Text

Proposed Pages to Tech Spec Change Request,Revising Accident & Transient Analyses Due to Steam Generator Tube Plugging
	ML18051A826
Person / Time
Site:	Palisades
Issue date:	03/16/1984
From:	; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	;
Shared Package
ML18051A825	List: ML18051A824; ML18051A826; ML20087J816; ML20087J819; ML20087J831;
References
	NUDOCS 8403230098
	Download: ML18051A826 (41)
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CONSUMERS POWER COMPANY PALISADES PLANT - DOCKET 50-255 - LICENSE DPR-20 PROPOSED TECHNICAL SPECIFICATIONS CHANGE REQUEST PROPOSED PAGES 38 Pages

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1 8403230098 840316 1

PDR ADOCK 05000255 P

PDR TS0384-0008B-NL02

2 D.

Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

3.

This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations:

10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50 and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A.

Maximum Power Levels B.

c.

The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2125.2 megawatts thermal (100% of the rated power level of the facility).

Technical Specifications The Technical Specifications contained in Appendices A and B (Environmental Protection Plan), as revised through Amendment No 77, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications, except as modified for Cycle 5 only by Paragraph 3.J of this license.

Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications and the Interim Special Technical Specifications.

D.

Records The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.

E.

The licensee may proceed with and is required to complete the modifications identified in Paragraphs 3.1.1 through 3.1.23 of the NRC's Fire Protection Safety Evaluation (SE) on the facility dated September 1, 1978, Supplement No 1 to the SE dated March 19, 1980, and Supplement No 2 to the SE dated February 10, 1981.

These modifications shall be completed as specified in Table 3.1 of the SE, as supplemented, in accordance with the schedule contained therein.

PROPOSED TS0384-0005A-NL02

TECHNICAL SPECIFICATIONS 1.0 DEFINITIONS 1.1 The following terms are defined for uniform interpretation of these Technical Specifications:

REACTOR OPERATING CONDITIONS Rated Power A steady state reactor core output of 2125.2 MWt.

Reactor Critical The reactor is considered critical for purpose of administrative control when the neutron flux logarithmitic range channel instrumentation indicates greater than 10-4% of rated power.

Power Operation Condition When the reactor is critical and the neutron flux power range instrumentation indicates greater than 2% of rated power.

Hot Standby Condition The reactor is considered to be in a hot standby condition if the average temperature of the primary coolant (T

) is greater than avg 525°F and any of the control rods are withdrawn and the neutron flux power range instrumentation indicates less than 2% of rated power

Hot Shutdown Condition When the reactor is subcritical by an amount greater than or equal to the margin as specified in Technical Specification 3.10 and T is avg greater than 525°F.

Refueling Shutdown Condition When the primary coolant is at refueling boron concentration and T avg is less than 210°F.

Cold Shutdown Condition When the primary coolant is at shutdown boron concentration and T avg is less than 210°F.

Refueling Operation Any operation involving movement of core components when the vessel head is unbolted or removed.

Shutdown Margin Shutdown margin shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming that all full-length control rods are fully inserted except for the single highest worth control rod which is assumed to be withdrawn.

1-1 PROPOSED TS0384-0006H-NL02

1.1 REACTOR OPERATING CONDITIONS (Continued)

Axial Off set Axial Offset equals the power in the lower half of the core minus the power in the upper half of the core divided by the sum of the powers in the lower half and upper half of the core

1-2a PROPOSED TS0384-0006Y-NL02

2.0 2.1 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS - REACTOR CORE Applicability This specification applies to the limiting combinations of reactor power, primary coolant system flow, temperature and pressure during operation.

Objective To maintain the integrity of the fuel cladding and prevent the release of significant amounts of fission products to the primary coolant.

Specifications The reactor power level shall not exceed the allowable limit for the pressurizer pressure and the cold leg temperatures (shown in Figure 2-1), for 4-pump operation.

The safety limit is exceeded if the point defined by the combination of primary coolant cold leg temperature and power level is at any time above the appropriate pressurizer pressure line.

The reactor power shall not exceed the allowable limits in Table 2.3.1 for 3-pump operation.

Basis To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under normal operation conditions.

This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature.

The upper boundary of the nucleate boiling regime is termed "departure from nucleate boiling" (DNB).

At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high-cladding temperatures and the possibility of cladding failure.

Although DNB is not an observable parameter during reactor operation, the observable parameters of thermal power, primary coolant flow, temperature and pressure, can ~I)related to DNB through the use of the "XNB Correlation."

The XNB Correlation has been developed to predict DNB and the location of DNB for axially uniform and nonuniform heat flux distributions.

The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.17.

A DNBR of 1.17 corresponds to a 95%

2-1 PROPOSED TS0384-0006G-NL02

2.1 SAFETY LIMITS - REACTOR CORE (Continued) probability at a 95% confidence level than DNB will not occur which is considered an appropriate margin to DNB for all operating conditions

The curves of Figure 2-1, represent the loci of points of thermal power, primary coolant system pressure and inlet temperature of 4-pump operation for which the DNBR is~ 1.17.

The area of safe operation is below these lines.

