ML18064A411

Proposed Tech Specs Re Pressure/Temp Limit

ML18064A411
Person / Time
Site:	Palisades
Issue date:	10/05/1994
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18064A409	List: ML18064A408 ML18064A411 ML18064A412
References
NUDOCS 9410140197
Download: ML18064A411 (44)

Text

a ATTACHMENT 1 Consumers Power Company Pali sades Pl ant Docket 50-255 PRESSURE/TEMPERATURE LIMIT TECHNICAL SPECIFICATIONS CHANGE REQUEST Proposed Pages 9410140197 941005 PDR ADDCK 05000255 p PDR

PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTION DESCRIPTION PAGE NO 1.0 DEFINITIONS 1-1 1.1 OPERATING CONDITIONS 1-1 1.2 MISCELLANEOUS DEFINITIONS 1-5 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 2.1 SAFETY LIMITS - REACTOR CORE 2-1 2.2 SAFETY LIMITS - PRIMARY COOLANT SYSTEM PRESSURE 2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS - RPS 2-1 Table 2.3.1 Reactor Protective System Trip Setting Limits 2-2 B2.1 Basis - Reactor Core Safety Limit B2-1 B2.2 Basis - Primary Coolant System Safety Limit B2-2 B2.3 Basis - Limiting Safety System Settings B2-3 3.0 LIMITING CONDITIONS FOR OPERATION 3-1 3.0 APPLICABILITY 3-1

3. I PRIMARY COOLANT SYSTEM 3-lb

3. I. I Operable Components 3-lb Figure 3-0 ASI Limit for Tinl t function 3-3a 3.1.2 Pressure - Temperature Limits 3-4 Figure 3-1 Pressure - Temperature Limits for Heatup 3-5 Figure 3-2 Pressure - Temperature Limits for Cooldown 3-6 3.1.3 Minimum Conditions for Criticality 3-12 3.1.4 Maximum Primary Coolant Radioactivity 3-17 3.1. 5 Primary Coolant System Leakage Limits 3-20 3.1.6 Maximum PCS Oxygen and Halogen Concentration 3-23 3.L7 Primary and Secondary Safety Valves 3-25 3 .1.8 Over Pressure Protection Systems 3-25a Figure 3-4 LTOP Limit Curve 3-25c 3 .1. 9 Shutdown Cooling 3-25h 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM 3-26 3.3 EMERGENCY CORE COOLING SYSTEM 3-29 3.4 CONTAINMENT COOLING 3-34 3.5 STEAM AND FEEDWATER SYSTEMS 3-38 3.6 CONTAINMENT SYSTEM 3-40 Table 3.6.1 Containment Penetrations and Valves 3-40b 3.7 ELECTRICAL SYSTEMS 3-41 3.8 REFUELING OPERATIONS 3-46 3.9 Deleted 3-49 Amendment No.

3.1 PRIMARY COOLANT SYSTEM

3. I. I Operable Components (continued)

h. Forced circulation (starting the first primary coolant pump) shall not be initiated unless one of the following conditions is met:

(I) PCS cold leg temperature (Tc) is> 430°F.

(2) S/G secondary temperature is ~ T~.

(3) S/G secondary temperature is < I00°F above Tc, and shutdown cooling is isolated from the PCS, and PCS heatup/cooldown rate is ~ I0°F/hour.

(4) S/G secondary temperature is < l00°F above Tc, and shutdown cooling is isolated from the PCS, and pressurizer level is ~ 57%.

i. When the PCS cold leg temperature is< 300°F and pressurizer level is> 57% the primary coolant pumps P-50A and P-50B shall not be operated simultaneously.

j. The PCS shall not be heated or maintained above 300°F unless a minimum of 375 kW of pressurizer heater capacity is available from both buses ID and IE. Should heater capacity from either bus ID or lE fall below 375 kW, either restore the inoperable heaters to provide at least 375 kW of heater capacity from both buses ID and IE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next I2 hours .

• Basis When primary coolant boron concentration is being changed, the process must be uniform throughout the pri.mary coolant system volume to prevent stratification of primary coolant at lower boron concentration which could result in a reactivity insertion. Sufficient mixing of the primary coolant is assured if one shutdown cooling or one primary coolant pump is in operation.< 1> The shutdown cooling pump will circulate the primary system volume in less than 60 minutes when operated at rated capacity. By imposing a minimum shutdown cooling pump flow rate of 28IO gpm, sufficient time is provided for the 5operator to terminate the boron dilution under asymmetric flow conditions.< > The pressurizer volume is relatively inactive, therefore will tend to have a boron concentration higher than rest of the primary coolant system during a dilution operation. Administrative procedures will provide for use of pressurizer sprays to maintain a nominal spread between the boron concentration in the pressurizer and the primary system during the addition of boron. <2 >

The limitation on operating P-50A and P-50B together with Tc below 300°F and pressurizer level above 57% allows the Pressure Temperature limits to be I9 psi higher than they would be without this limit. The 57% pressurizer level is not an analyti~al result, but simply a decision point between having and not having a bubble. It was chosen to agree with the maximum programmed level during power operation.

3-Id Amendment No. fH-, 8§., -H-7, HS, -l-3-1-, 6-l,

3.1 PRIMARY COOLANT SYSTEM Specification

• 3.1.2 PCS pressure, PCS temperature, and PCS heatup and cooldown rates shall be maintained within the following limits:

a. The primary coolant system (PCS) pressure shall be maintained within the limits of Figures 3-1 and 3-2.

b. The pressurizer heatup and cooldown rates be maintained

~ l00°F/hour **.

c. The primary coolant system (PCS) heatup and cooldown rates be maintained within the following limits:

Reactor Vessel Inlet Temperature (T) Heatup Rate Limit Cooldown Rate Limit T ~ 170°F 20°F/hour 40°F/hour 250 ~ T > 170°F 40°F/hour 40°F/hour 350 > T > 250°F G0°F/hour ** G0°F/hour T ~ 350°F l00°F/hour l00°F/hour

• • When shutdown cooling isolation valves M0-3015 and M0-3016 are open, PCS heatup rate shall be maintained ~ 40°F/hour and the

• pressurizer heatup rate shall be maintained ~ 60°F/hour.

Applicability Specification 3.1.2 applies at all times.

Action

a. If the limits of Specification 3.1.2 are exceeded:

1. Return to within limits within 30 minutes, and

2. Determine that the PCS condition is acceptable for continued operation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

b. If any action required by 3.1.2a is not met and the associated completion time has expired:

1. The reactor shall be placed in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and

2. The reactor shall be placed in a COLD SHUTDOWN with PCS pressure less than 270 psia, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

• 3-4 Amendment No. !8-, 4l, .§.§., 89, IN-, -H-7, 3-l,

Figure 3-1 Pressure-Temperature Limits for Heatup 2000

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750 o~_.._.._..._... ......_.._.._..._..........."'P"l'........"'P"l'"'P"l'................................"T"T""i 50 100 150 200 250 300 350 400 450 RV Inlet Temperature, F 3-5 Amendment No. ~' 41, .£5., 89, 91-, .}l-7, .i-3-1-,

Figure 3-2 Pressure-Temperature Limits for Cooldown 2500

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50 100 150 200 250 300 350 400 450 RV Inlet Temperature~ F 3-6 Amendment No. Y, 4+/-, .SS, 89, 9+, -H-7, 3+,

3.1 PRIMARY COOLANT SYSTEM Basis - Pressure Temperature Limits:

• The Primary Coolant System Pressure-Temperature limits are calculated for a reactor vessel wall fluence of 2.192 x 10 19nvt. Before the radiation exposure of the reactor vessel exceeds that fluence, Figures 3-1 and 3-2 shall be updated in accordance with the following criteria and procedure:

1. US Nuclear Regulatory Commission Regulatory Guide 1.99 Revision 2 has been used to predict the increase in transition temperature based on integrated fast neutron flux and surveillance test data. If measurements on the irradiated specimens show increase above this curve, a new curve shall be constructed such that it is above and to the left of all applicable data points.

2. Before the end of the integrated power period for which Figures 3-1 and 3-2 apply, the limit lines on the figures shall be updated for a new integrated power period. The total integrated reactor thermal power from start-up to the end of the new power period shall be converted to an equivalent integrated fast neutron exposure (E ~ 1 MeV). Such a conversion shall be made consistent with the dosimetry evaluation of capsule W-290< 12 >.

