ML20116F654

Tech Spec Change Request 135 Re Current Reactor Refueling for Cycle VI

ML20116F654
Person / Time
Site:	Crystal River
Issue date:	04/25/1985
From:	FLORIDA POWER CORP.
To:
Shared Package
ML20116F636	List: 3F0485-16, Application for Amend to License DPR-72 Consisting of Tech Spec Change Request 135 Re Current Reactor Refueling for Cycle Vi.Certificate of Svc Encl ML20116F654 ML20116F659
References
NUDOCS 8505010163
Download: ML20116F654 (46)

Text

. o 1

... FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-72 REQUEST NO.135, REVISION 0 CYCLE VI RELOAD REPORT

' LICENSE DOCUMENT INVOLVED: Technical Specifications PORTION: (See Attached Table)

DESCRIPTION OF REQUEST:

This submittal proposes changes to support operating Cycle VI for Crystal River Unit 3.

These changes include:

1. Reactor core safety limits and trip setpoints for reactor thermal power and axial power imbalance.

2. Minimum boric acid and borated water volumes.

3. Regulating and axial power shaping rod group insertion limits.

[

4. Axial power imbalance limits.

5. Reactor Protection System response time and testing requirements.

6. Deletion of certain Cycle V specific allowances.

t REASON FOR REQUEST:

!- For Cycle VI, Crystal River Unit 3 will operate with 60 fresh fuel assemblies. The Cycle VI core has been designed for a cycle lifetime of 425 3 10 effective full power days. As described in BAW-1860, Crystal River Unit 3 - Cycle 6 Reload Report, certain Technical

' Specifications require revision due to fuel and thermal characteristic changes.

EVALUATION OF REQUEST:

j The safety and licensing considerations for Cycle VI operation are described in BAW-1860.

These proposed changes will bring the Technical Specifications into agreement with the

! applicable portions of BAW-1860 and insure that the plant will continue to operate in a safe

! manner.

t i

i 8505010163 850425

, PDR ADOCK 05000302 P PDR l

I l

4 L

E

. c t TECHNICAL SPECIFICATION NO. PAGE REASON FOR CHANGE Figure 2.12 2-3 (1)

Figure 2.2-1 2-7 (1)

Bases B 2-1 (1)

Bases B 2-2 (1)

Bases B 2-3 (1)

Bases B 2-6 (1)

Bases B 2-8 (1) 3.1.2.8 3/4 1-14 (1) 3.1.2.9 3/4 1-16 (1) 3.1.3.1 3/4 1-19 (2) 3.1.3.6 3/4 1-23 (1)

Figure 3.1-1 3/4 1-27 (1)

Figure 3.1-la 3/41-27a (1)

Figure 3.1-2 3/4 1-28 (1)

Figure 3.1-2a 3/41-28a (2)

Figure 3.1-3 3/4 1-29 (1)

Figure 3.1-3a 3/41-29a (1)

Figure 3.1-4 3/4 1-30 (1)

Figure 3.1-4a 3/4 1-31 (2)

Figure 3.1-7 3/4 1-34 '(1) 3.1.3.9 3/4 1-37 (1)

Figure 3.1-9 3/4 1-38 (2)

Figure 3.1-9a 3/41-38a (2)

Figure 3.1-10 3/4 1-39 (2)

Figure 3.1-10a 3/4 1-40 (2) 3.2.1 3/42-1 (1)

Figure 3.2-1 3/42-2 (1)

Figure 3.2-la 3/42-2a (1)

Figure 3.2-2 3/42-3 (1)

Figure 3.2-2a 3/4 2-3a (1) 3.2.2 3/42-4 (1) 3.2.3 3/42-6 (1)

Table 3.3-2 3/4 2-11 (1)

U

. o 1 TECHNICAL SPECIFICATION NO. PAGE REASON FOR CHANGE Table 3.3-2 3/43-6 Response time of item 4 changed Table 4.3-1 3/43-7 Footnote to exclude CT & PT from testing Table 4.3-1 (Continued) 3/43-8 Same Table 4.3-2 (Continued) 3/4 3-19 (2)