For 3-pump operation, the operating limits presented in Table 2.3.1 were selected at sufficiently conservative values to ensure that DNB and hot leg saturation would not occur during steady-state operations, normal operational transients, and anticipated transients.

The core inlet temperature maldistribution due to a 3-pump operations will cause a significant core power tilt due to moderator temperature coefficient feedback.

Therefore, core power distribution monitoring would be more difficult.

Power operations with less than three primary coolant pumps is not allowed.

The reactor protection system is designed to prevent any anticipated combination of transient conditions for primary coolant system temperature, pressure and thermal power level that would result in a DNBR of less than 1.17. <2>

References (1)

(2)

XN-NF-621(P)

XN-NF-84-14 TS0384-0006A-NL02 2-2 PROPOSED f

2.2 SAFETY LIMITS - PRIMARY COOLANT SYSTEM PRESSURE Applicability Applies to the limit on primary coolant system pressure.

Objective To maintain the integrity of the primary coolant system and to prevent the release of significant amounts of fission product activity to the primary coolant.

Specification The primary coolant system pressure shall not exceed 2750 psia when there are fuel assemblies in the reactor vessel.

Basis The primary coolant system (l) serves as a barrier to prevent radionuclides in the primary coolant from reaching the atmosphere. In the event of a fuel cladding failure, the primary coolant system is the foremost barrier against the release of fission products.

Establishing a system pressure limit helps to assure the continued integrity of both the primary coolant system and the fuel cladding.

The maximum transient pressure allowable in the primary coolant system pressure vessel under the ASME Code,Section III, is 110% of design pressure.

The maximum transient pressure allowable in the primary coolant system piping, valves and fittings under ASA Section B31.1 is 120% of design pressure.

Thus, the safety limit of 2750 psia (110% of the 2500 psia design pressure) has been established. (2)

The settings and capacity of the secondary coolant system safety valves (985-1025 psig)(3), the reactor high-pressure trip(~ 2400 psia) and the primary safety valves (2500-2580 psia)(4) have been established to assure never reaching the primary coolant system pressure safety limit.

The initial hydrostatic test was conducted at 3125 psia (125% of design pressure) to verify the integrity of the primary coolant system.

Additional assurance that the nuclear steam supply system (NSSS) pressure does not exceed the safety limit is provided by setting the pressurizer power-operated relief valves at 2400 psia and the secondary coolant system steam dump and bypass valves References (1)

FSAR, Section 4.

(2)

FSAR, Section 4.3.

(3)

FSAR, Section 4.3.4.

(4)

FSAR, Section 4.3.9.

2-3 TS0384-0006N-NL02 at 900 psia.

Amendment No 25 March 11, 1977

2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM Applicability This specification applies to reactor trip settings and bypasses for instrument channels.

Objective To provide for automatic protective action in the event that the principal process variables approach a.safety limit.

Specification The reactor protective system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3.1.

2-4 TS0384-00060-NL02

I\\)

I Vl

1.

2.

3.

4.

5.

6.

7.

(1)

(2)

(3)

(4)

High Power Level(l)

Low Primary 2

Coolant Flow( )

High Pressurizer Thermal Mtzgi~1Low Pressure Low Steam Generator Water Level Low Steam Generator Containment High Pressure TAB

.1 Reactor Protective System Trip Setting Limits Four Primary Coolant Pumps Operating

~ 106.5% of Rated Power

~ 95% of Primary Coolant Flow With Four Pumps Operating

2255 Psia PT ~ Applicable Limits To Satisfy Figure 2-1 Not Lower Than the Center Line of Feed-Water Ring Which is Located 6'-0" Below Normal Water Level

~ 500 Psia

~ 5 Psig Three Primary Coolant Pumps Operating

~ 39% of Rated ~owerC 4

)

(Continuous Operation Not Permitted)

~ 71% of Primary Coolant Flow With Four Pumps Operating

2255 Psia Replaced by High Power Level Trip and 1750 Psia Minimum Low-Pressure Setting Not Lower Than the Center Line of Feed-Water Ring Which is Located 6'-0" Below Normal Water Level

~ 500 Psia

5 Psig Below 5% rated power, the trip setting may be manually reduced by a factor of 10.

Two Primary Coolant Pumps Operating (No Two-Pump Operation)

-4 May be bypassed below 10 % of rated power provided auto bypass removal circuitry is operable.

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

Th and TC in °F.

Minimum trip setting shall be 1750 psia for three-pump operation.

For four-pump operation, the minimum trip setting shall be 1650 psia for nominal operating pressures less than 1900 psia; and 1750 psia for nominal operating pressures 1900 psia and greater.

Operation with three pumps 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.

2.3 LIMITING SAFETY SYSTEM SETTINGS -*REACTOR PRPTECTIVE SYSTEM (Continued)

Basis 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.

High Power Level - A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding resulting from some reactivity excursions too rapid to be detected by pressure and temperature measurements. 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 actuatedcf' 112%, which was used for the purpose of.

safety analysis.

Provisions have been made to select a different high power level trip points for three primary coolant Plllf§)operation as described

below under "Low Primary Coolant Flow."

