ML20072Q329
| ML20072Q329 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/30/1994 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20072Q327 | List: |
| References | |
| NUDOCS 9409090238 | |
| Download: ML20072Q329 (24) | |
Text
.
2.
SAFETY LIMITS AND LIMITING SAFETY SYSTEM 3ETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow when the reactor is critical.
Objective To maintain the integrity of the fuel cladding.
Specification 2.1.1 The maximum local fuel pin centerline temperature shall be s 5080 - (6.5 x 10-3 x (Burnup, MWD /MTU) *F) for TACO 2 applications and 54642 - (5.8 x 10-3 x (Burnup, MWD /MTU)'F for TACO 3 applications.
Operation within this limit is ensured by compliance with the Axial Power Imbalance protective limits preserved by Table 2.3-1 " Reactor Protection System Trip Setting Limits," as specified in the COLR.
2.1.2 The departure from nucleate boiling ratio shall be maintained greater than the limits of 1.3 for the EAW-2 correlation and 1.18 for the BWC correlation.
Operation within this limit is ensured by compliance with Specification 2.1.3 and with the Axial Power Imbalance protective limits preserved by Table 2.3-1
" Reactor Protection System Trip Setting Limits," as specified in the COLR.
2.1.3 Reactor Coolant System (RCS) core outlet temperature and pressure shall be maintained above and to the left of the Safety Limit shown in Figure 2.1-1.
Bases To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB). At this point there is a sharp reduction of the heat transfer coefficient, which could result in high cladding temperatures and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of a critical heat flux (CHF) correlation. The BAW-2 (1) and BWC (2) correlations have been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The BAW-2 correlation applies to Mark-B fuel and the BWC correlation applies to Mark-BZ fuel.
The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual 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 (aAW-2) and 1.18 (BWC).
Amendment No, M,M3 7
9409090238 940830 PDR ADOCK 05000313 P
A DNBR of 1.30 (BAW-2) or 1.18 (BUC) corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occurs this is considered a conservative margin to DNB for all operating conoitions.
The difference between the actual core outlet pressure and the indicated reactor coolant system pressure for the allowable RC pump combination has been considered in determining the core protection safety limits.
The curve presented in Figure 2.1-1 represents the conditions at which the DNBR is greater than or equal to the minimum allowable DNBR for the limiting combination of thermal power and number of operating reactor coolant pumps. This curve is based on the following nuclear power peaking factors (3) with potential fuel densification effects:
F = 2.83; F[H=1.71; FN = 1.65.
The Axial Power Imbalance Protective Limits in the COLR are based on the l
more restrictive of two thermal limits and include the effects of potential fuel densification:
1.
TheDNBRlimitproducedbyanuclearpowerpeakingfactorofPk
= 2.83 or the combination of the radial peak, axial peak and position of the axial peak that yields no less than the DNBR limit.
2.
The combination of radial and axial peak that prevents central fuel melting at the hot spot as given in the COLR.
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 flow rates for curves 1, 2,
and 3 of Figure 2.1-3 correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.
The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump maximum thermal power combinations shown in Figure 2.1-3.
The I
curves of Figure 2.1-3 represent the conditions at which the DNBR limit is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation.
If the actual pressure / temperature point is i
below and to the right of the pressure / temperature line the Safety Limit is exceeded.
The local quality at the point of minimum DNBR is less than 22 percent ( BAW-2 ) (1) or 26 percent (BWC) (2).
l l
l i
e Amendment No. M, W, M, +H 8
l l
l l
THIS PAGE INTENTIONALLY LEFT BIRIK l
l Amendment lio. 6, M, M,4,62, G, 9b
%, %, M3,4-M
2. 3' LIMITZWG SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to instruments monitoring reactor power, reactor power imbalance, I
reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure.
Objective To provide automatic protection action to prevent any combination of process variables from exceeding a safety limit.
Specification 2.3.1 The reactor protection system trip setting lindts and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 cnd the Protection System Maximum Allowable Setpoints for Axial Power Imbalance as given in the COLR.
j Bases The reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the degree that a safety limit nay be reached.