Table 4.3-6 3/43-36 (2)

Table 4.3-7 3/4 3-39 (2) 3.4.3.2 3/4 4-4a (2) 3.4.5 3/44-9 (3) 3.5.2 3/45-4 (3) 3.5.2 3/45-5 (3) 3.7.1.1 3/47-2 (4) 3.7.9.1 3/47-25 (2)

Table 4.7-4 3/47-35 (2)

Bases B 3/41-2 (1)

Bases B 3/4 2-1 (1)

Bases B3/42-2 (1)

Bases B 3/4 2-3 (1)

Bases B3/47-1 (4) 5.3.2 5-4 (1)

NOTATION EXPLANATION (1) See Table 8-1 of BAW-1860, RELOAD REPORT.

(2) Deletion of non-Cycle VI specific technical specification allowance or waiver.

(3) Clarification.

(4) Correction to reflect Cycle VI nuclear overpower trip setpoint.

. e t FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-72 REQUEST NO.135, REVISION O SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION REQUEST:

Florida Power Corporation requests issuance of an amendment to Crystal River Unit 3 Technical Specifications for Fuel Cycle VI reload. The significant aspects of Cycle VI reload report (attached) include:

1. Use of cross flow methods in the thermal-hydraulic analysis.

2. A fuel cycle length of 425 1 10 EFPD.

3. Use of gray (inconel) axial power shaping rods during Cycle VI.

4. Use of Batch 8 fuel with an enrichment of 3.49 WT % U 235.

SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION:

(X) Amendment involves no significant hazards considerations.

() Amendment involves significant hazards considerations.

BASIS FOR DETERMINATION:

This amendment is not likely to involve a significant hazards consideration because it is a core reload that is not significantly different from previous core reloads (see Example lii of Federal Register 14864, April 6,1983). No significant changes have been made to the acceptance criteria for these Technical Specifications. A new analytical method, Noodle Code, was used to determine core physics parameters.

This code has been used by other utilities and found acceptable.

. - _ , . _ _._-_,,__r. , _ . , . . - _ - - _ . , , - - _ - _ - - . _ , _ _ . . , , . . _ _ _ - - , - - , . - - - _ --

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• 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 (31,89.6 (48.2,85.2)

(-48.5,77.2) .

3 & 4 Pump Opera tion i. - - 70 g

2 (48.2,62.8)

--- 60 i

-- . 50 b

E-- 40 D4 f- - 30 2

V 20 W

- - 10 m a a i I I I I I I i I

-60 -50 -40 -30 -20 -10 0 10 20 30 40 50 60 Axial Power Imbalance, %

CRYSTAL RIVER UNIT 3 2-3

_:A J

.. i FIGURE 2.2-1

'~

TRIP SETPOINT FOR NUCLEAR OVERPOWER BASED ON RCS FLOW AND AXIAL POWER IMBALANCE

(-17,108) - 110 (17,108)

M3 = 1.0

- - 100 Acceptable 4 Pump M2 = -1.815

(-34.7, 90.3 Operation - - 90

(-17,80.67) (17,80.67)

" CU (34.7,75.86)

Acceptable - - 70 3 & 4 Pump y

(-34.7.62.97) Operation o

-- 60

~ ~

(34.7,48.53)

. T

,%-- 40 a

g- - 30 3

n-s- - - 20 3o E

a: -

- 10 l = n i n i n e a n . .

-60 -50 -40 -30 -20 -10 0 10 20 30 40 50 60 Axial Power Imbalance, %

CRYSTAL RIVER UNIT 3 2-7 l

l

o i 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 bolling 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 DNS 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 limited 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 l coolant flow is 139.7 x 106 lbs/hr, which is 106.5% of the design flow rate for four operating l

reactor coolant pumps. This curve is based on the following nuclear power peaking factors

[ with potential fuel densification effects:

F = 2.82 F ~ * "

• dH i 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.

l I-l CRYSTAL RIVER - UNIT 3 B 2-1 L.-

p SAFETY LIMITS The reactor trip envelope appears to approach the safety limit more closely tnan 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.