If reactor operation at less than 10% of rated power is required for an extended period of time, provisions have been made to allow the operator to decrease the indicated power range by a factor of 10, which will also decrease the high power level t2~P point by a factor of 10 to 10.65% of indicated rated power.

Administrative procedures will allow this range change to be made during reactor start-up and also between 5% and 8% of rated power when the reactor power is reduced to that level.

2..

Low Primary Coolant Flow - A reactor trip is provided to protect the core against DNB should the coolant flow suddenly decrease significantly. Provisions are made in the reactor protective system to permit operation of the reactor at reduced power if one coolant pump is taken out of service.

These low-flow and high-flux settings have been derived in consideration of instrument errors and response times of equipment involved to assure that MDNBR ~ 1.17(gnd 1~tow stability will be maintai~~j during normal operation and anticipated transients.

For reactor operation with one coolant pump inoperative. the low-flow trip points and the overpower trip points must be manually changed to the specified values by means of a set point 2-6 PROPOSED TS0384-0006I-NL02

2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Continued)

Basis (Continued) selector switch.

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.

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 nominal value considering instrument errors. (3)

The high power level trip and low primary coolant flow trip are reduced to compensate for the corresponding core flow reduction experienced with three pumps in operation.

The trip points are shown in Table 2.3.1.

3.

High Pressurizer Pressure - A reactor trip for high pressurizer pressure is provided in conjunction with the primary and secondary safety valves to prevent primary system overpressure (Specification 3.1.7).

In the event of loss of load without reactor trip, the*

temperature and pressure of the primary coolant system would increase due to the reduction in the heat removed from the coolant via the steam generators.

The power-operated relief valves are set to operate concurrently with the high pressurizer pressure reactor trip.

This setting is at least 100 psi below the nominal safety valve setting (2500 psia) to avoid unnecessary operation of the safety valves.

This setting is consistent with the trip point assumed in the accident analysis. (ll)

4.

Thermal Margin/Low-Pressure Trip - A reactor trip is provided to prevent operation with an MDNBR of less than 1.17 or hot leg at saturation conditions.

The thermal and hydraulic safety limits shown on Figure 2-1, for four primary coolant pump operation defines the limiting values of primary coolant pressure, reactor inlet temperature, and core power level for which the criteria on MDNBR are met.

A thermal margin/low-pressure 2-7 PROPOSED TS0384-0006F-NL02 r

2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Continued)

Basis (Continued)

(TM/LP) trip will occur before these limits are reached. Reference 7 forms the basis for Figure 2-1 for 4-pump operation.

The trip is initiated whenever the pressurizer pressure drops below the minimum value given on Table 2.3.1, or a value computed as described below, whichever is higher.

The computed value is a function of reactor inlet temperature and reactor outlet temperature, and takes the form PT i

= ATH - BTC-C where A, B, and C are constants and TH and T are f nE measured hot and cold leg coolant temperatures, respectively.

The minimum value of reactor coolant flow and the maximum expected values of axial and radial peaking factors are assumed in generating this trip function.

The TM/LP trip set points are derived from the 4-pump operation core thermal limits (Figure 2-1) through application of appropriate allowances for measurement uncertainties and processing errors.

A maximum error of 165 psi is assumed to account for expected instrument drift and repeatability errors, process measurement uncertainties, flow stratification effects, and calibration errors.

As such, a maximum error in the calculated ~I~)point of -165 psi has been assumed in the accident analysis.

For three coolant pump operation, power is limited to 39% of rated power for a maximum of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The conservative high power level trip in conjunction with the TM/LP trip (minimum set point =

1750 psia) assure that the fuel thermal margin limits will not be violated.

5.

Low Steam Generation Water Level -

The low steam generation water level reactor trip protects against the loss of feed-water flow accidents and assures that the design pressure of the primary coolant system will not be exceeded.

The specified set point assures that there will be sufficient water inventory in the steam generator at the time of trip to provt§j a 15-minute margin before the auxiliary feedwater is required

2-8 PROPOSED TS0384-0006E-NL02

2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Continued)

Basis (Continued)

The setting listed in Table 2.3.1 assures that the heat transfer surface (tubes) is covered with water when the reactor is critical.

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 psi is sufficiently below the rated load operating point so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively(gf,g) steam flow.

The accident analysis assumes 500 psi or less.

7.

Containment High Pressure - A reactor trip on containment high pressure is provided to assure that the reactor is shut down upon the initiation of the safety injection system.

8.

Low Power Physics Testing - For low power physics tests, certain tests will require the reactor to be critical at low temperature

(~ 260°F) and low pressure (~ 415 psia).

For these certain tests only, the thermal margin/low pressure, 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 appro~Id written testing procedures.

At reactor power levels below 10

% of rated power, the thermal margin/low-pressure trip is 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)

FSAR, Section 4.1.

(2)

FSAR, Section 7.2.3.2.

(3) 'FSAR, Section 7.2.3.3.

(4)

XN-NF-84-18 (5)

FSAR, Section 3.3.3.