The trip setting limits for protection system instrumentation are listed in Table 2.3-1.
The safety analysis has been based upon these protection system instrumentation trip setpoints plus calibration and instrumentation l
errors.
Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.
j During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 104.9 percent of rated power. Adding to this the possible variation in trip setpoints due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which is the value used in the safety analysis.
A. Overpower Trip Based on Flow and Imbalance The power level trip set point produced by the reactor coolant i
system flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the design, the loss-of-coolant-flow accident from high power.
Analysis has demonstrated that the specified power-to-flow ratio is adequate to prevent a DNBR of less than 1.30 (BAW-2) or 1.18 (BWC) should a low flow condition exist due to any electrical malfunction.
i l
l I
Amendment 41, 43, 67 11
The' power level trip setpoint produced by the power-to-flow ratio provid'es 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 trip 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.
The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.
No penalty in reactor coolant flow through the core was taken for an open core vent valve because of the core vent valve surveillance program during each refueling outage.
For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.
The 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 reactor power imbalance (power in top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of the Protection System Maximmm Allowable Setpoints for Axial Power Imbalance in the COLR are produced. The power-to-flow ratio reduces the power level trip associated reactor power-to-reactor power imbalance boundaries by the value specified in the COLR for a 1 percent flow reduction.
l B.
Pump Monitors In conjunction with the power imbalance / flow trip, the pump monitors prevent the minimum core DNBR from decreasing below 1.30 (BAW-2) or 1.18 (BWC) by tripping the reactor due to the loss of reactor coolant l
l Ame ndme nt No. 24, M,44, 64,64, GG, 12 M4
i l
l I
THIS PAGE INTENTIONALLY LEFT BLANK l
I i
l i
i l
I l
Amendment 6, M, M, e, M, p,4, M, 14b 444,H4
Table 2.3-1 Reactor Protection System Trip Setting Limits One Reactor Coolant Pmo Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Loop (*)
l Operating (Nominal Operating (Nominal (Nominal Operating Shutdown Operating Power - 100%)
Operating Power, 75%)
Power, 49%)
Bypass Nuclear power, % of 104.9 104.9 104.9 5.0(a) l rated, max N0 clear Power based on Protection System Maximum Protection System Maximum Protection System Maximum Bypassed flovtb) and imbalance, Allowable Setpoints for Allowable Setpoints for Allowable Setpoints for
% of rated, max Axial Power Imbalance Axial Power Imbalance Axial Power Imbalance envelope in COLR envelope in COLR envelope in COLR Nuclear Power based on NA NA 55 Bypassed pump munitors, % of rated, max (c) l High RC system 2355 2355 2355 1720(a) l pressure, psig, max Low RC system 1800 1800 1800 Bypassed pressure, psig. min Variable low RC 13.89 Tout-6766(d) 13.89 Tout-6766(d) 13.89 Tout-6766(d)
Bypassed l
system pressure, psig, min RC temp, F,
max 618 618 618 618 High reactor building 4 (18.7 psia) 4(18.7 psia) 4(18.7 psia) 4(18.7 pressure, psig, max psia) i (a) Automatically set when other segments of the RPS (as specified) are bypassed.
(b) Reactor coolant system flow, %
(c)The pump monitors also produce a trip on (a) loss of two RC pumps in one RC loop, and (b) loss of one or two RC pumps during two-pump operation.
(d)T is in degrees Fahrenheit (F).
out
(*) operation with one Reactor Coolant Pump operating in each loop is limited to 24 hrs, with the reactor critical.
Amendment No. G,24,44,44,63, G,93,444,443 15 i
l 3.1 REACTOR COOLANT SYSTEM l
Applicability Applies to the operating status of the reactor coolant system.
Objective To specify those limiting conditions for operation of the reactor coolant system which must be met to ensure safe reactor operations.
3.1.1 Operational Components Specification 3.1.1.1 Reactor Coolant Pumps A.
Pump combinations permissible for given power levels shall be as shown in Table 2.3-1.
Operation with one Reactor Coolant Pump operating in each loop is limited to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with the reactor critical.
B.
The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.
