ML20011E570
| ML20011E570 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 01/31/1990 |
| From: | Jocelyn Craig Office of Nuclear Reactor Regulation |
| To: | Commonwealth Edison Co |
| Shared Package | |
| ML20011E562 | List: |
| References | |
| NPF-37-A-036, NPF-66-A-036 NUDOCS 9002160133 | |
| Download: ML20011E570 (18) | |
Text
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p ur jo UNITED STATES
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20065 1;
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COMMONWEALTH EDISON CONPANY l
DOCKET NO. 50 454 BYRON. STATION.. UNIT.1 AMENDMENT.TO. FACILITY OPERATING LICENSE Amendment No. 36 l
License No. NPF-37 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Comonwealth Edison Company (the licensee) dated July 31, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will o>erate in conformity with the application the provisions of tle Act, and the rules and regulations of the Comission; C.
Thereisreasonalleassurance(1)thattheactivitiesauthorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and i
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical L
Specifications as indicated in the attachment to this license l
amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby amended to read as follows:
i 21%h 0 I
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< (2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 36 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance.
FOR HE NUCLEAR REGULATORY C0lHISS10N 0
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[JbhnW.Craig, Director Project Directorate !!!-2 Division of Reactor Projects - III, IV, V and Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: January 31, 1990
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION
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w ASHINGTON. D. c. 20666
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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-455 BYRON STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING. LICENSE
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Amendment No. 36 License No. NPF-66 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Comonwealth Edison Company
}
(thelicensee)datedJuly 31, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR i
Chapter 1; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;-
C.
Thereisreasonableassurance(1)thattheactivitiesauthorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
i 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license l
amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby amended to read as follows:
l o l (2) Technical Specifications i
TheTechnicalSpecificationscontainedinAppendixA(NUREG-1113),
as revised through Amendment No. 36 and revised by Attachment 2 l
to NPF-60, and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-37, i
dated February 14, 1985, are hereby incorporated into this license.
- contains a revision to Appendix A which is hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance.
l ORTHENUCLEARREGULATORYCOMMISSION
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John W. Craig, Director Project Directorate III-2 i
Division of Reactor Projects - III, IV, V and Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: January 31, 1990 P
4
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.'..i ATTACHMENT T0 LICENSE AMENDMENT NOS. 36 AND 36 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66
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DOCKET NOS. 50-454 AND 50-455 r
Revise Appendix A as follows:
Remove Pages Insert Pages 2-8 2-8 B2-1 B2-1 B2-2 B2-2 3/4 1 - 4 3/4 1 - 4 3/4 1 - 5 3/4 1 - 5 3/4 1 - 19 3/4 1 - 19 3/4 2 - 4.
3/4 2 - 4 3/4 2 - 7 3/4 2 - 7 3/4 2 - 8 3/4 2 - 8 B 3/4 1 - 2 B 3/4 1 - 2 B 3/4 2 - 1 B 3/4 2 - 1 B 3/4 2 - 4 B 3/4 2 - 4 B 3/4 2 - 5 B 3/4 2 - 5 m. -m
~
Q TABLE 2.2-1 (Continued) l TABLE NOTATIONS (Continued)
NOTE 1:
(Continued)
C*
Time constant utilized in the measured T,yg lag compensator, is = 0 s, Ts
=
T'
$ 588.4'F (Nominal T at RATED-THERMAL POWER),
avg K
0.00134,
=
3 m
P
=
Pressurizer pressure, psig, P'
2235 psig (Nominal RCS operating pressure),
=
S
=
Laplace transform operator, s 1, and f (AI) is a function of the indicated difference between top and bottom detectors of the 3
power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:
7 (i) for qt 9b between -=% and +10% (Unit 1 Cycle 3 and Unit 2 Cycle 2), and -32% and +13%
(Unit 1 Cycle 4 and after; Unit 2 Cycle 3 and after), f (aI) = 0, where qt 8"d "b are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt*Ab I5 total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnitude of qt b exceeds +10% (Unit 1 Cycle 3 and Unit 2 Cycle 2),
~9 and +13% (Unit 1 Cycle 4 and after; Unit 2 Cycle 3 and after), the AT Trip Setpoint shall be automatically reduced by 2.0% (Unit 1 Cycle 3 and Unit 2 Cycle 2), and 1.74% (Unit 1 Cycle 4 and after; Unit 2 Cycle 3 and after) of its value at RATED THERMAL POWER.