(6)

XN-NF-84-18 Section 3.0 (7)

XN-NF-84-14, Section 3 2-9 PROPOSED TS0384-0006J-NL02

0 2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Continued)

References (Continued)

(8)

XN-NF-77-18, Section 3.8 (9)

XN-NF-84-18, Section 3.3.2 (10)

FSAR, Amendment No 17, Item 4.0 (11)

XN-NF-84-18, Table 2.1 (12)

XN-NF-84-18, Section 2.2 (13)

XN-NF-84-14, Section 2.0 2-10 TS0384-0006X-NL02 PROPOSED

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l. 2 fRAGTION Of 2125.2 MHT FIGURE. 2~1 BOUNDING SAFETY LIMIT LINES INCLUDING UNCERTAINTIES I\\)

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FIGURE 2-2 DELETED 2-12 PROPOSED OC0384-0006Y-NL02

FIGURE 2-3 DELETED 2-13 PROPOSED TS0384-0007B-NL02

3.0

3. 1 LIMITING CONDITIONS FOR OPERATION PRIMARY COOLANT SYSTEM Applicability Applies to the operable status of the primary coolant system.

Objective To specify certain conditions of the primary coolant system which must be met to assure safe reactor operation.

Specifications 3.1.1.

Operable Components

a.

b.

c.

At least one primary coolant pump or one shutdown cooling pump shall be in operation whenever a change is being made in the boron concentration of the primary coolant.

Four primary coolant pumps shall be in operation whenever the reactor is operated continually above 5% of rated power (exception to this specification is permitted as described in Table 2.3.1, Item 1).

The minimum flow for various power levels shall be as shown in Table 2.3.1.

The measured four primary coolant pumps operating reactor vessel flow (as determined by reactor coolant pump dif f eregtial pressures and pump performance curves) shall be 101.0 x 10 lb/h or greater, when corrected to 532°F.

In the event the measured flow is less than that required above, the limits specified on Figure 2-1 shall be reduced by 1.5°F in inlet temperature for each 1% of reactor flow deficiency.

Continuous operation at power shall be limited to four-pump operation.

With one pump out of service, return the pump to service (return to four-pump operation) within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in hot standby (or below) within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Start-up (above hot standby) with less than four pumps is not permitted.

d.

Both steam generators shall be capable of performing their heat transfer function whenever the average temperature of the primary coolant is above 325°F.

e.

Maximum primary system pressure differentials shall not exceed the following:

(1)

Maximum steam generator operating transient differential of 1534 psi.

3-1 PROPOSED TS0384-0006K-NL02

3.1 3.1.1 PRIMARY COOLANT SYSTEM (Continued)

Operable Components (Continued)

(2)

Hydrostatic tests shall be conducted in accordance with applicable paragraphs of Section XI ASME Boiler & Pressure Vessel Code (1974).

Such tests shall be conducted with sufficient pressure on the secondary side of the steam generators to restrict primary to secondary pressure differential to a maximum of 1380 psi. Maximum hydrostatic test pressure shall not exceed 1.1 Po plus 50 psi where Po is nominal operating pressure.

(3)

Primary side leak tests shall be conducted at normal operating pressure.

The temperature shall be consistent with applicable fracture toughness criteria for ferritic materials and shall be selected such that the differential pressure across the steam generator tubes is not greater than 1380 psi.

(4)

Maximum secondary hydrostatic test pressure shall not exceed 1250 psia.

A minimum temperature of 100°F is required.

Only ten cycles are permitted.

(5)

Maximum secondary leak test pressure shall not exceed 1000

psia.

A minimum temperature of 100°F is required.

(6)

In performing the tests identified in 3.1.1.e(4) and 3.l.l.e(5), above, the secondary pressure shall not exceed the primary pressure by more than 350 psi.

f.

Nominal primary system operating pressure shall not exceed 1990 psia or be less than 1910 psia.

g.

The reactor inlet temperature (indicated) shall not exceed the value given by the following equation at steady state 100% power operation:

Tinlet ~ 548.4 + (P-i970.3) [0.04 + 0.00015 (P-1970.3)] + 1.27 (W-97.6)

Where:

Tinlet = reactor inlet temperature in °F.

P = nominal operating pressure in psia.

W = total recirculating mass flow in 106 lb/h corrected to the operating temperature conditions.

h.

A reactor coolant pump shall not be started with one or more of the PCS cold leg temperature ~ 250°F unless 1) the pressurizer water volume is less than 700 cubic feet or 2) the secondary water temperature of each steam generator is less than 70°F above each of the PCS cold leg temperatures.

3-la PROPOSED TS0384-0006L-NL02

3.1 PRIMARY COOLANT SYSTEM (Continued)

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(y~e shutdown cooling or one primary coolant pump is in operation.

The shutdown cooling pump will circulate the primary system volume in less than 60 minutes when operated at rated capacity.

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(2) pressurizer and the primary system during the addition of boron.

Both steam generators are required to be operable whenever the temperature of the primary coolant is greater than the design temperature of the shutdown cooling system to assure a redundant heat removal system for the reactor.