With no reactor coolant pumps or decay heat removal pumps running, immediately suspend all operations involving a reduction of boron concentration in the reactor coolant system.
3.1.1.2 Steam Generator A.
Two steam generators shall be operable whenever the reactor coolant average temperature is above 280 F.
1 3.1.1.3 Pressurizer Safety Valves A.
Both pressurizer code safety valves shall be operable when the reactor is critical. With one pressurizer code safety valve inoperable, either restore the valve to operable status within 15 minutes or be in HOT l
SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B.
When the reactor is suberitical, at least one pressurizer code safety valve shall be operable if all reactor coolant system openings are closed, except for j
hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code,Section III.
The provisions of Specification 3.0.3 are not applicable.
3.1.1.4 Reactor Internals Vent Valves The structural integrity and operability of the reactor internals vent valves shall be maintained at a level censistent with the acceptance criteria in Specification 4.1.
The provisions of Specification 3.0.3 are not applicable.
3.1.1.5 Reactor Coolant Loops A.
With the reactor coolant average temperature above 280*F, the reactor coolant loops listed below shall be operable:
Amendment No. 3-1, M, M, -121 16
l 3.1.8 Low Power Physics Testing Restrictions Specification The following special limitations are placed on low power physics testing.
3.1.8.1 Reactor Protective System Requirements A.
Below 1720 psig, shutdown bypass trip setting limits shall apply in accordance with Table 2.3-1.
B.
Above 1800 psig, nuclear overpower trip shall be set at less than 5.0 percent.
Other settings shall be in accordance with Table 2.3-1.
3.1.8.2 Startup rate rod withdrawal hold ( 5) shall be in effect at all times.
3.1.8.3 During low power physics testing the minimum reactor coolant temperature for criticality shall be to the right of the criticality limit of Figure 3.1.2-2.
The shutdown margin shall be maintained greater than or equal to that specified in the COLR with the highest worth control rod fully withdrawn.
Bases The above specification provides additional safety margins during low power physics testing.
REFERENCES (1)
FSAR, Section 7.2.2.1.3.
Amendment No. M 31 s
3.5.2 Control Rod Group and Power Distribution Limits Applicability This specification applies to power distribution and operation of control rods during power operation, Objective To assure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion from a hypothetical control rod ejection, and to assure core subcriticality after a reactor trip.
l Specification i
3.5.2.1 The available shutdown margin shall be greater than or equal to that sfecified in the COLR with the highest worth control rod fult.y withdrawn. With the shutdown margin less than that required, immediately initiate and continue boration injection until the required shutdown margin is restored.
3.5.2.2 Operation with inoperable rods:
1.
Operation with more than one inoperable rod, as defined in Specification 4.7.1 and 4.7. 2. 3, in the safety or i
ragulating rod groups shall not be permitted.
2.
If a control rod in the regulating or safety rod groups is declared inoperable in the withdrawn position as defined in Specification 4.7.1.1 and 4.7.1.3, an evaluation shall be initiated immediately to verify the existence of an available shutdown margin greater than or equal to that specified in the COLR.
Boration may be initiated either to the worth of the inoperable rod or until the regulating and transient rod groups are withdrawn to the limits of Specification 3.5.2.5.3, whichever occurs first.
Simultaneously a program of exercising the remaining regulating and safety rods shall be initiated to verify operability.
l 3.
If within one (1) hour of deterndnation of an inoperable rod as defined in Specification 4.7.1, it is not determined that an available shutdown margin greater than or equal to that specified in the COLR exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to the Hot Standby condition until this margin is established.
4.
Following the determination of an inoperable rod as l
defined in Specification 4.7.1, all remaining rods shall be exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised weekly until the rod problem is solved.
i l
i 5.
If a control rod in the regulating or safety rod groups is declared inoperable per 4.7 l.2, power shall be reduced to 60% of the thermal power allowable for the reactor coolant pump combination.
Amendment No. 67 46
6.