3. The limit lines in Figures 3-1 and 3-2 are based on the requirements of Reference 9, Paragraphs IV.A.2 and IV.A.3.

All components in the primary coolant system are designed to withstand the effects of cyclic loads due to primary system temperature and pressure changes.<1> These cyclic loads are introduced by normal unit load transients, reactor trips and start-up and shutdown operation. During unit start-up and shutdown, the rates of temperature and pressure changes are limited. A maximum plant heatup and cooldown limit of l00°F per hour is consistent with the design number of cycles and satisfies stress limits for cyclic operation. <2 >

The reactor vessel plate and material opposite the core has been purchased to a specified Charpy V-Notch test result of 30 ft-lb or greater at an NDTT of+

l0°F or less. The vessel circumferential weld has the highest RTNor of plate, 10 weld and HAZ materials at the fluence to which the Figures 3-1 and 3-2 apply. < > The unirradiated RTN 1 has been determined to be -56°F. <11 > An RT Nor of -56°F is used as an unirra~1ated value to which irradiation effects are added. In addition, the plate has been 100% volumetrically inspected by ultrasonic test using both longitudinal and shear wave methods. The remaining material in the reactor vessel, and other primary coolant system components, meets the appropriate design code requirements and specific component function and has a maximum NOTT of +40°F.< 5 >

As a result of fast neutron irradiation in beltline region of the core, there will be an increase in the RTNor with operation. The integrated fast neutron (E >1 MeV) fluxes of the reactor vessel are contained in Reference 13. .

3-7 Amendment No. !8-, 4!, .§.§., 89, 9+, -H+, H+,

3.1 PRIMARY COOLANT SYSTEM Basis - Pressure Temperature Limits: (continued)

Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured transition shift from a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in calculated flux magnitude. The maximum exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calculated azimuthal neutron flux variation. The predicted RTNor shift for the base metal has10been predicted based upon surveillance data and the US NRG Regulatory Gui de. < > To compensate for any increase in the RT Nor caused by irradiation, limits on the pressure-temperature relationship are periodically changed to stay within the stress limits during heatup and cooldown.

Reference 7 provides a procedure for obtaining the allowable loadings for ferritic pressure-retaining materials in Class 1 components. This procedure is based on the principles of linear elastic fracture mechanics and involves a stress intensity factor prediction which is a lower bound of static, dynamic and crack arrest critical values. The stress intensity factor computed< 7 > is a function of RTNDT' operating temperature, and vessel wall temperature gradients.

Pressure-temperature limit calculational procedures for the reactor coolant pressure boundary are defined in Reference 8 based upon Reference 7. The limit lines of Figures 3-1 and 3-2 include an allowance to account for the fact that pressure is measured in the pressurizer rather than at the vessel beltline and to account for PCP discharge pressure. In addition, for calculational purposes, S°F and 30 psi was taken as measurement error allowances for calculation of criticality temperature. By Reference 7, reactor vessel wall locations at 1/4 and 3/4 thickness are limiting. It is at these locations that the crack propagation associated with the hypothetical flaw must be arrested. At these locations, fluence attenuation and thermal gradients have been evaluated. During cooldown, the 1/4 thickness location is always more limiting in that the RTNor is higher than that at the 3/4 thickness location and thermal gradient stresses are tensile there. During heatup, either the 1/4 thickness or 3/4 thickness location may be limiting depending upon heatup rate.

Figures 3-1 and 3-2 define stress limitations only from a fracture mechanics point of view.

Other considerations may be more restrictive with respect to pressure-temperature limits. For normal operation, other inherent plant characteristics may limit the heatup and cooldown rates which can be achieved. Pump parameters and pressurizer heating capacity tends to restrict both normal heatup and cooldown rates to less than 60°F per hour.

3-8 Amendment No. ~, 4-l-, .§.S, &9, 9+, H-7, -l-3-1-,

3.1 PRIMARY COOLANT SYSTEM Basis - Pressure Temperature Limits: (continued)

The revised pressure-temperature limits 19are applicable to reactor vessel inner wall fluences of up to 2.192 x 10 nvt. The application of appropriate fluence attenuation factors (Reference 10) at the 1/4 and 3/4 thickness locations results in RTNor shifts of 255°F and 191°F, respectively, for the limiting weld material.

The criticality condition which defines a temperature below which the core cannot be made critical (strictly based upon fracture mechanics' considerations) is 385°F. The most limiting wall location is at 1/4 thickness. The minimum criticality temperature, 385°F is the minimum permissible temperature for the inservice system hydrostatic pressure test.

That temperature is calculated based upon 2310 psig inservice hydrostatic test pressure.

The restriction of average heatup and cooldown rates to l00°F/hour when all PCS cold legs are ~ 350°F and the maintenance of a pressure-temperature relationship under Figures 3-1 and 3-2 ensures that the requirements of References 7, 8 and 9 are met. Calculation of average hourly cooldown rate must consider changes in reactor vessel inlet temperature caused by initiating shutdown cooling, by starting primary coolant pumps with a temperature difference between the steam generator and PCS, or by stopping primary coolant pumps with shutdown cooling in service.

The heatup and cooldown rate restrictions are consistent with the analyses performed for low temperature overpressure protection (Reference 14). Below 430°F, the Power Operated Relief Valves (PORVs) provide overpressure protection; at 430°F or above, the PCS safety valves provide overpressure protection.

I 3-9

• Amendment No. Y, 4l, .§-5, 89, 91-, -H-7, Hl,

3.1 PRIMARY COOLANT SYSTEM Basis - Pressure Temperature Limits: (continued}

References (1) FSAR, Section 4.2.2.

(2) ASME Boiler and Pressure Vessel Code,Section III, A-2000.

(3) Battelle Columbus Laboratories Report, "Palisades Pressure Vessel Irradiation Capsule Program: Unirradiated Mechanical Properties,"

August 25, 1977.

(4) Battelle Columbus Laboratories Report, "Palisades Nuclear Plant Reactor Vessel Surveillance Program: Capsule A-240," March 13, 1979, submitted to the NRC by Consumers Power Company letter dated July 2, 1979.

(5) FSAR, Section 4.2.4.

(6) (Deleted)

(7) ASME Boiler and Pressure Vessel Code,Section III, Appendix G, "Protection Against Non-Ductile Failure," 1974 Edition.

(8) US Atomic Energy Commission Standard Review Plan, Directorate of Licensing, Section 5.3.2, "Pressure-Temperature Limits."

(9) 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," May 31, 1983 as amended November 6, 1986.

(10) US Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, May, 1988.

(11) Combustion Engineering Report CEN-189, December, 1981.

(12) "Analysis of Capsules T-330 and W-290 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program," WCAP-10637, September, 1984.

(13) Consumers Power Company letter to NRC, June 10, 1993; 10CFR50.61 Pressurized Thermal Shock - Reactor Vessel Neutron Fluence - Additional Information.

(14) Consumers Power Company Engineering Analysis EA-A-PAL-92095-01, Rev O; "Pressure Temperature Curves and LTOP Limit Curve for Maximum Reactor Vessel Fluence of 2.192 x 10 19 Neutron/cm2" (Next Page 3-12) 3-10 Amendment No. !8-, 41, .§.§., 89, IJ+, -H-7, Bl,

3.1 PRIMARY COOLANT SYSTEM 3.1.3 Minimum Conditions for Crtticality a) Except during low-power physics test, the reactor shall not be made critical if the primary coolant temperature is below 525°F.

b) In no case shall the reactor be made critical if the primary coolant temperature is below 385°F.

c) When the primary coolant temperature is below the minimum temperature specified in "a" above, the reactor shall be subcritical by an amount equal to or greater than the potential reactivity insertion due to depressurization.

d) No more than one control rod at a time shall be exercised or withdrawn until after a steam bubble and normal water level are established in the pressurizer.

e) Primary coolant boron concentration shall not be reduced until after a steam bubble and normal water level are established in the pressurizer.

At the beginning of life of the initial fuel cycle, the moderator temperature coefficient is expected to be sliAhtly negative at operating temperatures with all control rods withdrawn.< However, the uncertainty of the calculation is such that it is possible that a slightly positive coefficient could exist.

The moderator coefficient at lower temperatures will be less negative or more positive than at operating temperature. <1 , 2 > It is, therefore, 3-12 Amendment No. 2:7-, 41-, .§.§., 89-, 91-, .f-l-7.,

3.1 PRIMARY COOLANT SYSTEM 3.1.3 Minimum Conditions for Criticality {Cont'd)

Basis {Cont'd) prudent to restrict the operation of the reactor when primary coolant temperatures are less than normal operating temperature {~525°F). Assuming the most pessimistic rods out moderator coefficient, the maximum potential reactivity insertion that could result from depressurizing the coolant from 2100 psia to saturation pressure at 525°F is 0.13 Ap.

