ML18057A965

Proposed Tech Specs Re Variable High Power Trip Setpoint

ML18057A965
Person / Time
Site:	Palisades
Issue date:	05/30/1991
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18057A964	List: ML18057A963 ML18057A965 ML18057A966
References
NUDOCS 9106180004
Download: ML18057A965 (8)

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* ATTACHMENT 1 Consumers Power Company Palisades Plant Docket 50-255 TECHNICAL SPECIFICATIONS CHANGE REQUEST VARIABLE HIGH POWER TRIP SETPOINT CHANGE PROPOSED CHANGED PAGES May 30, 1991 3 Pages

• TABLE 2.3.1 Reactor Protective Svstem Trip Setting Limits 1. Variable High Power<1> 2. Primary Cool ant Fl ow<2> 3. High Pressure Pressurizer

4. Thermal Low Pressure'2* > 5. Steam Generator Low Water Level 6. Steam Generator Low Pressure<2> 7. Containment High Pressure Four Primary Coolant Pumps Operating above core power, with a minimum setpoint of of rated power and a maximum of of rated power of Primary Coolant Flow With Four Pumps Operating Psia P1;ri1? Applicable L1m1ts Not Lower Than the Center Line of Feed-Water Ring Psia Psig Three Primary Coolant Pumps Op er at i ng<4>

above core power with a minimum setpoint of rated power and a maximum of of rated power of Primary ant Flow With Four Pumps Operating Psia Replaced by Variable High Power Trip and 1750 Psia Minimum Pressure Setting Not Lower Than the Center Line of Feed-Water Ring Psia Psig (1) The VHPT can be 30% of rated power for power levels 20% of rated power. (2) May be bypassed below 10-4% of rated power provided auto bypass removal circuitry is operable.

For low power physics tests, thermal margin/low pressure, primary coolant flow and low steam generator pressure trips may be bypassed until their react points are reached (approximately 1750 psia and 500 psia, respectively), provided automatic bypass removal circuitry at 10-1% rated power is operable.

(3) Minimum trip setting shall be 1750 psia. (4) Operation with three pumps for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is permitted to provide a limited time for repair/pump restart, to provide for an orderly shutdown or to provide for the conduct of reactor internals noise monitoring test measurements.

2-5 Amendment No. $J, TSCR9104.CTH

• * . 2.3 LIMITING SAFETY SYSTEM SETTINGS -REACTOR PROTECTIVE SYSTEM (Contd) The reactor protective system consists of four instrument channels to monitor selected plant conditions which will cause a reactor trip if any of these conditions deviate from a preselected operating range to the degree that a safety limit may be reached. 1. Variable High Power -The variable high power trip (VHPT} is incorporated in the reactor protection system to provide a reactor trip for transients exhibiting a core power increase starting from any initial power level (such as the boron dilution transient).

The VHPT system provides a trip setpoint no more than a Rredetermined amount above the indicated core power

> and also sets a maximum value. Operator action is required to increase the setpoint as core power is increased; the setpoint is automatically decreased as core power decreases.

Provisions have been made to select different set points for three pump and four pump operations.

2. TSCR9104.CTH During normal plant operation with all primary coolant pumps operating, reactor trip is initiated when the reactor power level reaches 106.5% of indicated rated power. Adding to this the possible variation in trip point due to calibration and instrument errors, the maximum actual steady state power at which a trip would be actuated is 115%, which was used for the purpose of safety analysis.

<4> Primary Coolant System Low Flow -A reactor trip is provided to protect the core against DNB should the coolant flow suddenly decrease significantly.<

3> Flow in each of the four coolant loops is determined from a measurement of pressure drop from inlet to outlet of the steam generators.