If a control rod in the regulating or axial powar shaping groups is declared inoperable per Specification 4.7.1.2 operation above 60 percent of the thermal power allowable for the reactor coolant pump combination may continue provided the rods in the group are positioned such that the rod that was declared inoperable is contained within allowable group average position limits of Specification 4.7.1.2 and the withdrawal limits of Specification 3.5.2.5.3.
3.5.2.3 The worth of single inserted control rods during criticality are limited by the restrictions of Specification 3.1.3.5 and the Control Rod Position Limits defined in Specification 3.5.2.5.
3.5.2.4 Quadrant Power Tilt:
1.
Except for physics tests, if quadrant power tilt exceeds the tilt limit set in the CORE OPERATING LIMITS REPORT, reduce power so as not to exceed the allowable power level for the existing reactor coolant pump combination less at least 2%
for each 1% tilt in excess of the tilt limit.
2.
Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall be reduced to less than the tilt limit except for physics tests, or the following adjustments in setpoints and limits shall be made:
a.
The Protection System Maximum Allowable Setpoints for Axial Power Imbalance in the COLR shall be reduced 2%
i in power for each 1% tilt in excess of the tilt limit.
b.
The control rod group and APSR withdrawal limits shall be reduced 2% in power for each 1% tilt in excess of the tilt limit.
c.
The reactor power imbalance setpoints shall be reduced 2% in power for each 1% tilt in excess of the tilt limit.
3.
If quadrant power tilt is in excess of 25%, except for physics tests or diagnostic testing, the reactor will be placed in the hot shutdown condition.
Diagnostic testing during power operation with a quadrant power tilt is permitted provided the thermal power allowable for the 1
reactor coolant pump combination is restricted as stated in 3.5.2.4.1 above.
4.
Quadrant power tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15% of rated power.
i Amendment !!o. 6, M, M,44, M, M6, 47 ble, M-7, %4
4 4
3 MARKUP OF CURRENT ANO-1 TECHNICAL SPECIFICATIONS (FOR INFO ONLY) l l
i
l l
i l
2.
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i
2.1 SAFETY LIMITS, REACTOR CORE l
Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow when the reactor is criticaldaring pcuer cperatien Of the planb.
Objective To maintain the integrity of the fuel cladding.
Specification 2.1.1 The maximum local fuel pin centerline temperature shall be s 5080 - ( 6. 5 x 10" x (Burnup, MWD /MTU)*F) for TACO 2 applicatiogg and s4642 - (5.8 x 10" x (Burnup, MWD /MTU)'F for TACO 3 applications.
Operation within this limit is ensured by compliance with the Axial Power Imbalance protective limits preserved by Table 2.3-1 " Reactor Protection System Trip Settina Limits," as specified in the COLR.
The combination of the reactcr syster prc ure and coolent t-empeeatwc chall not exceed the e fety limit : defined by the 1ccus cf point catablished in Figure 2.1-1.
If-the actual preocurc/ temperature peint is bcica and tc the r4 ht-ef--t-he 9
prc ure/tcmperatdre lin the cafety limit is encceded.
2.1.2 The departure from nucleate boilina ratio shall be maintained greater than the limits of 1.3 for the BAM-2 correlation and 1.18 for the BWC correlation.
Operation within this limit is gnsured by compliance with Specification 2.1.3 and with the Axial Power Imbalance _ protective limits preserved by Table 2.3-1
" Reactor Protection System Trip Settina Limits," as specified in the COLR.
The ccabination of Icccter thermal pcucr and reacter pcucr imbalance - {pcue r ir the top half cf the ccre minu the pcuer in t-he better half cf the cer: caprcaccd c: ; pcrecntage of the rated pcuc r } chall not encced the cafety limit c defined by the lecu cf peint f o r t h e s p e c14+ed-f4ew-eet--f e r t h i n Fi g u r e 2.1 -- 2.
If the actual reacter-thcrmal pcwcr,Meectcr pcuer imbalance point is above the line for the pccified ficw, the-eafety--l-i*i t is caeeeledr l
2.1.3 Reactor Coolant System (RCS) core outlet temperature and pressu,Le l
shall be maintained above and to the left of the Safety Limit shown in Ficure 2.1-1.
Bases To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB).