During physics tests, special operating precautions will be taken. In addition, the strong negative Doppler coefficient< 3 > and the small integrated Ap would limit the magnitude of a power excursion resulting from a reduction of moderator density. The requirement that the reactor is not to be made critical below 385°F provides increased assurance that the proper relationship between primary coolant pressure and temperature will be maintained relative to the RTNDT of the primary coolant system pressure boundary material. Heatup to this temperature will be accomplished by operating the primary coolant pumps.

If the shutdown margin required by Specification 3.10.l is maintained, there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure.

Normal water level is established in the pressurizer prior to the withdrawal of control rods or the dilution of boron so as to preclude the possible overpressurization of a solid primary coolant system.

References (1) FSAR, Table 3-2 (2) FSAR, Table 3-6 (3) FSAR, Table 3-3 (Next page is 3-17) 3-13 Amendment No . .§.§., 9+,

• Figure 3-4 LTOP Setpoint Limit 2500 .

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50 100 150 200 250 300 350 400 450 PCS Temperature, F 3-25c Amendment No. -+/--3+,

3.1.8 OVER PRESSURE PROTECTION SYSTEMS Basis 3.1.8.1 (continued)

Normally, during operation at HOT STANDBY and above, the PORV controls are in the CLOSE position, and the block valves are closed. The PORVs, block valves, and the associated manual controls must be operable. If either valve in a PORV flow path is inoperable, the other valve in the flow path must provide PCS integrity assurance. When a PORV is inoperable, the block valve must be closed; when a block valve is inoperable, the PORV must have its control in the "CLOSE" position.

If the inoperable valves cannot be restored to OPERABLE status within the specified completion time, the plant must be placed in HOT SHUTDOWN. The completion times allow the required action to be accomplished without undue haste, yet allow less time when more equipment is inoperable.

3.1.8.2 When PCS is below 430°F with the reactor vessel head installed, two PORVs are required to be operable to avoid pressures which might lead to failure of the reactor vessel. Pressure increases could be caused by sudden additions (or imbalances) of either mass or energy.

The allowable pressure limits are determined in accordance with 10 CFR 50, Appendix G, and are referred to as "Low Temperature Overpressure Protection" (LTOP) limits. The variable setpoint of the LTOP system is programmed and calibrated to ensure opening of the pressurizer PORVs when the PCS pressure is above the limit in Figure 3-4. The pressure limit for each temperature is developed from the heating or cooling limits for the PCS.

The limit in Figure 3-4 includes an allowance for pressure overshoot during the interval between the time pressurizer pressure reaches the limit, and the time a PORV opens enough to terminate the pressure rise.

LTOP is provided by two independent channels each consisting of measurement, control, actuation, and valves. Either channel is capable of providing full protection. The actual setpoint of PORV actuation for LTOP will be below the limit in Figure 3-4 to allow for potential instrument inaccuracies, and drift. This will ensure that at no time between calibration intervals will the PCS pressure exceed the limit of Figure 3-4 without PORV actuation.

Mass additions could come from the starting of pumps or from opening a Safety Injection Tank isolation valve. Only the charging pumps or high pressure safety injection pumps could cause the PCS pressure to exceed its limits.

Neither the shutoff head of the low pressure safety injection pumps nor the operating pressure of the safety injection tanks is above the cold PCS pressure limit. Specification 3.3.5 places limits on HPSI pump operability when the PCS is below 300°F to assure inadvertent starting does not cause overpressurization of the PCS.

3-25e Amendment No. -l-l-7-, +/-a+, .f.&G,

3.1.8 OVER PRESSURE PROTECTION SYSTEMS Basis 3.1.8.2 (continued)

Energy additions could come from either the steam generators or from the reactor core. Small energy addition could come from operation of the pressurizer heaters. Energy addition from the steam generators could occur if a primary coolant pump was started when the steam generator secondary temperature was significantly above the PCS temperature. Specification 3.1.1.h places limits on the starting of primary coolant pumps to avoid undesired energy additions from the steam generators. Energy addition from the reactor core could occur due to an inadvertent criticality or to an imbalance in decay heat removal. Specification 3.10.1 places limits on shutdown margin to avoid a rod withdrawal event causing a criticality and to provide sufficient time for operator action to terminate a dilution event prior to criticality.

The potential causes of a sudden PCS pressure increase which the LTOP system must be able to mitigate are imbalance in charging and letdown flow, starting of the HPSI pumps when above 300°F, and in an imbalance in decay heat (and pressurizer heat) addition and removal. A Safety Injection Signal (SIS) could both initiate flow from two HPSI pumps (when above 300°F) and three charging pumps, and isolate letdown. The PCS heatup from a loss of shutdown cooling event occurring 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown from a continuous full power run would generate less additional coolant volume than the starting of three charging pumps (Reference 5). The limiting event for the LTOP system would be an inadvertent SIS occurring during an established PCS heatup.

Analysis (Reference 1) has concluded that an SIS occurring, during a PCS and

• pressurizer heatup at the maximum allowable rates, either between 300°F and 430°F with the HPSI pumps, or below 300°F without the HPSI pumps, would not cause PCS pressure to exceed the Appendix G limit if either PORV opens when the set pressure is reached. With the PCS above 430°F, the pressurizer safety valves, required by Specification 3.1.7, provide adequate overpressure protection. Both PORVs are required to be operable to allow for a single failure.

If a PORV becomes inoperable when it is required for LTOP, it must be restored to operable status, or the plant must be cooled down, depressurized, and vented through a vent path with sufficient capacity to provide the necessary protection. Since the pressure response to a transient is greater if the pressurizer steam space is small or if PCS is solid, the allowed outage time for a PORV flow path out of service is shorter. The maximum pressurizer level at which credit can be taken for having a bubble (57%,

which provides about 700 cubic feet of steam space) is based on judgement rather than on analyses. This level provides the same steam volume to dampen pressure transients as would be available at full power. This steam volume provides time for operator action, if the PORVs failed to operate, between an inadvertent SIS and PCS pressure reaching the 10 CFR 50 Appendix G pressure limit. The time available for action would depend upon the existing pressure and temperature when the inadvertent SIS occurred.

3-25f Amendment No. -l-l-7, -l-3-1-, +/-6-G,

3.1.8 OVER PRESSURE PROTECTION SYSTEMS Basis 3.1.8.2 (continued)

Reference 1 has determined that any vent path capable of relieving 167 gpm at a PCS pressure of 315 psia is acceptable. The 167 gpm flow rate is based on an assumed charging imbalance due to interruption of letdown flow with three charging pumps operating, a 40°F per hour PCS heatup rate, a 60°F per hour pressurizer heatup rate, and an initially depressurized and vented PCS. The PCS heatup rate is limited to 40°F per hour by Specification 3.1.2c; the pressurizer heatup rate is limited to 60°F per hour by Specification 3.1.2b.

Neither HPSI pump nor PCP starts need to be assumed with the PCS initially depressurized, because Specification 3.3.5 requires both HPSI pumps to be incapable of injection into the PCS and operating procedures prohibit PCP operation.

The pressure relieving ability of a vent path depends not only upon the area of the vent opening, but also upon the configuration of the piping connecting the vent opening to the PCS. A long, or restrictive piping connection may prevent a larger vent opening from providing adequate flow, while a smaller opening immediately adjacent to the PCS would be adequate. The areas of multiple vent paths cannot simply be added to determine the necessary vent area.

The following vent path examples are acceptable:

1. Removal of the reactor vessel head,

2. Removal of a steam generator primary manway,

3. Removal of the pressurizer manway,

4. Removal of a PORV or pressurizer safety valve,

5. Both PORVs and associated block valves open,

6. Opening of both PCS vent valves PC-514 and PC-515.

Reference 2 determined that venting the PCS through PC-514 and PC-515 provided adequate flow area. The other listed examples provide greater flow areas with less piping restriction and are therefore acceptable. Other vent paths shown to provide adequate capacity could also be used.

One open PORV provides sufficient flow area to prevent excessive PCS pressure. However, if the PORVs are elected as the vent path, both valves must be used to meet the single failure criterion, since the PORVs are held open against spring pressure by energizing the operating solenoid.

When the shutdown cooling system is in service with M0-3015 and M0-3016 open, additional overpressure protection is provided by the relief valves on the shutdown cooling system. References 3 and 4 show that this relief capacity will prevent the PCS pressure from exceeding its pressure limits during any of the above mentioned events.

