ML20095L575
| ML20095L575 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/01/1992 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Cleveland Electric Illuminating Co, Duquesne Light Co, Ohio Edison Co, Toledo Edison Co |
| Shared Package | |
| ML20095L576 | List: |
| References | |
| NPF-73-A-046 NUDOCS 9205070140 | |
| Download: ML20095L575 (21) | |
Text
_
UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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t WASHINGTON, D. C. 20555 s.,..../
DV0VESNE LIGHT COMPAt{1
The Nuclear Regulatory Commission (the Commission) has found that*
A.
The application for amendment by Duquesne Light Company, et al.
(licensee) dated October 15, 1991, as supplemented by letters dated January 27. and February 25, 1992, complies with the standards and recuirements of the Atomic Energy Act of 1954, as amended (the Act) anc the commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the pruvisions of the Act, and the rules and regulations of the Commission; C.
Thereisreasonableassurance(i)thattheactivitiesauthorizedby
~
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 Commission's regulations; D.
The issuance of this amendment will not be inimical to the common 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 Commission's regulations and all applicable requirements have been satisfied.
9205070140 920501 PDR ADOCK 05000412 P
PDR L
l i
2.
/ccordingly, the license is amended by cb nges to the Technical Specifications as indicated in the attachment to this licenso amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-73 is hereby amended to read as follows:
(2) lechnical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 46 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. DLCO 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, to be impicmented within-60 days of issuance.
FOR THE NVCLEAR REGULATORY COMMISSION
^
,a ohn'F. Stolz, Directo rofectDirectorate(4 11/ision of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications-Date of Issuance:
May 1. 1992,
, +,
4 4
7 i
e ATTACHMENT TO LICENSE AMENDMENT NO. 46_
FAClllTY OPERATING LICENSE NO. NPF-73 DOCKET NO.50-41R Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change, Remove Insert B 2-1 B 2-1 B 2-2 B 2-2 B 2-3 8 2-3 B 2-4 B 2-4 B 2-5 B 2-5 B 2-6 8 2-6 8 2-7 B 2-7 B 2-B 3/4 1-23 3/4 1-23 8 3/4 2-i B 3/4 2-1 B 3/4 2-4 8 3/4 2-4 8 3/4 2-5 B 3/4 2-5 8 3/4 4-1 B 3/4 4-1
___m._
NPF-73 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding r rforation which would result in the release of fission products to the reactor coolant.
Overheating of the fuel cladding is prevented by reatricting fuel operation to within tbo 4
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 cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coef ficient.
DNB is not a directly measurable parameter during _ operation and thereforo THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the WRB-1 correlation.
The WRB-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat.
.. lux _ distributions.
The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, 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 fuel rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application).
The correlation DNBR limit is based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit (1.17 for the WRB-1 correlation).
Incorporating the peaking factor uncertainties in the correlation limit results in a DNBR design limit value of 1.21.
This DNBR value must be met in plant safety analyses using nominal lues of the input parameters that were included in the DNBR uncertainty evaluation.
In addition,. margin has been maintained in the design by me, ting a safety j
analysis DNBR limit of 1.33 in performing safety analyses.
The curve of Figure 2.1-1 shows the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature fer which the l minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
The curves are conservatively based on an enthalpy hot channel
- factor, F", of 1.62.
The Thermal-Hydraulic and non-LOCA analyses as that vere conducted for Unit 1 bounds the Unit 2 analyses (i.e.,
F" l
eB l
of 1.62).
The LOCA and Core Design licensing basis is 1.55.
These will bound actual plant operation which is restricted to an BEAVER VALLEY - UNIT 2 B 2-1 Amendment No. 46 l
l
liPF-73 SAFETY LIMITS BASES i
2.1.1 REACTOR CORE. continued F limit of 1.55 (see the CORE OPERATING LIMITS REPORT).
An allowance is included for an increase in F{ at reduced power based on the expression provided in the core Operating Limits Report (COLR).
There limiting heat flux condit ions are higher than those calculated for the range of all control roos fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f(AI) function of the overtemperature AT trip.
When the axial power imbalance is not within the tolerance, the axial powar imbalance effect on the overtemperature AT trip will reduce the setpoint to provide protection consistent with core safety limits.