At this point there is a sharp reduction of the heat transfer coefficient, which could result in high cladding temperatures and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron Amendment No. 21, 113 7
l l
power, reactor coolant flow, temperature, and pressure can be related to i
DNB through the use of a critical heat flux (CHF) correlation.
The BAW-2 (1) and BWC(2) correlations have been developed to predict DNB and the location of LNB for axially uniform and non-uniform heat flux distributions.
The BAW-2 correlation applies to Mark-B fuel and the BWC correlation applies to Mark-BZ fuel.
The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core
]
location to the actual heat flux, is indicative of the margin to DNB.
The j
minimum value of the DNBR, during steady-state operation, normal i
operational trancients, and anticipated transients is limited to 1.30 (BAW-2) and 1.18 (BWC).
i i
i 4
I
?
i 1
i l
i i
1 i
i 1
I J
l i
1 i
t i
i 4
E i
i.
i t
1 Amendment No. 24, 113 7
l l
f THIS PAGE INTENTIONALLY LEFT BLANKCor: Protection Safety I.i:niW Figure 2.1-2 l
l l
l Ther.01 Peuer Level 4-FP l
120 --
l
(-33.5,112.0)
(33.5,112) l n
(
i ACCEPT 39' E j
i4 PUMP j
(42.6,103.6) joPERATION 100 __
j j
(-49.7,96.9) j(-33.5,90.75)
(33.5,90.75)i l
l ACCEPTABLE i
f
' 3 & 4 PUMP i
! (42.6,82.35)
! OPERATION 80 --
i
(-49.7,96.9) j j
i i
5 l
!(-33.5,64.08)
(33.5,64.08)!
i I
i
[ ACCEPTABLE 60 --
!\\
2,3&4 PUMP N (42.6,55.68)
! OPERATION i
(-49.7,48.98).
j i
i i
i
?
e e
a l
40 --
i l
i i
i i
i i
i i
i j
j 20 --
j i
i i
i i
I i
i f
I i
i t '
i i
L 1
I n
l l
l 1
-60
-40
-20 0
20 40 60 80 l
Beeeter Power I-talance, 1
1 1
1 1
Amendment No. 4, M, M,44, M, M, 9b M,M,M3,137
l I
A DNBR of 1.30 (BAW-2) or 1.18 (BWC) corresponds to a 95 percent i
. probability at a 95 percent confidence level that DNB will not occurs this is considered a conservative margin to DNB for all operating conditions.
The difference between the actual core outlet pressure and the indicated l
reactor coolant system pressure for the allowable RC pump combination has been considered in determining the core protection safety lindts.
The curve presented in Figure 2.1-1 represents the conditions at which the i
(
DNBR is greater than or equal to the minimum allowable DNBR for the limiting combination of thermal power and number of operating reactor coolant pumps.
This curve is based on the following nuclear power peaking factors (3) with potential fuel densification effects:
FN = 1.65.
F = 2.83; F[H=1.71; The Axial Power Imbalance Protective Limits in the COLR eetvc cf Figurc 2.1-2 l
l are based on the more restrictive of two thermal limits and include the effects of potential fuel densification:
1.
The DNBR limit produced by a nuclear power peaking factor of Fq l
= 2.83 or the combination of the radial peak, axial peak and position of the axial peak that yields no less than the DNBR limit.
2.
The combination of radial and axial peak that prevents central fuel melting at the hot spot'as aiven in the coLR. The limit is 20.5 W/-f+.
Power peaking is not a directly observable quantity and therefore lindts have been established on the basis of the reactor power imbalance produced by the power peaking.
i The flow rates for curves 1, 2, and 3 of Figure 2.1-3 correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.
j The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump maximum thermal power combinations shown.in Figure 2.1-3.
The curves of Figure 2.1-3 represent the conditions at which the DNBR limit is l
predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation.
If the actual pressure / temperature point is below and to the richt of the pressure / temperature line the Safety Limit is exceeded.
The local quality at the point of minimum DNBR is less than 22 percent ( BAW-2 ) (1) or 26 percent (BWC) (2).
I 1
l Amendment No. M, M, M, 113 8
l l
2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure.