(iii) for each percent that the magnitude of g U exceeds -32%, the AT Trip 5etpoint shall be 3
t b
h automatically reduced by 1.67% of its value at RATED THERMAL POWER (Unit 1 Cycle 4 and after; g
Unit 2 Cycle 3 and after)
"z*
NOTE 2:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.9% of AT span.
M
2.1 SAFETY LIMITS I
o
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BASES i
2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result 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 cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive claddin from nucleate boiling (DNB)g temperatures because of the onset of departure 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.
This relation has been developed to predict the DNB flux and l
the location of DNB for axially uniform and nonuniform heat flux distributions.
The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.
The DNB design basis is as follows:
there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and 11 events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation for Optimized fuel Assembly (0FA) fuel and the WRB-2 correlation for VANTAGE 5 fuel in this application).
The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent 3robability witi 95 percent confidence that DNB will not occur when the minimum )NBR is at the correlation DNBR limit (1.17 for l
both the WRB-1 and WRB-2 correlations).
In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal 3arameters, and fuel fabrication parameters are considered statistically such t1at there is at least a 95 confidence that the minimum DNBR for the limiting rods is greater than or equal to the DNBR limit.
The uncer-tainties in the above plant parameters are used to determine the plant DNBR uncertainty.
This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analysis using values of input parameters without uncertainties.
The design DN.BR values are 1.34 and 1.32 for a typical cell and a thimble cell, respectively for 0FA fuel, and 1.33 for a typical cell and 1.32 for a thimble cell for the VANTAGE 5 fuel.
In addition margin has been maintained in both designs by meeting safety analysis DNBR limits of 1.49 for a typical cell and 1.47 for a thimble cell for 0FA fuel, and 1.67 and 1.65 for a typical cell and a thimble cell, respectively for the VANTAGE 5 fuel in performing safety analyses.
The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum design DNBR is no less than the design DNBR value, or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.
BYRON - UNITS 1 & 2 B 2-1 AMENDMENT NO. 36
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i SAFETY LIMITS BASES REACTOR CORE (Continued)
Thesecurvesarebasedonanenthalpyhotchannelfactor,FfH'OII'49 for 0FA fuel and 1.59 for VANTAGE 5 fuel.
An allowance is included for an l
increaseinFfg at reduced power based on the expression:
Ffg=1.49[1+0.3(1-P)]
i for OFA fuel FfH=1.59[1+0.3(1-P)]forVANTAGE5 fuel Where P is the fraction of RATED THERMAL POWER.
These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the fy (61) function of the Overtemperature trip.
When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtem-perature AT trips will reduce the Setpoints to provide protection consistent with core Safety Limits.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
-The reactor vessel, pressurizer, and the RCS piping, valves, and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure.
The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements.
The entire RCS is hydrotested at 3110 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.
BYRON - UNITS 1 & 2 B 2-2 AMENDMENT NO.36 t-
REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:
i a.
Less positive than 0 Ak/k/'F for the all rods withdrawn, hot zero l
THERMAL POWER condition, or b.
Less negative than -4.1 x 10 4 Ak/k/'F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.
APPLICABILITY:
Specification 3.1.1.3a. - MODES 1 and 2* only#
. Specification 3.1.1.3b. - MODES 1, 2, and 3 only#.
ACTION:
a.
With the MTC more positive than the limit of Specification 3.1.1.3a.
above, operation in MODES 1 and 2 may proceed provided:
1.
Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than 0 ok/k/ F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; 2.
The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and 3.
A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within'10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods r
withdrawn condition.
4.
The provisions of Specification 3.0.4 are not applicable.
l b.
With the MTC more negative than the limit of Specification 3.1.1.3b.
above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
"With K,ff greater than or equal to 1.
