ML19308D745

Tech Spec Change Request 31 Re Tech Spec 2.1,2.2 & 3.2, Addressing Removal of Orifice Rod Assemblies & Calculation of Resultant Changes to Core Parameters by BAW-2 DNB Correlation Per BAW-1490,Revision 1

ML19308D745
Person / Time
Site:	Crystal River
Issue date:	07/21/1978
From:	FLORIDA POWER CORP.
To:
Shared Package
ML19308D743	List: ML19308D742 ML19308D745 ML19308D746
References
NUDOCS 8003120915
Download: ML19308D745 (11)

Text

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- - '" TECHNICAL SPECIFICATION CHANGE REQUEST N0. 31 [ App;ndix A]  ;

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Replaces pages 2-3, B2-1, 2, 3, 4, 5, 6, 7, 3 4 2-6 ~ and B3/4 2-2 with the attached revised pages 2-3, B2-1, 2, 3, 4, 5, 6, 7, 3/4 2-6 and B3/4 2-2.

Proposed Change These channes address the removal of the Orifice Rod Assemblies (0RA's) from Crystil River Unit 3 and the calculation of the resultant changes to the core pa. ameters by the BAW-2 DNB correlation. This is supported in Revision 1 to BAW-1490 " Licensing Considerations for Continued Cycle 1 Operation Without Burnable Poison Rod Assemblies and Orifice Rod Assemblies."

These changes include: 1) increased Reactor Coolant Flow due to the removal of the BPRA's and the ORA's; 2) changing the nuclear enthalpy rise hot channel factor (Ff) to < . 71 [1 + 0.6 (1-P)]; 3) changing the DNB 2.57; 5) changing the local quality restriction at the point of minim,"umcor DNBR. to 22%; 6) changing the power assumed in the safety analysis to 112%;

and 7) changing the flux-flow ratio which reduces the power level trip and associated reactor power-reactor power imbalance boundaries to 1.043.

Reason for Proposed Change Crystal River Unit 3 will operate the remainder of the first fuel cycle with no ORA's in the reactor. As stated in Revision 1 to BAW-1490

" Licensing Considerations for Continued Cycle 1 Operation Without Burnable Poison Rod Assemblies and Orifice Rod Assemblies", certain Technical Specifications need to be revised because of this. All of the proposed changes are the result of this operational modification.

Safety Analysis Justifying Proposed Chance The Licensing Considerations for operation of Crystal River Unit 3 with the ORA's removed has been filed with the Comission in Revision 1 to BAW-1490

" Licensing Considerations for Continued Cycle 1 Operation Without Burnable Poison Rod Assemblies and Orifice Rod Assemblies." These proposed changes will bring the Technical Specifications into agreement with that filing.

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2 137.9 x 10 6 2 103.0 x 10 l Figure 2.12 Reactor Core Safety Lirmt 4

CRYSTAL ; RIVER - UNIT 3 2-3

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.2.1 SAFETY l.IMITS BASES 2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would recult in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladd'ag surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime would result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the BAW-2 DNB correlation. The DNB correlation has been i developed to predict the DNS flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value e 'NBR during steady state operation, normal operational transients, and a' ,..ed transients is limited to 1.30. This value corresponds to a % percent prcoability at a 95 percent confidence level that l DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.30 is predicted for the maximum possible thermal power 112% when the reactor coolant flow is 137.89 x 100 lbs/hr, which is 105% the design flow l rate for four operating reactor coolant pumps. This curve is based on the following nuclear power peaking factors with potential fuel densification e ffects:

FN = 2.57; FN = 1.71; FN = 1.50 Q M Z The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum allowable control rod withdrawal, and form the core DNBR design basis.

CRYSTAL RIVER - UNIT 3 B 2-1

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. SAFETY LIMITS BASES The reactor trip envelope appears to approach the safety limit more closely than it actually does because the reactor trip pressures are measured at a location where the inc"cated pressure is about 30 psi less than core outlet

' pressure, providing a more conservative margin to the safety limit.

The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and account for the effects of potential fuel densification and potential fuel rod bow:

1. The 1.30 DNBR limit produced by a nuclear power peaking factor of FJ = 2.57 or the combination of the radial peak, axial peak and l position of the axial peak that yields no less than a 1.30 DNSR.

2. The combination of radial and axial peak that causes central fuel melting at the hot spot. The limit is 19.7 kw/ft.

Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.

The specified flow rates for curves 1 and 2 of Figure 2.1-2 corrt'.pon to the expected minimum flow rates with four pumps and three pumps, res;ectively.