Objective To provide automatic protection action to prevent any combination of process variables from exceeding a safety limit.
Specification 2.3.1 The reactor protection system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and the Protection System Maximum Allowable Setpoints for Axial Power Imbalance as civen in the COLRFigure 2. 3 --2.
Bases The reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the degree that a safety limit may be reached.
The trip setting limits for protection system instrumentation are listed in Table 2.3-1.
The safety analysis has been based upon these protection system instrumentation trip setpoints plus calibration and instrumentation errors.
Nuclear Overpower j
i l
A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.
During normal plant operation with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level reaches 104.9 percent of rated power. Adding to this the possible variation in trip setpoints due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which is the value used in the safety analysis.
A. Overpower Trip Based on Flow and Imbalance l
The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the design, the loss-of-coolant-flow accident from high power. Analysis has demonstrated that the specified power-to-flow ratio is adequate to prevent a DNBR of less than 1.30 (BAW-2) or 1.18 (BWC) should a low flow condition exist due to any electrical malfunction.
Amendment 24,43,67 11
The po. war level trip setpoint produced by the pownr-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 trip 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.
The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor l
protective system receives a conservative indication of the RC flow.
No penalty in reactor coolant flow through the core was taken for an open core vent valve because of the core vent valve surveillance program during each refueling outage.
For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.
t The power-imbalance boundaries are established in order to prevent reactor thermal lindts from being exceeded. These thermal lindts are either power peaking kw/ft limits or DNBR limits. The reactor power imbalance (power in top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of the Protection System Maximgg j
Allowable Setpoints for Axial Power Imbalance in the COLRFigurc 2.2 2 are produced. The power-to-flow ratio reduces the power level trip associated reactor power-to-reactor power imbalance boundaries by the value specified in the COLRI.07 percent for a 1 percent flow reduction.
l l
B.
Pump Monitors i
In conjunction with the power imbalance / flow trip, the pump l
monitors prevent the minimum core DNBR from decreasing below 1.30 (BAW-2) or 1.18 (BWC) by tripping the reactor due to the l
loss of reactor coolant i
I l
i l
l l
i Amendment No. 24, M,43,62,64, GG,113 12 l
l I
u.._...:_.._-..,~.
,,_.._u..
k_ _ _ 2.. _.e.,..__
.e..__.r....o
.m.i,, A r.:_.. _ _...,..-
r..
l Thc-c.1 "cw : Levelv l
+-FP 1
_ 24
,,e.n,,n,,
s MO s
~_
e
,v e n e n v..u e, e
-F 3,...e,n,...
1 m
. ~
n n. e n..w.m. T. A s.,t 3
v u
w a
..e.,
g g,
,__,g GG e
0 1
1 (19.0,79.9)
,__,e.n._,.n.n, e
e s
e s
e Accs ' Ans s
....e, A M. P. n.R.P. T. AL..T s
r,
, _, e. e, e n.,.,,
e i
i i
e I
,, e. n., e n. e s, l
,., e. n., e.. c.,
s.
I e
e a
. n. n. e m. e.,, m, e 40 e
e.e.e.es g
- 9.. S..P #.