2.1.2 REACT _OR COOLA11T SYSTEM PRESEllEE The restriction of this safety limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents tho i
release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure 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 Reactor coolant System is hydrotested at 3107 psig to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP INSTRUMENTATION JJTPOIN"'S The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor Trips are set for each functional unit.
The Trip Setpoints have been selected to ensure that the peactor core and Reactor Coolant System are prevented from exceeding their saf'ety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuatio.i System in mitigating the consequences of accidents.
The setpoint for a reactor trip system or interlock function is considered to be adj 2sted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy.
.To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, Allowable Values for the reactor trip setpoints have been specified in Table 2.2-1.
Operation with setpoints less conservative than the Trip Setpoint but within the Allowable Value is LEAVER VALLEY - UNIT 2 B 2-2 Amendment s.
46
NPT-73 LIMITTNG_ SAFETY SYSTEM SETTINGS BASES e....,.;.....;.-..
22211 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS. continued acceptable since an allowance has been made in the safety analysis forll to '
accomm0date this error.
An optional provision has been included determining the OPERABILITY of a channel when its trip setpoint is found to exceed the Allowable Value.
The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack.:
and sensor components in conjunction with a statistical combination of the other uncertainties in calibrating the instrumentation.
In Equation 2.2-1, Z + R + S S TA, tho interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered.
Z, as specified in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis exc,1uding those Essociated with the sensor and rack drift and the accuracy of their measuromont.
fA or Total Allowance is the difference, in percent span, between the trip setpoint and the value used in the analysis for reactor trip.
R or Rack Error is the "as measured" deviation, in percent span, for the affected channel from the specified trip setpoint.
S or Sensor Drift is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent span, 1
from the analysis assumptions.
Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.
The methodology to derive the trip setpoints is based upon combining all of the uncertainties in the channels.
Inherent to the determination of the trip setpoints are the magnitudes of these channel uncertainties.
l Sensors and other instrumentation utilized in these channels are expected l
to be capable of operating within the allowances of these uncertainty l
magnitudes.
Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.
Being that there is a small statistical chance that this will happen, an infrequent excessive drif t is expected.
Rack or sensor drif t, in excess of the a)Wrance that l
1s more than occasional, may be indicative of more serious problems and should warrant further investigation.
Manual Reactor Trio The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
Egwer Rance. Neutron Flux The Power Range, Neutron Flux channel high setpoint provides reactor l
core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.
.The low setpoint provides redundant protection in the power range for a power excursion beginning from low power.
The trip BEAVER VALLEY - UNIT 2 B 2-3 Amendment No. 46 i
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NPF-73 LIMITING SAFETl SYSLEM SETTINGS BASES acso iated with the low setpoint may be manually bypassed when P-10 is activa (two of the four pouer range channels indicate a power level of above approximately 10 percent of RATED THERMAL POWER) and is automatically reinstated when P-10 becomes inactive (three of the four channels indicato a power level below approximately 10 percent of RATED THERMAL POWER).
Power Rance, Neutron Flux, High..PJLtag The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level.
Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.
The Power Range Negative Rate trip provides protection to enstre that the minimum DNBR is maintained above the design DNBR limit for control rod drop accidents.
At high power a multiple rod drop accident could cause local flux peaking which, when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconservative local DNBR to exist.
The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor.
No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the design DNBR limit.
Intermediate and Source Rance, Nuclear Flux The Intermediate and Source Range, Nuclear ilux trips prvvide reactor core prncoction during reactor startup to mitigate the consequences of an uncont:011ed rod cluster control assembly bank withdrawal from a subcritical condition.
These trips provide redundant protection to the low setpoint trip of the Power Range, Neutron Flux channels.
The Sctrce Range Channels will initiate a reactor trip at about 10" counts per second unless manually blocked when P-6 becomes active.
The intermediate range channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active.
Although no explicit credit was taken for operation of the Source Range Channels in the accident
- analyses, operability requirements in the Technical Specificaticns will ensure that the Source Range Channels are available to mitigate the consequences of an inadvertent control bank withdrawal in MODES 3, 4 and S.
o Overtemperature AT The overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit, thermowell, and RTC responFS time delays BEAVER VALLEY - JNIT 2 B 2-4 Amendment No. 46
l NPF-73 l
LYMITIN_G SAFETY SYSTEM SEITL1193 l
BASES secbds),
from the core to the temperature detectors (about 4 and pressure is within the range between the High and Low pressure reactor trips.