Calculations have been performed to demonstrate that a pressure differential of 1380 psi can be withstood by a tube uniformly thinned to 36% of its original nominal wall thickness (64% degradation), while maintaining:

(1) A factor of safety of.three between the actual pressure differential.and the pressure differential required to cause bursting.

(2) Stresses within the yield stress for Inconel 600 at operating temperatures.

(3) Acceptable stresses during accident conditions.

The maximum transient steam generator differential pressure is expected to occur during the excess load transient. The excess load transient initiated from hot full power operating conditions and assuming a high flux trip of 2380 MWt is analyzed in Reference 3.

Results of this analysis indicate that the maximum steam generator differential pressure is less than 1534 psi.

The 1534 psi limit on transient pressure differential is approximately 11% greater than that 3-2 PROPOSED TS0384-0006M-NL02

3.1 PRIMARY COOLANT SYSTEM (Continued) allowed during normal operation, so that substantial safety margin exists between this pressure differential and the pressure differential required for tube rupture

Secondary side hydrostatic and leak testing requirements are consistent with ASME BPV Section XI

The differential maintains stresses in the steam generator tube walls within code allowable stresses.

The minimum temperature of 100°F for pressurizing the steam generator secondary side is set by the NDTT of the manway cover of + 40°F.

The limiting transient was the control rod drop from 100% power as shown in Reference 4.

These transient6analyses were performed assuming a vessel flow at full power of 99 x 10 lb/h minus 3% to account for flow measurement uncertainty.

A 50 psi allowance on pressure was added to account for the control band inQsteady state operations and the inlet temperature was increased by 7°F to account for cold leg temperature measurements and asymmetries in the core inlet due to asymmetric plug-ging of the steam generator.

In addition to the measurement and control uncertainties, three changes due to the transient were included.

These included a reduction of pressurizer pressure of 20.3 psi, an increase in flow, due to an increase in the cold leg water density, of 1.38 Mlb/h and 8°F reducti~~)in cold leg temperature at the time when minimum DNB is calculated.

Fitting a functional form to the limiting transient analysis data results in the following equation for limiting reactor inlet temperature:

Inlet ~ 548.4 + (P-1970.3) [0.04 + 0.00015 (P-1970.3)] + 1.27 (W-97.6)

The nominal full power inlet temperature is set 2°F less than the value given in Section 3.1.1.g to allow for drift within the temperature control band.

The limits of validity of this equation are:

1910 ~ Pressure ~ 1990 psia 93 X 10 6~ Vessel Flow = 112 X 10 6~ lb/hr 6

If the vessel flo~ is determined to be greater than 112 X 10 lb/hr, a value of 112 X 10 lb/hr shall be used to determine the limiting reactor inlet temperature.

The restrictions on starting a Reactor Coolant Pump with one or more PCS cold legs ~ 250°F are provided to prevent PCS pressure transients, caused by energy additions from the secondard systems, which would exceed the limits of Appendix G to 10 CFR Part 50.

The PCS will be protected against over-pressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam gentsf tor is less than 70°F above each of the PCS cold leg temperatures.

References (1)

FSAR, Sections 6.1.2.2 and 14.3.2.

(3)

XN-NF-84-18.

(2)

FSAR, Section 4.3.7.

(4)

XN-NF-84-14 (5)

"Palisades Plant Overpressurization Analysis," June, 1977, and "Palisades Plant Primary Coolant System Overpressurization Subsystem Description," October, 1977.

TS0384-0006B-NL02 3-3 PROPOSED

FIGURE 3-0 DELETED 3-3a PROPOSED OC0384-0003L-NL02

3.11 POWER DISTRIBUTION INSTRUMENTATION 3.11.1 INCORE DETECTORS LIMITING CONDITION FOR OPERATION The incore detection system shall be operable:

a.

With at least 50% of the incore detectors and 2 incores per axial level per core quadrant.

b.

With the incore alarming function of the datalogger operable and alarm set points entered into the datalogger.

APPLICABILITY (1) Item a. above is applicable when the incore detection system is used for:

Measuring quadrant power tilt, Measuring radial peaking factors, Measuring linear heat rate (LHR), or Determining target Axial Offset (AO) and excore monitoring allowable power level.

(2) Items a. and b. above are applicable when the incore detection system is used for monitoring LHR with automatic alarms.

(Incore Alarm System.)

ACTION 1:

With less than the required number of incore detectors, do not use the system for the measuring and calibration functions under (1) above.

ACTION 2:

With the alarming function of the datalogger inoperable, do not use the system for automatic monitoring of LHR (Inoperable Incore Alarm System).

3-65 TS0384-0006P-NL02 Amendment No 68 December 8, 1981

3.11 POWER DISTRIBUTION INSTRUMENTATION 3.11.1 INCORE DETECTORS LIMITING CONDITION FOR OPERATION ACTION 2:

(Continued)

Operation may continue using the excore monitoring system as specified in 3.11.2 or by meeting the requirements of 3.23.1.

Basis The operability of the incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial nuetron flux distribution of the reactor core.

The operability of the incore alarm system depends on the availability of the datalogger as well as the operability of a minimum number of incore detectors.