The curve of Figure 2.1-1 is the most restrictive of all possibie reactor collant pump-maximum thermal power combinati?M shown in BASES rigure 2.11 The curves of BASES Figure 2.1 represent the conoif. ions at which a ainimum DNBR of 1.30 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to 22%, whichever conditiort is more restrictive. l These curves include the potential effects of fuel rod bow and fuel densi-fication. l The DNBR as calculated by the BAW-2 DNB correlation continuall:/ increases from I point of minimum DNBR, so that the exit DNBR is always higher. Extrapolation of the correlation beyond its published quality range of 22% is f Jstified on the I basis of experimental data.

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I SAFETY LIMITS l BASES For each curve of BASES Figure 2.1, a pressure-temperature point

above and to the left of the curve would result in a DNBR greater than 1.30 or a local quality at the point of minimum DNBR less than 22% for l that particular reactor coolant pump situation. The 1.30 DNBR curve for-four pump operation is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the lef t of the four pump curve will be above and to the left of the other curves.

2.1.3 REACTOR COOLANT SYSTEM PRESSURE l

The restriction of this Safety Limit protects the integrity of the

Reactor Coolant System from overpressurization and thereby prevents the release of Padionuclides contained in the reactor coolant from reaching the containment atmosphere.

l The reactor pressure vessel and pressurizer are designed to Section III of the ASME Boiler and Pressure Vessel Code which permits a maximum transient pressure of 110%, 2750 psig, of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to USAS B 31.7, February,1968 Draft Edition, which permits a maximum transient pressure i of 110%, 2750 psig, of component design pressure. The Safety Limit of

( 2750 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3125 psig, 125%

of design pressure, to demonstrate integrity prior to initial operation.

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2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS t

The Reactor Protection System Instrumentation Trip Setpoint specified in Table 2.2-1 are the values at which the Reactor -Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip setpoint less conse"vative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

The Shutdown Bypass provides for bypassing certain functions of the Reactor Protection System in order to permit control rod drive tests, zero power PHYSICS TESTS and certain startup and shutdown procedures.

The purpose of the Shutdown Bypass RCS Pressure-High trip is to prevent normal operation with Shutdown Bypass activated. This high pressure trip setpoint is lower than the normal low pressure trip setpoint so that the reactor must be tripped before the bypass is initiated. The Nuclear Overpower Trip Setpoint of < 5.0% prevents any significant reactor power from being produced. Sufficient natural circulat' ion would be available to remove 5.0% of RATED THERMAL POWER if none of the reactor coolant pumps were operating. .

Manual Reactor' Trio The Manual Reactor Trip is a redundant channel to the automatic Reactor Protection SystE.i instrumentation channels and provides manual reactor trip capability. -

Nuclear Overoower A Nuclear Overpower trip at high power level (neutron flux) provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.

During normal station operation, reactor trip is initiated when the reactor power level reaches 105.5% of rated power. Due to calibration and. instrument errors, the maximum actual power at which a trip would be actuated could be.ll2%, which was used in the safety analysis. l CRYSTAL RIVER - UNIT 3 8 2-4 ww-c- --

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LIMITING SA'FETY SYSh.A SETTINGS BASES RCS Outlet Temperature - High The RCS Outlet Temperature High trip 1619'F prevents the reactor outlet temperature from exceeding the design limits and acts as a backup trip for all power excursion transients.

Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE The power level trip setpoint produced by the reactor coolant system flow is based on a flux-to-flow ratio which has been established to accomodate flow decreasing transients from high power.

The power level trip setpoint produced by the power-to-flow ratio provides both higher power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level setpoint produced by the power-to-flow ratio provides overpower DNB protection for all . modes. of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.2-1 are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 1104.3% and reactor flow rate is 100%, or flow rate is l 195.9% and power level is 100%. i

2. Trip would occur when three reactor coolant pumps are operating if power is 177.9% and reactor flow rate is 74.7%, or flow rate is 171.9% and power is 75%.

3. Trip would occur when one reactor ecolant pump is operating in each loop (total of two pumps operating) if the power if 161.3% and reactor flow rate is 49.2% or flow rate is 147.9% and the power level is 50.0%

For safety calculations the maximum calibration and instrumentation errors for the power level were used.

1 CRYSTAL RIVER - UNIT 3 8 2-5

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LIMITING SAFETY SYSTEM SETTINGS BASES The AXIAL POWER IMBALANCE boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kw/ft limits or OflBR limits. The AXIAL POWER IMBALANCE reduces the power level trip produced by the flux-to-flow ratio such that the boundaries of Figure 2.2-1 are produced. The flux-to-flow ratio reduces the power level trip and associated reactor power-reactor power-imbalance boundaries by 1.043% for a 1" flow reduction. l RCS Pressure - Low, High and Variable Low The High and Low trips are provided to limit the pressure range in which reactor operation is permitted.

