ML13331B146

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Proposed Tech Specs Re Reactor Coolant Pump Bus Undervoltage Trip
ML13331B146
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Site: San Onofre Southern California Edison icon.png
Issue date: 03/11/1989
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SOUTHERN CALIFORNIA EDISON CO.
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ML13331B145 List:
References
NUDOCS 8903170018
Download: ML13331B146 (9)
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Attachment 2 Proposed Technical Specifications 8 90D17001890 PDR ADOCK 05000:206 P

PDIC

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 SAN ONOFRE -

UNIT 1 2-1

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.

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 1.30.(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 the event that steam/feedflow mismatch trip cannot be credited due to single failure considerations, the pressurizer high level trip is provided. In order to meet acceptance criteria for the Loss of Main Feedwater and Feedline Break transients, the pressurizer high level trip must be set at 20.8 ft. (50%) or less.

2. Variable Low Pressure Loss of Flow and Nuclear Overpower Trips These settings are established to accommodate the most severe transients upon which the design i's based, e.g.,

loss of coolant flow, rod withdrawal at power, control rod SAN ONOFRE -

UNIT 1 2-2

ejection, 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)(7)

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 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. Reactor Coolant Pump Breaker Open The Reactor Coolant Pump (RCP) Breaker Open reactor trip provides a redundant trip to the low flow trip. The overcurrent trip of the RCP breakers protects the core following a locked rotor and the undercurrent trip of the RCP breakers protects the core following a sheared shaft.

The trip settings are selected to meet the analysis assumptions that rods begin to drop 6.1 seconds after the initiating event. The Reactor Protection System Permissives change the trip on RCP breaker open to 2/3 loops instead of 1/3 loops at power levels below 50%.

In loss of forced coolant flow events caused by loss of RCP bus, the undervoltage trip provides redundancy to the low flow trip. This is consistent with assumptions in the accident analysis in the UFSAR Section 15.7.1.

SAN ONOFRE -

UNIT 1 2-3

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) NIS Safety Review Report, April 1988 (8) Reload Safety Evaluation, Cycle 10, Revision 1, March 1989, by Westinghouse, editor J. Skaritka SAN ONOFRE -

UNIT 1 2-3a

TABLE 2.1 MAXIMUM SAFETY SYSTEM SETTINGS Three Reactor Coolant Pumps Operating

  • 1. Pressurizer

< 20.8 ft. above bottom of pressurizer High Level when steam/feedflow mismatch trip is not credited, or

< 27.3 ft. above bottom of pressurizer when steam/feedflow mismatch trip is credited

2. Pressurizer

< 2220 psig Pressure: High

3. Nuclear Overpower
a. High Setting**

< 109% of indicated full power

b. Low Setting

< 25% of indicated full power

      • 4. Variable Low Pressure

> 26.15 (0.894 WT+T avg.) -

14341

      • 5. Coolant Flow

> 85% of indicated full loop flow

a. Overcurrent

< 2900 amps at 4160 volts

b. Undercurrent

> 110 amps at 4160 volts

c. Undervoltage

> 60% of rated bus voltage Credit can be taken for the steam/feedflow mismatch trip when this system is modified such that a single failure will not prevent the system from performing its safety function.

The nuclear overpower trip is based upon a symmetrical power distribution.

If an asymmetric power distribution greater than 5% should occur, the nuclear overpower trip on all channels shall be reduced one percent for each percent above 5%.

May be bypassed at power levels below 10% of full power.

SAN ONOFRE - UNIT 1 2-4 03630

TABLE 3.5.1-1 REACTOR TRIP SYSTEM INSTRUMENTATION 0 Z 0

MINIMUM

%j TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTION UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 2

1 2

1, 2 2

2 3*,4*,5*

7

2. Power Range, Neutron Flux, 4

2 3

I, 2 2#

Overpower Trip

3. Power Range, Neutron Flux, 4

1**

4 1,2 28#

Dropped Rod Rod Stop

4. Intermediate Range, Neutron 2

2 1###, 2 3

Flux W

5. Source Range, Neutron Flux A. Startup 2

1**

2 2##

4 B. Shutdown 2

I**

2 3*, 4*, 5*

7 C. Shutdown 2

0 1

3, 4, and 5 5

6. NIS Coincidentor Logic 2

I 2

1, 2 29 3*, 4*, 5*

7

7. Pressurizer Variable 3

2 2

iM0 6#

Low Pressure

8. Pressurizer Fixed High 3

2 2

1, 2 6#

Pressure

9. Pressurizer High Level 3

2 2

I 6#

Z TABLE 3.5.1-1 (Continued) 0 z

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTION UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION z

10. Reactor Coolant Flow A. Single Loop I/loop I/loop in any I/loop in each I

6#

(Above 50% of Full Power) operating loop operating loop B. Two Loops I/loop I/loop in two I/loop in each li***

6#

(Below 50% of Full Power) operating loops operating loop II. Steam/Feedwater Flow Mismatch 3

2 2

1,2 6#

I

12. Turbine Trip-Low Fluid a

Oil Pressure 3

2 2

19#f*

6#

13.

Reactor Coolant Pump Breaker Position A. Single Loop I/loop I/loop in any I/loop in each I

6#

(Above 50% of Full Power) operating loop operating loop B. Two Loops I/loop I/loop in two I/loop in each 1####

6#

(Below 50% of Full Power) operating loops operating loop 03670

TABLE 4.1.1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING DEVICE CHANNEL CHANNEL CHANNEL OPERATIONAL ACTUATION FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST C

Z I. Manual Reactor Trip N.A.

N.A.

N.A.

R N.A.

0 ma

2.

Power Range, Neutron Flux S

D (2,3)

M N.A.

N.A.

R (3,4)

3.

Power Range, Neutron Flux, N.A.

N.A.

M N.A.

N.A.

Dropped Rod Rod Stop

4.

Intermediate Range, S

R (3,4)

S/U (I),

N.A.

N.A.

Neutron Flux M

5.

Source Range, Neutron Flux S

R (3)

S/U (1),

N.A.

N.A.

M

6.

NIS Coincidentor Logic N.A.

N.A.

N.A.

N.A.

M (5)

7.

Pressurizer Variable Low S

R M

N.A.

N.A.

Pressure

8.

Pressurizer Pressure S

R M

N.A.

N.A.

9.

Pressurizer Level S

R M

N.A.

N.A.

10.

Reactor Coolant Flow S

R Q

N.A.

N.A.

II. Steam/Feedwater Flow S

R M

N.A.

N.A.

Mismatch

12.

Turbine Trip-Low Fluid N.A.

N.A.

N.A.

S/U (1,6)

N.A.

Oil Pressure

13.

Reactor Coolant Pump Breaker S

R R

N.A.

N.A.

Position*

  • Appllcable to Item 6 in Table 2.1
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