Text

Proposed Tech Specs,Reflecting New Values of Limiting Core Parameters Resulting from Reanalysis of LOCAs Due to Core Reload
	ML13316B997
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Site:	San Onofre
Issue date:	03/04/1989
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ATTACHMENT 1 EXISTING TECHNICAL SPECIFICATIONS 8903090211 090304 P

PDC

2.1 REACTOR 00RE Limiting Combination of Powet Pressure, and Temperature APPLICABILIlY:

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 60 shall not exceed the locus of points established for the 6/8/81 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.

117 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 (1101 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 SAN ONOFRE - UNIT 1 2-1 Revised: 12/21/88

for a given set of conditions to the actual heat flux experienced at a point is the 06 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 ON 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 DNS ratios greater than 1.30.(1)

The DN8 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 49 conditions at which a minimum 0NB ratio of 1.30 or 7/19/79 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 97 pressurizer high level trip is provided. In order to meet 4/7/86 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 Trmi 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 SAN ONOFRE - UNIT 1 2-2 Revised:

12/21/88

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 1

as a result of transionts.(3)(4)(5)(7) 112/13/88 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 117 above a value of 1.30 throughout design transients 12/13/88 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)

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 113/88 SAN ONOFRE -

UNIT 1 2-3 Revised: 12/21/88

TABLL.1 MAXIMUM SAFETY SYSTEM SETTINGS 117

12/13/88 Three Reactor Coolant Pumos Onaeratina

1. Pressurizer 1 20.8 ft. above bottom of pressurizer High Level when steam/feedflow mismatch trip is not.

credited, or s 27.3 ft. above bottom of pressurizer when steam/feedflow mismatch trip J credited

2. Pressurizer j 2220 psig Pressure: High

3. Nuclear Overpower 117

a. High Setting**

1 109% of indicated full power 12/13/88

b. Low Setting

s. 25% of indicated full power

4. Variable Low Pressure 2 26.15 (0.894 AT+T avg.) -

14341

- - 5. Coolant Flow 2 85% of indicated full loop flow 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 117 each percent above 51.