The total flow through the reactor core is measured by summing the loop pressure drops across the steam generators and correlating this pressure sum with the pump calibration flow curves. The percent of normal core flow is shown in the following table: 4 Pumps 3 Pumps 100.0% 74.7% During four-pump operation, the low-flow trip setting of 95% insures that the reactor cannot operate when the flow rate is Less than 93% of the nominal value considering instrument errors. <4> 2-6 Amendment No. pJ, Jpl

• 2.3 LIMITING SAFETY SYSTEM SETTINGS -REACTOR PROTECTIVE SYSTEM (Contd) Basis (Contd) 6. Low Steam Generator Pressure -A reactor trip on low steam generator secondary pressure is provided to protect against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the primary coolant. The setting of 500 psia is sufficiently below the rated load operating point of 739 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This setting was used in the accident analysis.

7. Containment High Pressure -A reactor trip on containment high pressure is provided to assure that the reactor is shutdown before the initiation of the safety injection system and containment spray. <10> 8. Low Power Phvsics Testing -For low power physics tests, certain tests will require the reactor to be critical at low temperature and low pressure psia). For these certain tests only, the thermal margin/low pressure, primary coolant flow and low steam generator pressure trips may be bypassed in order that reactor power be increased for improved data acquisition.

Special operating precautions will be in effect during these tests in accordance with approved written testing procedures.

At reactor power levels below 10-1% of rated power, the thermal margin/low-pressure trip and low flow trip are not required to prevent fuel rod thermal limits from being exceeded.

The low steam generator pressure trip is not required because the low steam generator pressure will not allow a severe reactor cooldown, should a steam line break occur during these tests. References (1) ANF-91-181, Section 15.0.7 (2) deleted (3) Updated FSAR, Section 7.2.3.3. (4) ANF-90-078, Section 15.0.7 (5) XN-NF-86-9l(P)

(6) deleted (7) deleted (8) ANF-90-078, Section 15.1.5 (9) ANF-87-150(NP), Volume 2, Section 15.2.7 (10) Updated FSAR, Section 7.2.3.9. (11) ANF-90-078, Section 15.2.1 2-9 Amendment No. $1, J$7 TSCR9104.CTH

• ATTACHMENT 2 Consumers Power Company Palisades Plant Docket 50-255 TECHNICAL SPECIFICATIONS CHANGE REQUEST VARIABLE HIGH POWER TRIP SETPOINT CHANGE EXISTING PAGES WITH PROPOSED CHANGES MARKED May 30, 1991 3 Pages

• TABLE. 2.3.1 Reactor Protective System Trip Setting Limits Four Primary Coolant Pumps Operating Three Primary Coolant Pumps

1. Variable High Power'1> \5%

above core power, with a minimum setpoint of of rated power and a maximum of of rated power core power with a minimum setpoint of rated power and a maximum of of rated power 2. Primary Cool ant Fl ow<2>

of Primary Coolant Flow With Four Pumps Operating of Primary ant Flow With Four Pumps Operating

3. High Pressure Pressurizer Psia Psia 4. Thermal Marginf, Low Pressure<2* > Ptrlp Limits Replaced by Variable High Power Trip and 1750 Psia Minimum Pressure Setting 5. Steam Generator Low Water Level Not Lower Than the Center Line of Feed-Water Ring Not Lower Than the Center Ltne of Feed-Water Ring 6. Steam Generator Low Pressure<2>

Psia Psia 7. (1) (2) (3) (4) Containment High Pressure Psig ** . .:*** .. 'l -. ; The VHPT can be 30% of rated for power levels 20% of rated power. . -* * .. * * , ** -May be bypassed below 10-43 of rated power provided auto bypass removal circuitry is operable.

For low power physics tests, thermal margin/low pressure, primary coolant flow and low steam generator pressure trips may be bypassed until their react points: .. reached .. (approx*imately 1750 psi a and 500 psi respectively), provided automatic bypass removal circuitry at 10-1% rated power* is

• Minimum trip setting shall be 1750 psia. Operation with three pumps for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is permitted to provide a limited time for repair/pump restart, to provide for an orderly shutdown or to provide for the conduct of reactor internals noise monitoring tesf measurements.