- See Special Test Exceptions Specification 3.10.3.
BYRON - UNITS 1 & 2 3/4 1-4 AMENDMENT NO. 36
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:
a.
The MTC shall be measured and compared to the predicted MTC to establish administrative rod withdrawal limits, as necessary, to assure that the limit of Specification 3.1.1.3a., above, is met throughout core life, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading, and b.
The MTC shall be measured at any THERMAL POWER and compared to
-3.2 x 10 4 Ak/k/"F (all rods withdrawn, RATED THERMAL POWER condi-tion) within 7 EFPD after reaching an equilibrium boron concen-tration of 300 ppm.
In the event this comparison indicates the MTC is more negative than -3.2 x 10 4 Ak/k/*F the MTC shall be remeasured, andcomparedtotheEOLMTClimitofSpecIfication3.1.1.3b.,at least once per 14 EFPD during the remainder of the fuel cycle.
4 BYRON - UNITS 1 & 2 3/4 1-5 AMENDMENT NO. 36
REACTIVITY CONTROL SYSTEMS R00DROPTIM LIMITING CONDITION FOR OPERATION 3.1. 3. 4 The individual full-length shutdown and control rod drop time from the fully withdrawn position shall be less than or equal to 2.4 seconds (Unit 1 Cycle 3 and Unit 2 Cycle 2), and 2.7 seconds (Unit 1 Cycle 4 and af ter; Unit 2 Cycle 3 and after) from beginning of decay of stationary gripper coil voltage to dashpot entry with:
T,yg greater than or equal to 550 F, and a.
b.
All reactor coolant pumps operating.
APPLICABILITY:
MODES 1 and 2.
ACTION:
With the rod drop time of any full-length rod determined to exceed a.
the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
b.
With the rcd drop time within limits but determined with three reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 66% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full-length rcds shall be demonstrated through measurement prior to reactor criticality:
q For all rods following each removal of the reactor vessel head, a.
b.
For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and c.
At least once per 18 months.
BYRON - UNITS 1 & 2 3/4 1-19 AMENDMENT NO. 36
7 L ;c' i
POWER DISTRIBUTION LIMITS l
1 1
3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z) f i
LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:
9 F (Z) $ [2.32) [K(Z)] for P > 0.5*, and A
P F (Z) 5 [4.64) [K(Z)] for P $ 0.5*.
9 F (Z) $ [2.50] [K(Z)] for P > 0.5**,
O P
F (Z) $ [5.00) [K(Z)] for P $ 0.5**,
g Where:
p ~ THERMAL POWER RATED THERMAL POWER and f((Z) is the function obtained from Figure 3.2-2 for a given core height location.
APPLICABILITY:
MODE 1.
ACTION:
i With F (Z) exceeding its limit:
n a.
Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit 9
within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds the limit; and 0
b.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a,, above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit.
0
- Unit 1 Cycle 3 and Unit 2 Cycle 2
- ** Unit 1 Cycle 4 and after; Unit 2 Cycle 3 and after BYRON. UNITS 1 & 2 3/4 2-4 AMENDMENT NO. 36
n,s-r--
_J.
POWER DISTRIBUlION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
C RTP 2)
When the F is less than or equal to the F limit for the x
x appropriate measured core plane, additional power distribution RTP maps shall be taken and F compared to F and F at least x
7 x
once per 31 EFPD.
e.
The F limits for RATED THERMAL POWER (FRTP) shall be within the xy x
limits prt"41ded in the 0?ERATING LIMITS REPORT for all core planes containing Bank "0" control rods and for all unrodded core rianes; f.
The F,y limits of Specification 4.2.2.2e., above, are not applicable in the following c ue planes regions as measured in percent of core height from the bottom of the fuel:
1)
Lower core region from 0 to 15%, inclusive, 2)
Upper core region from 85 to 100%, inclusive, 3)
Within i 2% of grid plane regions (except VANTAGE 5 assembly Intermediate Flow Mixer Grids) such that no more than 20% of the total core height in the center core region is affected, and 4)
Core plane regions within i 2% of core height (1 2.88 inches) about the bank demand position of the Bank "D" control rods.
g.