12/13/88

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

SAN ONOFRE - UNIT 1 2-4 Revised: 12/21/88

~~~~Figure 2.1.1 Safety Limits i

-. =7:

Temperature, Power, Pressure I-%

4j-7 D

PF..n.......

.~

RCS Pressure a

IAI0 _

_OPSI 65 i=

TT eTetfNoia-Pwr(.....................................

S.

a........-

.z.:rt

3.3.3 MINIMUM WATER VOLUME AND BORON CONCENTRATION IN THE REFUELING WATER STORAGE TANK APPLICABILITY: Applies to the inventory of borated refueling water.

OBJECTIVE:

To ensure immediate availability of safety injection and containment spray water of required quality.

SPECIFICATION: When the Safety Injection System or the Containment Spray System is required to be operable, the refueling water tank shall be filled to at least elevation 50 feet with water 34 having a boron concentration of not less than 3750 ppm and not 4/1/77 greater than 4300 ppm.

BASIS:

The refueling water storage tank serves two purposes; namely:

(1) As a reservoir of borated water for accident mitigation

purposes, (2) As a reservoir of borated water for flooding the refueling cavity during refueling.

Approximately 220,000 gallons of borated water is required to provide adequate post-accident core cooling and containment spray to maintain cali lated post-accident doses below the limits of 10 CFR 100'.

The refueling water storage tank filled to elevation 50 feet represents in excess of 240,000 gallons.

34 4/1/77 A boron concentration of 3750 ppm is requireJ 25o meet the requirements of postulated steam line break.

A maximum boron concentration of 4300 ppm ensures that the post-accident contiinment sump water is maintained at a pH between 7.0 and 7.5.

The refueling tank capacity of 240,000 gallons is based on refueling volume requirements.

Sustained temperatures below 320F do not occur at San Onofre. At 320F, boric acid is soluble up to approximately 34 4650 ppm boron. Therefore, no special provisions for 4/1/77 temperature control to avoid either freezing or boron precipitation are necessary.

References:

(1) Enclosure 1 "Post-Accident Pressure Reanalysis, San Onofre Unit 1" to letter dated January 19, 1977 in Docket No. 50-206.

(2) "Steam Line Break Accident Reanalysis, San Onofre Nuclear Generating Station, Unit 1, October 1976" submitted by letter dated December 30, 1976 in Docket No. 50-206.

(3) Additional information, San Onofre, Unit 1 submitted by letter dated March 24, 1977 in Docket No. 50-206.

Typo Revised:

7/9/82 3-35 Revised:

5/14/77

3.5 INSTRUMENTATION AND CONTROL 3.5.1 REACTORTRIP SYSTE1 INSTRUMENTATION APPLICABILITY:

As shown in Table 3.5.1-1.

OBJECTIVE:

To delineate the conditions of the Plant instrumentation and 84 safety circuits necessary to ensure reactor safety.

11/2/84 SPECIFICATION:

As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.5.1-1 shall be OPERABLE.

ACTION:

As shown in Table 3.5.1-1.

BASIS:

During plant operations, the complete instrumentation systems will normally be in service.(1) Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits.(2) Safety is not compromised, however, by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design.(1)(3) This Standard outlines limiting conditions for operation necessary to preserve the effectiveness of the

.2/13/88 reactor control and protection system when any one or more of the channels is out of service.

References:

(1) Final Engineering Report and Safety Analysis, Section 6.

(2) Final Engineering Report and Safety Analysis, Section 6.2.

117 (3) NIS Safety Review Report, April 1988 12/13/88 SAN ONOFRE - UNIT 1 3-44 Revised:

12/21/88

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, 2I 2

I 2

3*, 4*, 5*

7

2.

Power Range, Neutron Flux, 4

2 3

1, 2 2#

Overpower Trip

3.

Power Range, Neutron Flux, 4

i**

4 1, 2 28#

Dropped Rod Rod Stop

4. Intermediate Range, Neutron 2

I 2

1##, 2 3

W Flux LR

5. Source Range, Neutron Flux A. Startup 2

1**

2 219 4

B. Shutdown 2

1**

2 3O, 40, 5*

7 C. Shutdown 2

0 I

3, 4, and5

6. NIS Coincidentor Logic 2

I 2

1, 229 3*, 4*, 5*

7

7. Pressurizer Variable 3

2 2

101#

60 Low Pressure

8. Pressurizer Fixed High 3

2 2

1, 2 6#

Pressure

9. Pressurizer High Level 3

2

.2 I

6 (CD N.

cm c

-C

_N N

4 0-11 12/13/8

-C

.C 3-6RVSD1/18

TABLE 3.5.1-1 (Continued)

TABLE NOTATION 83 With the reactor trip system breakers in the closed position, the 11/2/84 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.

117 12/13/88 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 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 -

With the number of OPERABLE channels one less than the Total Number a

of Channels, STARTUP and/or POWER OPERATION may proceed provided 11/2/84 the following conditions are met:

a. The inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

b. The Minimum Channels OPERABLE requirement is met; however, the 117 inoperable channel may be returned to the untripped condition 12/13/88 for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 83 Setpoint.

84 11/2/84

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for 117 surveillance testing per Specification 4.1, provided the other 12/13/88 channel is OPERABLE.

83 ACTION 4 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement suspend all operations involving 11/2/84 positive reactivity changes.

Typo Revision:

2/17/89 SAN ONOFRE -

UNIT 1 3-47 Revised: 12/21/88

ACTION 5 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, verify compliance with the SHUTON MARGIN requirements of Specification 3.5.2 as applicable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

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 83 the inoperable channel is placed in the tripped condition within 11/2/84 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 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 manual mode.

117 ACTION 29 - With the number of OPERABLE channels one less than the Minimum L2/13/88 Channels OPERABLE requirements, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.1, provided the other channel is OPERABLE.

SAN ONOFRE - UNIT 1 3-48 Revised:

12/21/88

3.5.2 CONTROL ROD INSERTION LIMITS APPLICABILIII:

MODES l and 2 111 10/21/88:

OBJECTIVE:

This specification defines the insertion limits for the control rods in order to ensure (1) an acceptable core power distribution during power operation, (2) a limit on potential reactivity insertions for a hypothetical control rod ejection, and (3) core subcriticality after a reactor trip.

SPECIFICATION:

A. Except during low power physics tests or surveillance testing pursuant to Specification 4.1.1.6, the Shutdown Groups and Control Group 1 shall be fully withdrawn, and 111 the position of Control Group 2 shall be at or above the 10/21/88 21-step uncertainty limit shown in Figure 3.5.2.1.

B. The energy weighted average of the positions of Control Group 2 shall be at least 90% (i.e. > Step 288) withdrawn after the first 20% burnup of a core cycle. The average shall be computed at least twice every month and shall 21 consist of all Control Group 2 positions during the core 5/13/75 cycle.

ACTION:

A. With the control groups inserted beyond the above.

insertion limits either:

1. Restore the control groups to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

2. Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the above figure, or 111

3. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

10/21/88 B. With a single dropped rod from a shutdown group or control group, the provisions of Action A are not applicable, and retrieval shall be performed without increasing THERMAL POWER beyond the THERMAL POWER level prior to dropping the rod. An evaluation of the effect of the dropped rod shall be made to establish permissible THERMAL POWER levels for continued operation. If retrieval is not successful within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months from the time the rod was dropped, appropriate action, as determined from the evaluation, shall be taken. In no case shall operation longer than 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months be permitted if the dropped rod is worth more than 0.4% a k/k.

BASIS:

During Startup and Power Operation, the shutdown groups and control group 1 are fully withdrawn and control of the reactor is maintained by control group 2. The control group insertion limits are set in consideration of maximum specific SAN ONOFRE -

UNIT 1 3-49 Revised:

11/10/88

power, shutdown capability, and the rod ejection accident.