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:

1. The 1.30 DNBR limit produced by a nuclear power peaking factor of F N - 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.

2. The combination of radial and axial peak that causes central fuel melting at -

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 or, the basis of the reactor power imbalance produced by the power peaking.

The specified flow rates for curves I 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 published quality range of 22% is justified on the basis of experimental data.

- CRYSTAL RIVER - UNIT 3 B 2-2

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• V 4

LIMITING SAFETY SYSTEM SETTINGS i BASES RCS Outlet Temperature - High The RCS Outlet Temperature High trip less than or equal to 618.80F prevents the reactor outlet temperature from exceeding the design limits and acts as a backup trip for all power excursion transients.

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.

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 l 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 i less than or equal to 95.88% and power level is 100%. I

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.

CRYSTAL RIVER - UNIT 3 . 2-5 i

m --

. .- 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/ft limits or DNBR limits. The AXIAL POWER IMBALANCE reduces the power level trip.

produced by the flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced.

The flux-to-flow ratio reduces the power level trip and associated reactor power-reactor 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.

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.

N CRYSTAL RIVER - UNIT 3 B 2-6 l

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

E U

=

b E

s 2000 -

CURVE 1 o 4 PUMP W

8 1800 -

1 I 580 600 620 640 REACTOR OUTLET TEMPERATURE, F REACTOR COOLANT FLOW PUMPS OPERATING CURVE FLOW (% DESIGN) 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 8 2-8

I.

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.8 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 105 F.

b. The borated water storage tank (BWST) with:

1. A minimum contained borated water volume of 13,500 gallons,

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

3. A minimum solution temperature of 40 F.

APPLICABILITY: MODES 5 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 concentration of the water,

2. Verifying the contained borated water volume of the tank, and l

)

l CRYSTAL RIVER - UNIT 3 3/4 1-14 l

\

l

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,

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

3. A minimum solution temperature of 105 F.

b. The borated water storage tank (BWST) with:

1. A minimum contained borated water volume of 415,200 gallons,

2. Between 2,270 and 2,450 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; 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 <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,

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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

CRYSTAL RIVER - UNIT 3 3/4 1-16

REACTIVITY CONTROL SYSTEMS ACTION: (Continued) c) A power distribution map is obtained from the incore detectors and F

O and Fp are verified to be within their limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and d) The THERMAL POWER level is reduced to d60% of the THERMAL POWEls allowable for the reactor coolant pump combination within one hour and within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the Nuclear Overpower Trip Setpoint is reduced toe 70% of the THERMAL POWER allowable for the reactor coolant pump combination, or e) The remainder of the rods in the group with the inoperable rod are aligned to within + 6.5% of the inoperable rod within one hour while maintaining the rod sequence, insertion and overlap limits of Figures 3.1-1, 3.1-2, 3.1-3, 3.1-4, 3.1-5 and 3.1-6; the THERMAL POWER level shall be resricted pursuant to Specification 3.1.3.6 during subsequent operation.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each control rod shall be determined to be within the group average height limits by verifying the individual rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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.

CRYSTAL RIVER - UNIT 3 3/41-19

REACTIVITY CONTROL SYSTEMS REGULATING ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The regulating rod groups shall be limited in physical insertion as shown on i 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 + 5% l between sequential withdrawn groups 5 and 6, and 6 and 7.

APPLICABILITY: MODES 1* 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

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

• See Special Test Exceptions 3.10.1 and 3.10.2.

1. With Keff greater than or equal to 1.0.

CRYSTAL RIVER - UNIT 3 3/4 1-25 4

FIGURE 3.1-1 REGULATING ROD GROUP INSERTION LIMITS FOR FOUR PUMP OPERATION FROM 0 EFPD TO 200 + 10 EFPD l 110

( 3, 0 (300,102) 100 .