2-5 Amendment No.

138 February 22, 1991

• 2.3 LIMITING SAFETY SYSTEM SETTINGS -REACTOR PROTECTIVE SYSTEM (Contd) The reactor protective system consists of four instrument channels to monitor selected plant conditions which will cause a reactor trip if any of these conditions deviate from a preselected operating range to the degree that a safety limit may be reached. 1. Variable High Power -The variable high power trip (VHPT) incorporated in the reactor protection system to provide a reactor trip for transients exhibiting a core power increase starting from any initial power level (such as the boron dilution transient).

The VHPT system provides a trip setpoint no more (I) than a predetermined amount above the indicated core Ai-lC AL?o I Operator action is required to increase the setpoint as core A power is increased; the setpoint is automatically decreased as core power decreases.

Provisions have been made to select different set points for three pump and four pump operations.

During normal plant operation with all primary coolant pumps ope rat i n*g, reactor trip is initiated when the reactor power level reaches 106.5% of indicated rated power. Adding to this the possible variation in trip point due to calibration and instrument errors, the maximum actual steady state power at which a trip would be actuated is 115%, which was used for the purpose l of safety analysis.-<-1+-

Cq) 2. Primary Coolant System Low Flow -A reactor trip is provided to protect the core against DNB should the coolant flow suddenly decrease significantly.<

3> Flow in each of the four coolant loops is determined from a measurement of pressure drop from inlet to outlet of the steam generators.

The total flow through the reactor core is measured by summing the loop pressure drops across the steam generators and correlating this pressure sum with the pump calibration flow curves. The percent of normal core flow is shown in the following table: 4 Pumps 3 Pumps 100.0% 74.7% During four-pump operation, the low-flow trip setting of 95% insures that the reactor cannot operate when the flow rate is Less than 93% of the nominal value considering instrument errors. <4> 2-6 Amendment No. $J, 137 February 20, 1991

2.3 LIMITING

SAFETY SYSTEM SETTINGS -REACTOR PROTECTIVE SYSTEM (Contd) Basis (Contd) 6. Low Steam Generator Pressure -A reactor trip on low steam generator secondary pressure is provided to protect against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the primary coolant. The setting of 500 psia is sufficiently below the rated load operating point of 739 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This setting was used in the accident analysis.

ca> 7. Containment High Pressure -A reactor trip on containment high pressure is provided to assure that the reactor is shutdown before the initiation of the safety injection system and containment spray. <10> . 8. Low Power Physics Testing -For low power physics tests, certain tests will require the reactor to be critical at low temperature and*low pressure psia). For these certain tests only, the thermal margin/low pressure, primary coolant flow and low steam generator pressure trips may be bypassed in order that reactor power can be increased for improved data acquisition.

Special operating precautions will be in effect during these tests in accordance with approved written testing procedures.

At reactor power levels below 10-1% of rated power, the thermal margin/low-pressure trip and low flow trip are not required to prevent fuel rod thermal limits from being exceeded.

The low steam generator pressure trip is not required because the low steam generator pressure will not allow a severe reactor cooldown, should a steam line break occur during these tests. References ANF-C\t-1e1>

5ec.T1oto1 1c;.o.1 (1) ANF 98 878, Tahle 16.8.7 1 (2) deleted (3) Updated FSAR, Section 7.2.3.3. (4) AHF 98 878, Sectiou 15.8.1 1 i\NF-90:--016i SEc..'T10H.

\S.0 .. 1 (5) XN-NF-86-9l(P)

(6) deleted (7) deleted (8) ANF-90-078, Section 15.1.5 (9) ANF-87-150(NP), Volume 2, Section 15.2.7 {10) Updated FSAR, Section 7.2.3.9. {11) ANF-90-078, Section 15.2.1 2-9 Amendment No.

137 February 20, 1991