The considerations associated with each of these quantities are as follows:

1. The initial design maximum value of specific power is 15 kW/ft. The values of FJH and FQ total associated with this specific power are 1.75 and 3.23, respectively.

A more restrictive limit on the design value of specific power, FH and F? is applied to operation in 60 accordance with the current safety analysis including 6/8/8L fuel densification and ECCS performance. The values of the specific power, FWH and FQ are 13.7 kW/ft, 1.57 and 2.89, respectively. At partial power, the F H maximum values (limits) increase according to 11/23/84 the following equation, F H (P) = 1.57 (1 + 0.2 (1-P)], where P is the 111 fraction of RATED THERMAL POWER. The control group 10/21/88 insertion limits in conjunction with Specification B prevent exceeding these values even assuming the most adverse Xe distribution.

6/8/81

2. The minimum shutdown capability required is 1.25% Ap at 21 80L, 1.9% Ap at EOL and defined linearly between these 5/13/75 values for intermediate cycle lifetimes.

The rod

/

insertion limits ensure that the available SHUTDOWN MARGIN is greater than the above values.

111/88

3. The worst case ejected rod accident (8) covering HFP-BOL, HZP-BOL, HFP-EOL shall satisfy the following accident safety criteria:

a) Average fuel pellet enthalpy at the hot spot below 54 225 cal/gm for nonirradiated fuel and 220 cal/gm for 5/29/80 irradiated fuel.

b) Fuel melting is limited to less than the innermost 10% of the fuel pellet at the hot spot.

Low power physics tests are conducted approximately one to four times during the core cycle at or below 10% RATED 1

THERMAL POWER. During such tests, rod configurations 110/21/8 different from those specified in Figure 3.5.2.1 may be employed.

It is understood that other rod configurations may be used during physics tests. Such configurations are permissible based on the low probability of occurrence of steam line break or rod ejection during such rod configurations.

SAN ONOFRE -

UNIT 1 3-50 Revised:

11/10/88

Operation of the reactor during cycle stretch out is conservative relative to the safety considerations of the control rod insertion limits, since the positioning of the rods during stretch out results in an increasing net available SHUTDOWN MARGIN.

Compliance with Specification B prevents unfavorable axial power distributions due to operation for long intervals at deep control rod insertions.

The presence of a dropped rod leads to abnormal power distribution in the core. The location of the rod and its worth in reactivity determines its effect on the temperatures of nearby fuel.

Under certain conditions, continued operation could result in fuel damage, and it is the intent of ACTION 8 to avoid such damage.

111

References:

(1) Final Engineering Report and Safety Analysis, revised July 28, 1970.

(2) Amendment No. 18 to Docket No. 50-206.

(3) Amendment No. 22 to Docket No. 50-206.

(4) Amendment No. 23 to Docket No.90-206.

(5) Description and Safety Analysis, Proposed Change No. 7, dated October 22, 1971.

0I(6)

Description and Safety Analysis Including Fuel Densification, San Onofre Nuclear Generating Station, Unit 1, Cycle 4, WCAP 8131, May, 1973.

(7) Description and Safety Analysis Including Fuel Densification, San Onofre Nuclear Generating Station, Unit 1, Cycle 5, January, 1975, Westinghouse Non-Proprietary Class 3.

(8) An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods, WCAP-7588, Revision 1-A, January, 1975.

SAN ONOFRE -

UNIT 1 3-51 Revised:

11/10/88

CONTROL GROUP INSERTION LIMITS FULLY WITHDRAWN 320

.300 261 25042 oz 200.240 O 100 OP 35 35 FULLY INSERTED 0

0 10 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.5.2.1 SAN ONOFRE - UNIT 1 3-52 Revised:

11/10/88

3.10 INCORE INSTRUMENTATION 112 S

APPLICABILTY:

MODE 1 above 90% RATED THERMAL POWER 0/28/88 OBJECTIVE:

To specify the type and frequency of incore measurements used to verify linear power density values.

SPECIFICATION:

a. A power distribution measurement shall be performed every 30 effective full power days (EFPD) and after attainment of equilibrium xenon upon return to power 1112 following a refueling shutdown.

LO/28/88

b. The incore instrumentation system shall be used to accomplish the Correlation Verification-of incore versus excore data for the axial offset monitoring system prior to exceeding 90% of RATED THERMAL POWER r

following each refueling and at least once per 180 112 effective full power days (EFPD) thereafter.

10/28/88 Subsequent to the Correlation Verification and for the duration of each cycle, incore instrumentation shall be used to perform a Correlation-Check of the axial offset monitoring system every 30 EFP0.

c. A core thermocouple map shall be taken every 30 EFPO 112 and after attainment of equilibrium xenon-upon return 110/28/88 to power following a refueling shutdown.

ACTION:

A. If the correlation check, power distribution measurement or core thermocouple map described above 112 cannot be made within the prescribed time, a maximum 10/28/88 of 15-EFPD will be allowed for equipment correction.

B. In the event that Specification a, b and c cannot be met during the 15 EFPD allowed for corrective action, within one hour action shall be taken such that THERMAL POWER is restricted to less than or equal to 90% of RATED THERMAL POWER until these specifications 112 can be met.

0/28/88 8ASIS:

The flux mapping system is used to measure the core power distribution and to correlate incore versus excore data for the axial offset system. Measurements made with the flux mapping system every 30 effective full power days and upon return to power following a refueling shutdown will monitor the core power distribution to confirm that the maximum linear power density remains below allowable values. The SAN ONOFRE -

UNIT 1 3-91 Revised:

10/28/88

axial offset system will monitor the axial core power distribution in a continuous manner. If the Correlation Verification or Correlation Check is not performed, the 90%

of full thermal power restriction assures safe operation of the reactor. In addition, core thermocouples provide an independent means of measuring the balance of power among the core quadrants.

The flux mapping system and the thermocouple system are not integral parts of the Reactor Protection System. These systems are, rather, surveillance systems which may be required in the event of an abnormal condition such as a power tilt or a control rod misalignment. Since such a condition cannot be predicted, it is prudent to have the surveillance systems in an operable state. The 90% of full power restriction, used when these measurements cannot be taken as scheduled, is applied to minimize the probability of exceeding allowed peaking factors.

Operation for a 180 effective full power day period prior to reperforming the correlation verification is acceptable 112 on the basis that the allowed incore axial offset limits LO/28/88 are reduced by the amount in percent of incore axial offset that the monthly correlation check differs from the correlation.

0 SAN ONOFRE - UNIT I 3-92 Revised: 10/28/88

3.11 CONTINUOUS PWER DISTRIBUTION MONITORING 112 0O/28/88 APPLICABILITY:

MODE 1 above 90% RATED THERMAL POWER OBJECTIVE:

To provide corrective action in the event that the axial offset monitoring system limits are approached.

SPECIFICATION:

The incore axial offset limits shall not exceed the functional relationship defined by:

2.89/P - 2.1225 For positive offsets:

IAO FCC 0.03021 112 2.89/P -

2.1181 10/28/88 For negative offsets:

IAO.----------------

FCC

-.03068 where IA0 -

incore axial offset P -

fraction of RATED THERMAL POWER FCC -

The larger of 3.0 or the value in percent of incore axial offset by which the current 112 correlation check differs from the 10/28/88 incore-excore correlation.

ACTION:

A. With IAO exceeding the limit defined by the specification, within I hour action shall be taken to reduce THERMAL POWER until IAO is within specified limits or such that THERMAL POWER is restricted to less than 90% of RATED THERMAL POWER.

B. With one or both excore axial offset channel(s) inoperable, as an alternate, one OPERABLE NIS channel 12/13/88 for each inoperable excore axial offset channel, shall be logged every two hours to determine IAO.

C. With no method for determining IAO available, within I hour action shall be taken such that THERMAL POWER is reduced to less than 90% of RATED THERMAL POWER until a method of determining axial offset is restored.

