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Proposed Tech Specs Changes Associated W/Conversion of Narrow Range Transmitters
	ML13331B129
Person / Time
Site:	San Onofre
Issue date:	02/17/1989
From:	; SOUTHERN CALIFORNIA EDISON CO.
To:	;
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ML13331B128	List: ML13331B127; ML13331B129;
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	NUDOCS 8902210476
	Download: ML13331B129 (14)
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Attachment 2 Supplemental Changes To Proposed Change No. 184

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3.1.2 OPERATIONAL COMPONENTS APPLICABILITY:

Applies to the operating status of the reactor coolant system equipment and related equipment.

For the applicable surveillance requirements, see Table 4.1.2.

OBJECTIVE:

To identify those conditions of the reactor coolant system necessary to ensure safe reactor operation.

SPECIFICATIONS: A. At least one pressurizer safety valve shall be operable or open when the reactor head is on the vessel, except for hydrostatic tests.

B. The reactor shall not be made critical or maintained critical unless both pressurizer safety valves are operable.

C. During Modes 1 and 2 and in Mode 3 with reactor trip breakers closed, all three reactor coolant loops and their associated steam generators and reactor coolant pumps shall be in operation. With less than the above required coolant loops in operation, be in at least Hot Standby with reactor trip breakers open within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , except as modified by Specification 0 below.

D. The limitations of Specification C may be suspended as follows:

1. During Modes 1 and 2, operation may be conducted with 0, 1, 2 or 3 reactor coolant pumps operating at less than 5% of full power for purposes of conducting low power physics testing.

2. During Modes 1 and 2 and in Mode 3 with reactor trip breakers closed, operation may be conducted for less than 24 consecutive hours with one or two reactor coolant pumps operating if reactor power is less than 10% of full power.

E. During Mode 3 with the reactor trip breakers open, the following specifications shall apply:

1. At least two of the reactor coolant loops listed below shall be operable:

a. Reactor coolant loop A and its associated steam generator and reactor coolant pump.

b. Reactor coolant loop B and its associated steam generator and reactor coolant pump.

c. Reactor coolant loop C and its associated steam generator and reactor coolant pump.

2. At least one of the above reactor coolant loops shall be in operation.*

3. With less than the above required reactor coolant loops operable, restore the required loops to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system and immediately initiate corrective action to return the required reactor coolant loop to operation.

F. During Mode 4, the following specifications shall apply:

1. At least two of the reactor coolant loops/residual heat removal (RHR) trains listed below shall be operable:

a. Reactor coolant loop A and its associated steam generator and reactor coolant pump.

b. Reactor coolant loop B and its associated steam generator and reactor coolant pump.

c. Reactor coolant loop C and its associated steam generator and reactor coolant pump.

d. Residual heat removal (RHR) pump G-14A and one associated RHR train.

e. Residual heat removal (RHR) pump G-14B and one associated RHR train.

2. At least one of the above loops/trains shall be in operation.**

All reactor coolant pumps may be de-energized for up to one hour provided (a) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (b) core outlet temperature is maintained at least 40OF below saturation temperature.

All reactor coolant pumps and residual heat removal pumps may be deenergized for up to one hour provided (a) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (b) core outlet temperature is maintained at least 40*F below saturation temperature.

3. With less than the above required loops/trains operable immediately initiate corrective action to return the required loops/trains to operable status as soon as possible; if the remaining operable loop/train is an RHR train, be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4. With no loop or train in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system and immediately initiate corrective action to return one required loop or train to operation.

G. During Mode 5 with reactor coolant loops filled, the following specifications shall apply:

1. At least one residual heat removal (RHR) train shall be operable and in operation*, and either

a. One additional RHR train shall be operable,** or

b. The secondary side water level of at least two steam generators shall be greater than or equal to 256 inches of actual level.

2. With less than the above required loops/trains operable, or with less than the required steam generator level, immediately initiate corrective action to return the required loops/trains to operable status or to restore the required level as soon as possible.

3. With no RHR train in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system and immediately initiate corrective action to return the required RHR train to operation.

