ML13316B304

Amend 121 to License DPR-13,revising Reactor Trips for Pressurizer High Level & Steam/Feedwater Flow Mismatch

ML13316B304
Person / Time
Site:	San Onofre
Issue date:	04/04/1989
From:	Knighton G Office of Nuclear Reactor Regulation
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ML13316B303	List: ML13316B302 ML13316B304 ML13316B305 ML13316B306
References
DPR-13-A-121 NUDOCS 8904200078
Download: ML13316B304 (10)

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-UNITED STATES o1 0NUCLEAR REGULATORY COM WASHINGTON, D..C. 20555 SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY DOCKET NO. 50-206 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. 1 AMENCMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 121 License No. DPR-13

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendmeit by Southern California Edison.

Company and San Diego Gas and Electric Company (the licensee) dated November 11, 198E, as supplemented February 14, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by.this.amendrient can.be conducted without endancering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will rct be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

~oo6~ a o~b _

2. Accordingly, the license is amended by changes to the Technical Speci fications as indicated in the attachment to this license amendment, and paragraph 3.B. of Provisional Operating License No. DPR-13 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 121, are hereby incorporated in the license. Southern California Edison Company shall operate the facility in accordance with the Technical Specifications, except where otherwise stated in specific license conditions.

3. This license amendment is effective as of the date of its issuance and must be fully implemented no later than 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION George W. Knighton, Director Project Directorate V Division of Reactor Projects -

III, IV, V and Special Projects

Attachment:

Chariges to the Technical Specifications Date of Issuance:- April 4, 1989

ATTACHMENT TO LICENSE AMENDMENT.NO. 121 PROVISIONAL OPERATING LICENSE NO. DPR-13 DOCKET NO. 50-206 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 2

2 3

3 4

4 5

5 29 29 30 30 30a 30a 30b

- 2 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 ODerating 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.

Amendment No.

7,777.121

-3 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 1.30.11)

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 Trips 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, control rod 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 roducts 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 Amendment No. 43. 177,121

-4 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) NIS Safetey Review Report, April, 1988 (8) SCE to NRC letter November 20, 1987, Engineered Safety Features Single Failure Analysis Amendment'No. 07,;;7121

TABLE 2.1 MAXIMUM SAFETY SYSTEM SETTINGS Three Reactor Coolant Pumps Operatina

1. Pressurizer 50% Pressurizer Narrow Range Level High Level

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 2 26.15 (0.894 T+T avg.) - 14341

• • 5.

Coolant Flow 2 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 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.

May be bypassed at power levels below 50% of full power.

High and Low feedwater flow relative to steam flow Amendment No. 717,121

TABLE 3.5.1-1 REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTION UNIT OF CHANNELS TO TRIP OPERABLE MO0ES 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 2V

3.

Power Range, Meutron Flux, Dropped 4

1**

4 1, 2 28#

Rod Rod Stop

4.

Intermediate Range, Neutron 2

I 2

19ff, 2 3

Flux

5.

Source Range, Neutron Flux A. Startup 2

1**

2 20 4

B. Shutdown 2

I**

2 3', 4*9 5*

7 C. Shutdown 2

0 I

3, 4, and 5 5

6.

NIS Coincident Logic 2

I 2

1, 2 611

7.

PressurIzer Variable 3

2 2

I###

60 Low Pressure

8. Pressurizer Fixed High 3

2 2

1, 2 66 Pressure

9.

Pressurizer High Level 3

2 2

I 66

3

10.

Reactor Coolant Flow I/loop I/loop in any I/loop In each I

60 A. Single Loo operating loop operating loop (Above 50% of Fl oe) 0 B.

Two Loops I/loop I/loop In two I/loop In each 1###

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

2 2

1####

61

12. Turbine Trip-Low Fluid Oi Pressure 3

2 2

l000 61

TABLE 3.5.1-1 (Continued)

TABLE NOTATION Wi th 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 Defeat) Setpoint.

Above the P-7 (At Power Reactor Trip Defeat) Setpoint.

1. 1. 1. 1. 1. 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 I hours.

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be returned to the untripped condition for up tdo 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.

However, one 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 per Specification 4.1, provided the other channel is OPERABLE.

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 I hour 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.

Amendment No. tAA.8.83.7,121

• gp 30a ACTION 6 -

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required OPERATIONAL TEST provided 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 />.

ACTION 7 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the reactor trip breakers within the next hour.

ACTION 28 - With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirements, within one hour reduce THERMAL POWER such that Tave is less than or equal to 551.5*F, and place the rod control system in the manual mode.

ACTION 29 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirements, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be removed from service 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 per Specification 4.1, provided the other channel is OPERABLE.

Amendment No. $3,777,121