During a slow reactivity insertion startup accident from low power or a slow reactivity insertion from high power, the RCS Pressure-High setpoint is reached before the Nuclear Overpower Trip Setpoint. The trip setpoint for RCS Pressure-High, 2355 psig, has been established to maintain the system pressure below the safety limit, 2750 psig, for any design transient. The RCS Pressure-High trip is backed up by the pressurizer code safety valves for RCS over pressure protection, and is therefore set lower than the set pressure for these valves, 2500 psig. The RCS Pressure-High trip also backs up the Nuclear Overpower trip.

The RCS Pressure-Low, 1800 psig, and RCS Pressure-Variable Low, (16.25 T *F-7838) psig, Trip Setpoints have been established to maintain ?$e DNB ratio greater than or equal to 1.30 for those design accidents that result in a pressure reduction. It also prevents reactor operation at pressures below the valid range of DNB correlation limits, protecting against DNB.

Due to the calibration and instrumentation errors, the safety analysis used a RCS Pressure-Variable Low Trip Setpoint of (16.25 Tout F-

/878) psig.

Reactor Containment Vessel Pressure - High The Reactor Containment Vessel Pressure-High Trip Setpoint < 4 psig, provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the containment vessel or a loss-of-coolant accident, even in the absence of a RCS Pressure -Low trip.

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Reactor Coolant Flow  ;

Curve (LBS/HR) Power Pumps Operating (Type of Limit) 1 1 137.9x106 005 %) 112% Four Pumps (DNBR Limit) l 2 103.0 x106(74.7%) 85 . 9 *, Three Pumps (DNBR Limit)

Pressure / Temperature Limits at Maximum Allowable Power for Minimum DNBR 1

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NUCLEAR ENTHALPY, RISE HOT CHANNEL FACTOR - F H

. LIMITING CONDITION FOR'0PERATION N

~3.2.3 F aH shall be limited by the following relationship:

N F

aH 1 1.71 [1 + 0.6(1-P)]

iHERMAL POWER where P = RATED THERMAL POWER f

and P 1'l.0 APPLICABILITY: MODE 1.

ACTION:

With F H exceeding its limit:

a. Reduce THERMAL POWER at least 1" for each 1", that F N exceeds the limit within 15 minutes and similarly . reduce the NuSYear Overpower Trip Setpoint and Nuclear Overpower.~ based on RCS Flow and AXIAL POWER IMBALANCE Trip Setpoint within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. Demonstrate through in-core mapping that F N is within its limit within24hoursafterexceedingthelimitSkreduceTHERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

-c. Identify and correct the cause of the out of l_imit condition prior.to increasing THERMAL POWER above the reduced limit required by a or b, aboge; subsequent POWER OPERATION may proceed provided that F - is demonstrated through in-core mappingtobewithinit$Hlimit at a nominal 50", of RATED -

THERMAL POWER prior to exceeding this THERMAL POWER, at a nominalv75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.

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CRYSTAL ilIVER - UNIT'3l - 3/4 2-6 m

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p POWER DISTRIBUTION LIMITS BASES N

F . Nuclear Enthalpy Rise Hot Channel Factor, is defined as the AH ratio of the integral of linear power along the rod on which minimum DNBR occurs to the average rod power.

It has been determined by extensive analysis of possible operating power shapes that the design limits on nuclear power peaking and on minimum DNBR at full power are met, provided:

F g1 3.12; F H 1 I'71 Power Peaking'is not a directly observable quantity and therefore limits have been established on the bases of the AXIAL POWER IMBALANCE produced by the power peaking. It has been determined that the above hot channel factor limits will be met provided the following conditions are maintained.

1. Control rods in a single group move together with no individual rod insertion differing by more than + 6.5% (indicated position) from the group average height. -

2. Regulating rod groups are sequenced with overlapping groups as required in Specification 3.1.3.6.

3. The regulating rod insertion limits of Specification 3.1.3.6 are maintained.

4. AXIAL POWER IMBALANCE limits are maintained.

The AXIAL POWER IMBALANCE is a measure of the difference in power between the top and bottom halves of-the core. Calculations of core average axial peaking factors for many plants and measurements from operating plants under a variety of operating conditions have been correlated with AXIAL. POWER IMBALANCE. 1 9 correlation shows that the design power shape is not exceeded if the AXIAL POWER IMBALANCE is maintained between + 15 percent and - 17 percent at RATED THERMAL POWER.

The design limit power peaking factors are the most restrictive calculated _at full power for the. range from all control rods fully withdrawn to minimum allowable control rod insertion and are the core ONBR design basis. Therefore, for operaticn at a fraction cf RATED THERMAL POWER, the design limits' are met. When usjng incore detectors to make power distribu-tion maps to determine Fq and F'H a eas

a. The measurement of total peaking factor, F , shall be increased by 1.4 percent to account for mahufacturing tolerances and further increased by 7.5 percent to account for measurement error.

CRYSTAL RIVER - UNIT 3 B 3/4 2-2 6 '