References

1. Consumers Power Company Engineering Analysis, EA-A-PAL-92095-01

2. Consumers Power Company Engineering Analysis, EA-TCD-91-01-01.

3. Consumers Power Company Engineering Analysis, EA-PAL-89-040-1

4. Consumers Power Company Corrective Action Document, A-PAL-91-011

5. Consumers Power Company Engineering Analysis, EA-AG-93-02 tit 3-25g Amendment No . .f-l-7., .f-3-1-, -l-6-G,

3.3 EMERGENCY CORE COOLING SYSTEM (Continued) 3.3.3 Prior to returning to the Power Operation Condition after every time the plant has been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and testing of Specification 4.3.h has not been accomplished in the previous 9 months, or prior to returning the check valves in Table 4.3.l to service after maintenance, repair or replacement, the following conditions shall be met:

a. All pressure isolation valves listed in Table 4.3.1 shall be functional as a pressure isolation device, except as specified in

b. Valve leakage shall not exceed the amounts indicated.

b. In the event that integrity of any pressure isolation valve specified in Table 4.3.1 cannot be demonstrated, at least two valves in each high pressure line having a non-functional valve must be in and remain in, the mode corresponding to the isolated condition. en 1

Motor-operated valves shall be placed in the closed position and power supplies deenergized.

3.3.4 Two HPSI pumps shall be operable when the PCS temperature is >325°F.

a) One HPSI pump may be inoperable provided the requirements of Section 3.3.2.c are met.

3.3.5 Two HPSI pumps shall be rendered incapable of injection into the PCS when PCS temperature is <300°F, if the reactor vessel head is installed.

Note: Specification 3.3.5 does not prohibit use of the HPSI pumps for emergency addition of makeup to the PCS.

3-30 Amendment No . .§+, .J:.Gl., -l-1-7, ~, -l-6-1-,

3.3 EMERGENCY CORE COOLING SYSTEM Basis (continued) demonstrate that the maximum fuel clad temperatures that could occur over the break size spectrum are well below the melting temperature of zirconium (3300°F).

Malfunction of the Low Pressure Safety Injection Flow control valve could defeat the Low Pressure Injection feature of the ECCS; therefore, it is disabled in the 'open' mode (by isolating the air supply) during plant operation. This action assures that it will not block flow during Safety Injection.

The inadvertent closing of any one of the Safety Injection bottle isolation valves tn conjunction with a LOCA has not been analyzed. To provide assurance that this will not occur, these valves are electrically locked open by a key switch in the control room. In addition, prior to critical the valves are checked open, and then the 480 volt breakers are opened. Thus, a failure of a breaker and a switch are required for any of the valves to close.

Insuring both HPSI pumps are incapable of injecting into the PCS when the PCS tempe.rature is <300°F eliminates PCS mass additions due to inadvertent HPSI pump starts. Both HPSI pumps starting in conjunction with a charging/letdown imbalance may cause 10CFRSO Appendix G limits to be exceeded when the PCS temperature is <300°F. A note is provided to assure that this specification does not cause hesitation in the use of a HPSI pump for PCS makeup if it is needed due to a loss of shutdown cooling of a loss of PCS inventory.

Rendering the HPSI pumps "incapable of injection" means to assure that a single event cannot cause overpressurization of the PCS due to operation of the HPSI pump. Typical methods of accomplishing this are the pulling of the HPSI pump breaker control power fuses, racking out the HPSI pump circuit breaker, or closing the manual discharge valve.

The requirement to have both HPSI trains operable above 325°F provides added assurance that the effects of a LOCA occurring under LTOP conditions would be mitigated. If a LOCA occurs when the primary system temperature is less than or equal to 300°F, the pressure would drop to the level where low pressure safety injection can prevent core damage. When the PCS temperature is ~300°F and ~325°F operation of the HPSI system would not cause the 10CFRSO Appendix G limits to be exceeded nor is HPSI system operation necessary for core cooling.

References (1) FSAR, Section 9.10.3; (2) FSAR, Section 6.1, (3) FSAR, Section 14.17 (4) Letter, H.G.Shaw (ANF) to R.J.Gerling (CPCo), "Standard Review Plan Chapter 15 Disposition of Events Review for Changes to Technical Specifications Limits for Palisades Safety Injection Tank Liquid Levels", April 11, 1990.

3-33 Amendment No. 1-, .§.I., 9-:l-, H-7, +3-1-, +/-a-e, l,

4 .1 OVERPRESSURE PROTECTION SYSTEM TESTS Surveillance Requirements In addition to the requirements of Specification 4.0.5, each PORV flow path shall be demonstrated OPERABLE by:

1. Testing the PORVs in accordance with the inservice inspection requirements for ASME Boiler and Pressure Vessel Code,Section XI, Section IWV, Category B valves.

2. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months.

3. When the PORV flow path is required to be OPERABLE by Specification 3.1.8.1:

(a) Performing a complete cycle of the PORV with the plant above COLD SHUTDOWN at least once per 18 months.

(b) Performing a complete cycle of the block valve prior to heatup from COLD SHUTDOWN, if not cycled within 92 days.

4. When the PORV flow path is required to be OPERABLE by Specification 3.1.8.2:

(a) Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, at least once per 31 days .

• 5.

(b) Verifying the associated block valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Both High Pressure Safety Injection pumps shall be verified incapable of injection into the PCS at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, unless the reactor head is removed, when either PCS cold leg temperature is <300°F, or when both shutdown cooling suction valves, M0-3015 and M0-3016, are open.

With the reactor vessel head installed when the PCS cold leg temperature is less than 300°F, or if the shutdown cooling system isolation valves M0-3015 and M0-3016 are open, the start of one HPSI pump could cause the Appendix G or the shutdown cooling system pressure limits to be exceeded; therefore, both pumps are rendered inoperable.

Amendment No. 13Q, 149, 16Q, ~'

4-6

ATTACHMENT 2 Consumers Power Company Palisades Plant Docket 50-255 PRESSURE/TEMPERATURE LIMIT TECHNICAL SPECIFICATIONS CHANGE REQUEST Existing Pages Marked to Show Proposed Changes

1.0 DEFINITIONS 1-1 1.1 OPERATING CONDITIONS 1-1 1.2 MISCELLANEOUS DEFINITIONS 1-5 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 2.1 SAFETY LIMITS - REACTOR CORE 2-1 2.2 SAFETY LIMITS - PRIMARY COOLANT SYSTEM PRESSURE 2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS - RPS 2-1 Table 2.3.1 Reactor Protective System Trip Setting Limits 2-2 B2.1 Basis - Reactor Core Safety Limit B2-1 B2.2 Basis - Primary Coolant System Safety Limit B2-2 B2.3 Basis - Limiting Safety System Settings B2-3 3.0 LIMITING CONDITIONS FOR OPERATION 3-1 3.0 APPLICABILITY 3-1 3.1 PRIMARY COOLANT SYSTEM 3-lb 3.1.1 Operable Components 3-lb Figure 3-0 ASI Limit for Tinlet function 3-3a

- Figure 3-2 Figure 3 3 3.1.3 3.1.4 3.1.5 Pressure - Temperature Limits for Cooldown Pressure Tem~erature Limits for Hydro Minimum Conditions for Criticality Maximum Primary Coolant Radioactivity Primary Coolant System Leakage Limits 3-6 3 6 3-12 3-17 3-20 3.1.6 Maximum PCS Oxygen and Halogen Concentration 3-23 3.1.7 Primary and Secondary Safety Valves 3-25 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM 3-26 3.3 EMERGENCY CORE COOLING SYSTEM 3-29 3.4 CONTAINMENT COOLING 3-34 3.5 STEAM AND FEEDWATER SYSTEMS 3-38 3.6 CONTAINMENT SYSTEM 3-40 Table 3.6.1 Containment Penetrations and Valves 3-40b 3.7 ELECTRICAL SYSTEMS 3-41 3.8 REFUELING OPERATIONS 3-46 3.9 Deleted 3-49

3.I PRIMARY COOLANT SYSTEM (CoRt'd)

3. I . I Operab1e Components (CO flt Id r:(@~~:~:!:~:g:~!:l.:

h. Forced circulation (starting the first primary coolant pump) shall not be initiated unless one of the following conditions is met:

(1) Primary cool aRtPP.11 cold 1eg temperature :Iffiii'l:::::11i s > 430° F.

(2) PCS celd leg teiiiperat1:H'e is .s 430°f" aAd S/G....secon~tii.rY temperature is less thaFI PCS eel d leg temf)erature'@flh=.