Incore alarm set points must be updated periodically based on measured power distributions.

The incore detector Channel Check is normally performed by an off-line computer program that correlates readings with one another and with computed power shapes in order to identify inoperable detectors.

3-66 TS0384-0006Q-NL02 Amendment No 68 December 8, 1981

3.11 POWER DISTRIBUTION INSTRUMENTATION 3.11.2 EXCORE POWER DISTRIBUTION MONITORING SYSTEM LllHTING CONDITION FOR OPERATION The excore monitoring system shall be operable with:

a.

The target Axial Offset (AO) and the Excore Monitoring Allowable Power Level (APL) determined within the previous 31 days using the incore detectors, and the measured AO not deviated from the target AO by more than 0.05 in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.. The AO measured by the excore detectors calibrated with the AO measured by the incore detectors.

c.

The quadrant tilt measured by the excore detectors calibrated with the quadrant tilt measured by the incore detectors.

APPLICABILITY:

(1)

Items a., b. and c. above are applicable when the excore detectors are used for monitoring LHR.

(2)

Item c. above is applicable when the excore detectors are used for monitoring quadrant tilt.

ACTION 1:

With the excore monitoring system inoperable, do not use the system for

-~

monitoring LHR.

ACTION 2:

If the measured quadrant tiit has not been calibrated with the incores, do not use the system for monitoring quadrant tilt.

Basis The excore power di~tribution monitoring system consists of Power Range Detector Channels 5 through 8.

The operability of the excore monitoring system ensures that the assumptions employed in the PDC-II analysis (l) for determining.AO limits that ensure operation within allowable LHR limits are valid.

3-66a PROPOSED TS0384-0006R-Nt02

3.11 POWER DISTRIBUTION INSTRUMENTATION 3.11.2 EXCORE POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION Basis (Continued)

Surveillance requirements ensure that the instruments are calibrated to agree with the incore measurements and that the target AO is based on the current operating conditions.

Updating the Excore Monitoring APL ensures that the core LHR limits are protected with the +/- 0.05 band on AO.

The APL considers both LOCA and DNB based LHR limits, and factors are included to account for changes in radial power shape and LHR limits over the calibration interval.

The APL is determined from the following:

LHR(Z)TS APL = --------------- ] Min x Rated Power Where:

LHR(Z) Max x V(Z) x 1.02 (1)

LHR(Z)TS is the limiting LHR vs Core Height (from Section 3.23.1),

(2)

LHR(Z)Max is the measured peak LHR including uncertainties vs Core Height, (3)

V(Z) is the function (shown in Figure 3.11-1),

3-66b PROPOSED TS0384-0006S-NL02

3.11 POWER DISTRIBUTION INSTRUMENTATION 3.11.2 Excore Power Distribution Monitoring System LIMITING CONDITION FOR OPERATION Basis (Continued)

(4)

The factor of 1.02 is an allowance for the effects of upburn, (5)

The quantity in brackets is the minimum value for the entire core at any elevation (excluding the top and bottom 10% of core) considering limits for peak rods and interior fuel rods.

If the quantity in brackets is greater than one, the APL shall be the rated power level.

Reference (1)

XN-NF-80-4 7 3.11.3 Excore Axial Offset Monitoring System The Axial Offset (AO) measured by the excore monitoring system shall be within the range of -.07 to +.12.

The excore AO shall be calibrated with the incore AO for the purpose of this measurement.

APPLICABILITY

1)

Power operation above 40% of rated power after completing the calibration required by 4.18.3.1.

2)

The most limiting requirements of section 3.11.2 and this section shall apply when the excores are used to monitor LHR limits.

Action 1:

If the measured AO falls outside the allowed range of values, bring the AO within the allowed range within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or failing to do this, be at less than 40% of rated power within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Action 2:

If the excore AO monitoring system becomes inoperable, at operating power level less than 90%, adjust the high power trip down to within 10% (of rated power) of the operating power level. If the high power trip cannot be adjusted within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the operating power level shall be limited to 40%

_ rated power within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Basis The limitation upon AO assures that the core axial power profile is within the assumed core axial power sha~I' for the DNB analysis of anticipated transients protected by the TM/LP trip References (1)

XN-NF-84-26 3-66c PROPOSED TS0384-0006C-NL02

3.23 POWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE (LHR)

LIMITING CONDITION FOR OPERATION The LHR in the peak power fuel rod at the peak power elevation Z shall not exceed the value in Table 3.23-1 times FA(Z) [the function FA(Z) is shown in Figure 3.23-1]. The LHR at the peak power elevation in any interior fuel rod shall not exceed the value in Table 3.23-1 times Applicability:

Power Operation above 40% of rated power.

ACTION 1:

When using the incore alarm system to monitor LHR, and with four or more coincident incore alarms, to reduce the LHR to within the limits and restore the incore readings to less than the alarm setpoints within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or failing this, be at less than 40% of rated power within the following 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

ACTION 2:

When using the excore monitoring system to monitor LHR and with the AO deviating from the target AO by more than 0.05, discontinue using the excore monitoring system for monitoring LHR.