H. During Mode 5 with reactor coolant loops not filled, the following specifications shall apply:

The RHR pump may be de-energized for up to one hour provided (a) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (b) core outlet temperature is maintained at least 40*F below saturation temperature.

One RHR train may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing, provided the other RHR train is operable and in operation.

3.5.6 ACCIDENT MONITORING INSTRUMENTATION APPLICABILITY:

MODES 1, 2 and 3.

OBJECTIVE:

To ensure reliability of the accident monitoring instrumentation.

SPECIFICATION:

The accident monitoring instrumentation channels shown in Table 3.5.6-1 shall be OPERABLE.

ACTION:

A. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.5.6-1, except as noted in ACTIONS B and C, either restore the inoperable channel(s) to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B. With one or more channels of Auxiliary Feedwater Flow Rate or Steam Generator Water Level or RCS Loop Delta-T indication inoperable, restore the inoperable channel(s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

C. With channels from more than one type of Auxiliary Feedwater Flow Indication inoperable, restore the inoperable channel(s) to OPERABLE status such that no more than one type of indication has an inoperable channel(s) within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 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 in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

D. With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.5.6-1, except as noted in ACTIONS B and C, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

E. The provisions of Specification 3.0.4 are not applicable for Specifications A and D above.

BASIS:

The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident,"

December 1975 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."

The Auxiliary Feedwater flow indication is subject to the more restrictive ACTION requirements for the AFW system. In order to satisfy decay heat removal requirements and minimize

the potential for exceeding water hammer flow limits for a main feedwater line break upstream of the in-containment check valve, the OPERABILITY of AFW Train B is subject to the ability to equalize flow to the steam generators.

Verification of equalization is provided by the AFW flow transmitters. If the capability to equalize flow or the ability to verify equalization is not available, Train A would be utilized to provide the necessary decay heat removal capability. AFW Train A provides adequate flow for this scenario without reliance on operator action to equalize flow.(3) The steam generator wide range level indicators and the RCS loop delta-T indicators provide backup means for verification of auxiliary feedwater flow to the steam generators, and also provide alternate means for identification of a broken loop.

References:

(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.

(2) NRC letter dated November 1, 1983, from 0. G. Eisenhut to all Pressurized Water Reactor Licensees, NUREG-0737 Technical Specification (Generic Letter No. 83-37).

(3) SCE letter dated November 6, 1987, from M. 0. Medford to NRC Document Control Desk.

TABLE 3.5.6-1 ACCIDENT MONITORING INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS INSTRUMENT OF CHANNELS OPERABLE Pressurizer Water Level 3

2 Auxiliary Feedwater Flow Indication*

o Auxiliary Feedwater Flow Rate 1/steam generator 1/steam generator o Steam Generator Water Level (Wide Range) 1/steam generator 1/steam generator S

o Reactor Coolant System Loop Delta-T Indication 1/loop 1/loop Reactor Coolant System Subcooling Margin Monitor 2

1 PORV Position Indicator (Limit Switch) 1/valve 1/valve PORV Block Valve Position Indicator (Limit Switch) 1/valve 1/valve Safety Valve Position Indicator (Limit Switch) 1/valve 1/valve Containment Pressure (Wide Range) 2 1

Refueling Water Storage Tank Level 1

1 Containment Sump Water Level (Narrow Range)**

2 1

Containment Water Level (Wide Range) 2 1

Neutron Flux (Wide Range) 2 1

Auxiliary feedwater flow indication for each steam generator is provided by one channel of auxiliary feedwater flow rate (Train B), one channel of environmentally qualified steam generator wide range level (Train A), and one channel of RCS Loop Delta-T indication. These comprise the three types of indication of auxiliary feedwater flow for each steam generator.

- Operation may continue up to 30 days with one less than the total number of channels OPERABLE.

