ML20212K816
| ML20212K816 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/05/1987 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20212K772 | List: |
| References | |
| 3289F, NUDOCS 8703090424 | |
| Download: ML20212K816 (8) | |
Text
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2.2 SAFETY LIMITS - REACTOR SYSTEM PRESSURE Applicability Applies to the limit on reactor coolant. system pressure.
Objective To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity.
Specification 2.2.1 The reactor coolant system pressure shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel.
Bases The reactor coolant system (1) serves as a barrier to prevent radionuclides in the reactor coolant from reaching the atmosphere.
In the event of a fuel cladding failure, the reactor coolant system is a barrier against the release of fission products. Establishing a system pressure limit helps to assure the integrity of the reactor coolant system. The maximum transient pressure all,wable in the reactor coolant system pressure vessel under the ASME Code,Section III, is 110% of design pressure (2). The maximum transient pressure allowable in the reactor coolant system piping, valves, and fittings under ANSI Section B31.7 is 110% of design pressure. Thus, the safety limit of 2750 psig (110% of the 2500 psig design pressure) has been established (2). The maximum settings for the reactor high pressure trip (2355 psig) and the pressurizer code safety valves (2500 psig)(3) have been established in accordance with ASME Boiler and Pressure Vessel Code,Section III, Article 9, Winter, 1968 to assure that the reactor coolant system pressure safety limit is not exceeded. The initial hydrostatic test was conducted at 3125 psig (125% of design pressure) to verify the integrity of the reacter coolant system. Additional assurance that the reactor coolant system pressure does not exceed the safety limit is provided by the presence of a pressurizer electromatic relief valve (4).
References (1) FSAR, Section 4 (2) FSAR, Section 4.3.10.1 (3) FSAR, Section 4.2.4 (4) FSAR, Table 4-1 8703090424 870305 gDR ADOCK 05000289 PDR 2-4 3289f
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R;;ctor coolant cystca prt22ura During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip setpoint is reached before the nuclear overpower trip setpoint. The trip setting limit shown in Figure 2.3-l'for high reactor coolant system pressure ensures that the system pressure is maintained below the safety limit (2750 psig) for any design transient (6). Due to calibration and instrument errors, the safety analysis assumed a 45 psi pressure error in the high reactor coolant system pressure trip setting.
As part of the post-TMI-2 accident modifications, the high pressure trip setpoint was lowered from 2390 psig to 2300 psig.
(The FSAR' Accident Analysis Section still uses the 2390 psig high pressure trip setpoint.)
The lowering of the high pressure trip setpoint and raising of the setpoint for the Powec Operated Relief Valve (PORV), from 2255 psig to 2450 psig, has the effect of reducing the challenge rate to the PORV while maintaining ASME Code Safety Valve capability.
A B&W analysis completed in September of 1985 concluded that the high reactor coolant system pressure trip setpoint could be raised to 2355 psig with negligible impact on the frequency of opening of the PORV during anticipated overpressurization transients (8). The high pressure trip setpoint was subsequently raised to 2355 psig. The potential safety benefit of this action is a reduction in the frequency of reactor trips.
The low pressure (1800 psig) and variable low pressure (11.75 Tour-5103) trip setpoint were initially established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction (3,4).
The B&W generic ECCS analysis, however, assumed a low pressure trip of 1900 psig and, to establish conformity with this analysis, the low pressure trip setpoint has been raised to the-more conservative 1900 psig. Figure 2.3-1 shows the high pressure, low pressure, and variable low pressure trip setpoints.
d.
Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (619 F) shown in Figure 2.3-1 has been established to prevent excessive core coolant temperature in the operating range.
The calibrated range of the temperature channels of the RPS is 520' to
.620*F.
The trip setpoint of the channel is 619'F.
Under the worst case environment, power supply perturbations, and drift, the accuracy of the trip string is l'F.
This accuracy was arrived at by summing the worst case accuracies of each module. This is a conservative method of error analysis since the normal procedure is to use the root mean square method.
Therefcre, it is assured that a trip will occur at a value no higher than 620*F even under worst case conditions. The safety analysis used a high temperature trip setpoint of 620*F.