SAN ONOFRE - UNIT 1 3-93 Revised:

12/21/88

BASIS:

The percent full power axial offset limits are conservatively established considering the core design peaking factor, analytical determination of the relationship between core peaking factors and incore axial offset considering a wide range of maneuvers and core conditions, and actual measurements relating incore axial offset to the axial offset monitoring systems. The axial offset limit established from the incore versus excore data have been reduced by an amount equivalent to FCC to allow 112 for burnup and time dependent differences between the 10/28/88 periodic.correlation verification and the monthly correlation check. Correcting the allowed incore axial offset limits by an amount equal to FCC maintains plant operation within the original safety analysis assumptions.

Should a specific cycle analysis establish that the analytical determination of the relationship between core peaking factors and incore axial offset has changed in a manner warranting modification to the existing envelope of peaking factor (1,2), then a change to functional relationship of the specification shall be submitted to the Commission. The incore-excore data correlation is checked or verified'periodically as delineated in Specification 3.10, INCORE INSTRUMENTATION.

Reducing power in cases when limits are approached or exceeded, will assure that design limits which were set in consideration of accident conditions are not exceeded. In the event that no method exists for determining IAO, actions are specified to reduce THERMAL POWER to 90% of 117 RATED THERMAL POWER. However, if axial offset channel(s) 12/13/88 are inoperable, hand calculational methods of determining IAO from OPERABLE NIS channels can be employed until OPERABILITY of the axial offset channel(s) is restored.

References:

(1) Supporting Information on Periodic Axial Offset Monitoring. San Onofre Nuclear Generating Station, Unit 1, September, 1973 (2) Supporting Information on the Continuous Axial Offset Monitoring System, San Onofre Nuclear Generating Station, Unit 1, July, 1974 (3) Description and Safety Analysis, Including Fuel Densification, San Onofre Nuclear Generating Station, Unit 1 Cycle 5, January, 1975, Westinghouse Non-Proprietary Class 3.

SAN ONOFRE - UNIT 1 3-94 Revised:

12/21/88

4.1.1 OPERATIONAL SAFETY ITEMS gnljcabijIU:

Applies to surveillance requirements for items directly related to Safety Standards and Limiting Conditions for Operation.

Qbj~ac.tjn:

To specify the minimum frequency and type of surveillance to be applied to plant equipment and conditions.

Scifiscation:

A. Reactor Trip System instrumentation shall be checked, 117 tested, and calibrated as indicated in Table 4.1.1.

12/13/88 B. Equipment and sampling tests shall be as specified in Table 4.1.2.

C. The specific activity and boron concentration of the reactor coolant shall be determined to be within the 38 limits by performance of the sampling and analysis 12/20/77 program of Table 4.1.2., Item la.

0. The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.1.2.,

Item lb.

E. All control rods shall be determined to be above the rod insertion limits shown in Figure 3.5.2.1 by verifying that each analog detector indicates at least 21 steps above the rod insertion limits, to account for the instrument inaccuracies, at least once per shift during Startup conditions with Keff equal to or greater than one.

61 F. The position of each rod shall be determined to be within the group demand limit and each rod position indicator shall be determined to be OPERABLE by verifying that the rod position indication system (Analog Detection System) and the step counter indication system (Digital Detection System) agree within 35 steps at least once per shift during Startup and Power Operation except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the rod position indication system (Analog Detection System) and the step counter indication system (Digital Detection System) at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

G. During MODE 1 or 2 operation each rod not fully inserted 83 in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at 11/2/84 least once per 31 days.

H. Instrumentation shall be checked, tested, and calibrated 117 as indicated in Table 4.1.3.

2/13/88 SAN ONOFRE - UNIT 1 4-3 Revised: 12/21/88

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

1.

Manual Reactor Trip N.A.

N.A.

R M.A.

2.

Power Range, Neutron Flux S

D (2.3)

N 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 (1).

N.A.

Neutron Flux M

5. Source Range, Neutron Flux S

R (3)

S/U (1),

N.A.

M

6.

NIS Coincidentor Logic N.A.

N.A.

N (5)

7.

Pressurizer Variable Low S

R M

N.A.

Pressure

8.

Pressurizer Pressure S

R M

N.A.

9. Pressurizer Level S

R M

N.A.

10. Reactor Coolant Flow S

R Q

N.A.

11.

Steam/Feedwater Flow S

R M

N.A.

Mismatch

12.

Turbine Trip-Low Fluid N.A.

N.A.

S/U (1.6)

N.A.

Oil Pressure

(This page intentionally blank) 117 12/13/88 SAN ON0FRE -UNIT 1 4-5 Revised:

12/21/88

TABLE 4.1.1 (Continued)

TABLEATION (1) -

If not performed in previous 31 days.

(2) -

Heat balance only, above 151 of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent.

117 12/13/88 (3) -

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(4) -

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(5) -

Each train shall be tested at leastevery 62 days on a STAGGERED TEST BASIS.

(6) -

Setpoint verification is not applicable.

0 SAN ONOFRE -

UNIT 1 4-6 Revised:

12/21/88

TABLE 4.1.2 MINIMUM EUIPMENT CHECK AND SAMPLING FREOUENCY Check FreauencY la. Reactor Coolant 1. Gross Activity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

96 Samples Determination Required during Modes 3/5/87 1, 2, 3 and 4.

2. Isotopic Analysis 1 per 14 days. Required for DOSE EQUIVALENT only during Mode 1.

1-131 Concentration

3. Spectryscopic 1 per 6 months(2) for E('i Required only during Determination Mode 1.

74 12/6/84

4. Isotopic Analy-a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,( 3) sis for Iodine whenever the specific Including 1-131, activity exceeds 1-133, and I-135.

1.0 pCi/gram DOSE EQUIVALENT 1-131 or 100/ E (1) pCi/gram.

b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change 38 exceeding 15 percent of 12/20/77 the RATED THERMAL POWER within a one hour period.

5. Boron concentration Twice/Week (1) E is defined in Section 1.0.

117 (2) Sample to be taken after a minimum of 2 EFPO and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

(3) Until the specific activity of the reactor coolant system is restored within its limits.

SAN ONOFRE - UNIT 1 4-7 Revised:

12/21/88

TABLE 4.1.2 (continued)

Check Frequency 1.b Secondary

1. Gross Activity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Coolant Determination Required only during Samples Modes 1, 2, 3 and 4.

2. Isotopic Analy-a) 1 per 31 days, whenever sis for DOSE the gross activity EQUIVALENT 1-131 determination indicates Concentration iodine concentrations greater than 10% of the 74 allowable limit.

12/6/83 Required only during Modes 1, 2, 3 and 4.

b) 1 per 6 months, whenever the gross activity determination indicates iodine concentrations below 1O% of the allow able limit. Required only during Modes 1, 2, 3, and 4.

4-8 Revised:

12/30/83 Typo Revision:

1/23/84

TABLE 4.1-.2 (continued)

Check Frequency

2. Safety

a. Boron Concentration Monthly when the reactor is Injection critical and prior to return Water Samples of criticality when a period 12 of subcriticality extends 9/17/73 the test beyond 1 month

3. Control Rod

a. Verify that all rods At each refueling shutdown Drop move from full out to full in, in less than 101 2.44 seconds 4/26/88

4. (Deleted) 61 6/11/81

5. Pressurizer

a. Pressure Setpoint At each refueling shutdown Safety Valves

6. Main Steam

a. Pressure Setpoint At each refueling shutdown Safety Valves

7. Main Steam

a. Test for Operability At each refueling shutdown Power Operated Relief Valves

8. Trisodium

a. Check for system At each refueling shutdown Phosphate availability as Additive delineated in Technical Speci fication 4.2

9. Hydrazine

a. Hydrazine concentra-Once every six months when Tank Water tion the reactor is critical and Samples prior to return of critica lity when a period of sub-34 criticality extends the test 4/1/77 interval beyond six months

10. Transfer

