ML20209B296
| ML20209B296 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/15/1987 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20209B222 | List: |
| References | |
| NUDOCS 8704280407 | |
| Download: ML20209B296 (25) | |
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2.1 SAFETY LIMITS
. BASES j
2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit preven't 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 l
heat transfer coefficient is large and the cladding surface -temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary.of the nucleate-boiling regime would result in excessive cladding temperature -because of the. onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer-coefficient.
DNB is not a directly measurable parameter during operation. but THERMAL POWER and Reactor Coolant Temperature and Pressure can be related to DNB using a Critical Heat Flux (CHF) correlation.
The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux is indicative of the martin to DNB.
The curve presented in Figure 2.1-1 represents the conditions at which the minimum allowable DNBR or greater is predicted for the thermal power and number of operating reactor coolant pumps.
This curve is based on the following nuclear power peaking factors with potential fuel densification effects:
N N
N F
= 2.82 F = 1.71 F = 1.65 0
AH Z
The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum allowable control rod withdrawal, and form the core DNBR design basis.
l 2
i d
d h
CRYSTAL RIVER - UNIT 3 B 2-1
SAFETY LIMITS BASES 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.
The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and consider the effects of potential fuel densification and potential fuel rod bow:
1.
The DNBR limit produced by a nuclear power peaking factor of F
= 2.82 I
or the combination of the radial peak, axial peak and positkon of the axial peak that yields no less than the DNBR limit.
2.
The combination of radial and axial peak that causes central fuel melting at the hot spot.
l 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 the DNBR limit l
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.
CRYSTAL RIVER - UNIT 3 B 2-2
I SAFETY LIMITS i
BASES For each curve of BASES Figure 2.1, a pressure-temperature point above and to the left of the curve would result in a DNBR within the limit or.a local quality-4 at the point of minimus DNBR within the limit for that particular reactor coolant pump situation.
The curve for three pump operation is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the left of the three pump curve will be above and to the left of the other curves.
2.1.3 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel and pressurizer are designed to Section III of the ASME Boiler and Pressure Vessel Code which permits a maxinua transient pressure of 110%, 2750 psig, of design pressure.
The Reactor Coolant System piping, valves and fittings, are designed to USAS B 31.7, February, 1968 Draft. Edition, which permits a maximum transient pressure of 110%, 2750 psig, of component design pressure. The Safety Limit of 2750 psig is therefore consistent with.the design criteria and associated code requirements.
The entire Reactor Coolant system is hydrotested at 3125 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.
t I
1 i
r CRYSTAL RIVER - UNIT 3 B 2-3
REACTIVITY CONTROL SYSTEMS REGULATING R0D INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3 6 The regulating rod groups shall be limited in physical insertion to the acceptable operation region as shown on Figures 3.1-1, 3.1-2, 3.1-3, and 3.1-4, with a rod group overlap of 25 i 5\\ between sequential withdrawn groups 5 and 6, and 6 and 7.
APPLICABILITY: MODES 1* and 2*#
ACTION:
a.
With the regulating rod groups inserted in the unacceptable operation region, immediately initiate and continue boration at greater than or equal to 10 GPM of 11,600 ppm boric acid solution or its equivalent, until out of the unacceptable operation region.
Additionally, either:
- 1. Restore the regulating groups to within the acceptable region limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of exceeding the acceptable operation region, or
- 2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figures within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of exceeding the acceptable operation region, or
- 3. 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 /> of exceeding the acceptable operation region.
b.
With the regulating rod groups inserted in the restricted operation region or with any group sequence or overlap outside the specified
- limits, except for surveillance testing pursuant to Specification 4.1.2.1.2, either:
- 1. Restore the regulating groups to within the acceptable region limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of exceeding the acceptable operation region, or
- 2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figures within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of exceeding the acceptable operation region, or
- 3. 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 /> of exceeding the acceptable operation region.
- See Special Test Exceptions 3.10.1 and 3.10.2.
- With Kefg greater than or equal to 1.0.
CRYSTAL RIVER - UNIT 3 3/4 1-25
Figure 3.1-1 Regulating Rod Group Insertion Limits for Four-Pump Operation From 0 to 500+10 EFPD l
110 (200,102)
(300,102) 100
-(270,102)
(270,92) i UNACCEPTABLE RESTRICTED (250,80) y 80 OPERATION OPERATION g
70 SHUTDOWN c)
MARGIN N
LIMIT 60
?