90 - (267,92 h 80 -

2 (250,80)

- 70 -

.E UNACCEPTABLE

• 60 .

OPERATION 7 "

% 50 (175,50) x

• • 40 .

C l 30 - ACCEPTABLE OPERATION 20 -

( ,15) 10 .

8 8 I I I 0

O 50 100 150 200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25' 50 75 100 l

t 1 i I i t 1 i i 1 Group 5 Group 7 l

0 25 50 75 100 t 1 i 1 J

Group 6 CRYSTAL RIVER UNIT 3 3/4 1-27

FIGURE 3.1-la

. REGULATING R0D GROUP INSERTION LIMITS FOR FOUR PUMP OPERATION FROM 200 110 TO 400 110 EFPD 110 (273,102 100 (300,102) l 90 -

(267,92) u y 80 -

2 (250,80)

- - UNACCEPTABLE

@ 70 OPERATION

$ 60 -

?

e 50 -

E 175,50) 40 -

s 30 -

20 - ACCEPTABLE 58,15) OPERATION 10 -

0 0 50 100 150 200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25 50 75 100 t I 1 a  :  :

Group 5 Group 7 0 25 50 75 100 L i i i 1 Group 6 CRYSTAL RIVER UNIT S 3/4 1-27a

FIGURE 3.1-2 REGULATING ROD GROUP INSERTION LIMITS FOR FOUR PUW OPERATION AFTER 400 +10 EFPD (265.102) (300,102) 100 -

90 - (260,92) -

' 80 ~

I 250,80)

E UNACCEPTABLE 70 -

OPERATION i 60 -

r T 50 -

,% (175,50) a 40 -

u g 30 -

O ACCEPTABLE 20 -

OPERATION 58,15) 10 -

0,5) 0 8 ' ' ' 8 0 50 100 150 200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25 50 75 100 1 1 I I I i l I i l Group 5 Group 7 0 25 50 75 100 t i I I e Group 6 CRYSTAL RIVER UNIT 3 3/4 1-28

g

-DELETED-CRYSTAL RIVER UNIT 3 3/41-28a

FIGURE 3.1-3 l REGULATING R0D GROUP INSERTION LIMITS FOR THREE PUMP OPERATION FROM 0 TO 200 +10 EFPD

(

110 100< -

90 -

80 , -

(275,77)

(300,77)

~

UNACCEPTABLE (267,69)

OPE W ION 60 .

y 250,60)

,% 50 < -

u 40 -

l 2 30 -

175,37.5) 20 - -

ACCEPTABLE

10 - (32,11.75) OPERATION 0,4.25) 0 50 100 150 200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25 50 75 100 t I i i I i i i I f Group 5 Group 7 0 25 50 75 100 t i 1 1 1 Group 6 CRYSTAL RIVER UNIT 3 3/4 1-29

FIGURE 3.1-3a REGULATING R0D GROUP INSERTION LIMITS FOR THREE PUMP OPERATION FROM 200 110 to 400 110 EFPD 110-100 -

90 -

80 -

(275,77)

(300,77)

" 70 -

UNACCEPTABLE (267,69)

OPERATION g

60 -

c i (250,60)

= 50 -

40 -

". 175,37.5) g30 -

20 .

ACCEPTABLE 10 - 58,11.75) OPERATION (0,4.25) 0 ' ' ' ' '

0 50 100 150 200 250 300 Rod Index, % Withdrawn 0 25 50 75 100 0 25 50 75 100 t I e i e a i i t i Group 5 Group 7 l

,0 25 j0 7,5 1,00 Group 6 t

CRYSTAL RIVER UNIT 3 3/4 1-29a 1 _ - - , . , - _ _ _ _ . . - . - _ ,

FIGURE 3.1-4 REGULATING R00 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

$ (250,60)

] 50 -

UNACCEPTABLE OPERATION E

w 40 -

h30 *

• 2 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 I I t i e i Group 5 Group 7 0 25 50 75 100 i i l i I Group 6 CRYSTAL RIVER UNIT 3 3/4 1-30