a. Verify that the fuse Monthly, when the reactor is Switch No. 7 block for breaker critical and prior to 8-1181 to MCC 1 is returning reactor to criti removed cal when period of subcriti cality extended the test interval beyond one month 4-9 Revised:

5/26/88

TABLE 4.1.2 (continued)

Check Frequency

11.

MOV-LCV-1100 C

a. Verify that the fuse Same as Item 10 above Transfer Switch block for either breaker 8-1198 to MCC 1 or breaker 42-12A76 to MCC 2A is removed.

12. Emergency Siren

a. Verify that the fuse Same as Item 10 above Transfer Switch block for either breaker 34 8-1145 to MCC 1 or 4/1/77 breaker 8-1293A to MCC 2 is removed

13. Communication

a. Verify that the fuse Same as Item 10 above Power Panel block for either Transfer Switch breaker 8-1195 to MCC I or breaker 8-12938 to MCC 2 is removed 14a. Spent Fuel Pool Verify water level per

a. Once every seven days Water Level Technical Specification when spent fuel is being 3.8 stored in the pool.

b. Refueling Pool

b. Within two hours prior Water Level to start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during 9/25/78 movement of fuel assemblies or RCC's.

15. Reactor

a. Per Technical Specifi-

a. Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Coolant Loops/

cations 3.1.2.C and Residual Heat 3.1.2.0, in Mode 1 Removal Loops and Mode 2 and in Mode 3 with reactor trip 104 breakers closed, verify 6/9/8 that all required reactor coolant loops are in operation and circulating reactor coolant.

b. Per Technical Specifi cation 3.1.2.E, in Mode 3 with the reactor trip breakers open, verify SAN ONOFRE -

UNIT 1 4-10 Revised: 7/1/88

TABLE 4.1.2 (Continued)

Check Frequency

1. At least two required

1. Once per 7 days reactor coolant pumps are operable with correct breaker align ments and indicated power availability.

2. The steam generators

2. Once per 12 associated with the two hours required reactor coolant pumps are operable with secondary side water level

> 256 inches of narrow range on cold calibrated scale.

3. At least one reactor

3. Once per 12 coolant loop is in hours operation and circulating reactor coolant.

c. Per Technical Specification 80 3.1.2.F, in Mode 4 verify 10/4/84

1. At least two required

1. Once per 7 days (RC or RHR) pumps are operable with correct breaker alignments and indicated power availability.

2. The required steam

2. Once per 12 generators are operable hours with secondary side water level > 256 inches of narrow range on cold calibrated scale.

3. At least one reactor

3. Once per 12 coolant loop/RHR train hours is in operation and circulating reactor coolant.

d. Per Technical Specifications 3.1.2.G and 3.1.2.H, in Mode,5 verify, as applicable:

4-11 Revised:

10/04/84

TABLE 4.1.2 (Continued)

Check Frequency

1. -At least one RHR train

1. Once per 12 is in operation and hours circulating reactor coolant.

2. When required, one

2. Once per 7 additional RHR train is days operable with correct pump breaker alignments and indicated power availability.

3. When required, the

3. Once per 12 secondary side water level hours of at least two steam generators is > 256 inches of narrow range on cold calibrated scale.

e. Per Technical Specification

e. Once per 12 80 3.8.A.3, in Mode 6, with water hours 10/4/84 level in refueling pool greater than elevation 40 feet 3 inches, verify that at least one method of decay heat removal is in operation and circulating reactor coolant at a flow rate of at least 400 gpm.

f. Per Technical Specification 3.8.A.4, in Mode 6, with water level in refueling pool less than elevation 40 feet 3 inches, verify

1. At least one decay heat

1. Once per 12 removal method is in hours operation and circulating reactor coolant.

2. One additional decay heat

2. Once per 7 removal method is operable days with correct pump breaker alignments and indicated power availability.

4-12 Revised:

10/04/84

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 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 fraivte set of itions to the actual heat flux for a given set of ondis 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 1251 of rated power. Thus, with the expected power distribution and peaking factors, no significant release of fission products togthe reactor coolant system should occur at DNB ratios greater than 1.30.()

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)

Maxmu Sfet Sstm Setting IIIl and Hih Presure 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 pres Isure 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/feedflo w 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. Oa r FiR E L w P ssu e Loss of Flow and N -ve rNo 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 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%.

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 SAN ONOFRE - UNIT 1 2-3

(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 I 2-3a

TABLE 2.1 MAXIMUM SAFETY SYSTEM SETTINGS Three Reactor Coolant Pumps Operating

1. Pressurizer i 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 i credited

2. Pressurizer

< 2220 psig Pressure: High

3. Nuclear Overpower

a. High Setting**

1 109% of indicated full power

b. Low Setting i 25% of indicated full power

- - 4. Variable Low Pressure

> 26.15 (0.894 AT+T avg.) -

14341

- - 5. Coolant Flow 2 85% of indicated full loop flow

6. Reactor Coolant Pump Breaker Open

- - - a. Overcurrent

< 2900 amps at 4160 volts

- - - b. Undercurrent

> 110 amps at 4160 volts 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.

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

SAN ONOFRE -

UNIT 1 2-4

Figure 2.1.1.

Safety Limits

?

-~F~.!

~.. s. Temperature, Power, Pressure

o.

= 30 bi ff M nRCS Pressure goo PSI 150

_7

... 6........

.6

~--8

~05 E77-ercentg-f-hominal-Power-((1347 t't).t4-2-S 6/8/81 Revised:

6/23/81

3.3.3 MINIMUM BORON CONCENTRATION IN THE REFUELING WATER STORAGE TANK (RWST)

AND SAFETY INJECTION (SI) LINES AND MINIMUM RWST WATER VOLUME APPLICABILITY: MODES 1, 2, 3 and 4; or as described in Specification 3.2.

OBJECTIVE:

To ensure immediate availability of borated water from the RWST for safety injection, containment spray or emergency boration.

SPECIFICATION: a. The RWST shall be OPERABLE with a level of at least plant elevation 50 feet of water having a boron concentration of not less than 3750 ppm and not greater than 4300 ppm.

b. The safety injection (SI) lines from the RWST to MOV 850 A, B, and C, with the exception of lines common to the feedwater system, shall be OPERABLE with a boron concentration of not less than 1500 ppm and not greater than 4300 ppm.

ACTION:

A. With the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

B. With one or both SI lines inoperable due to boron concentration of less than 1500 ppm, restore the SI lines to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

BASIS:

The refueling water storage tank serves three purposes; namely:

(1) As a reservoir of borated water for accident mitigation

purposes, (2) As a reservoir of borated water for flooding the refueling cavity during refueling.

(3) As a deluge for fires in containment.

Approximately 220,000 gallons of borated water is required to provide adequate post-accident core cooling and containment spray to maintain calculated post-accident doses below the limits of 10 CFR 100(1). The refueling water storage tank filled to a plant elevation 50 feet represents in excess of 240,000 gallons.

A boron concentration of 3750 ppm in the RWST and 1500 ppm in the SI lines is required to meet the requirements of a postulated steam line break.( 2)(4) A maximum boron concentration of 4300 ppm ensures that the post-accident containment sump water is maintained at a pH between 7.0 and 7.5(3)

SAN ONOFRE -

UNIT 1 3-35

The refueling tank capacity of 240,000 gallons is based on refueling volume requirements and includes an allowance for water not usable because of tank discharge line location.

Sustained temperatures below 32*F do not occur at San Onofre.

At 32*F, boric acid is soluble up to approximately 4650 ppm boron. Therefore, no special provisions for temperature control to avoid either freezing or boron precipitation are necessary.

References:

(1) Enclosure 1 "Post-Accident Pressure Reanalysis, San Onofre Unit 1" to letter dated January 19, 1987 in Docket No.

