ML13316B324

Amend 122 to License DPR-13,revising Safety Analysis for Increase in Steam Generator Tube Plugging from 15% to 20%, Including Reanalysis of Loca,Reactor Coolant Pump Locked Rotor & Shaft Break & Addl Delay in Safety Injection Lines

ML13316B324
Person / Time
Site:	San Onofre
Issue date:	04/14/1989
From:	Knighton G Office of Nuclear Reactor Regulation
To:
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ML13316B323	List: ML13316B322 ML13316B324 ML13316B325 ML13316B326
References
DPR-13-A-122 NUDOCS 8904280289
Download: ML13316B324 (29)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION o

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 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No.122 License No. DPR-13

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

A. The application for amendment by Southern California Edison Company and San Diego Gas and Electric Company (the licensee) dated January 11, 1989, as supplemented January 27, March 4 and 11, and April 1, 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 amendment can be conducted without endangering 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 not 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.

8-90428028 890414 PDR ADOCK 0500020A PDC

-2

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. 122, 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

. Knighto Director Project Directora e V Division of Reactor Projects -

III, IV, V and Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 14, 1989

ATTACHMENT TO LICENSE AMENDMENT NO.122 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 5a 25 24a 25 28 28 29 29 29a 31 31 32 32 32a 32a 39a 39a 39b 39b 39c 39c 39d 39d 40b 40b 41 41 42 42 43 43 43a 43a 43b 43b 43c 43c 43d 43d 43d-1

-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 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 AMENDMENT NO.: #M,07,17,122

-3 of determining the probability of DNB occurrence. The ratio of the heat flux at which DNB is expected to occur for a given set of conditions to the actual heat flux experienced at a point is the DNB ratio and reflects the probability that DNB will actually occur.

It has been determined that under the most unfavorable conditions of power distribution expected during core lifetime and if a DNB ratio of 1.44 should exist, not more than 7 out of the total of 28,260 fuel rods would be expected to experience DNB. These conditions correspond to a reactor power of 125% of rated power. Thus, with the expected power distribution and peaking factors, no significant release of fission products to the reactor coolant system should occur at DNB ratios greater than 1.30.(0) The DNB ratio, although fundamental, is not an observable variable. For this reason, limits have been placed on reactor coolant temperature, flow, pressure, and power level, these being the observable process variables related to determination of the DNB ratio. The curves presented in Figure 2.1.1 represent loci of conditions at which a minimum DNB ratio of 1.30 or greater would occur. (1)(2)(3)

Maximum Safety System Settings

1. Pressurizer High Level and High Pressure In the event of loss of load, the temperature and pressure of the Reactor Coolant System would increase since there would be a large and rapid reduction in the heat extracted from the Reactor Coolant System through the steam generators. The maximum settings of the pressurizer high level trip and the pressurizer high pressure trip are established to maintain the DNB ratio above 1.30 and to prevent the loss of the cushioning effect of the steam volume in the pressurizer (resulting in a solid hydraulic system) during a loss-of-load transient.(3)(4) In order to meet acceptance criteria for certain secondary side transients, the pressurizer high level trip must be set at 50% narrow range level or less.( 8)

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

loss of coolant flow, rod withdrawal at power, 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.

AMENDMENT NO.: M,77,122

-4 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. 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 )

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

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

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

AMENDMENT NO.: 97,777,122

-5

References:

(1) Amendment No. 10 to the Final Engineering Report and Safety Analysis, Section 4, Question 3 (2) Final Engineering Report and Safety Analysis, Paragraph 3.3 (3) Final Engineering Report and Safety Analysis, Paragraph 6.2 (4) Final Engineering Report and Safety Analysis, Paragraph 10.6 (5) Final Engineering Report and Safety Analysis, Paragraph 9.2 (6) Final Engineering Report and Safety Analysis, Paragraph 10.2 (7) NIS Safety Review Report, April 1988 (8) SCE to NRC letter November 20, 1987, Engineered Safety Features Single Failure Analysis (9) Reload Safety Evaluation, Cycle 10, Revision 1, March 1989, by Westinghouse, editor 3. Skaritka AMENDMENT NO.:

117,122

5a TABLE 2.1 MAXIMUM SAFETY SYSTEM SETTINGS Three Reactor Coolant Pumps Operating

1. Pressurizer 1 50% Pressurizer Narrow Range Level High Level

2. Pressurizer s 2220 psig Pressure: High

3. Nuclear Overpower

a. High Setting' 1 109% of indicated full power

b. Low Setting

, 25% of indicated full power

• • 4.

Variable Low Pressure 1 26.15 (0.894 HT+T avg.) -

14341

• • 5. Coolant Flow I 85% of indicated full loop flow

• • • 6. Steam/Feedwater Flow Mismatch

a. Low+ Setting:

Steam Flow - Feedwater Flow I

0.25 Feedwater Flow @ 100% Power

b. High+ Setting:

Feedwater Flow - Steam Flow I

0.25 Feedwater Flow @ 100% Power

• • 7.

Reactor Coolant Pump Breaker Open

a.

Overcurrent 1 2400 amps

b.

Undercurrent 110 amps

c.

Undervoltage 2 60% of rated bus voltage 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 High and Low feedwater flow relative to steam flow AMENDMENT NO.: 122

24a 3.3.3 MINIMUM BORON CONCENTRATION IN THE REFUELING WATER STORAGE TANK (RWST)

AND SAFETY INJECTION (SI) LINES AND MINIMUM RWST HATER 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 I hour or be in at least 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 /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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 I hour or 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 /> and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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).