TABLE 3.5.7-2 AUXILIARY FEEDWATER INSTRUMENTATION SETPOINTS FUNCTIONAL UNIT SETPOINT ALLOWABLE VALUES

a. Manual Actuation Not Applicable Not Applicable

b. Automatic Actuation Logic Not Applicable Not Applicable

c. Steam Generator Water Level-Low

> 238 inch level of wide

> 233 inch level of range instrument span wide range instrument for each steam generator span for each steam generator

d. AFW Train Interlocks

i. Decreasing Flow in Train B/

Start Train A Flow 23 gpm*

> 10 gpm ii. Increasing Flow in Train B/

Stop Train A Flow 37 gpm*

< 48 gpm

Each flow switch monitoring Train B flow utilizes its set and reset points for permissive signals for starting and stopping Train A.

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 actual level.

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 actual level.

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:

A*

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 actual level.

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.

TABLE 4.1.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION Pressurizer Water Level M

R Auxiliary Feedwater Flow Indication*

o Auxiliary Feedwater Flow Rate M

R o Steam Generator Water Level (Wide Range)

M R

o Reactor Coolant System Loop Delta-T Indication M

R Reactor Coolant System Subcooling Margin Monitor M

R PORV Position Indicator M

R PORV Block Valve Position Indicator M

R Safety Valve Position Indicator M

R Containment Pressure (Wide Range)

M R

Refueling Water Storage Tank Water Level M

R Containment Sump Water Level (Narrow Range)

M R

Containment Water level (Wide Range)

M R

Neutron Flux (Wide Range)

M R**

See footnote of Table 3.5.6-1.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

AI AUXILIARY FEEDWATER SYSTEM FLOW TEST PURPOSE:

Testing will be performed in Mode 1 on the Auxiliary Feedwater System, Train A to provide flow rate data for the steam driven AFW pump, G-10, acting alone and in combination with the motor driven AFW pump, G-10S.

This test is being performed to verify that the system performance with AFW pump G-10 alone, and with pump G-10 plus G-10S, meets or exceeds the analyzed flow requirements for these pump combinations. Analyses have been completed (Reference 1) which show that acceptable AFW system flowrates to the steam generators are achieved for all AFWS design basis events. Mode 5 testing of the Train A motor driven AFW pump, G-10S, and the Train B motor driven AFW pump, G-10W, will be completed prior to Mode 4 entry to verify that the auxiliary feedwater flow rates provided by these pumps and the modified piping system meets or exceed the analyzed flow requirements.

DESCRIPTION:

This test is being performed in Mode 1, with reactor power level at less than 25% of full power (nominally at 15% to 20% of full power).

The control rods will be automatically controlled and the turbine generator will be on-line and connected to the switchyard. The main feedwater flow control valves are expected to be in automatic, but may be manually controlled. The 15% to 20% reactor power level is optimal in that the reactor is in a stable condition, and the automatic control systems are available to minimize plant perturbations resulting from initiation, changes in, and termination of auxiliary feedwater system flow. Since data for the Train A steam driven pump alone is required, the steam driven pump will be placed in automatic, and the motor driven pump will be in the manual position. This will allow the steam driven pump to start when the AFWS is manually initiated from the control room. The motor driven pump will then be started by placing the pump control in automatic to obtain combined pump flow rates.

Train B of the AFWS will be in the automatic mode, enabling it to respond to a valid AFW actuation signal.

If a valid AFW actuation occurs during the test, the Train B pump will start, interlocks will close the Train A discharge valves, and the breaker to the Train A motor driven pump will open, as designed. All other engineered safety systems will be in their normal, operable configurations during this test.

-2 ENGINEERING EVALUATION:

The AFW system Train A test in Mode 1 below 25% power was assessed to determine the transient RCS response. Initial conditions include the plant at 25% power with the turbine generator on-line and connected to the switchyard, and the control rods in automatic. The MFW flow control valves would either be automatically or manually controlled. Per SONGS 1 Drawing No. 568793, Heat Balance Diagram at 112,548 KW Gross, MFW temperature would be about 308 0F and flow rate would be approximately 3250 gpm. Startup of AFW Train A (AFW pumps G-10 and G-10S) would simultaneously increase feedwater flow and decrease feedwater temperature. The MFW flow control valves would be controlled to compensate for the increase in feedwater flow. Assuming MFW flow at approximately 3000 gpm and 3080F at 25% power, AFW flow at approximately 300 gpm and 600F, the effective (mixed) feedwater temperature is 2830 F, or a reduction in feedwater enthalpy of 25 BTU/1b. A decrease in feedwater temperature causes a decrease in the temperature of the reactor coolant, resulting in an increase in reactor power due to the negative moderator temperature coefficient, and a decrease in the RCS and steam generator pressures. Without control system action, the reactor would reach equilibrium at a higher power level.