M. T fi. t.
g
,. S e. e, S.S.,.1 A M. P B.P.P. T Al.f B
5 8
8 e
20 i
i i
e i
e i
e e
e a
8 8
s I
I I
B I
0 t'
I I
I 1
1 t
1 l
l
-60
-40
-20 0
20 40 60 i
l 1
Reactor Power Imbalance, %
1 l
I 1
l 4
i I
t l
l i
l i
l'
- cendme nt 6, el, M,44, Ge, G,4, ee, 14b 443, 137
i
(
t Table 2.3-1 Reactor Protection System Trip Setting Limits s
One Reactor Coolant Pump Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Loop.L 1 l
Operating (Nominal Operating (Nominal (Nominal Operating Shutdown Operating Power - 100%)
Operating Power, 75%)
Power, 49%)
D ass Nuclear power, % of 104.9 104.9 104.9 5.0181 l
rated, max Protection System Maximum Protection System Maximum Protection System Maximum Nuclear Power based on 1.07 tincs flou rdnus 1.07 tiscs flor minus 1.07 tiscs flce minus Bypassed flowib1 and imbalance, Allowable Setpoints for Allowable Setpoints for Allowable Setpoints for j
% of rated, max reduction duc to rcduction duc to rcduction duc to i
Axial Power Imbalance Axial Power Imbalance Axial Power Imbalance Ebclancc;c) im5cicnce(c) imicicncc(c)
~
, envelope in COLR envelope in COLR envelope in COLR Nuclear Power based on NA NA 55 Bypassed pump monitors, % of t
rated, maxic1 l
l l
High RC system 2355 2355 2355 17201a1 l
pressure, psig, max Low RC system 1800 1800 1800 Bypassed pressure, psig. min 3
Variable low RC 13.89 Tout-67661d1 13.89 Tout-67661d1 13.89 Tout-67661d1 Bypassed l
i system pressure, l
psig, min RC temp, F,
max 618 618 618 618 High reactor building 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) j pressure, psig, max da1 Automatically set when other segments of the RPS (as specified) are bypassed.
Jb1 Reactor coolant system flow, %
dc1The pump monitors also produce a trip on (a) loss of two RC pumps in one RC loop, and (b) loss of one or two RC pumps during two-pump operation.
Id1T is in degrees Fahrenheit (F).
out
}*IOperation with one Reactor Coolant Pump operating in each loop is limited to 24 hrs. with the reactor critical.
Amendment 3, 24, 43, 49, 63, 63, GG,444,113 15 i
3.1 REACTOR COOLANT SYSTEM l
Applicability Applies to the operating status of the reactor coolant system.
Objective To specify those limiting conditions for operation of the reactor coolant system which must be met to ensure safe reactor operations.
3.1.1 Operational Components Specification 3.1.1.1 Reactor Coolant Pumps A.
Pump combinations permissible for given power levels shall be as shown in Table 2.3-1.
Operation with one Reactor Coolant Pumg_gperatina in each loop is limited to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with the_zg_ actor critical.
B.
The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.
With no reactor coolant pumps or decay heat removal pumps running, immediately suspend all operations involving a reduction of boron concentration in the reactor coolant system.
3.1.1.2 Steam Generator I
A.
Two steam generators shall be operable whenever the reactor coolant average temperature is above 280 F.
3.1.1.3 Pressurizer Safety Valves A.
Both pressurizer code safety valves shall be operable when the reactor is critical. With one pressurizer code safety valve inoperable, either restore the valve to operable status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
1 B.
When the reactor is suberitical, at least one
]
pressurizer code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code,Section III.
The provisions of l
Specification 3.0.3 are not applicable.
3.1.1.4 Reactor Internals Vent Valves The structural integrity and operability of the reactor
)
internals vent valves shall be maintained at a level consistent with the acceptance criteria in Specification 4.1.
The provisions of Specification 3.0.3 are not applicable.
3.1.1.5 Reactor Coolant Loops A.
With the reactor coolant average temperature above 280 F, the reactor coolant loops listed below shall be operable:
i Amendment No. M, %, M,44 16 l
a 3.1.8, Low Powar Physics Testing Restrictions Specification The following special limitations are placed on low power physics testing.
3.1.8.1 Reactor Protective System Requirements A.
Below 1720 psig, shutdown bypass trip setting limits shall apply in accordance with Table 2.3-1.
B.
Above 1800 psig, nuclear overpower trip shall be set at less than 5.0 percent Other settings shall be in accordance with Table 2.3-1.
3.1.8.2 Startup rate rod withdrawal hold (1) shall be in effect at all times.
3.1.8.3 During low power physics testing the minimum reactor coolant temperature for criticality shall be to the right of the A A-mittimum shutdown margin criticality limit of Figure 3.1.2-2.
T cf it Ak/k shall be maintsined greater than or equal to that specified in the COLR with the highest worth control rod l
fully withdrawn.
Bases l
The above specification provides additional safety margins during low power physics testing.