AMENDMENT NO.:

122

25 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 3. Skaritka, Revision 1, Westinghouse, March, 1989 AMENDMENT NO.: ?5,00,122

-28 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.(l)(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 AMENDMENT NO.:

3,77,122

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

1 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

II 4

1, 2 28#

Dropped Rod Rod Stop

4. Intermediate Range, Neutron 2

I 2

I##, 2 3

Flux

5. Source Range, Neutron Flux A. Startup 2

1**

2 2ff 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

I 2

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

7

7. Pressurizer Variable 3

2 2

1 ###

6#

Low Pressure

8. Pressurizer Fixed High 3

2 2

I, 2 69 Pressure

9. Pressurizer High Level 3

2 2

1 6#

tIl Z0 t=1 0

TABLE 3.5.1-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

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

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

6#

(Above 50% of Full Power) operating loop operating loop B. Two Loops I/loop I/loop in two I/loop in each 1ff 6*

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

2 2

1999ff 6#

12.

Turbine Trip-Low Fluid Oil Pressure 3

2 2

I99#9

13.

Reactor Coolant Pump Breaker Position A.

Single Loo I/loop I/loop in any I/loop In each I

6#

(Above 505 of Full Power) operating loop operating loop B.

Two Loops I/loop I/loop in two I/loop In each 16#

(Below 50% of Full Power) operating loops operating loop tII z 0

-31 3.5.2 CONTROL ROD INSERTION LIMITS APPLICABILITY:

MODES 1 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 1 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

2. Reduce THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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 <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 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 <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> be permitted if the dropped rod is worth more than 0.4% H 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 AMENDMENT NO.:

32 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 FNH 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, FNH and FO is applied to operation in accordance with the current safety analysis including fuel densification and ECCS performance. The values of the specific power, FNH and FO are 13.2 kW/ft, 1.57 and 2.78, respectively (8).

At partial power, the FNH maximum values (limits) increase according to the following equation, FNH (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% Hp at BOL, 1.9% Np 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.

• AMENDMENT NO.:

,777,1l22

32a 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, 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) 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.

AMENDMENT NO.:

77,122

39a 3.10 INCORE INSTRUMENTATION APPLICABILITY:

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 POWER 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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 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.

AMENDMENT NO.: $,117,122

39b 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 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.

AMENDMENT NO.: 7,77?,122

39c 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

AMENDMENT NO.:

122

39d 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 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 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. 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.

References:

(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 Monitoring, San Onofre Nuclear Generating Station, Unit 1, September, 1973 (3) Supporting Information on the Continuous Axial Offset j Monitoring System, San Onofre Nuclear Generating Station, Unit 1. July, 1974 (4) Description and Safety Analysis, Including Fuel Densification, San Onofre Nuclear Generating Station, Unit 1 Cycle 5, January, 1975, Westinghouse Non-Proprietary Class 3.

AMENDMENT NO.: 7,7,77,122

40b 4.1.1 OPERATIONAL SAFETY ITEMS Applicability:

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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

G. During MODE 1 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.

AMENDMENT NO.:

00,M117,122

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.

N.A.

R N.A.

2.

Power Range, Neutron Flux S

D (2,3)

M N.A.

N.A.

R (3,4)

3.

Power Range, Neutron Flux, N.A.

N.A.

M N.A.

N.A.

Dropped Rod Rod Stop

4.

Intermediate Range, S

R (3,4)

S/U (1),

N.A.

N.A.

Neutron Flux M

5.

Source Range, Neutron Flux S

R (3)

S/U (1),

N.A.

N.A.

M

6.

NIS Coincidentor Logic N.A.

N.A.

N.A.

N.A.

M (5)

7.

Pressurizer Variable Low S

R M

N.A.

N.A.

Pressure

8.

Pressurizer Pressure S

R M

N.A.

N.A.

9.

Pressurizer Level S

R M

N.A.

N.A.

10. Reactor Coolant Flow S

R Q

N.A.

N.A.

II. Steam/Feedwater Flow S

R M

N.A.

N.A.

Mismatch

12. Turbine Trip-Low Fluid N.A.

N.A.

N.A.

S/U (1,6)

N.A.

Oil Pressure

13.

Reactor Coolant Pump Breaker S

R R

N.A.

N.A.

Position*

N)bl I

42 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.

AMENDMENT NO.: 7,,7,122

43 TABLE 4.1.2 MINIMUM EQUIPMENT CHECK AND SAMPLING FREQUENCY Check Frequency la. Reactor Coolant 1. Gross Activity At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

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

2. Isotopic Analysis 1 per 14 days.

Required for DOSE EQUIVALENT only during Mode 1.

1-131 Concentration

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

Required only during Determination Mode 1.

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

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

b) One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

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

AMENDMENT NO.:

70,70,90,17,122

43a TABLE 4.1.2 (continued)

Check Frequency 1.b Secondary

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

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.

AMENDMENT NO.: M00,7030,100,122

43b 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 RWST Water of criticality when a period Samples of subcriticality extends the test beyond I 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 Water 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 to MCC I is returning reactor to criti removed cal when period of subcriti cality extended the test interval beyond one month AMENDMENT NO.: H,70,122

43c 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 8-1293B 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter during movement of fuel assemblies or RCC's.

15.

Reactor

a. Per Technical Specifi-

a. Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 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 AMENDMENT NO.:

,1,70,122

43d 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 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 1 256 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:

AMENDMENT NO.: 77,122

43d -1 TABLE 4.1.2 (continued)

Check Frequencv

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 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 > 50 ft. plant

a. Monthly when the Contained elevation reactor is critical Water Volume and prior to return of criticality when a period of subcriticality extends the surveillance beyond I month AMENDMENT NO.:

122