9 9

-6

-DELETED-CRYSTAL RIVER UNIT 3 3/4 1-31

FIGURE 3.1-7

, CONTROL R0D LOCATIONS AND GROUP DESIGNATIONS FOR CRYSTAL RIVER 3 CYCLE 6 I X

Fuel Transfer

[ hul  ;

A B 1 6 1 C 2 5 5 2 D 7 8 7 8 7 E 2 5 4 4 5 2 F 1 8 6 3 6 8 1 G 5 4 3 3 4 5 N W- 6 7 3 3 7 6

-Y K 5 4 3 3 4 5 L 1 8 6 3 6 8 1 M e 2 5 4 4 5 2 N l 7 8 7 8 7 0 l l 2 5 5 2 P l l l 1 6 1 R I t Z

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 X Group NAer Group 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

REACTIVITY CONTROL SYSTEMS AXIAL POWER SHAPING ROD INSdRTION 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:

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

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

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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• With Keff 31.0.

CRYSTAL RIVER - UNIT 3 3/4 1-37

O e e

-DELETED-l CRYSTAL RIVER UNIT 3 3/4 1-38

@ 9 4

-DELETED-CRYSTAL RIVER UNIT 3 3/41-38a

O O 6

-DELETED-CRYSTAL RIVER UNIT 3 3/4 1-39

-DELETED-

! CRYSTAL RIVER UNIT 3 3/4 1-40 l

l l

0.

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.

l 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when above 40% 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.

CRYSTAL RIVER - UNIT 3 3/42-1

FIGURE 3.2-1 AXIAL POWER IPEALANCE ENVELOPE FOR FOUR PUMP OPERATION FROM 0 TO 400 +10 EFPD i

• - 110

(-23,102)

, ,m (15.102)

(-25,92)

. . 90 (15,92)

(-26.80) - 80 (20,80)

- 70

. 60 n.

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

. . 40 ACCEPTABLE E UNACCEPTABLE OPERATION "-- 30 OPERATION i

. . 20

. 10 1

I i i i f l 1

-40 -30 -20 -10 0 10 20 30 40 Axial Power Imbalance, %

CRYSTAL RIVER UNIT 3 3/4 2-2

FIGURE 3.2-la

. AXIAL POWER IpBALANCE ENVELOPE FOR FOUR-PUW OPERATION AFTER 400 +10 EFPD 110

(-21.102) . 4vu (15,102)

(-25,92 (15,92)

(-28,80) -

- 80 (20,80)

- 70 60 I'-

a.

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

?

7- - 40 ACCEPTABLE 3 UNACCEPTABLE OPERATION OPERATION i

E- 20

. 10 1 1 e a l i ) a 1

-40 -30 -20 -10 0 10 20 30 40 Axial Power Imbalance, %

CRYSTAL RIVER UNIT 3 3/4 2-2a

FIGURE 3.2-2 AXIAL POWER I W LANCE ENVELOPE FOR THREE PUMP OPERATION FROM 0 TO 400 +10 EFPD '

.,100

. .90

. .80

(-20.67) (15,77)

(-25,69) (15,69)

- 60 (20,60)

(-26,60)

I'

. .50

(-28,37.5,l l g..40 (20,37.5)

T, 3 .30 w

- 20 l

ACCEPTABLE a. UNACCEPTABLE I

' OPERATION OPERATION

- 10 i

i i i f i i t a a

-40 -30 -20 -10 0 10 20 ~ 30 40 Axial Power Imbalance, %

4 9

i

. CRYSTAL RIVER UNIT 3 3/4 2-3

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)

L I

2-- 50

(-28,37.5) g-- 40 (20,37,5)

To j-- 30 5-

. 20 ACCEPTABLEOPERATION]. UNACCEPTABLE m OPERATION

- - 10 l

a 1 i I e l e  :

-40 -30 -20 -10 0 10 20 30 40 l Axial Power Imbalance, %

• l l

l l CRYSTAL RIVER UNIT 3 3/4 2-3a 1

4 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:

FQ 6~32 P where P = THERM AL POWER and P 6 1.0

RATED THERMAL POWER APPLICABILITY

MODE 1 ACTION:

With FQ exceeding its limit:

a.