3 50 (120,50)
(175,50)
==
$ 40 x
ACCEPTABLE E
OPERATION 30 i
20 (60,15) 10 (0,5) 0 50 100 150 200 250 300 Rod Index, % Withdrawn 0
25 50 75 100 0
25 50 75 100 l
l l
t t
i i
i E
I Group 5 Group 7 0
25 50 7,5 100 e
i I
I I
Group 6 Note: This Figure shall be used up to complete APSR withdrawal per Specification 3.1.3.9 3/4 1-27
-DELETED-CRYSTAL RIVER UNIT 3 3/4 1-27a
Figure 3.1-2 Regulating Rod Group Insertion Limits for Four-Pump Operation After 500f_10 EFPD 110 (208,102)
(300,102)
(265,102)_r 100 (260,92),
90 UNACCEPTABLE CPERATION RESTRICTED (250,80) 80
'y OPERATION O
SHUTDOWN 70 g
MARGIN 2
LIMIT 60 of 50 (127,50)
(175,50) a ACCEPTABLE 40 b
OPERATION 5
30 20 (64,15) 10 0
0 50 100 150 200 250 300 Rod Index, % Withdrawn O
25 50 75 100 0
25 50 75 100 t
t I
I i
l I
I I
I Group 5 Group 7 0
25 50 75 100 e
i I
i I
Group 6 l
Note: Tfils Figure shall be used after complete APSR withdrawal per Specification 3.1.3.9 3/4 1-28
Figure 3.1-3 Regulating Rod Group Insertion Limits for Three-Pump Operation From 0 to 500+10 EFPD l
110 100 90 (300,77) 80 b
UNACCEPTABLE (200,77)
(270,77) g 0PERATION (270,69.5) 70 i
Ey 60 RESTRICTED (250,60.5)
OPERATION i
SHUTDOWN g
50 MARGIN LIMIT a
40 (120,38) g (175,'33) c-30 ACCEPTABLE 20 OPERATION 10 (60,11.75)
(0,4.25]
0 50 100 150 200 250 300 Rod Index, % Withdrawn 0
25 50 76 100 0
25 50 75 100 e
t i
I I
I I
I I
I Group 5 Group 7 0
25 50 75 100 t
l E
t i
Group 6 Note: This Figure shall be used up to complete APSR withdrawal per Specification 3.1.3.9 3/4 1-29
-DELETED-CRYSTAL RIVER UNIT 3 3/4 1-29a
I Figure 3.1-4 Regulating Rod Group Insertion Limits for Three-Pump Operation After 500+10 EFPD 110 100 90 UNACCEPTABLE j
OPERATION (300,77) 80 g
(208,77.)
(265,77) -
x2 70 (260,69.5)
RESTRICTED (250,60.5) 60 O M ATION 5
SHUTDOWN 7
MARGIN g
50 LIMIT c:
40
~
b (175,38)
ACCEPTABLE OPERATION 20 (64,11.75) 10 0
0 50 100 150 200 250 300 Rod Index, % Withdrawn 50 75 100 0
25 50 75 100 J,5 0
t i
Group 5 Group 7 0
25 50 75 100 e
i e
i I
Group 6 Note: This Figure shall be used after complete APSR withdrawal per Specification 3.1.3.9 3/4 1-30
Fiqure 3.1-7 Control Rod Locations and Group Designations for Crystal River 3 Cycle 7 1
X Fuel Transfer Canal A
B 1
6 1
C 3
5.
5 3
D 7
8 7
8 7
E 3
5 4
4 5
3 F
1 8
6 2
6 8
1 l
G 5~
4 2
2 4
5 1
H W-6 7
2 2
7 6
-Y K
5 4
2 2
4 5
l L
1 8
6 2
6 8
1 i
l it 3
5 4
4 5
3 N
l 7
8 7
8 7
O l
3 5,
5 3
i P
l l
1 6
1 R
i l
l Z
1 2
3 4
5
.6 7
8 9
10 11 12 13 14 15 l
X Group Number Grouo No. of Rods Function 1
8 Safety 2
8 Safety 3
8 Safety 4
8 Safety 5
12 Control i
6 8
Control 7
8 Control 8
8 APSRs Total 68 l
3/4 1-34
REACTIVITY CONTROL SYSTEMS AXIAL POWER SHAPING ROD INSERTION LIMITS LIMITIN$'CONDITIONFOROPERATION A
~
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 - 490 EFPD,' the APSR's - may be
-l positioned as necessary.
' The APSR's shall be completely withdrawn o
(100%) by 510 EFPD.
Between-490 andc 510 EFPD, the APSR's may be b,'
withdrawn.
However, once withdrawn - during this period, the APSR's shall not be reinserted.
APPLICABILITY: MODES 1 and 2*.