With F exceeding F, the effects of F on F (Z) shall be x
xy 9
evaluated to determine if F (Z) is within its limits.
9 4.2.2.3 When F (Z) is measured for other than F determinations, an overall q
xy measured F (Z) shall be obtained from a power distribution map and increased 9
by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.
I BYRON - UNITS 1 & 2 3/4 2-7 AMENDMENT NO. 36 l
7 e
POWER DISTRIB!l TION _ LIMITS E4. 2. 3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR I
LIMITING CONDITION FOR OPERATION IndicatedReactorCoolantSystem(RCS)totalflowrateandFfH 3.2.3 shall be maintained as follows for four loop operation.
1 a.
RCS Total Flowrate > 390,400 gpm, and Ffg51.55(1.0+0.3(1.0-P))for0FAfuel b.
F g 51.65 [1.0 + 0.3 (1.0-P)] for VANTAGE 5 fuel where:
MeasuredvaluesofFfg are obtained by using the movable incore detectors.
An appropriate uncertainty of 4% (nominal) or greater shallthenbeappliedtothemeasuredvalueofFfg before it is compared to the requirements, and THERMAL POWER
'3, RATED THERMAL POWER q
RPLICABILITY:
MODt' 1.
ACTION:
With RCS total flow rate or Ffg outside the region of acceptable operation:
a.
Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:
4 1.
Restore RCS total flow rate and F to within the above limits, H
or 2.
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the l
next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
)
)
BYRON - UNITS 1 & 2 3/4 2-8 AMENDMENT NO. 36
9 1
REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)
The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions.
These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MW) to an all rods withdrawn condition and, a conversion for the rate of chat of moderator density with temperature at RATED THERMAL POWER conditions.
I....; value of the MDC was then transformed into the limiting MTC value -4.1 x 10-4 ok/k/'F.
The MTC value of -3.2 x 10 4 Ak/k/*F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppta equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -4.1 x 10 4 Ak/k/* F.
The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC can be maintained within its limits.
The BOL MTC measurement, combined with the nredicted MTC throughout core life, will be used to impose administrative limits on rod withdrawal, as required during core life to ensure that MTC will always be less positive than 0 AK/K/*F.
This coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.
3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average te:pecture less than 550 F.
This limitation is required to ensure: (1) the inoderator temperature coefficient is within its analyzed temperature range, (2' the trip instrumentation is within its normal operating range, (3) the press.irizer is capable of being in an OPERABLE status with a steam bubble, (4) the reactor vessel is above its minimum RT temperature, and (5) the plant is above the cooldown steam dump NDT permissive, P-12.
3/4.1.2 BORATION SYSTEMS TheBoronInjectionSystemensuresthatnegativereactivitycontrolis available during each MODE of facility operation.
The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.
l With the RCS average temperature above 350'F, a minimum of two boron f
injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable.
The boration capability of either flow path is sufficient to provide a SHUTDOWN l
MARGIN from expected operating conditions of 1.3% Ak/k after xenon decay and cooldown to 200 F.
The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 15,780 gallons of 7000 ppm borated water from the boric acid storage tanks or 70,450 gallons of 2000 ppm borated water from the refueling water storage tank.
t I
BYRON - UNITS 1 & 2 B 3/4 1-2 AMENDMENT N0 36
}
i 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (dormal Operation) and II (Incidents of Moderate Frequency) events by:
(1) maintaining the minimum DNBR in the core greater than or l
equal to the appropriate DNBR limit (See Bases 2.1) during normal operation and l
in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria.
In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for t
the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded.
The definitions of certain hot channel and peaking factors as used in these specifications are as follows:
f F (Z)
Heat Flux Hot Channel Factor, is defined as the maximum local 9
heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; FfH Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power; and Fxy(Z)
Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.
3/4.2.1 AXIAL FLUX DIFFERENCE t
The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper bound 9
envelope of the F Limit times the normalized axial peaking factor is not exceeded l q
during either normal operation or in the event of xenon redistribution following power changes.