If the incore alarm system is inoperable, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months be at 85% (or less) of rated thermal power and follow the procedure in ACTION 3 below

3-103 PROPOSED TS0384-0006T-NL02

3.23 POWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE (LHR)

LIMITING CONDITION FOR OPERATION ACTION 3:

If the incore alarm system is inoperable and the excore monitoring system is not being used to monitor LHR, operation at less that or equal to 85% of rated power may continue provided that incore readings are recorded manually.

Readings shall be taken on a minimum of 10 individual detectors per quadrant (to include 50% of the total number of detectors in a 10-hour period) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and at least every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thereafter. If readings indicate a local power level equal to or greater than *the alarm setpoints, the action specified in ACTION 1 above shall be taken.

Basis The limitation on LHR ensures that, in the event of a LOCA, the peak temperature of the cladding will not exceed 2200°F.(l)

In addition, the limi~*ation on LHR for the interior fuel rod ensures that the minimum DNBR will be maintained above 1.17 during anticipated transients; and, that fuel damage during Condition IV events such as locked rotor wil~ not exceed acceptable limits. (Z)(J)

The inclusion of the axial power distribution tenn ensures t~at the operating power distribution is enveloped by the design power distributions.

Either of the two core power distribution monitoring systems (the incore alarm system or the excore monitoring system) provides adequate monitoring of the core power distribution and is cap~ble of verifying that the LHR does not*exceed its limits.

The incore alarm system performs this 3-104 PROPOSED TS0384-0006U-NL02

\\.

POW'ER DISTRIBtrrION L!~!TS J.23.l LI?-."'EAR H!AT RATE: (!.HR)

LIMITING CD~IT!ON ~OR OPERATION Basis (Contd) function by con~i:uously Qonitoring the local power at many points th=oughout the c~=e and c~mparins the measurements to predetermined setpoi.nts above which tbe limit en I.HR could be exceeded. The excore monitoring system perfcr.:s this function by providing cocparison of ~e measured core AO with prede'te:rmined AO l:ii:its based en incore

. :easurements.

An Excore Monitoring Allowable Power Level (APL), t.:hich may be less than :rated power, is appl.ied when using 'the ex.core moni-:oring system to ensure that the AO U.mits adequately :restrict the I.HR to. less than the limiting values.

(4)

If the incore ala::m system and the ~cc=e mo~!~cri!lg system are"both i:loperable, pow;er will be reduced to provide cargi.n bet:we~ the actuz.l.

peak I.KR and the *r..-s licits and the ~core readings will' be msnully collected at the ter.:inal blocks in the control room util:i.%ing a suitahle sig::.al detector. If *this is not feasible with the manpower available. the reactor power will be_ reduced to a point belcn..- which it is improbable that the I.KR limits could be exceeded.

The* time :interval *of. 2 hou:s. and the ci:lil:um of 10 detectc=s per quadrant are ~uff icient to maintai.:i adequate su:veille..:ice of the core power distribution to detect sig:nificm:it ch~ges until the monitcring syste:s ~re retU-"'lled to service.

To ensure :hat the design =argin cf safety is maintained, the determina:io:c. of both :he inc:ore alarm setpoinu and the APL takes b-:o accoun~ a measuremen: uncertainty fac:or of 1.10. an engi.neeril::lg 3-105

.Amendment No 68 December 8, 1981

POWER DISTRIBUTION LIMITS 3.23.l LINEAR HEAT RATES (LHR)

LIMITING CONDITIONS OF OPERATION Basis (Continued) uncertainty factor of 1.03, a thermal power measurement uncertainty factor of 1.02 and allowance for quadrant tilt.

References (1)

XN-NF-84-24 (2)

XN-NF-84-18 (3)

XN-NF-84-14 (4)

XN-NF-80-47 TS0384-0007E-NL02 3-106 PROPOSED

TABLE 3.23-1 Lih"tAR HEAT RATE LH1ITS Fuel Rod Type No of Fuel Rods 208 Peak Rod 12,84 InteriOr _Rod 12.84 TABLE 3.23-2 in Assembly 216 12.84 12.84 RADIAL PEAKil\\G FACTOR LnlITS, FL Peaking Fai:o;o:-

No of Fuel Rods in Assembly 208 216 Assembly F'

l,45 1,45 r

Peak Rod FT r

1.77 1,84 Ini;erior Rod

..AH 1.64 le70 rr 3-107 PROPOSED

w I

0 en

~

3 0-9 -*

x =-x x -<

x u..

0 0.f) -

x C>

I-0

~

u.. -

CY. :s o. 7 -

U.I

...J IO

~ 0.65 -

ACCErTADLE OPERATIOtl DREAK.POIHTS:

I.. (

II!), I

0)

2. (. 7 ~

91 )

3. (l.O,.67}

C>

~

fL~~-1*~~~il ____ ---1IL.._ __ ~~*~~~Jl_~-----'I------..&-~---'-----__..----__.

O.I 0,2 0.3 O.ij Oi6 o *. G 0.1 0.8 LOCATION OF AXIAL POWER PEAK (FRACTION OF ACTIVE FUEL llEIOllT)

ALLOWADLE UIR AS A FUHCTIOH OF PEAK POWER LOCATION.