3

( ) shutdown coo  ;,~~~~~,,9~'!1!~ffi,,!!1::;::~~:~w1~~e:g:1!11:1!!;:1::::::;~;i'\!~*:: at ure

( )i~s\!!::~.:-gher tha*

4 Shutdown cooling is isolated from the PCS AND PCS cold lbg temf)erature 1

1s1:~g~e! :~~.*~!~:::!r!:i;:~;'~::mi-;::

temf)erature is ~ 120°F aAd < 170°F AND S/G secoAdary temf)erature is less thaA l00°F higher thaR PCS cold leg

~j The PCS shall not be heated or maintained above 300°F unless a minimum of

• • • 375 kW of pressurizer heater capacity is available from both buses ID and IE. Should heater capacity from either bus ID -aOO i.ir IE fall below 375 kW, either restore the inoperable heaters to provid.it. . at least 375 kW of heater c.CiP.ii.~Jt.Y ... ftqllJ. both buses ID and IE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in ft&t.

shutdo~mRO.TJ:S.f:UJ.:ffU.O.W.N within the next I2 hours.

Basis When primary coolant boron concentration is being changed, the process must be uniform throughout the primary coolant system volume to prevent stratification of primary coolant at lower boron concentration which could result in a reactivity insertion. Sufficient mixing of the primary coolant is assured if one shutdown cooling or one primary coolant pump is in operation. 111 The shutdown cooling pump wi 11 circulate the primary system volume in less than 60 minutes when operated at rated capacity. By imposing a minimum shutdown cooling pump flow rate of 28IO gpm, sufficient time is provided for the operator to terminate the boron dilution under asymmetric flow conditions.~ 1 The pressurizer volume is relatively inactive, therefore will tend to have a boron concentration higher than rest of the primary coolant system during a dilution operation. Administrative procedures will provide for use of pressurizer sprays to maintain a nominal spread between the boron concentration in the pressurizer and the primary system during the addition of boron. 121

• Amendment No. 6J, ~' H-7, HS, rn, !l!:!:,

0 0

PROPOSED NEW PAGE 3-4

FIGURE 3-1 PALISADES PRrssURE AND TEMPERATURE.LIMITS FOR

• \

• - --- . I

- . ---t-1

. . --+ I

--*--+

r --- --~

I I I. I. ----+

I I

.., .-L 450 TEMPERATURE DEGREES F

• Figure 3-1 Pressure-Temperature Limits for Heatup 2000

-+-OF/Hr

- 20F/Hr

<( 1750 _.._ 40 F/Hr Ul a.. -e- 60 F/Hr

...::::s Cl) 1500

............. 80 F/Hr Cl)

Cl) -+-100 F/Hr Cl) a..

Cl) 1250 N

• c::

s Cl)

Cl)

Cl) a..

1000 50 100 150 ~ 200 250 300 350 400 450 RV Inlet Temperature, F 3-5 Amendment No. !8-, 4!, 5, 89, 91, H-7, -l-3-l-,

• -* -LJSAOES PRESSURE FIGURE 3-2 LIMllS FOR COOL DOWN l 'J 2 t' ~ l .11 l< lU 11/c,. (Nu He.al:iUH!llli!nt Uu~t!rtain~v-:J,nci,~_~dl

'\ *,*:......*'"':'

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S7!IO f.*:f 1500 40 20 LO

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.... r

l

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- ID

.~* i:s rt fJOO

tpl'tii;, , - ' -_

~ ........... ...~""~- ... ~ .,::.1i 1s * -*-1ocf ._ ,,..125 150 S7D I

?25 *250 275 375 TENPERAT~ ~GREES f

Figure 3-2 Pressure-Temperature Limits for Cooldown 2000

-+- 0 F/Hr

-20F/Hr

__..,... 40 F/Hr 1750

~ 60F/Hr

<t - SOF/Hr (i)

Q.a 1500 -+-100, F/Hr Q)

s en en Q)

~ 1250 Q)

N

• c:

s en 1000 en Q) a.

750 o ............................................._.._~ .............-..P"""P"t-._...................

50 100 150 200 250 300 350 400 450 RV Inlet Temperature,*F 3-6 Amendment No. !8-, 41-, ~' ~' 9+, -H-7, -H-1-,

,,, ! i FIGURE ~-l *(

,'_-_f,'

PALISADES PRESSU~*il  !\NO TEMPERATURE

' ,~*:

-* ' --.* '* - . ' . t~.IMITS FOR HYDRO r**,

19 2 f,

• 1.8 X 10 n/c*

--~' ' .' . .,

r, ..

r. t *,**1 J-"'( ,.... -

• . f'.

, 1 ..,.'

• .*/,,...,..I.11-.,........;.;.a..;.;;..;:ii.;,;-....-.;;.;;.;.,.;.;,;;...1----1----1-----tR~-.,:--+.---t----t---H,.--+~-l-f----ill+.i~YU~-+-.,.....  ;: *£

)

[..'::~ '::-: -

...._,;.-....j~.,......;.,;.:.;a...;~io-.o'~,.---11---f---++--+--~1---+----++~-...t~,.--l-+--,,...,. .~._,..-+----I-l****

S7IO...._~.....;..a._;__;,.-1---..-1-~--1~~.f--~M--;,.----t-.---.f--~-+----1t~~~.,_.'"lll-i~..._.----+----t----+

• . *' J,1:

sooo-11-~--1-~--1~~1-----1----1---~~~......--1~~.,.~~~~~~---.-1----1----+--- -.~.

l~~~1~*~-!~~.~~~~~~::+--::1 ii 2'P :c:~~---+--+-~~-t-f-UM> 175 376 ..oo

~edl

• Before the radiation exposure of the reactor vessel exceeds the

e~li~;Mi1i1ini.:;:~:;~: !~~ 1fi6~r~~d:~:~y,~\i!:~:~~:~~,j~!~ ~~ f~r~~e 3 f~i1 ~w~ ng cr i t er i *a*****a"ri"d". . ."J)'rofed ure :

1. US Nuclear Regulatory Commission Regulatory Guide 1.99 Revision 2 has been used to predict the increase in transition temperature based on integrated fast neutron flux and surveillance test data. If measurements on the irradiated specimens show increase above this curve, a new curve shall be constructed such that it is above and to the left of all applicable data points.

2. Before the end of the i n.t.~.9r..~t..~.~....P.9.Wer period for which Figures 3 1, 3 2 aRd 3 3:l!!i!!UllrH:l.f!~IMg apply, the l i mi t l i nes on the figures sha 11 be updafea-*****r<n~********a*. . . new integrated power period. The total integrated reactor thermal power from start-up to the end of the new power period shall be converted to an equivalent integrated fast neutron exposure (E~l MeV). Such a conversion shall be made consistent with the dosimetry evaluation of capsule W-290 1121

• The limit lines in Figures 3 1, 3 2 aAd 3 31.!Hl:~::Ii.!nl:I~H~ are 3.

based on the requirements of Reference 9, P"afagFapli*s******1v. A. 2 and IV.A.3. These liRes reflect a preserviee hydrestatie test pressure ef 2400 psip aAd a vessel flaAge material refereAce temperature ef 60 °F18

• Basis All components in the primary coolant system are designed to withstand the effects of cyclic loads due to primary system temperature and pressure changes. 111 These cyclic loads are introduced by normal unit load transients, reactor trips and start-up and shutdown operation. During unit start-up and shutdown, the rates of temperature and pressure changes are limited. A maximum plant heatup and cooldown limit of l00°F per hour is consistent with the design number of cycles and satisfies stress limits for .eye l i c operation. 121 The reactor vessel plate and material opposite the core has been purchased to a specified Charpy V-Notch test result of 30 ft-lb or greater at an NDTT of + 10°F or less. The vessel ¢.'iU~b.'ifufiifenfimiJ.l!\weld has the highest RTNor of plate, weld and HAZ materiars"'"al"."UWt'*'*tllieWEe to which the Figures 3 1, 3 2 aAd 3 3~it!:l!i!i!iiifill!~~il apply.' 101 The unirradiated RTNor has been determined to be -S6°F.mf An RTNor of -56°F is used as an unirradiated value to which irradiation effects are added. In addition, Amendment No. !1:1-, 4.J., S, 89, 9+, -H-7, -l-3-l,

~.

the plate has been 100% volumetrically inspected by ultrasonic test using both longitudinal and shear wave methods. The remaining material in the reactor vessel, and other primary coolant system components, meets the appropriate design code requirements and specific component function and has a maximum NDTT of +40°F. 151 As a result.of fast neutron irradiation in ~j.1fi[i!I!fiilil!li1iihi region of the core, there wi 11- be an increase in the RTNttt wi th****'aperiff<frf~ The 1

!~~~~r:!:~ ~:~!g:Qlii:i~:l.@~[ijj~ji; R:~~~e~~~x~{*<f~t!~~ z~~:c6g~ ~~r~!d:r~i th the SAi LOR set ef cr.e"S"s***se"C"tloi:ts.

Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured transition shift from a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in calculated flux magnitude. The maximum exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calculated azimuthal neutron flux variation. The predicted RTNor shift

• for the base -metal has been predicted based upon surveil 1 ance data and the US NRC Regulatory Gui de .11°1 To compensate for any increase in the RT caused by irradiation, limits on the pressure-temperature relationship are periodically changed to stay within the stress limits during heatup and cool down.

Reference 7 provides a procedure for obtaining the allowable loadings for ferritic pressure-retaining materials in Class 1 components. This procedure is based on the principles of linear elastic fracture mechanics and involves a stress intensity factor prediction which is a lower bound of static, dynamic and crack arrest critical values. The stress intensity factor computed 171 is a fun ct ion of RT Non operating temperature, and vessel wall temperature gradients.

Pressure-temperature limit calculational procedures for the reactor coolant pressure boundary are defined in Reference 8 based upon Reference

7. The JJ~.H. .J..i.r.i~s of Figures 3-1 threugh 3 3i.ni.I:l.Ei ceRsider a §4 13si 13ressurei~!il.lm!lli!i!!il! a11 o':'ance to account for th.ii . . .facr**lhat p~essure is measured in the pressurizer rather than at the vessel beltline and to account for.PCP discharge pressure. In addition, for calculational purposes, 5°F !P.~iiiiiil~jjjjj[:p;!i'=j[ was taken as measurement error a11 owances for ca lcul at ion of" crlficality temperature. By Reference 7, reactor vesse 1 wall locations at 1/4 and 3/4 thickness are limiting. It is at these locations that the crack propagation associated with the hypothetical flaw must be arrested. At these locations, fluence attenuation and thermal gradients have been Amendment No. ~' 41, .&&, 89, g+, .J-1-7., 3+,

evaluated. During cooldown, the 1/4 thickness location is always more limiting in that the RTNoT is higher than that at the 3/4 thickness location and thermal gradient stresses are tensile there. During heatup, either the 1/4 thickness or 3/4 thickness location may be limiting depending upon heatup rate.

Figures 3-1 thro1:1gh 3 a:ii,4.l!lii!I define stress limitations only from a fracture mechanics poi nr****crr*vlew.

Other considerations may be more restrictive with respect to pressure-temperature limits. *For normal operation, other inherent plant characteristics may limit the heatup and cooldown rates which can be achieved. Pump parameters and pressurizer heating capacity tends to restrict both normal heatup and cooldown rates to less than 60°F per hour.

The revised pressure-temperature limits are19 applicable to reactor vessel inner wall fluences of up to -!-.-&~:~:::~ill: x 10 nvt. The application of appropriate fl uence attenuation facf(J"rs ( Referenc~... J.9) at the V.4. and 3/ 4 thickness locations results in RTNDT shifts of ~~:§:§°F and m1:11:°F, respectively, for the limiting weld material. The******-c-riticality**:-c-cfr1dition, which defines a temperature below which the core cannot be made critical (strictly based upon fracture mechanics' considerations) is 37llii°F. The most limiting wa]J. ...location is at 1/4 thickness. The minimum c'Ffficality temperature, ~lill°F is the minimum permi ssi bl e temperature for the i nservi ce system***nydrostat i c pressure test. That temperature is calculated based upon 2310 psig inservice hydrostatic test pressure.

The restriction of average heatup and cooldown rates to l00°F/hji:U.r when all PCS cold legs are~ 350°F and the maintenance of a pressure*:=*=*=*=*=*=*=*