50-206 (2) "Main Steamline Break Analysis, San Onofre Nuclear Generating Station, Unit 1, August 1988" (3) Additional information, San Onofre, Unit 1 submitted by letter dated March 24, 1987 in Docket No. 50-206 (4) Reload Safety Evaluation, San Onofre Nuclear Generating Station, Unit 1, Cycle 10, edited by J. Skaritka, Revision 1, Westinghouse, March, 1989 SAN ONOFRE -

UNIT 1 3-35a

3.5 INSTRUMENTATION AND CONTROL 3.5.1 REACTOR TRIP SYSTEM INSTRUMENTATION APPLICABILITY:

As shown in Table 3.5.1-1.

OBJECTIVE:

To delineate the conditions of the Plant instrumentation and safety circuits necessary to ensure reactor safety.

SPECIFICATION:

As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.5.1-1 shall be OPERABLE.

ACTION:

As shown in Table 3.5.1-1.

BASIS:

During plant operations, the complete instrumentation systems will normally be in service.(1) Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits.(2) Safety is not compromised, however, by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design.(1)(3) This Standard outlines limiting conditions for operation necessary to preserve the.effectiveness of the reactor control and protection system when any one or more of the channels is out of service.

References:

(1) Final Engineering Report and Safety Analysis, Section 6.

(2) Final Engineering Report and Safety Analysis, Section 6.2.

(3) NIS Safety Review Report, April 1988 SAN ONOFRE -

UNIT 1 3-44

TABLE 3.5.1-1 o

REACTOR TRIP SYSTEM INSTRUMENTATION 0

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTION UNIT OF CHANNELS TO TRIP OPERABLE MODES ION 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#

Overpower Trip

3.

Power Range, Neutron Flux, 4

1**

4 1, 2 28#

Dropped Rod Rod Stop

4. Intermediate Range, Neutron 2

I 2

1##, 2 3

Flux

5. Source Range, Neutron Flux

.n A. Startup 2

1**

2 211 4

B. Shutdown 2

1**

2 3*, 4*, 5 7

C. Shutdown 2

0 I

3, 4, and 5 5

6. NIS Coincidentor Logic 2

1 2

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

7

7.

Pressurizer Variable 3

2 2

IS#S 69 Low Pressure

8. Pressurizer Fixed High 3

2 2

1, 2 6#

Pressure

9.

Pressurizer High Level 3

2 2

I 6#

TABLE 3.5.1-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION 0 z MINIMUM 0m TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTION UNIT OF CHANNELS TO TRIP OPERABLE MO0ES ACION

10. Reactor Coolant Flow A. Sin le L I/loop I/loop in any I/loop in each I

0 (AbovenO of 7ull Power) operating loop operating loop B. Two Loops I/loop Ioop in two I/loop in each f###

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

2 2

1,2 61

12. Turbine Trip-Low Fluid Oil Pressure 3

2 2

li###

611

13.

Reactor Coolant Pump Breaker 1/loop I/loop I/loop I60 Position (Above 50% of Full Power) 4-1

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 Defeat) Setpoint.

Above the P-7 (At Power Reactor Trip Defeat) 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 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be returned to the untripped condition for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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.

SAN ONOFRE -

UNIT 1 3-47

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 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 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.1, provided the other channel is OPERABLE.

SAN ONOFRE -

UNIT 1 3-48

3.5.2 CONTROL ROD INSERTION LIMITS APPLICABILITY:

MODES I and 2 OBJECTIVE:

This specification defines the insertion limits for the control rods in order to ensure (1) an acceptable core power distribution during power operation, (2) a limit on potential reactivity insertions for a hypothetical control rod ejection, and (3) core subcriticality after a reactor trip.

SPECIFICATION:

A. Except during low power physics tests or surveillance testing pursuant to Specification 4.1.1.G, the Shutdown Groups and Control Group I shall be fully withdrawn, and the position of Control Group 2 shall be at or above the 21-step uncertainty limit shown in Figure 3.5.2.1.

B. The energy weighted average of the positions of Control Group 2 shall be at least 90% (i.e. > Step 288) withdrawn after the first 20% burnup of a core cycle. The average shall be computed at least twice every month and shall consist of all Control Group 2 positions during the core cycle.

ACTION:

A. With the control groups inserted beyond the above insertion limits either:

1. Restore the control groups to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

2. Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the above figure, or

3. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

B. With a single dropped rod from a Shutdown Group or Control Group, the provisions of Action A are not applicable, and retrieval shall be performed without increasing THERMAL POWER beyond the THERMAL POWER level prior to dropping the rod. An evaluation of the effect of the dropped rod shall be made to establish permissible THERMAL POWER levels for continued operation. If retrieval is not successful within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months from the time the rod was dropped, appropriate action, as determined from the evaluation, shall be taken. In no case shall operation longer than 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months be permitted if the dropped rod is worth more than 0.4% A k/k.

BASIS:

During Startup and Power Operation, the Shutdown Groups and Control Group 1 are fully withdrawn and control of the reactor is maintained by Control Group 2. The Control Group insertion limits are set in consideration of maximum specific SAN ONOFRE -

UNIT 1 3-49

power, shutdown capability, and the rod ejection accident.

The considerations associated with each of these quantities are as follows:

1. The initial design maximum value of specific power is 15 kW/ft. The values of FJH and FQ total associated with this specific power are 1.75 and 3.23, respectively.

A more restrictive limit on the design value of specific power, FJH and FO is applied to operation in accordance with the current safety analysis including fuel densification and ECCS performance. Thevalues of the specific power, FJH and F0 are 13.2 kW/ft, 1.57 and 2.78, respectively (8). At partial power, the F H maximum values (limits) increase according to t e following equation, F H (P) - 1.57 [1 + 0.2 (-P)], where P is the fraction of RATED THERMAL POWER. The Control Group insertion limits in conjunction with Specification B prevent exceeding these values even assuming the most adverse Xe distribution.

2. The minimum shutdown capability required is 1.25% Ap at BOL, 1.91 Ap at EOL and defined linearly between these values for intermediate cycle lifetimes. The rod insertion limits ensure that the available SHUTDOWN MARGIN is greater than the above values.

3. The worst case ejected rod accident (9) covering HFP-BOL, HZP-BOL, HFP-EOL shall satisfy the following accident safety criteria:

a) Average fuel pellet enthalpy at the hot spot below 225 cal/gm for nonirradiated fuel and 220 cal/gm for irradiated fuel.

b) Fuel melting is limited to less than the innermost 10% of the fuel pellet at the hot spot.

Low power physics tests are conducted approximately one to four times during the core cycle at or below 10% RATED THERMAL POWER. During such tests, rod configurations different from those specified in Figure 3.5.2.1 may be employed.

It is understood that other rod configurations may be used during physics tests. Such configurations are permissible based on the low probability of occurrence of steam line break or rod ejection during such rod configurations.

SAN ONOFRE - UNIT 1 3-50

Operation of the reactor during cycle stretch out is conservative relative to the safety considerations of the control rod insertion limits, since the positioning of the rods during stretch out results in an increasing net available SHUTDOWN MARGIN.

Compliance with Specification B prevents unfavorable axial power distributions due to operation for long intervals at deep control rod insertions.

The presence of a dropped rod leads to abnormal power distribution in the core. The location of the rod and its worth in reactivity determines its effect on the temperatures of nearby fuel. Under certain conditions, continued operation could result in fuel damage, and it is the intent of ACTION B to avoid such damage.

References:

(1) Final Engineering Report and Safety Analysis, revised July 28, 1970.

(2) Amendment No. 18 to Docket No. 50-206.

(3) Amendment No. 22 to Docket No...50-206.

(4) Amendment No. 23 to Docket No.90-206.

(5) Description and Safety Analysis, Proposed Change No. 7, dated October 22, 1971.