Target flux difference is determined at equilibrium xenon conditions.
The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels.
The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.
The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.
l BYRON - UNITS 1 & 2 B 3/4 2-1 AMENDMENT NO. 36
d POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) c.
The control rod insertion limits of Specification 3.1.3.6 are maintained, and d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
Ffg will be maintained within its limits provided the Conditions a. through
- d. above are maintained.N The combination of the RCS flow requirement (390,400 gpm) and the requirement on F will be met.
3g guarantee that the DNBR used in the safety analysis Margin between the safety analysis limit DNBRs (1.49 and 1,47 for the OFA fuel typical and thimble cells, respectively and 1.67 and 1.65 for the VANTAGE 5 typical and thimble cells) and the design limit DNBRs (1.34 and 1.32 for the OFA fuel typical and thimble cells, and 1.33 and 1.32 for the VANTAGE 5 fuel typical and thimble cells, respectively) is maintained.
A fraction cf this margin is utilized to accommodate the transition core DNBR penalty (maximum of 12.5%) and the appropriate fuel rod bow DNBR penalty (less than 1.5% per WCAP-8691, Revision 1).
The rest of the margin between design and safety analysis DNBR limits can be used for plant design flexibility.
The RCS flow requirement is based on the loop minimum measured flow rate of 97,600 gpm which is used in the Improved Thermal Design Procedure described in FSAR 4.4.1 and 15.0.3.
A precision heat balance is performed once each cycle and is used to calibrate the RCS flow rate indicators.
Potential fouling of the feedwater venturi, which might not be detected, could bias the results from the I
precision heat balance in a non-conservative manner.
Therefore, a penalty of 0.1% is assessed for potential feedwater venturi fouling.
A maximum measurement uncertainty of 2.2% has been included in the loop minimum measured flow rate to account for potential undetected feedwater venturi fouling and the use of the RCS flow indicators for flow rate verification.
Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant aerformance parameters.
If detected, action shall be i
taken, before performing su) sequent precision heat balance measurements, i.e.,
either the effect of fouling shall be quantified and compensated for in the RCS flow rate measurement, or the venturi shall be cleaned to eliminate the fouling.
Surveillance Requirement 4.2.3.4 provides adequate monitoring te etect possible flow reductions due to any rapid core crud buildup.
Surveillance Requirement 4.2.3.5 specifies that the measurement instrumen-tation shall be calibrated within seven days prior to the performance of the calorimetric flow measurement.
This requirement is due to the fact that the drif t effects of this instrumentation are not included in the flow measurement uncertainty analysis.
This requirement does not apply for the instrumentation whose drift effects have been included in the uncertainty analysis.
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BYRON - UNITS 1 & 2 B 3/4 2-4 AMENDMENT NO. 36 L
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A POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)
The limits of Section 3.2.3 for F do not assume any specific uncertainty l
H onthemeasuredvalueofFfg.
An appropriate uncertainty of 4% (nominal) or greaterisaddedtothemeasuredvalueofFfH before it is compared with the requirement.
When an F measurement is taken, an allowance for both experimental error q
and manufacturing tolerance must be made.
An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.
The Radial Peaking Factor, Fxy(Z) is measured periodically to provide assurance that the Hot Channel F (Z) remains within its limit.
The F*Y limit for RATED THERMAL POWER (F RTP) Q as provided in Specification 3.2.2 was xy determined from expected power control maneuvers over the full range of burnup conditions in the core.
The 12-hour periooic surveillance of indicated RCS flow is sufficient to detect flow degradation which could lead to operation outside the acceptable limit.
3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power dis-tribution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during startup testing and periodically during power operation.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generat' ion rate protection with x-y plane power tilts.
A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod.
In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by redue q
ing the maximum allowed power by 3% for each percent of tilt in excess of 1.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that i the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.
The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.
The two sets of four sym-metric thimbles is a unique set of eight detector locations.
These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.
BYRON - UNITS 1 & 2 B 3/4 2-5 AMENDMENT N0.36
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