Reduce THERMAL POWER at least 1% for each 1% FQ exceeds the limit within 15 minutes and similarly reduce the Nuclear Overpower Trip Setpoint o and Nuclear Overpower based on RCS Flow and AXIAL POWER IMBALANCE Trip Setpoint within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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

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

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 FQ is.

demonstrated through in-core mapping to be within its limit at a nominal 50%

of RATED THERM AL 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.

SURVEILLANCE REOUIREMENTS l

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 CRYSTAL RIVER - UNIT 3 3/42-4 c

POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FIH LIMITING CONDITION FOR OPERATION 3.2.3 F shall be limited by the following relationship:

H 1.71 h + 0.3 (1-Pf Thermal Power P=

Rated Thermal Power and P (1.0

. APPLICABILITY: MODE 1.

ACTION:

With F[H exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% that Fh exceeds the limit within 15 minutes and similarly reduce the Nuclear Overpower Trip Setpoint and Nuclear Overpower based on RCS Flow and the AXIAL POWER IMBALANCE Trip Setpoint within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,

b. Demonstrate through in-core mapping that F$ g is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c. Identify and correct the cause of the out of limit condition prior to increasing THf.RMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that Fh is i

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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.

l i

CRYSTAL RIVER - UNIT 3 3/42-6 l

l

(

i l TABLE 3.2-2 y QUADRANT POWER TILT LIMITS

.I s STEADY STATE TRANSIENT MAXIMUM LIMIT LIMIT LIMIT QUADRANT POWER TILT as Measured by:

Symmetrical Incore Detector System 3.20 9.08 20.0 Power Range Channels 1.61 6.96 20.0 Minimum incore Detector System 1.73 4.40 20.0 e

t CRYSTAL RIVER - UNIT 3 3/42-11

!^

TABLE 33-2 O REACTOR PROTECTION SYSTEM INSTRUMENTATION RESPONSE TIMES m

d 5!

r- Functional Unit Response Times 3

sW

l. Manual Reactor Trip Not Applicable C

h 2. Nuclear Overpower

• 6 0.266 seconds u

3. RCS Outlet Temperature - High Not Applicable

4. Nuclear Overpower Based on RCS Flow and AX1AL POWER IMBALANCE

• d 1.842 seconds

5. RCS Pressure - Low d 0.44 seconds

6. RCS Pressure - High 5 0.44 seconds

7. Variable Low RCS Pressure Not Applicable

8. Pump Status Based on RCPPMs** 51.44 seconds

9. Reactor Containment Pressure - High Not Applicable

• Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

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

TABLE 4.3-1 g REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS x ,

d

-1 CHANNEL MODES IN WHICH

% CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE x FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED E

C 1. Manual Reactor Trip N.A. N.A. S/U(1) N.A.

{

w

2. Nuclear Overpower S D(2) and Q(7) M 1, 2

3. RCS Outlet Temperature--High 5 R M 1, 2

4. Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE S(4) M(3) and Q(7, 3) M 1, 2

5. RCS Pressure--Low S R M 1, 2

6. RCS Pressure--High S R M 1, 2

7. Variable Low RCS Pressure S R M 1, 2,

$ 8. Reactor Containment Pressure--High S R M 1, 2

{ 9. Intermediate Range, Neutron Flux and Rate S R(7) S/U(IX5) I,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**,5**

Pressure--High

14. Reactor Coolant Pump Power Monitors S R(9) M 1, 2 l

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 (TP) is above 30% of RATED THERMAL POWER (RTP), compare out-of-core measured AXIAL POWER IMBALANCE (APlo) to incore measured AXIAL POWER IMBALANCE (APIg) as follows:

(APlo - APII ) = Imbalance Error TP Recalibrate if the absolute value of the Imbalance Error is equal to or greater than 3.5%

(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 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.