The calibrated range of the channel is that portion of the span of indication which has been qualified with regard to drift, linearity, repeatability, etc.
This does not imply that the equipment is restricted to operation within the calibrated range. Additional testing has demonstrated that in fact, the temperature channel is fully operational approximately 10% above the calibrated range.
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Sinc 2 it han b :n s2t:blieh;d th:t th2 chtnn:1 will trip et e v21u2 of RC outlet temperature no higher than 620*F even in the worst case, and since the channel is fully operational approximately 10% above the calibrated range and exhibits no hysteresis or foldover characteristics, it is concluded that the instrument design is acceptable.
e.
Reactor building pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.
f.
Shutdown bypass In order to provide for control rod drive tests, zero power physics testings, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in Table 2.3-1.
Two conditions are imposed when the bypass is used:
1.
By administrative control the nuclear overpower trip setpoint must be reduced to value 6. 5.0 percent of rated power during reactor shutdown.
2.
A high reactor coolant system pressure trip setpoint of 1720 psig is automatically imposed.
The purpose of the 1720 psig high pressure trip set point is to prevent normal operation with part of the reactor protection system bypassed.
This high pressure trip set point is lower than the normal low pressure trip set point so that the reactor must be tripped before the bypass is initiated. The overpower trip setpoint of d[ 5.0 percent prevents any significant reactor power from being produced when performing the physics tests.
Sufficient natural circulation (5) would be available to remove 5.0 percent of rated power if none of the reactor coolant pumps were operating.
References (1) FSAR, Section 14.1.2.3 (2) FSAR, Section 14.1.2.2 (3) FSAR, Section 14.1.2.7 (4) FSAR, Section 14.1.2.8 (5) FSAR, Section 14.1.2.6 (6) Technical Specification Change Request No. 31, January 16, 1976, and Technical Specification Change Request No. 84, June 23, 1978.
(7) "ECCS Analysis of B&W's 177-FA Lowered Loop NSS," BAW-10103, Rev. 2, Babcock and Wilcox, April 1976.
(8) " Justification for Raising Setpoint for Reactor Trip on High Pressure",
BAW-1890, Rev. O. Babcock and Wilcox, September 1985.
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_________________________.________________J
REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS 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 (5) on pump monitors, Max. % of rated power 4.
High reactor coolant sys-2355 2355 2355 1720(4) tem pressure, psig max.
h 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-5103)(1)
Bypassed coolant system pres-sure, psig min.
7.
Reactor coolant temp.
619 619 619 619 F., Max.
8.
High Reactor Building 4
4 4
4 pressure, psig max.
(1) Tout is in degrees Fahrenheit (F).
(2) Reactor coolant system flow, %.
(3) Administrative 1y 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
2500 2300 P = 2355 os ia 3
ACCEPTABLE f
OPERATION T = 619'F 3
/
$ 2100 s/
E s o-4 2
+
b1900 P = 1900 ps ig Sj E
U UNACCEPTABLI:
j OPERATION 1700 1500 540 560 580 600 620 640 Reactor Outlet Temperature, 'F TMI-1 PROTECTION SYSTEM MAXIMUM ALLOWABLE SET POINTS Figure 2.3-1
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I raictor coolent temparature instrument channels, four rarctor coolant flow instrument channels, four reactor coolant pressure instrument channels, four pressure-temperature instrument channels, four flux-imbalance flow instrument channels, four power-number of pumps instrument channels, and four high reactor building pressure instrument channels. The reactor trip, on loss of feedwater may be bypassed below 7% reactor power.
The bypass is automatically removed when reactor power is raised above 7%. The reactor trip, on turbine trip, may be bypassed below 45% reactor power (2). The safety features actuation system must have two analog channels functioning correctly prior to startup. The anticipatory reactor trips on loss of feedwater pumps and turbine trip have been added to reduce the number of challenges to the safety valves and power operated relief valve but have not been credited in the safety analyses.
Operation at rated power is permitted as long as the systems have at least the redundancy requirements of Column "B" (Table 3.5-1).
This is in agreement
^
with redundancy and single failure criteria of IEEE 279 as described in FSAR Section 7.