(6) Description and Safety Analysis Including Fuel Densification, San Onofre Nuclear Generating Station, Unit 1, Cycle 4, HCAP 8131, May, 1973.

(7) Description and Safety Analysis Including Fuel Densification, San Onofre Nuclear Generating Station, Unit 1, Cycle 5, January, 1975, Nestinghouse Non-Proprietary Class 3.

(8) Reload Safety Evaluation, San Onofre Nuclear Generating Station, Unit 1, Cycle 10, edited by J. Skaritka, Revision 1, Westinghouse, March, 1989 (9) An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods, WCAP-7588, Revision 1-A, January, 1975.

SAN ONOFRE - UNIT 1 3-51

CONTROL GROUP INSERTION LIMITS FULLY WITHDRAWN 320 300 224 O20 0

CL

I CL 10/21/88 8100 c

35 FULLY 0

1020304050607080 901002 PERCENT OF RATED THERMAL POWER FIGURE 3.5.2.1 SAN ONOFRE - UNIT 1 3-52 Revised: 11/10/88

3.10 INCORE INSTRUMENTATION APPLICABILTY:

MODE 1 OBJECTIVE:

To specify the type and frequency of incore measurements used to verify linear power density values.

SPECIFICATION:

a. A power distribution measurement shall be performed every 30 Effective Full Power Days (EFPDs) and after attainment of equilibrium xenon upon return to power following a refueling shutdown.

b. The incore instrumentation system shall be used to accomplish the Correlation Verification of incore versus excore data for the axial offset monitoring system prior to exceeding 90% of RATED THERMAL POHER following each refueling and at least once per 180 EFPDs thereafter. Subsequent to the Correlation Verification and for the duration of each cycle, incore instrumentation shall be used to perform a Correlation Check of the axial offset monitoring system every 30 EFPDs.

c. A core thermocouple map shall be taken every 30 EFPDs and after attainment of equilibrium xenon upon return to power following a refueling shutdown.

ACTION:

A. If the correlation check, power distribution measurement or core thermocouple map described above cannot be made within the prescribed time, a maximum of 15 EFPDs will be allowed for equipment correction.

B.

In the event that Specification a, b and c cannot be met during the 15 EFPDs allowed for corrective action, be in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

BASIS:

The flux mapping system is used to measure the core power distribution and to correlate incore versus excore data for the axial offset system. Measurements made with the flux mapping system every 30 EFPDs and upon return to power I

following a refueling shutdown will monitor the core power distribution to confirm that the maximum linear power density remains below allowable values. The axial offset system will monitor the axial core power distribution in a continuous manner. In addition, core thermocouples provide an independent means of measuring the balance of power among the core quadrants.

SAN ONOFRE - UNIT 1 3-91

The flux mapping system and the thermocouple system are not integral parts of the Reactor Protection System. These systems are, rather, surveillance systems which may be required in the event of an abnormal condition such as a power tilt or a control rod misalignment. Since such a condition cannot be predicted, it is prudent to have the surveillance systems OPERABLE. If the measurements cannot be taken as specified, the plant will be placed in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months as specified by the actions..

Operation for a 180 EFPD period prior to reperforming the correlation verification is acceptable on the basis that the allowed incore axial offset limits are reduced by the amount in percent of incore axial offset that the monthly correlation check differs from the correlation.

SAN ONOFRE - UNIT 1 3-92

3.11 CONTINUOUS POWER DISTRIBUTION MONITORING APPLICABILITY:

MODE 1 OBJECTIVE:

To provide corrective action in the event that the axial offset monitoring system limits are approached.

SPECIFICATION:

The incore axial offset limits shall not exceed the functional relationship defined by:

2.78/P -

2.10 For positive offsets:

IAO -

FCC 0.033 2.78/P - 2.10 For negative offsets:

IAO -

+ FCC

-0.033 where IAO -

Incore Axial Offset P = fraction of RATED THERMAL POWER FCC -

The larger of 3.0 or the value in percent of IAO by which the current correlation check differs from the incore-excore correlation.

ACTION:

A. With IAO exceeding the limit defined by the specification, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action shall be taken to reduce THERMAL POWER until IAO is within specified limits.

B. With one or both excore axial offset channel(s) inoperable, as an alternate, one OPERABLE NIS channel for each inoperable excore axial offset channel, shall be logged every two hours to determine IAO.

C. With no method for determining IAO available, be in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SAN ONOFRE - UNIT 1 3-93

DASS:

The percent full power axial offset limits are conservatively established considering the core design peaking factor, analytical determination of the relationship between core peaking factors and IAO considering a wide range of maneuvers and core conditions, and actual measurements relating IAO to the axial offset monitoring systems (1).

The axial offset limit established from the incore versus excore data have been reduced by an amount equivalent to FCC to allow for burnup and time dependent differences between the periodic correlation verification and the monthly correlation check.

Correcting the allowed IAO limits by an amount equal to FCC maintains plant operation within the original safety analysis assumptions. Should a specific cycle analysis establish that the analytical determination of the relationship between core peaking factors and IAO has changed in a manner warranting modification to the existing envelope of peaking factor (1,2), then a change to functional relationship of the specification shall be submitted to the Commission. The incore-excore data correlation is checked or verified periodically as delineated in Specification 3.10, INCORE INSTRUMENTATION.

Reducing power until IAO is within the specified limits in i cases when limits are exceeded, will assure that design limits which were set in consideration of accident conditions are not exceeded. In the event that no method exists for determining IAO, actions are specified to place the plant in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . However, if axial offset channel(s) are inoperable, hand calculational methods of determining IAO from OPERABLE NIS channels can be employed until OPERABILITY of the axial offset channel(s) is restored.

(1) Reload Safety Evaluation, San Onofre Nuclear Generating Station, Unit 1, Cycle 10, edited by J.

Skaritka, Revision 1, Westinghouse, March, 1989 (2) Supporting Information on Periodic Axial Offset I

Monitoring, San Onofre Nuclear Generating Station, Unit 1, September, 1973 (3) Supporting Information on the Continuous Axial Offset I Monitoring System, San Onofre Nuclear Generating Station, Unit 1, July, 1974 (4) Description and Safety Analysis, Including Fuel I

Densification, San Onofre Nuclear Generating Station, Unit 1 Cycle 5, January, 1975, Westinghouse Non-Proprietary Class 3.

SAN ONOFRE - UNIT 1 3-94

4.1.1 OPERATIONAL SAFETY ITEMS Appcability:

Applies to surveillance requirements for items directly related to Safety Standards and Limiting Conditions for Operation.

Objective:

To specify the minimum frequency and type of surveillance to be applied to plant equipment and conditions.

Specification:

A. Reactor Trip System instrumentation shall be checked, tested, and calibrated as indicated in Table 4.1.1.

B. Equipment and sampling tests shall be as specified in Table 4.1.2.

C. The specific activity and boron concentration of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.1.2., Item la.

D. The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.1.2.,

Item lb.

E. All control rods shall be determined to be above the rod insertion limits shown in Figure 3.5.2.1 by verifying that each analog detector indicates at least 21 steps above the rod.insertion limits, to account for the instrument inaccuracies, at least once per shift during Startup conditions with Keff equal to or greater than one.