REFERENCES i
I (1)
FSAR, Section 7.2.2.1.3.
l l
l l
l i
I l
l l
l t
Amendment No. 28 31 I
.a 3.5.2.
Control Rod Group r.nd Power Distribution Limits Applicability i
This specification applies to power distribution and operation of control l
rods during power operation.
Objective i
To assure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion from a hypothetical control rod ejection, and to assure core subcriticality after a reactor trip.
Specification 3.5.2.1 The available shutdown margin shall be greaternet 1::: than or equal to that specified in tre COLRi? ak/t with the highest worth control rod fully withdrawn. With the l
shutdown margin less than that reauiredle-Ak/h, immediately initiate l
l and continue boration injection until the required shutdown margin is restored.
3.5.2.2 Opera. tion with inoperable rods:
1.
Operation with more than one inoperable r: a, as defjned in Specification 4.7.1 and 4.7.2.3, in the safety or regulating rod groups shall not be permitted.
2.
If a control rod in the regulating or safety rod groups is declared inoperable in the withdrawn position as defined i
i in Specification 4.7.1.1 and 4.7.1.3, an eveluation shall be initiated immediately to verify the existence of ante Ak/k available shutdown margin areater than or equal to that specified in the COLR.
Boration may be initiated either to the worth of the inoperable rod or until the regulating and transient rod groups are withdrawn to the limits of Specification 3.5.2.5.3, whichever occurs first.
l Simultaneously a program of exercising the remaining regulatiLs and safety rods shall be initiated to verify operability.
3.
If within one (1) hour of determination of an inoperabl rod as defined in Specification 4.7.1, it is not determineo that an 15 A'/4 available shutdown margin areater than or equal j
to that specified in the COLR exists combining the worth of the inoperable rod with each of the other I
rods, the reactor shall be brought to the Hot Standby condition until this margin is established.
4.
Following the determination of an inoperable rod as defined in Specification 4.7.1, all remaining rods shall be exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised weekly until the rod problem is solved.
]
5.
If a control rod in the regulating or safety rod groups is declared inoperable per 4.7.1.2, power shall be reduced to 60s of the thermal power allowable for the reactor coolant pump combination.
Amendment No. 57 46
4 6.
If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification 4.7.1.2 operation above 60 percent of the thermal power allowable for the reactor coolant pump ccmbination may continue provided the rods in the group are positioned such that the rod that was declared inoperable is contained within allowable group average position limits of Specification 4.7.1.2 and the withdrawal limits of Specification 3.5.2.5.3.
3.5.2.5 The worth of single inserted control rods during criticality are lindted by the restrictions of Specification 3.1.3.5 and the Control Rod Position Limits defined in Specification 3.5.2.5.
)
3.5.2.4 Quadrant Power Tilt:
1.
Except for physics tests, if quadrant power tilt exceeds the tilt limit set in the CORE OPERATING LIMITS REPORT, reduce power so as not to exceed the allowable power level for the existing reactor coolant pump combination less at least 2%
for each 1% tilt in excess of the tilt limit.
2.
Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall be reduced to less than the tilt limit except for physics tests, or the following adjustments in setpoints and limits shall be made:
The Pprotection Seystem Mmaximum 3 allowable {eatpoints for a.
Axial Power Imbalance in the COLR (Figure 2.2 21 shall be reduced 2% in power for each 1%
tilt in excess of the tilt limit.
b.
The control rod group and APSR withdrawal limits shall be reduced 2% in power for each 1% tilt in excess of the tilt limit.
The reactor power imbalance retpoints shall be reduced c.
2% in power for each 1% tilt in excess of the tilt limit.
3.
If quadrant power tilt is in excess of 25%, except for physics tests or diagnostic testing, the reactor will be placed in the hot shutdown condition.
Diagnostic testing during power operation with a quadrant power tilt is permitted provided the thermal power allowable for the reactor coolant pump combination is restricted as stated in 3.5.2.4.1 above.
4.
Quadrant power tilt shall be monitored on a minimum frequency of once every two hours during power operation I
above 15% of rated power.
l Amendment No. 6, M, M,43, IG, M6, 47 443,447,159 i
.,