Thermal Power in Ibs/ hour 6,160,000 lbs/hr).

NOTS = Nuclear Overpower Trip Setpoint specified in Table 2.2.1.

CRYSTAL RIVER - UNIT 3 B3/47-1 Amendment No. S A, 77

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l PLANT SYSTEMS BASES 3/4.7.1.2 EMERGENCY FEEDWATER SYSTEMS The OPERABILITY of the emergency feedwater systems ensures that the Reactor Coolant system can be cooled down to less than 230 F from normal operating conditions in the event of a t'otal loss of offsite power.

Each emergency feedwater pump is capable of delivering a total feedwater flow of 740 gpm at a pressure of 1144 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature.to less than 230 F where the Decay Heat Removal System may be placed into operation.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available for cooldown of the Reactor Coolant System to less than 230*F in the event of a total loss of offsite power or of the main feedwater system. The minimum water volume is sufficient to maintain the RCS at HOT STANDBY conditions for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with steam discharge to atmosphere concurrent with loss of offsite power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 ACTIVITY The limitations on secondary systern specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures thr.: no more than one i

steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the CRYSTAL RIVER - UNIT 3 B3/47-2 Amendment flo. 64

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i DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The Reactor Containment building is designed and shall be maintained for a maximum internal pressure of 55 psig and a temperature of 2810F.

5.3 REACTOR CORE FUEL ASSEMBLIES l

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 1

uranium. The initial core loading shall have a maximum enrichment of 2.83 weight i

percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.50 weight percent U-235.

CONTROL RODS 3.3.2 The reactor core shall contain 60 safety and regulating and 8 axial power shaping l

(ASPR) control rods. The safety and regulating control rods shall contain a nominal 134 inches of absorber materia!. The nominal values of absorber material shall be 80 percent silver,15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. The APSRs shall contain a nominal 63 inches of absorber material at their lower ends. The absorber material for the APSRs shall be 100% Inconel.

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CRYSTAL RIVER - UNIT 3 5-4 Amendments No.s M. M, 72, 77

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