F. The position of each rod shall be determined to be within the group demand limit and each rod position indicator shall be determined to be OPERABLE by verifying that the rod position indication system (Analog Detection System) and the step counter indication system (Digital Detection System) agree within 35 steps at least once per shift during Startup and Power Operation except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the rod position indication system (Analog Detection System) and the step counter indication system (Digital Detection System) at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

G. During MODE I or 2 operation each rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

H. Instrumentation shall be checked, tested, and calibrated as indicated in Table 4.1.3.

SAN ONOFRE - UNIT 1 4-3

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

1. Manual Reactor Trip N.A.

N.A.

R N.A.

2. Power Range, Neutron Flux S

0 (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 (1),

N.A.

Neutron Flux M

5.

Source Range, Neutron Flux S

R (3)

S/U (1),

N.A.

M

6.

NIS Coincidentor Logic N.A.

N.A.

M (5)

7. Pressurizer Variable Low S

R M

N.A.

Pressure

8.

Pressurizer Pressure S

R M

N.A.

9.

Pressurizer Level S

R M

N.A.

10. Reactor Coolant Flow S

R Q

N.A.

II.

Steam/Feedwater Flow S

R M

N.A.

Mismatch

12.

Turbine Trip-Low Fluid N.A.

N.A.

S/U (1,6)

N.A.

Oil Pressure

13.

Reactor Coolant Pump Breaker S

R R

N.A.

Position

(This page intentionally blank)

SAN ONOFRE -

UNIT 1 4-5

TABLE 4.1.1 (Continued)

TABLE NOTATION (1) -

If not performed in previous 31 days.

(2) -

Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent.

(3) -

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(4) -

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(5) -

Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(6) -

Setpoint verification is not applicable.

SAN ONOFRE - UNIT 1 4-6

TABLE 4,1.2 MINIMUM EOUIPMENT CHECK AND SAMPLING FREQUENCY Check FrequencY la. Reactor Coolant 1. Gross Activity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Samples Determination Required during Modes 1, 2, 3 and 4.

2. Isotopic Analysis I per 14 days. Required for DOSE EQUIVALENT only during Mode 1.

1-131 Concentration

3. Spectroscopic I per 6 months( 2) for E(1)

Required only during Determination Mode 1.

4. Isotopic Analy-a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,( 3) sis for Iodine whenever the specific Including 1-131, activity exceeds 1-133, and 1-135.

1.0 pCi/gram DOSE EQUIVALENT 1-131 or 100/ E (1) pCi/gram.

b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period.

5. Boron concentration Twice/Week (1) E is defined in Section 1.0.

(2) Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

(3) Until the specific activity of the reactor coolant system is restored within its limits.

SAN ONOFRE - UNIT 1 4-7 Revised:

12/21/88

TABLE 4.1.2 (continued)

Check FrequencY 1.b Secondaryl

1. Gross Activity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Coolant Determination Required only during Samples Modes 1, 2, 3 and 4.

2. Isotopic Analy-a) 1 per 31 days, whenever sis for DOSE the gross activity EQUIVALENT 1-131 determination indicates Concentration iodine concentrations greater than 10% of the allowable limit. Required only during Modes 1, 2, 3 and 4.

b) 1 per 6 months, whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit. Required only during Modes 1, 2, 3, and 4.

SAN ONOFRE - UNIT 1 4-8 AMENDMENT:

TABLE 4.1.2 (continued)

Check Frequency

2. Safety

a. Boron Concentration Monthly when the reactor is Injection Line critical and prior to return and RNST Hater of criticality when a-period Samples of subcriticality extends the test beyond 1 month

3. Control Rod

a. -Verify that all rods At each refueling shutdown Drop move from full out to full in, in less than 2.44 seconds

4. (Deleted)

5. Pressurizer

a. Pressure Setpoint At each refueling shutdown Safety Valves

6. Main Steam

a. Pressure Setpoint At each refueling shutdown Safety Valves

7. Main Steam

a. Test for Operability At each refueling shutdown Power Operated Relief Valves

8. Trisodium

a. Check for system At each refueling shutdown Phosphate availability as Additive delineated in Technical Specification 4.2

9. Hydrazine

a. Hydrazine concentra-Once every six months when Tank Hater tion the reactor is critical and Samples prior to return of critica lity when a period of subcriticality extends the test interval beyond six months

10. Transfer

a. Verify that the fuse Monthly, when the reactor is Switch No. 7 block for breaker critical and prior to 8-1181 to MCC 1 is returning reactor to criti removed cal when period of subcriti cality extended the test interval beyond one month SAN ONOFRE -

UNIT 1 4-9 AMENDMENT:

TABLE 4.1.2 (continued)

Check Frequency

11.

MOV-LCV-1100 C

a. Verify that the fuse Same as Item 10 above Transfer Switch block for either breaker 8-1198 to MCC 1 or breaker 42-12A76 to MCC 2A is removed.

12. Emergency Siren

a. Verify that the fuse Same as Item 10 above Transfer Switch block for either breaker 8-1145 to MCC 1 or breaker 8-1293A to MCC 2 is removed

13. Communication

a. Verify that the fuse Same as Item 10 above Power Panel block for either Transfer Switch breaker 8-1195 to MCC 1 or breaker 8-12938 to MCC 2 is removed 14a. Spent Fuel Pool Verify water level per

a. Once every seven days Hater Level Technical Specification when spent fuel is being 3.8 stored in the pool.

b. Refueling Pool

b. Hithin two hours prior Hater Level to start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during movement of fuel assemblies or RCC's.

15. Reactor

a. Per Technical Specifi-

a. Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Coolant Loops/

cations 3.1.2.C and Residual Heat 3.1.2.D, in Mode 1 Removal Loops and Mode 2 and in Mode 3 with reactor trip breakers closed, verify that all required reactor coolant loops are in operation and circulating reactor coolant.

b. Per Technical Specifi cation 3.1.2.E, in Mode 3 with the reactor trip breakers open, verify SAN ONOFRE - UNIT 1 4-10 Revised:

7/1/88

TABLE4.1.2 (continued)

Check FrequencY

1. At least two required

1. Once per 7 days reactor coolant pumps are operable with correct breaker align ments and indicated power availability.

2. The steam generators

2. Once per 12 associated with the two hours required reactor coolant pumps are operable with secondary side water level

> 256 inches of narrow range on cold calibrated scale.

3. At least one reactor

3. Once per 12 coolant loop is in hours operation and circulating, reactor coolant.

c. Per Technical Specification 3.1.2.F, in Mode 4 verify

1. At least two required

1. Once per 7 days (RC or RHR) pumps are operable with correct breaker alignments and indicated power availability.

2. The required steam

2. Once per 12 generators are operable hours with secondary side water level > 256 inches of narrow range on cold calibrated scale.

3. At least one reactor

3. Once per 12 coolant loop/RHR train hours is in operation and circulating reactor coolant.

d. Per Technical Specifications 3.1.2.G and 3.1.2.H, in Mode 5 verify, as applicable:

SAN ONOFRE - UNIT 1 4-11 AMENDMENT:

TABLE 4.1.2 (continued)

Check Freguency

1. At least one RHR train

1. Once per 12 is in operation and hours circulating reactor coolant.

2. When required, one

2. Once per 7 additional RHR train is days operable with correct pump breaker alignments and indicated power availability.

3. When required, the

3. Once per 12 secondary side water level hours of at least two steam generators is 1 256 inches of narrow range on cold calibrated scale.

e. Per Technical Specification

e. Once per 12 3.8.A.3, in Mode 6, with water hours level in refueling pool greater than elevation 40 feet 3 inches, verify that at least one method of decay heat removal is in operation and circulating reactor coolant at a flow rate of at least 400 gpm.

f. Per Technical Specification 3.8.A.4, in Mode 6, with water level in refueling pool less than elevation 40 feet 3 inches, verify

1. At least one decay heat

1. Once per 12 removal method is in hours operation and circulating reactor coolant.

2. One additional decay heat 2. Once per 7 removal method is operable days with correct pump breaker alignments and indicated power availability.

16.

RWST

a. Verify volume 2 50 ft. plant

a. Monthly when the Contained elevation reactor is critical Water Volume and prior to return of criticality when a period of subcri ti cali ty extends the surveillance beyond I month SAN ONOFRE - UNIT 1 4-12 AMENDMENT:

ATTACHMENT 3 ACCIDENT ANALYSES

ML13316B997

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