There are four reactor protection channels. Normal trip logic is two out of four. Required trip logic for the power range instrumentation channels is two out of three. Minimum trip logic on other instrumentation channels is one out of two.
The four reactor protection channels were provided with key operated bypass switches to allow on-line testing or maintenance on only one channel at a time during power operation. Each channel is provided alars and lights to indicate when that channel is bypassed. There will be one reactor protection system bypass switch key permitted in the control room.
Each reactor protection channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypass switch is being used.
Power is normally supplied to the control rod drive mechanisms from two separate parallel 460 voit sources. Redundant trip devices are employed in each of these sources.
If any one of these trip devices fails in the untripped state on-line repairs to the failed device, when practical, will be made, and the remaining trip devices and in many cases make on-line repairs.
Eight hours is ample time to test the remaining trip devices and in many cases make on-line repairs.
REFERENCE (1)
FSAR, Section 7.1 (2)
" Basis for Raising Arming Threshold for Anticipatory Reactor Trip on Turbine Trip", BAW-1893, Rev. O, Babcock and Wilcox, October 1985.
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e T2ble 3.5-1 Continued INSTRUMENTS OPERATING CONDITIONS (A)
(B)
(C)
Minimum Operable Minimum Degree Operator Action if Conditions of Functional Unit Channels of Redundancy Column A and B Cannot Be Met A.
Reactor Protection system (Cont'd.)
9.
Power / number of pumps 2
1 Maintain hot shutdown instrument channels 10.
High reactor building 2
1 Maintain hot shutdown pressure channels (a) For channel testing, calibration, or maintenance, the minimum number of operable channels may be two and a degree of redundancy of one for a maximum of four hours.
(b) When 2 of 4 power range instrument channels.are greater than 10 percent full power, hot shutdown is not required.
(c) When 1 of 2 intermediate range instrument channels is greater than 10-'* amps, or 2 of 4 power range instrument channels are greater than 10 percent full power, hot shutdown is not required.
B.
Other Reactor Trips 1.
Loss of Feedwater 2(a)(c) 1(a)(c)
Maintain less than 7% indicated reactor power.
2.
Turbine Trip 2(b)(c) 1(b)(c)
Maintain less than 45% indicated reactor power.
(a) Bypass of the feedwater pump trip signal may be placed in effect when indicated reactor power is less than 7%.
The bypass will be removed when reactor power is raised above 7%.
(b) The main turbine trip bypass may be placed in effect when indicated reactor power is less than 45%. The bypass will be removed when reactor power is raised above 45%.
(c) Trip may be defeated during low power physics tests.
Table 4.1-1 (Continued)
CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS 38.
Steam Generator Water Level W
NA R
39.
Turbine Overspeed Trip NA R*
NA 40.
BWST/NaOH Differential NA NA R
Pressure Indicator 41.
Sodium Hydroxide Tank Level NA NA R
Indicator 42.
Diesel Generator Protective NA NA R
Relaying 43.
4 KV ES Bus Undervoltage Relays (Diesel Start) j l
a.
Degraded Grid NA M(1)
R (1) Relay operation will be checked j
by local test pushbuttons b.
Loss of Voltage NA M(1)
R (1) Relay operation will be checked by local test pushbuttons 44.
Reactor Coolant Pressure S(l)
M R
(1) When reactor coolant system is DH Valve Interlock Bistable pressurized above 300 psig or Taves is greater than 200*F.
45.
Loss of Feedwater Reactor Trip S(l)
M(1)
R (1) When reactor power exceeds 7% power l
46.
Turbine Trip / Reactor Trip S(l)
M(1)
R (1) When reactor power exceeds 45%
power.
47.
a.
Pressurizer Code Safety Valve S(l)
R (1) When T.v. is greater than 525*F and PORV Tailpipe Flow Monitors b.
PORV - Acoustic / Flow NA M(1)
R (1) When T.v.
is greater than 525*F 48.
PORV Setpoints NA M(1)
R (1) Per Specification 3.1.12 excluding valve operation.
- Test to be performed prior to exceeding 20% power during Cycle 5 startup only.