The consequences of the AFW system test are bounded by the excess feedwater event analyzed in the SONGS 1 UFSAR. The UFSAR feedwater event resulted in an RCS temperature cooldown rate of less than 1oF per minute. Based on a comparison of the feedwater enthalpy changes, the RCS cooldown rate due to the addition of AFW (25 BTU/lb) would be less than or equal to the UFSAR feedwater event cooldown rate. At an initial test power level less than 25%, a somewhat greater reduction in enthalpy would result, but the cooldown rate would remain bounded by the 1oF per minute UFSAR cooldown rate. The Unit 1 UFSAR, Section 7.4, excess feedwater event resulted in a power increase of less than 5% per minute assuming an end of life moderator temperature coefficient of -3.5 x 10(-4) delta k/oF. At beginning of life, the moderator temperature coefficent is estimated at -1.0 x 10(-4) delta k/oF, so that the power increase for the AFW test would be significantly less. Control rod motion would restore the primary average temperature. If the reactor control system were unable to maintain plant conditions within the protection limits during the transient, the overpower or variable low pressure protection (floor value) will cause a reactor trip. However, the power transients are expected to be very small, and a reactor trip is not anticipated. (RCS pressure decrease for the UFSAR analysis is less than 20 psi.)

The acceptability of consequences for the limiting design basis transients (feedline breaks, loss of normal feedwater) at a power level of 25% was also evaluated. AFW flow requirements for the FWLB-U and LONF events at 50% power are 125 gpm and 185 gpm, respectively (Reference 1).

(The Train A AFW flow of 250 gpm for

A*

-3 the FWLB-D event will be verified with pump G-10S alone during testing to be completed in Mode 5.)

At 25% power, AFW flow requirements would be approximately one-half these values, or 63 gpm (FWLB-U) and 93 gpm (LONF).

These values are approximately 48% and 36%, respectively, of the predicted combined AFW flow from Train A AFW pumps G-10 and G-IOS.

Hence, AFW flow requirements for design basis events at 25% power are satisfied with less than one-half the predicted AFW system flows.

Since the expected values from AFW pump G-10S alone are 40 gpm (FWLB-U) and 204 gpm (LONF), it is anticipated that the combined flow from pumps G-10 and G-10S will exceed AFW flow requirements at 25% power. AFW pump G-10 is definitely expected to provide the minimal additional flow of 23 gpm required for the FWLB-U event.

CONCLUSION:

All safety systems, including the auxiliary feedwater system, will be operable during the performance of this test. As discussed in the Engineering Evaluation above, the transient system responses were evaluated and are bounded by existing Unit 1 UFSAR analyses. The probability of a turbine trip is not increased as a result of this test. Manual control of main feedwater flow (based on steam generator level) at approximately 15% power or less is part of normal plant operation. No accidents or malfunctions of equipment previously analyzed in the UFSAR could result from or be caused by performance of this test.

Technical Specification changes associated with the installation of the third auxiliary feedwater pump are included in Amendment Application No.

158. These Technical Specification changes will be reviewed and approved by the NRC prior to Mode 4 entry. Performance of this test to verify flow rates of the AFW Train A motor driven pump, G-10S, does not impact or reduce the margin of safety as defined in the basis for the Technical Specifications associated with the auxiliary feedwater system, or any other Technical Specification.

REFERENCE:

1.

Letter from M. 0. Medford (SCE) to the NRC, Engineered Safety Features Single Failure Analysis, SONGS 1; dated November 20, 1987.

03270

ML13331B129

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