ML20072F010
| ML20072F010 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/20/1983 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20072F003 | List: |
| References | |
| NUDOCS 8306270234 | |
| Download: ML20072F010 (4) | |
Text
o a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits.
The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip set points to correspond to the elevated location where the pressure is actually measured.
The curve presented in Figure 2.1-1 represents the ccnditions at which a minimumDNBRof1.3ispredictedforthemaximumposgblethermalpower(112 percent) when the reactor coolant flow is 139.8 x 10 lbs/h, which is less than the actual flow rat'e for four operating reactor coolant pumps. This curve is based on the following nuclear power peaking factors (2) with potential fuel densification and fuel rod bowing effects; F
= 2.57, F
= 1.71; F
= 1.50 q
4H z
The 1.5 axial peaking factor associated with the cosine flux shape provides a lesser margin to a DNBR of 1.3 than the 1.7 axial peaking factor associated with a lower core flgx distribution.
For this reason the cosine flux shape and the associated F = 1.50 is more limiting and thus the more conservative z
assumption.
The 1.50 cosine axial flux shape in conjunction with FAH = 1.71 define the reference design peaking condition in the core for operation at the maximum overpower.
Once the reference peaking condition and the associated thermal-hydraulic situation has been established for the hot channel, then all other conbinations of axial flux shapes and their accompanying radials must result in a condition which will not violate the previously established design criteria on DNBR. The flux shapes examined include a wide range of positive and negative offset for steady state and transient conditions.
These design limit power peaking factors are the most restrictive calcuated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion, and form the core DNBR design basis.
i The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod j
bowing; a.
The 1.3 DNBR iimit produced by a nuclear power peaking N
factor of F = 2.57 of the combination of the radial peak, axial peak, Snd position of the axial peak that yields no l
l less than 1. 3 DNBR.
b.
The combination of radial and axial peak that prevents central fuel melting at the hot spot.
The limit is 20.15 kW/ft.
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.
l The specified flow rates for curves 1, 2, and 3 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps, three pumps, and one l
pusp'in each Loop, respectively.
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i TABLE.2.3-l
. REACTOR PROTECTION SYSTEM TRIP. SETTING LIMITS i^
~*
Four Reactor Coolant Three Reactor Coolant One Reactor Coolant -
Pumps Operating ~
. Pumps Operating Pump Operating in
-(Nominal Operating (Nominal Operating Each Loop (Nominal
. Shutdown Power - 100%)
Power - 75%)-
. Operating Power 49%)
Bypass
' 1.
Nuclear power, Max.
105.5 105.5 105.5 5.0(3)
% of-rated power 2.
Nuclear power based on 1.08 times flow 1.08 times flow.
1.08 times flow minus Bypassed
. flow (2) and imbalance minus reduction due minus reduction due reduction due to max. of rated power
-to imbalance to imbalance' imbalance 3.
Nuclear power based NA NA 55%
Bypassed l,
(5) on pump monitors, Max. % of rated power 4.
High reactor coolant sys-2300 2300 2300' 1720(4) tem pressure, psig max.
Y 5.
Low reactor coolant sys-1900 1900 1900 Bypassed tem pressure, psig min.
6.
Variable low reactor' (11.75 Tout-5103)(1)
(11.75 Tout-5103)(1)
(11.75 Tout-5130)(1)
Bypassed coolant system pres-sure psig, min.
7.
Reactor coolant temp.
619 619 619 619 F., Max.
8.
liigh Reactor Building 4
4 4
4 pressure, psig, max.
(1)' Tout is in degrees Fahrenheit (F)
(2) Reactor coolant system flow,'%
(3) Administratively controlled reduction set only during reactor shutdown (4) Automatically set when other segments of the RPS (as specified) are bypassed.
(5) The' pump monitors also produce a trip on:
(a) loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operation.
(6) Trip settings limits are setting limits on the setpoint side of the protection system bistable connectors.
)
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f.
If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification 4.7.1.2., operation may continue provided the rods in the group are positioned such that the rod was declared inoperable is maintained within allowable group
[ '
average position limits of Specification 4.7.1.2.
i g.
If the inoperable rod in Paragraph "e" above is in groups 5, 6, 7, or 8, the other rods ~in the group may be trimmed to the same position.
Normal operation of 100_ percent of the thermal power allowable for the reactor coolant pump combination may then continue provided that the rod that was declared inoperable is maintained within allowable group average position limits in 3.5.2.5.
1.
L 3.5.2.3 The worth of single inserted control rods during criticality are 1-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 tilt:
Except for physics tests the quadrant tilt shall not exceed a.
+3.52% as determined using the full incore derector system.
b.
When the full incore detector system is not available and except for physics tests quadrant tilt shall not exceed +1.96% as determined using the power range channels displayed on the console each quadrant (out of core detection system).
When neither detector system above is available and, except for c.
physics tests, quadrant tilt shall not exceed +1.90% as determined using the minimum incore detector system.
d.
Except for physics tests if quadrant tilt exceeds the tilt limit power shall be reduced immediately to below the power level cutoff (see Figure 3.5-2A, and 3.5-2B). Moreover, the power level cutoff value shall be reduced 2 percent for each 1 percent tilt in excess of the tilt limit.
For less than four pump operation, thermal power shall be reduced 2 percent of the thermal power allowable for the reactor coolant pump combination for each 1 percent tilt in excess of the tilt limit.
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 e.
l reduced to less than the tilt limit except for physics tests, or the following adjustments in setpoints and limits shall be made:
1.
The protection system reactor power / imbalance envelope trip setpoints shall be reduced 2 percent in power for each 1 percent tilt.
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