temperature relationship under the heat1:1p, cooldowR aRd iRservice test

~h our

~=v~ly!qcoo

~ frelFmdiown

~~~~sora 3tfe~~,,f,;fu,,;,,n,~,;,;,,,i=:~,:t~,,~,,,,,,,;,,;~~,:,;,,:,,:,~,;,,:,~~~i=~-~-~if,!~!!!'l~l~e~

lllU'$::1ii'*=*=*eQ.'tt:s:-:1=u*1nr=*=*t:1:iang*es.=*=*=*:ltr'*'r::e:cnih;;tn~r*=ve*s.s.e=* *=*=*:*=, , =md!flii v!~:!e

• Fat=r=*=*5=Ka=*11 be oRly from the time the lower cooldowR rate is req1:1ired. "

The core operatioRal limit applies oRly wheR the reactor is critical.

~!:i:I Amendment No. 2:+, 4-l-, -&&, 89, W-, -H-7-, +/-3-l,

The heatup and cooldown rate restrictions are consistent with the analyses performed for low temperature overpressure protection (LTOP) (Reference~

~ 14 aAe 1§). Below 430°F, the Power Operated Relief Valves (PORVs) provide overpressure protection; at 430°F or above, the PCS safety valves provide overpressure protection.

The eritieality temperatijre is eetermiAee per RefereAee 8 aAe the eere eperatieAal eijrves aehere te the re§ijiremeAts ef RefereAee 9. The iAserviee test eijrves iAeerperate alle~~aAees fer the thermal graeieAts

~~:::i~~;~e:i:~fj~; ~;:;~~e~~=~ee~~~:st:Af!t!~~hi~:;;~!iet!e:!r:r=si=~e.

primary membraAe stress. Dije te the shifts iA RT1d01 , NDTT reqijiremeAts asseeiatee ~1ith AeAreaeter vessel materials are, fer all praetieal pijrpeses, Ae leAger limitiAg.

References ll

• FSAR, Section 4.2.2.

I3 2 ASME Boiler and Pressure Vessel Code,Section III, A-2000.

Battelle Columbus Laboratories Report, "Palisades Pressure Vessel Irradiation Capsule Program: Unirradiated Mechanical Properties,"

August 25, 1977.

(4) Battelle Columbus Laboratories Report, "Palisades Nuclear Plant Reactor Vessel Surveillance Program: Capsule A-240," March 13, 1979,

• submitted to the NRC by -Consumers Power Company 1etter dated July 2, 1979.

FSAR, Section 4.2.4.

(Deleted)

ASME Boiler and Pressure Vessel Code,Section III, Appendix G, "Protection Against Non-Ductile Failure," 1974 Edition.

(8) US Atomic Energy Commission Standard Review Plan, Directorate of Licensing, Section 5.3.2, "Pressure-Temperature Limits."

(9) 10 CFR Part 50,. Appendix G, "Fracture Toughness Requirements," May 31, 1983 as amended November 6, 1986.

(10) US Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, May, 1988.

(1 1) Combustion Engineering Report CEN-189, December, 1981.

( 12 ) "Analysis of Capsules T-330 and W-290 from the Consumers Power Company Palisades Reactor Vessel Radiation Surveillance Program,"

WCAP-10637, September, 1984.

(13) "AAalysis ef Fast NeijtreA Expesijre ef the Palisaees Reaeter Pressijre Vessel" by WestiAgheijse Eleetrie GerperatieA, Mareh 1989.

3.1.3 Minimum Conditions for Criticality a} -Except during low-power physics test, the reactor shall not be made critical if the primary coolant temperature is below 525°F.

b} In no case shall the reactor be !11.~.ge critical if the primary cool ant temperature is below 3111,~1,1° F.

c} When the primary coolant temperature is below the minimum temperature specified in "a" above, the reactor shall be subcritical by an amount equal to or greater than the potential reactivity insertion due to depressurization.

d} No more than one control rod at a time shall be exercised or withdrawn until after a steam bubble and normal water level are established in the pressurizer.

e} Primary coolant boron concentration shall not be reduced until after a steam bubble and normal water level are established in the pressurizer.

At the beginning of life of the initial fuel cycle, the moderator temperature coefficient is expected to be sli~htly negative at operating temperatures with all control rods withdrawn. 1 However, the uncertainty of the calculation is such that it is possible that a slightly positive coefficient could exist.

The moderator coefficient at lower temperatures will be less negative or more positive than at operating temperature. 11 *21 It is, therefore, 3-12

• Amendment No. !B, 41-, .§.£, 89, 9+, H+,

3.1.3 Minimum Conditions for Criticality (Cont'd)

Basis (Cont'd) prudent to restrict the operation of the reactor when primary coolant temperatures are less than normal operating temperature (~525°F).

Assuming the most pessimistic rods out moderator coefficient, the maximum potential reactivity insertion that could result from depressurizing the coolant from 2100 psia to saturation pressure at 525°F is 0.1% Ap.

During physics tests, special operating precautions will be taken. In addition, the strong negative Doppler coefficient~' and the small integrated Ap would limit the magnitude of a power excursion resulting from a reduction of moderator density~*** The requirement that the reactor is not to be made critical below ~3.IS.°F provides increased assurance that the proper rel at i onshi p between*=*=*=*µ*rimary cool ant pressure and temperature will be maintained relative to the RTNoT of the primary coolant system pressure boundary material. Heatup to this temperature will be accomplished by operating the primary coolant pumps.

If the shutdown margin required by Specification 3.10.l is maintained, there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure.

Normal water level is established in the pressurizer prior to the withdrawal of control rods or the dilution o~boron so as to preclude the possible overpressurization of a solid primary coolant system .

References (1) FSAR, Table 3-2 (2) FSAR, Table 3-6 (3) FSAR, Table 3-3

~:111~11:::1~1i1::::::1=:~~ml!11:1::t::1:

3-13 Amendment No. £, 9J-,

(Next ~age is 3 17)

"'. .

• fl} (H

' ~-~~I ~:_:),'

• * *I

~ *. t!;.:'..,..;q:

' "b.~ ~1*:

. . . . . . . . ... ~ . .. -. . . . .. . *. . . .

. ~ ' -.. : . ~'  : ".. -

' Cia .............. . * . . . . . . . . . . . .. . . . . * . . . . . * *' . . . .

• 'I:.!,*

~ **t *.; * *::: -~ *.. . . . . .

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• • * . '1!'

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• .*. "< i 1.200 1.'**

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t R

o -~~~~~~~~~....--.......--r---.-~----~-...---.--..----r--.--.r-i.--.--,_.....:-r-....--.----.----.-r.-*

50 1 00 1 50 200 250 300 350 PCS DAgrees F

• Figure 3-4 LTOP Setpoint Limit 2500 2250 II 2000 1750

<(

ena.. .. )

1500 oi .

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Iv Q)

N .

• c:

s .

1000 ti)

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750 j

500

/ I

-~

250 0 . . .. . . . . . . ... . .. . . . . . ... . ...

s.o 1~o 150 200 250 300 350 400 450 PCS Temperature, F 3-25c .

Amendment No. 3+,

Basis 3.1.8.1 (continued)

Normally,. during operation at HOT STANDBY and above, the PORV controls are in the CLOSE position, and the block valves are closed. The PORVs, block valves, and the associated manual controls must be operable. If either valve in a PORV flow path is inoperable, the other valve in the flow path must provide PCS integrity assurance. When a PORV is inoperable, the block valve must be closed; when a block valve is inoperable, the PORV must have its control in the "CLOSE" position.

If the inoperable valves cannot be restored to OPERABLE status wit.hin the specified completion time, the plant must be placed in HOT SHUTDOWN. The completion times allow the required action to be accomplished without undue haste, yet allow less time when more equipment is inoperable.

Bi11:m1 3

• i. a. 2 When PCS is below 430°F with the reactor vessel head installed, two PORVs are required to be operable to avoid pressures which might lead to failure of the reactor vessel. Pressure increases could be caused by sudden additions (or imbalances) of either mass or energy.

The allowable pressure limits are determined in accordance with 10 CFR 50, Appendix G, and are referred to as "Low Temperature Overpressure Protection" (LTOP) limits. The variable setpoint of the LTOP system is programmed and calibrated to ensure opening of the pressurizer PORVs when the PCS pressure is above the limit in Figure 3-4. The pressure limit for each temperature is developed from the heating or cooling limits for the PCS.

The limit in Figure 3-4 includes an allowance for pressure overshoot during the interval between the time pressurizer pressure reaches the limit, and the time a PORV opens enough to terminate the pressure rise.

LTOP is provided by two independent channels each consisting of measurement, control, actuation, and valves. Either channel is capable of providing full protection. The actual setpoint of PORV actuation for LTOP will be below the limit in Figure 3-4 to allow for potential instrument inaccuracies, and drift.

This will ensure that at no time between calibration intervals will the PCS pressure exceed the limit of Figure 3-4 without PORV actuation.

Mass additions could come from the starting of pumps or from opening a Safety Injection Tank isolation valve. Only the charging pumps or high pressure safety injection pumps could cause the PCS pressure to ex9..~.g-~L. . i ts limits. Neither the shutoff head of the low pressure safety injection ~ijfilli nor the operating pressure of the .~..!!.fg_!y injection tanks is above the cold PtS.... jfressure limit. Specification 3.3.2.gl:S.MSMS places limits on HPSI pump operability when the PCS is below

-2W=lil PCS":*. .-.. . . .

0 r***1ci""*"assure inadvertent starting does not cause overpressuri zat ion of the 3-25e Amendment No. -l-l-7, +/-3-1-, +/-6G

Basis 3.1.8.2 (continued)

Energy additions could come from either the steam generators or from the reactor core. Small energy addition could come from operation of the pressurizer heaters.

Energy addition from the steam generators could occur if a primary coolant pump was started when the steam generator secondary temperature was significantly above the PCS temperature. Specification 3.1.1.h places limits on the starting of primary coolant pumps to avoid undesired energy additions from the steam generators. Energy addition from the reactor core could occur due to an inadvertent criticality or to an imbalance in decay heat removal. Specification 3.10.1 places limits on shutdown margin to avoid a rod withdrawal event causing a criticality and to provide sufficient time for operator action to terminate a dilution event prior to criticality.

The potential causes of a sudden PCS pressure increase which the LTOP system must be able to mitigate are im~~lance in charging and letdown flow, starting of the HPSI pumps when above ~jpj°F, and in an imbalance in decay heat (and pressurizer heat) addition and removaL.... A Safety InJ~.~.tion Signal (SIS) could both initiate flow from two HPSI pumps (when above ~IP:l°F) and three charging pumps, and isolate 1etdown. The PCS heatup from a ro*5*5* of shutdown cooling event occurring 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown from a continuous full power run would generate less additional coolant volume than the starting of three charging pumps (Reference 5).