Pallsados Technical SpoQlflcallons o.o

1. 0 f IQURE 3.23"1

. *----.-.... *--------~*..:__ _______. ______ _

FIGURE 3.23-2 DELETED 3-109 PROPOSED TS0384-0006Z-NL02

FIGURE 3.23-3 DELETED 3-110 PROPOSED TS0384-0006ZZ-NL02

3.23 POWER DISTRIBUTION LIMITS 3.23.2 RADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION The radial peaking factors ~' F;, F6~ shall be less than or equal to the value in Table 3.23-2 times the following quantity.

The quantity is [1.0 + 0.3 (1 - P)] for P ~.5 and the quantity is 1.15 for P <.5.

P is the core thermal power in fraction of rated power.

Applicability:

Power operation above 40% of rated power.

ACTION:

With any radial peaking factor exceeding its limit within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , reduce thermal power to less than the lowest value of:

(1 - 3.33 (

F r 1 )

x Rated Power Where F is the measured value of either FA r

r the corresponding limit from Table 3.23-2.

Basis A

T 6 H The limitations on Fr* Fr and F r are provided to ensure that assumptions used in the analysis for establishing DNB margin, LHR and the thermal margin/low-pressure and high-power trip set points remain valid during operation.

Data from the incore detectors are used for determining the measured radial peaking factors.

The periodic surveillance requirements for determining the measured radial peaking factors provide assurance that they remain within prescribed limits.

Determining the measured radial peaking factors after each fuel loading prior to exceeding 40% of rated power provides additional assurance that the core is properly loaded.

3-111 PROPOSED TS0384-0006V-NL02

4.18 POWER DISTRIBUTION INSTRUMENTATION 4.18.1 Incore Detectors SURVEILLANCE REQUIREMENTS 4.18.1.1.

4.18.1.2 The incore detection system shall be demonstrated operable:

a.

By performance of a Channel Check prior to its use following a core alteration and at least once per 7 days during power operation when required for the function listed in Section 3.11.1.

b.

At least once per refueling by performance of a Channel Calibration which exempts the neutron detectors but includes electronic components.

The incore alarm system is demonstrated operable through use of the datalogger program out-of-sequence alarm.

The out-of-sequence alarm is demonstrated operable once per refueling by performance of a Channel Check

4-81 Amendment No 68 December 8, 1981 TS0384-0006D-NL02

POWER DISTRIBUTION INSTRUMENTATION 4.18.2 EXCORE MONITORING SYSTEM SURVEILLANCE REQUIREMENTS 4.18.2.1 At least every 31 days of power operation:

a.

A target AO and excore monitoring allowable power level shall be determined using excore and incore detector readings at steady state near equilibrium conditions.

b.

The excore measured AO shall be compared to the.incore measured AO.

If the difference is greater than 0.02, the excore monitoring system shall be recalibrated.

c.

The excore measured Quadrant Power Tilt shall be compared to the incore measured Quadrant Power Tilt. If the difference is greater than 2%, the excore monitoring system shall be recalibrated

4-82 Amendment No 68 December 8, 1981 TS0384-0007D-NL02

.. '1

(~.,,..

(,,

4.18 POWER DISTRIBUTION INSTRUMENTATION 4.18.3 Excore Axial Offset System SURVEILLANCE REQUIREMENTS 4.18.3.l 4.18.3.2 Prior to power operation above 50% rated power following a refueling outage, verify that the AO has been calibrated with the incores.

After the initial calibration has been completed in accordance with 4.18.3.1, verify that the AO is calibrated with the incores as described in item b of 4.18.2.1 for power operation above 40% rated power

4-82a PROPOSED TS0384-0007A-NL02

4. 19 POWER DISTRIBUTION LIMITS 4.19.l LINEAR HEAT RATES SURVEILLANCE REQUIREMENTS 4.19.1.1 When using the incore alarm system to monitor LHR, prior to operation above 40X of rated power and every 7 days of power operation thereafter, incore alarms shall be set based on a measured power distribution.

4.19.1.2 When using the excore monitoring system to monitor LHR:

a.

Prior to use, verify that the measured AO has not deviated from the target AO by more than 0.05 in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

Once per day, verify that the measured Quadrant Power Tilt is less than or equal to 3~.

c.

Once per hour, verify that the power is less than or equal to the APL and not more than 10% of rated power greater than the power level used in determining the APL.

d.

Once per hour, verify that the measured AO is within 0.05 of the established target AO.

4-83 PROPOSED

4.19 POWER DISTRIBUTION LIMITS 4.19.2 RADIAL PEAKING FACTORS SURVEILLANCE REQUIREME!\\'TS 4.19.2.1 A

T 6H N

The measured radial peaking factors (Fr, Fr, Fr and Fr) obtained by using the incore detection system, shall be determined to be less than or equal to the values stated in the LCD at the following intervals:

a.

After each fuel loading prior to operation above 40% of rated power, and

b.

At least once per week of power operation.

4-84 PROPOSED

ML18051A826

Text

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