CRYSTAL RIVER - UNIT 3 3/43-8

TABLE 4.3-2 (Continued) o ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION SURVEILLANCE REQUIREMENTS

o d CHANNEL MODES IN WHICH y CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE r- FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED 3

3. REACTOR BUILDING SPRAY C a. Reactor Building Pressure E High-High coincident with

[ HPI Signal S R M(4) 1,2,3

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

4. OTHER SAFETY SYSTEMS

a. Reactor Building Purge Exhaust g Duct Isolation on High Radioactivity ge 1. Gaseous S Q M All Modes G

b. Steam Line Rupture Matrix

1. Low SG Pressure N/A R N/A 1,2,3  !

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

c. Emergency Feedwater

, 1. MFW Pump Turbines A and B Control Oil Low S R N/A 1,2,3

3. OTSG A or B Level Low-Low S R N/A 1,2,3,4

s o g TABLE 4.3.6 m

j REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 9

r-N q CHANNEL CHANNEL g INSTRUMENT CHECK CALIBRATION e

C 2 1. Reactor Trip Breaker Indication M N.A.

-i

• 2. Reactor Coolant Temperature-Th M R

3. Reactor Coolant Pressure M R

4. Pressurizer Level M R  !

$ 5. Steam Generator Level M R

6. Steam Generator Pressure M R l

7. Decay Heat Removal M R Temperature

8. Motor Driven Emergency M R Feedwater Pressure

9. Nuclear Services Closed M R Cycle Cooling Pumps Discharge Pressure

10. Nuclear Services Cicud M R Cycle Cooling Cooler Outlet Temperature

TABLE 4.3-7 POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS .

o N

6, CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION 9

r-

% 1. P'wer Range Nuclear Flux M Q*

2. Reactor Building Pressure M R e

3. Source Range Nuclear Flux M R*

C z 4. Reactor Coolant Outlet Temperature M R

[ 5. Reactor Coolant Total Flow Rate M R

6. RC Loep Pressure M R l 7. Pressurizer Level M R

3. Steam Generator Outlet Pressure M R R 9. Steam Generator Level M R b

i (Primary EFW Flow Detector)

O 10. Borated Water Storage Tank Level M R

11. Startup Feedwater Flow Rate M R

12. Reactor Coolant System Subcooling Margin Monitor M R l 13. PORV Position Indicator (Primary Detector) M R

14. PORV Position Indicator (Backup Detector) M R I

15. PORY Block Valve Position Indicator M R

16. Safety Valve Position Indicator (Primary Detector) M R

17. Safety Valve Position Indicator (Backup Detector) M R

18. Emergency Feedwater Ultrasonic Flow Indicator M R (Backup EFW Flow Detector)

I

• Neutron detectors may be excluded from CHANNEL CALIBRATION

r-

' REACTOR COOLANT SYSTEM POWER OPERATED RELIEF VALVES LIMI!!NG 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 PORY inoperable, within I hour either restore the PORV 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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 PORV and remove power from the associated solenoid valve; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. The provisions of Specification 3.0.4 are not applicable.

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.

I l

l l

CRYSTAL RIVER - UNIT 3 3/4 4-4a l

l

e -

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 occuring on either inside or outside of a tube.

3. Degraded Tube means a tube containing imperfections &20% 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 means in 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. Plugging 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 struct. ural 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 f ar as possible.

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).

f 4.4.5.5 Reports

, a. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.

t 4 >

CRYSTAL RIVER - UNIT 3 3/44-9 l

_ . _ . , _ , _ _ _ , _ _ , _ , _ ~- , _ _ _ _ _ _ . _

e . .

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.21 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power operated 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.

c. By verifying the correct position of each mechanical position stop for the following HPI stop check valves prior to restoring the HPI system to OPERABLE status following periodic valve stroking or maintenance on the valves.

