ML13331B098
| ML13331B098 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/11/1988 |
| From: | SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML13331B095 | List: |
| References | |
| NUDOCS 8811160256 | |
| Download: ML13331B098 (7) | |
Text
Attachment 2 PROPOSED TECHNICAL SPECIFICATIONS 2.1 REACTOR CORE -
Limiting Combination of Power, Pressure, and Temperature APPLICABILITY: Applies -to reactor power, system pressure, coolant temperature, and flow during operation of the plant.
OBJECTIVE:
To maintain the integrity of the reactor coolant system and to prevent the release of excessive amounts of fission product activity to the coolant.
SPECIFICATION: Safety Limits (1) The reactor coolant system pressure shall not exceed 2735 psig with fuel assemblies in the reactor.
(2) The combination of reactor power and coolant temperature shall not exceed the locus of points established for the RCS pressure in Figure 2.1.1. If the actual power and temperature is above the locus of points for the appropriate RCS pressure, the safety limit is exceeded.
Maximum Safety System Settings The maximum safety system trip settings shall be as stated in Table 2.1 BASIS:
Safety Limits
- 1. Reactor Coolant System Pressure The Reactor Coolant System serves as a barrier which prevents release of radionuclides contained in the reactor coolant to the containment atmosphere. In addition, the failure of components of the Reactor Coolant System could result in damage to the fuel and pressurization of the containment. A safety limit of 2735 psig (110% of design pressure) has been established which represents the maximum transient pressure allowable in the Reactor Coolant System under the ASME Code,Section VIII.
- 2. Plant Operating Transients In order to prevent any significant amount of fission products from being released from the fuel to the reactor coolant, it is necessary to prevent clad overheating both during normal operation and while undergoing system transients. Clad overheating and potential failure could occur if the heat transfer mechanism at the clad surface departs from nucleate boiling. System parameters which affect this departure from nucleate boiling (DNB) have been correlated with experimental data to provide a means of determining the probability of DNB occurrence. The ratio of the heat flux at which DNB is expected to occur for a given set of conditions to the actual heat flux experienced at a point is the DNB ratio and reflects the probability that DNB will actually occur.
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It has been determined that under the most unfavorable conditions of power distribution expected during core lifetime and if a DNB ratio of 1.44 should exist, not more than 7 out of the total of 28,260 fuel rods would be expected to experience DNB. These conditions correspond to a reactor power of 125% of rated power. Thus, with the expected power distribution and peaking factors, no significant release of fission products to the reactor coolant system should occur at DNB ratios greater than I.30.(1)
The DNB ratio, although fundamental, is not an observable variable. For this reason, limits have been placed on reactor coolant temperature, flow, pressure, and power level, these being the observable process variables related to determination of the DNB ratio. The curves presented in Figure 2.1.1 represent loci of conditions at which a minimum DNB ratio of 1.30 or greater would occur. (1)(2)(3)
Maximum Safety System Settings
- 1. Pressurizer High Level and High Pressure In the event of loss of load, the temperature and pressure of the Reactor Coolant System would increase since there would be a large and rapid reduction in the heat extracted from the Reactor Coolant System through the steam generators. The maximum settings of the pressurizer high level trip and the pressurizer high pressure trip are established to maintain the DNB ratio above 1.30 and to prevent the loss of the cushioning effect of the steam volume in the pressurizer (resulting in a solid hydraulic system) during a loss-of-load transient.( 3)(4) In order to meet acceptance criteria for certain secondary side transients, the pressurizer high level trip must be set at 50% narrow range level or less.( 8)
- 2. Variable Low Pressure, Loss of Flow, and Nuclear Overpower These settings are established to accommodate the most severe transients upon which the design is based, e.g., loss of coolant flow, rod withdrawal at power, inadvertent boron dilution and large load increase without exceeding the safety limits.
The settings have been derived in consideration of instrument errors and response times of all necessary equipment. Thus, these settings should prevent the release of any significant quantities of fission products to the coolant as a result of transients.(3)(4)(5)
In order to prevent significant fuel damage in the event of increased peaking factors due to an asymmetric power distribution in the core, the nuclear overpower trip 2-2
setting on all channels is reduced by one percent for each percent that the asymmetry in power distribution exceeds 5%. This provision should maintain the DNB ratio above a value of 1.30 throughout design transients mentioned above.
The response of the plant to a reduction in coolant flow while the reactor is at substantial power is a corresponding increase in reactor coolant temperature. If the increase in temperature is large enough, DNB could occur, following loss of flow.
The low flow signal is set high enough to actuate a trip in time to prevent excessively high temperatures and low enough to reflect that a loss of flow conditions exists.