N ACTION:
w l
With the axial power shaping' red group outside the above insertion limits, either:
I-
- 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 1
1
- 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 REQUIREMENTS 4.1.3.9 The position of the axial power shaping rod group shall be determined to be within the insertion limits at least-once'every'12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- With Regg 1 1.0.
i 4
1 t
i
[
CRYSTAL RIVER - UNIT 3 3/4 1-37 -Amendments Nos. 76,79,546,'64,77
,. ~ 4 <_ -,_
3/4.2 POWER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE F
LIMITING CONDITION FOR OPERATION 3.2.1 AXIAL POWER IMBALANCE shall be maintained within the limits shown on
^ Figures 3.2-1, and 3.2-2.
l APPLICABILITY: MODE 1 above 40% of RATED THERMAL POWER *.
ACTION:
With AXIAL POWER IMBALANCE exceeding the limits specified above, either:
- a. Restore the AXIAL POWER IMBALANCE to within its limits within 15 minutes, or
- b. Be in at least HOT STANDBY within 2 hours2.314815e-5 days <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 3/4 2-1 j
Figure 3.2-1 Axial Power Imbalance Invelope for Four-Pump Operation From 0 EFPD to E0C l
110--
(-20,102)
(15,102) 100-(-25,92)
(15,92) 90-
,(20,80)
(-28,80)<
80-70- -
b 4 60 -
c.
(-28,50)o 2 50-
'(20,50) m ACCEPTABLE UNACCEPTABLE 40 -
OPERATION OPERATION 7
a:
30 a
b' p 20 -
10 -
0
-50
-40
-30
-20
-10 0
10 20 30 40 50 Axial Power Imbalance, %
3/4 2-2
-DELETED-CRYSTAL RIVER UNIT 3 3/4 2-2a
Figure 3.2-2 Axial Power Imbalance Invelope for Three-Pump -
Operation From 0 EFPD to E0C l
110--
100- -
90-
~~
(-20,77 ) -
(15,77)
-(-25,69.5) 70-(15,69.5)
(-28,60.5)<
60 i(20,60.5) 5x 2 50- -
E 40..
E
(-28,38') i
'(20,38) i ACCEPTABLE UNACCEPTABLE 330- 0PERATI0fl 0PERATION e
"'20..
s' I
E 10- -
i i
i o
i i
i i
-50
-40
-30
-20
-10 0
10 20 30 40 50
^
Axial Power Imba1ance, %
3/4 2-3
-DELETED-l l
l i
l CRYSTAL RIVER UNIT 3 3/4 2-3a
TABLE 3.2-2 OUADRANT POWER TILT LDtIt3 STEADY STATE TRANSIENT MAXIMUM LIMIT LIMIT LIMIT QUADRANT POWER TILT as Measured by:
Symmetrical Incore Detector System 4.12 10.03 20.0 Power Range Channels 1.96 6.96 20.0 Minimum Incore Detector System 1.9 4.40 20.0 j
CRYSTAL RIVER - UNIT 3 3/4 2-11
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 (APIo) to incore measured AXIAL POWER IMBALANCE (APIj) as follows:
RTP (APIo - APIj) = Imbalance Error TP Recalibrate if the absolute value of the Imbalance Error is equal to or greater than 2.5%.
l (4) - AXIAL POWER IMBALANCE and loop flow indications only.
(5) - Verify at least one decade overlap if not verified in previous 7 days.
(6) - Each train tested every other month.
(7) - Neutron detectors may be excluded from 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.
+
J CRYSTAL RIVER - UNIT 3 3/4 3-8
f REACTIVITY CONTROL SYSTEMS J
BASES j
3/4.1.1.4 NININUM TEMPERATURE FOR CRITICALITY l
This specification ensures that' the reactor will not-be made critical with f
the Reactor Coolant system average temperature less than 525'F.
This limitation-is required to ensure.that
( 1) the moderator temperature j
coef ficient ' is within : its ; analyzed. temperature 5 range, (2)tthe protective instrumentation is within its normal operating range,-(3) the pressuriser is capable of being in an OPERABLE status with a steam bubble, and (4).the reactor pressure vessel is above its'sinimus RT t
- NDT temperature, i
3/4.1.2 BORATION SYSTEMS 4
i The boron injection system ensures that negative reactivity control - is j
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.
a
~
i
{
with the RCS average temperature above 200*F, a :minimua 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--
i inoperable.
Allowable out-of-service periods ensure that ainor. component repair or corrective-action may be completed without undue risk to overall j
facility safety from injection system failures during the repair period.
I l
The boration capability of either system is sufficient to provide a SHUTDOWN i
j' MARGIN from all operating conditions of 1.0L A k/k af ter xenon decay and -
~
l cooldown to 200*F.
The maximum boration capability requirement occurs from j
full power equilibrium xenon conditions and requires either 5,300 gallons of 11,600 ppa boric acid solution from the boric acid ~ storage tanks or 39,000 gallons of 2,270 ppe 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 i
the RCS: to the desired level.
The specified quantity of borated -water is j
consistent with the ECCS requirements of Specification 3.5.4.