1. MUV-2

2. MUV-6

3. MUV-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-111

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 3/45-4

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

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 structural distress or corrosion.

5. Verifying a total leak rate less than or equal to 6 gallons per hour for the 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 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 signal, as appropriate.

2. Verifying that each HPl and LPI pump starts automatically upon receipt l of a high pressure or low pressure safety injection test signal, as appropriate.

g. Following completion of HPl 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:

HPl System - Single Pump 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 the flow rate for all combinations of 3 gpm.

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

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 effect.

l CRYSTAL RIVER - UNIT 3 3/45-5

TABLE 3.7-1 MAXIMUM ALLOWABLE NUCLEAR OVERPOWER TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES 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 81.95 3 67.5 I

CRYSTAL RIVER - UNIT 3 3/47-2

{

t__

o . o 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 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 propylene or other materials demonstrated compatible with the operating environment and approved as such by the NRC, shall be determined OPERABLE at least once after 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 include, but are not necessarily limited to, inspection of the hydraulic 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 previously 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 hydraulic fluid reservoirs, fluid connections, and linkage connections to the piping and anchors.

CRYSTAL RIVER - UNIT 3 3/4 7-25 i

k O

o

• d

-t r TABLE 4.7-4 2

r1 HYDRAULIC SNUBBER INSPECTION SCHEDULE

o E NUMBER OF SNUBBERS FOUND INOPERABLE NEXT REQUIRED q DURING INSPECTION OR DURING INSPECTION INTERVAL (*) INSPECTION INTERVAL **

w 0 18 months + 25%

1 12 months ~ 25 %

2- 6 months _7 25%

3 or 4 124 daysi 25 %

5, 6, or 7 62 days 1 25%.

Greater than or equal to 8 31 days 1 25 %

4 c-

?

i d

o 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.

l

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 525 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 The boron injection system ensures that negative reactivity control is available during each mode of facility 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 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 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 cooidown 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 minumum 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 l Specification 3.1.2.9 is conservative for this cycle.

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 200oF is sufficient to provide a SHUTDOWN MARGIN of 1.0% A k/k after xenon decay and cooldown from 200oF to 1400F. This condition requires either 390 gallons of 11,600 ppm boron from the boric acid 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 volumes of 13,500 gallons and 600 gallons, respectively are specified.

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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 the minimum DNBR in the core 21.30 during normal operation and during short term transients, (b) maintaining the peak linear power density g 18.0 kW/ft during normal operation, and (c) maintaining the peak power density g 20.5 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 linear 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 .Jtuation 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-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 application of the above peaking augmentation 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 these specifications are as follows:

FQ 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.

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

AH integral of linear power along the rod on which minimum 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 minimum DNBR at full power are met, provided:

Fq 4 3.13; F H S 1*7l 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 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 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 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 FNH:

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

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POWER DISTRIBUTION LIMITS BASES

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

For Condition II events, the core is protected from exceeding 20.5 kW/ft 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.

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 si 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 of 1.30 or greater throughout each analyzed transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 i the RCS total flow rate is adequate to detect flow degradation and ensure correlation of 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

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.- 3/4.7- PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE ~

3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary

- system pressure will 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 III of the ASME Boiler and Pressure Vessel Code,1971 Edition. The total relieving 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 11.0 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 following bases:

SP = X-AY x NOTS XI where: SP = reduced Nucelar Overpower Trip Setpoint in percent of Rated Thermal Power.

X = total actual relieving 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 (845,759 lbs/hr).

X l = total Ratedrequired Thermalrelieving Power in capacity (of each steam Ibs/ hour 6,160,000 lbs/hr). generator for 112%

NOTS = Nuclear Overpower Trip Setpoint specified in Table 2.2.1. .

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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 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 3.50 weight percent U-235.

CONTROL RODS 5.3.2 The reactor core shall contain 60 safety and regulating and 8 axial power shaping (ASPR) control rods. The safety and regulating control rods shall contain a nominal l 134 inches of absorber material. 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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