Since coolant loop flow is either full on or full off, any loss of flow would mean a reduction of the initial flow (100%) to zero.( 3)(6)
- 3. Steam/Feedwater Flow Mismatch A significant mismatch of steam flow and feedwater flow to the steam generators occurs at greater than 50% power in the event of LONF and FLB. In the event of these transients, the 2 out of 3 mismatch trip logic will result in reactor trip on the order of 1 second after the initiating event. The safety analysis conservatively assumed that reactor trips would occur at 5 seconds and 10 seconds for LONF and FLB, respectively. The high and low settings assure that regardless of the type of mismatch occurring for individual loops, a protective reactor trip is provided, which satisfy the single failure criterion for the postulated events.( 8)
References:
(1) Amendment No. 10 to the Final Engineering Report and Safety Analysis, Section 4, Question 3 (2) Final Engineering Report and Safety Analysis, Paragraph 3.3 (3) Final Engineering Report and Safety Analysis, Paragraph 6.2 (4) Final Engineering Report and Safety Analysis, Paragraph 10.6 (5) Final Engineering Report and Safety Analysis, Paragraph 9.2 (6) Final Engineering Report and Safety Analysis, Paragraph 10.2 (7) (Reserved for Proposed Change No. 180)
(8) SCE to NRC letter November 20, 1987, Engineered Safety Features Single Failure Analysis 2-3
TABLE 2.1 MAXIMUM SAFETY SYSTEM SETTINGS Three Reactor Coolant Pumps Operating
- 1.
Pressurizer
< 50% Pressurizer Narrow Range Level High Level
- 2.
Pressurizer
< 2220 psig Pressure:
High
- 3. Nuclear Overpower
< 109% of indicated full power
- 4.
Variable Low Pressure
> 26.15 (0.894 T+T avg.) -
14341
- 5.
Coolant Flow
> 85% of indicated full loop flow
- 6.
Steam/Feedwater Flow Mismatch
- a. Low+ Setting:
Steam Flow -
Feedwater Flow 0.25 Feedwater Flow @ 100% Power
- b. High+ Setting:
Feedwater Flow - Steam Flow 0.25 Feedwater Flow @ 100% Power The nuclear overpower trip high setting is based upon a symmetrical power distribution. If an asymmetric power distribution greater than 10% should occur, the nuclear overpower trip on all channels shall be reduced one percent for each percent above 10%.
May be bypassed at power levels below 10% of full power.
May be bypassed at power levels below 50% of full power.
High and Low feedwater flow relative to steam flow 2-4
TABLE 3.5.1-1 REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTION UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION I. Manual Reactor Trip 2
I 2
1, 2 I
2 I
2 3*, 4*, 5*
7
- 2. Power Range, Neutron Flux 4
2 3
1, 2 2#
- 3. (Reserved for Proposed Change 180)
- 4. Intermediate Range, Neutron 2
I 2
1###,
2 3
Flux
- 5. Source Range, Neutron Flux A. Startup 2
1**
2 2##
4 B. Shutdown 2
1**
2 3*, 4* 5*
7 C. Shutdown 2
0 I
3, 4, and 5 5
- 6. (Reserved for Proposed Change 180)
- 7. Pressurizer Variable 3
2 2
I****
6#
Low Pressure
- 8. Pressurizer Fixed High 3
2 2
1, 2 6#
Pressure
- 9. Pressurizer High Level 3
2 2
1 6#
- 10. Reactor Coolant Flow I/loop I/loop in any I/loop in each I
6#
A. Single Loop operating loop operating loop (Above 50% of Full Power)
B.
Two Loops I/loop I/loop in two I/loop in each 1###
6#
(Below 50% of Full Power) operating loops operating loop II.
Steam/Feedwater Flow Mismatch 3
2 2
1####
6#
- 12.
Turbine Trip-Low Fluid Oil Pressure 3
2 2
1####
6#
TABLE 3.5.1-1 (Continued)
TABLE NOTATION With the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal.
A "TRIP" will stop all rod withdrawal.
The provisions of Specification 3.0.4 are not applicable.
Below the Source Range High Voltage Cutoff Setpoint.
Below the P-7 (At Power Reactor Trip's Active) Setpoint.
Above the P-7 (At Power Reactor Trip's Active) Setpoint.
Above the P-8 Setpoint ACTION STATEMENTS ACTION 1 -
With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in 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 />.
ACTION 2 -
With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are met:
- a. The inoperable channel is placed in the tripped condition within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing of other channels per Specification 4.1.
ACTION 3 -
With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
- a. Below the Source Range High Voltage Cutoff Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the Source Range High Voltage Cutoff Setpoint.
- b. Above the Source Range High Voltage Cutoff Setpoint but below 10 percent of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10 percent of RATED THERMAL POWER.
ACTION 4 -
With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement suspend all operations involving positive reactivity changes.
ACTION 5 -
With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.5.2 as applicable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
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