Therefore,'the larger volume-of borated water is specified.
Also the 6,000 gallons ainlaus 3
i BAST requirement per Specification 3.1.2.9 is conservative for this cycle.
With the RCS temperature below 200*F, one ' injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the~ single injection system becomes~ inoperable.
i l
l The boron capability required below 200*F is sufficient to provide a SHUTDOWN j
MARGIN of 1.0%
a k/k af ter xenon decay and cooldown from. 200'F to 140'F.
i.
This condition requires either 471 gallons of 11,600 ppe boron from the boric.
j acid storage system or 2,405 gallons ' of 2,270 ppa boron from the borated -
1-water storage tank.
To envelop future cycle BWST and BAST contained borated j
water volume requirements, a mirtimum volume of.13,500 gallons and 600 gallons, respectively, are specified, n
. CRYSTAL RIVER
. UNIT 3 B 3/4 1-2 r
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 minimus DNBR in the core ). within the limit during
.l normal operation and during short term transients, (b) maintaining the peak linear power density 1 18.0 kW/ft during normal operation, and (c) maintaining the. peak power density 120.5 kW/f t during short term transients.
In addition, the above criteria 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, and 3.2-2 and the insertion limit curves, Figures 3.1-1, 3.1-2, 3.1-3, and 3.1-4, are based on LOCA analyses which have defined the maximum linear heat rate such that the maximum clad temperature will not exceed the Final Acceptance criteria of 2200*F 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 limits, 'as defined by Figures 3.1-1, 3.1-2, 3.1-3, and 3.1-4, and if the steady state limit QUADRANT POWER TILT exists.
Additional conservatism is introduced by application of:
a.
Nuclear uncertainty factors, b.
Thermal calibration uncertainty, c.
Fuel densification effects.
d.
Hot rod manufacturing tolerance factors.
The conservative application of the above peaking augmentation factors considers I
the potential peaking penalty due to fuel rod blow.
The ACTION statements which permit limited variations from. the basic requirements are accompanied by additional restrictions which ensure that the i
original criteria are met.
l The definitions of the design limit nuclear power peaking factors as used in these specifications are as follows:
Fg 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
POWER DISTRIBUTION LIMITS
. BASES 1
FN Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of AH the 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:
Fg 1 3.13; FN g 1,7j AH Power Peaking is not a directly observable quantity and therefore limits have been established on the bases of the AXIAL POWER IMBALANCE produced by the power peaking.
It has been determined that the above hot channel factor limits will be met provided the following conditions are maintained.
1.
Control rods in a single group move together with no individual rod insertion differing by more than i 6.5% (indicated position) from the group average height.
2.
Regulating rod groups are sequenced with overlapping groups as required in Specification 3.1.3.6.
f 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.
I 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, and 3.2-2.
l t
l 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 t
allowable control rod insertion and are the core DNBR design basis. Therefore, i
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 Fg andFfH a.
The measurement of total peaking factor, FgMeas, shall be increased i
by 1.4 percent to account for manufacturing tolerances and further increased by 7.5 percent to account for measurement error.
CRYSTAL RIVER - UNIT 3 B 3/4 2-2 i
_m 4
4 POWER DISTRIBUTION LIMITS 4
BASES y.
b.
The measurement. of enthalpy rise hot channel factor, F{H, shall be increased by 5 percent to account for measurement error, 4'
i.
For Condition II events, the core is protected from exceeding allowable fuel -
. melt limit locally, and from going below minimum allowable DNBR s by automatic j
protection on - power, AXIAL POWER IMBALANCE, pressure and temperature.
Only conditions 1 through 3, above,~are mandatory since'the AXIAL POWER IMBALANCE is
.an. explicit input to the Reactor protection' System.
~
1 The QUADRANT POWER TILT : limit assures-that the. radial power distribution satisfies the design values used in the power capability 'nalysis. Radial power a
j 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'4.92, for Transient' State is 11.07, and~
l i
for the Maximum Limit is 20.0.
4 The QUADRANT POWER TILT limit at which corrective action is required provides i
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 3
j uncertainty on Fg is ' reinstated by reducing the power by 2. percent for each j
percent of tilt in excess of the limit.
I I
3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.
The limits are consistent with the FSAR initial assumptions and have been analytically demonstrated adequate to maintain
{
.a DNBR within the limit throughout each analyzed transient.
l 1
]
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 j
is sufficient to ensure that the parameters are restored within their limits i
following load changes and other expected transient operation.
The 18 month periodic measurement of the RCS total flow rate is adequate to detect ' flow degradation and ensure correlation of the flow indication channels with 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.
1 i
i 1
i 5
-CRYSTAL RIVER - UNIT 3
.B 3/4 2-3 4
.