The limiting event for the LTOP system would be an inadvertent SIS occurring during an established PCS heatup.

Analysis*(Reference 1) has concluded that an SIS occurring, during a PCS and pressurizer heatup at the maximum a11 ow.ilb..l..e rates, either between -26G:lii° F and 430 °F with the HPS I pumps, or below ~!2@° F without the HPS I pumps, wci"U"l d not cause PCS pressure to exceed the Appendix G limit if either PORV opens when the set pressure is reached. With the PCS above 430°F, the pressurizer safety valves, required by Specification 3.1.7, provide adequate overpressure protection. Both PORVs are required to be operable to allow for a single failure.

If a PORV becomes inoperable when it is required for LTOP, it must be restored to operable status, or the plant must be cooled down, depressurized, and vented through a vent path with sufficient capacity to provide the necessary protection.

Since the pressure response to a transient is greater if the pressurizer steam space is small or if PCS is solid, the allowed outage time for a PORV flow path out of service is shorter. The maximum pressurizer level at which credit can be taken for having a bubble (57%, which provides about 700 cubic feet of steam space) is based on judgement rather than on analyses. This level provides the same steam volume to dampen pressure transients as would be available at full power. This steam volume provides time for operator action, if the PORVs failed to operate, between an inadvertent SIS and PCS pressure reaching the 10 CFR 50 Appendix G pressure limit. The time available for action would depend upon the existing pressure and temperature when the inadvertent SIS occurred.

3-25f Amendment -l-l-7, -l-3-1-, &G,

Basis 3.1.8.2 (continued)

Reference 1 has determined that any vent path capable of relieving 167 gpm at a

• PCS pressure of 315 psia is acceptable. The 167 gpm flow rate is based on an assumed charging imbalance due to interruption of letdown flow with three charging pumps operating, a 40°F per hour PCS heatup rate, a 60°F per hour pressurizer heatup rate, and an initially depressurized and vented PCS. The PCS heatup rate is limited to 40°F per hour by Specification 3.1.2ai; the pressurizer heatup rate

!~a~;~~~=~~ t~o Fa~.~,HW~,~.u~i ~h. ~h:c ~~~

6 g; i;*

ci~ ~~~a~ ~:~res~~~1~:d, H~~~'~'~'~,~~~''~'~':,,,,,,,:,:.P

lfii;iii:~i:ilii!~::iil~!ili:f::ll§'=:~ffdr~~~~ ~i ~ n~ ~~o~~~ ~ r~~m~ ~o~~ b~i ~~ ~P~~:~! i":~~:~Pi.ll:~:t!:i 0

The pressure relieving ability of a vent path depends not only upon the area of the vent opening, but also upon the configuration of the piping connecting the vent opening to the PCS. A long, or restrictive piping connection may prevent a larger vent opening from providing adequate flow, while a smaller opening immediately adjacent to the PCS would be adequate. The areas of multiple vent paths cannot simply be added to determine the necessary vent area.

The following vent path examples are acceptable:

1. Removal of the reactor vessel head,

2. Removal of a steam generator primary manway,

3. Removal of the pressurizer manway,

4. Removal of a PORV or pressurizer safety valve,

5. Both PORVs and associated block valves open,

6. Opening of both PCS vent valves PC-514 and PC-515.

Reference 2 determined that venting the PCS through PC-514 and PC-515 provided adequate flow area. The other listed examples provide greater flow areas with less piping restriction and are therefore acceptable. Other vent paths shown to provide adequate capacity could also be used.

One open PORV provides sufficient flow area to prevent excessive PCS pressure.

However, if the PORVs are elected as the vent path, both valves must be used to meet the single failure criterion, since the PORVs are held open against spring pressure by energizing the operating solenoid.

When the shutdown cooling system is in service with M0-3015 and M0-3016 open, additional overpressure protection is provided by the relief valves on the shutdown cooling system. References 3 and 4 show that this relief capacity will prevent the PCS pressure from exceeding its pressure limits during any of the above mentioned events.

References I. Consumers Power Company Engineering An.~JY.~.t~.L . ~.f.i.... F..G.... ~.9..~ 13, Rev I ii.i~ER#.Mt::~:4.19:!S.EQ:l

2. Consumers Power Company Engineering An*nyrrs. ;*********~A::::rtil:*91-01-01

3. Consumers Power Company Engineering Analysis, EA-PAL-89-040-1

4. Consumers Power Company Corrective Action Document, A-PAL-91-011

5. Consumers Power Company Engineering Analysis, EA-AG-93-02

• 3-25g Amendment No. -l-l-7, -l-3-1-, GG,

3.3 EMERGENCY CORE COOLING SYSTEM (Continued)

g. HPSI pijmp operability shall be as follows:

1) If the reactor head is iRstalled, both HPSI pijmps shall be reRdered iRoperable wheR:

a. The PCS temperatijre is <260°F, or

b. ShijtdewR ceoliRg iselatioR valves MO 3915 aRd MO 3016 are

&pefh-

. 2) Twe HPSI pijmps shall be operable wheR the PCS te~peratijre is >

399°F.

3) 0Re HPSI pijmp may be made iRoperable wheR the reactor is Sijbcritical provided the reqijiremeRts of SectioR 3.3.2.c are met.

4) ~~~6 0 ~ij;~o!i~!~R~h:a~P~~ :~:;o::=~aYh~7s!~:r:;sv!~:~er;tijre is closed.

3.3.3 Prior to returning to the Power Operation Condition after every time the plant has been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and testing of Specification 4.3.h has not been accomplished in the previous 9 months, or prior to

• I returning the check valves in Table 4.3.1 to service after maintenance, I repair or replacement, the following conditions shall be met:

a. All pressure isolation valves listed in Table 4.3.1 shall be functional as a pressure isolation device, except as specified in

b. Valve leakage shall not exceed the amounts indicated.

b. In the event that integrity of any pressure isolation valve specified in Table 4.3.1 cannot be demonstrated, at least two valves in each high pressure line having a non-functional valve must be in and remain in, the mode corresponding to the isolated condition. m Motor-operated valves shall be placed in the closed position and power supplies deenergized.

1m1::~::1m:i11:Iitmw1:i1a§ltE:tirna!11::::1:b:11;1::::::::11:[1tii11~:11::~1;,:1:::::m1:1:i11:1:::1:11m1iii~1:11I:::1::!:::::::11~g:1:t~1::~::

3-30 Amendment No. ~J, )~J, ))7, .J..3-1., -l-6-l-,

3.3 EMERGENCY CORE COOLING SYSTEM Basis (continued)

• demonstrate that the maximum fuel clad temperatures that could occur over the break size spectrum are well below the melting temperature of zirconium (3300°F).

Malfunction of the Low Pressure Safety Injection Flow control valve could defeat the Low Pressure Injection feature of the ECCS; therefore, it is disabled in the 'open' mode (by isolating the air supply) durin~ plant operation. This action assures that it will not block f1ow during Safety Injection.

The inadvertent closing of any one of the Safety Injection bottle isolation valves in conjunction with a LOCA has not been analyzed. To provide assurance that this will not occur, these valves are electrically locked open by a key switch in the control room. In addition, prior to critical the valves are checked open, and then the 480 volt breakers are opened. Thus, a failure of a breaker and a switch are required for any of the valves to close.

!~~~rl~~ ~g~h t~~~!r~~~~~ i ~e <~§:ii~~l :r!iiif=l!l:iff!l!l!!~!Jiil!l!'=f=i!il!fi!flf:!l,j!j!!i!i!iil;§

• 1al'les are e~eA eliminates PCS m*irs*s* additions due to inadvertent HPSI pump starts. Bot HPSI pumps starting in conjunction with a charging/letdown imbalance may cause... .l.O.CFR50 Appendix G limits to be exceeded when the PCS temperature is <-26G:iQQ°F. WheA the PCS temperatl:Jre is ~ 430°F, the pressurizer safety Va:l:v.e.s. ....e..l'l.~.ij_re. ... t.h.a.t....t.h.e ... .P..C.S....@xe..s..s..li.r..e... J~.i.ll ...l'l.e.t.... e.x.c..e..e.d.......... . "'

• • 1 I

The requirement to have both HPSI trains operable above 300.!..F l.l!Sltll provides added assurance that the effects of a LOCA occurring under LTO}f'Eohditions would be mitigated. If a.,. .J.PCA occurs when the primary system temperature is 1ess than or equa 1 to a.GG.3:25.° F, the pressure would drop to the 1eve1 where 1ow pressure safety i nJ~.c.Yfon can preyg,.n._t core damage. Therefore, When the PCS temperature is ~~$.Qj!)° F and s300'i.:?.?.° F operation of the HPS I system would not cause the 10CFR50 A'Jl"p'ifodix G 1imft"f""to be exceeded nor is HPSI system operation necessary for core cooling.

HPSI fJl-JlftfJ testiAg with the HPSI f)l-JlftfJ maAlial discharge valve clesed is References ll FSAR, Section 9.10.3;

!2 FSAR, Section 6.1, 3 FSAR, Section 14.17 4 Letter, H.G.Shaw (ANF) to R.J.Gerling (CPCo), "Standard Review Plan Chapter 15 Disposition of Events Review for Changes to Technical Specifications Limits for Palisades Safety Injection Tank Liquid Levels", April 11, 1990.

Amendm8A!3No. l, .§.+/-, -!Gl, -l-l-7-, -l-3-l-, ~' +/-6-1-,

1m1::::::rn~w.1!:1:1:11~~~~~~:~1I1!!.~!!.~m~!~:t:1~!~~:1:1:1m~!:!~1

~11.!~~:~m1~~~:1~:~~~s:1.!'~~~?~:i:

In addition to the requirements of Specification 4.0.5, each PORV flow path shall be demonstrated OPERABLE by:

1. Testing the PORVs in accordance with the inservice inspection requirements for ASME Boiler and Pressure Vessel Code,Section XI, Section IWV, Category B valves.

2. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months.

3. When the PORV flow path is required to be OPERABLE by Specification 3.1.8.1:

(a} Performing a complete cycle of the PORV with the plant above COLD SHUTDOWN at least once per 18 months.

(b} Performing a complete cycle of the block valve prior to heatup from COLD SHUTDOWN, if not cycled within 92 days.

4. When the PORV flow path is required to be OPERABLE by Specification 3.1.8.2:

(a} Performance of a CHANNEL. FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, at least once per 31 days.

(b) Verifying the associated block valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

5.

!~~~~:!~~~:~t:r~~ff~~~~c~f~ t~g temperature is< ~P.QQ°F, or when both shutdown cooling suction valves, M0-3015 and Mtr::**3016, are open.

With the react.9r vessel head installed when the PCS cold leg temperature is less than ~$.!QQ°F, or if the shutdown cooling system isolation valves M0-3015 and M0-3-Cffb are open, the start of one HPSI pump could cause the Appendix G or the shutdown cooling system pressure limits to be exceeded; therefore, both pumps are rendered inoperable.

Amendment No. -l-3G, -149, -!6G, ~'

• ATTACHMENT 3 Consumers Power Company Palisades Plant Docket 50-255 PRESSURE/TEMPERATURE LIMIT TECHNICAL SPECIFICATIONS CHANGE REQUEST Engineering Analysis