This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic compensation for transport, thermowell, and RTD response time delays from the core to RTD output indication.
With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown on Figure 2.1-1.
If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notation in Tab'o 2.2-1.
Overoower AI The overpower AT reactor trip provides assurance of fuel integrity, e.g.,
no melting, under all possible overpower conditions, limits the
(
required range for overtemperature AT protection, and provides a backup to the High Neutren Flux Trip.
The setpoint includes corrections for l
changes in density and heat capacity of water with temperature, and I
dynamic compensation for transport, thermowell, and RTD response time l
delays from the core to RTD output indication.
The Overpower AT trip provides protection to mitigate the consequences of various size steam i
line t reaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secoauary Steam Release."
I' Pressuriggr_PfA.ggyIg, The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted.
The High Pressure trip is backed up by the pressurizer code safety valves for RCS l
overpressure protection, and is therefore set lower than the set pressure for those valves (2485 psig).
The Low Pressure trip protects against low l
pressure which could lead to DNB by tripping the reacter in the event of i
a loss of reactor coolant pressure.
On decreasing power the Low Pressure trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER or i
turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.
l Pf34syrlzer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level BEAVER VALLEY - UNIT 2 B 2-5 Amendment No. 46
NPF-73 LIMITING SAFETY SYSTEM SETTINGS BASES to a volume sufficient to retain a steam bubble and prev 6nt water relief through the pressurizer safety valves.
On decreasing power, the pressurizer high water level trip is aucomatically blocked by P-7 (a e
power level of approximately 10 percent of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10 percent of full power equivalent); and on increasing power, automatically reinstated by P-7.
No credit was taken for operation of this trip in the accident analyses;
- however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.
Loss of.. Flow The Loss of Flow trips provide core protection to prevent DNB in the event of a loss or one or more reactor coolant pumps.
Above 10 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 90 percent of nominal full loop flow.
Above 30 percent (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90 percent of nominal full loop flow.
Beam Generator Water Level The Steam Generator water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity.
The specified setpoint provides allowance that here will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system.
4D_prvoltace a_Dd Underfrequency - Reactor Coohnt Pumo Busses d
The Undervoltage and Underflaquency Reactor Coolant Pump bus trips I
provide reactor core protection against DNB as a result of loss of l
voltage or undnefrequency to more than one reactor coolant pump.
The l
~
specified setpoints assure a reactor trip signal is generated before the low flow trip setpoint is reached.
Time delays are incorporated in the underfrequency and undervoltage trips to prevent spurious reactor trips from momentary electrical power transients.
For undervoltage, the delay l
is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds.
For underfrequency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip setpoint is reached chall not exceed 0.6 seconds.
On decreasing power, the Undervoltage
.e.id Underfrequency Reactor Coolant Pumn bus trips are automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with a BEAVER VALLEY - UNIT 2 B 2-6 Amendment No. 46
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NPF-73 LIMITING SAFETY SYSTEM SETTINGS BASES turbine impulse chamber pressure at approximately 10 percent of full power equivalent); and on increasing power, reinstated automatically by P-7.
.s.
Turbine Trio-A Turbine Trip causes a direct reactor trip when operating above P-9.
Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient.
No credit was taken in the accident analyses for operation of these trips.
Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System.
Safety Iniection Inout from ESP If a reactor trip has not already been generated by the reactor protective instrumentation, the ESF automatic actuation logic channels will initiate a reactor trip upon any signal which initiates a safety injection.
This trip is provided to protect the core in the event of a LOCA.
The ESF instrumentation channels which initiate a safety injection signal are shown in TABLE 3.3-3.
Reactor coolaqi Pumo Breaker Position Trio The Reactor Coolant Pump Breaker Position Triss are anticipatory trips which provide roanter core protection against DNB resulting from the opening of two or note pump breakers above P-7.
These trips are blocked below P-7.
The open/close position trips assure a reactor trip signal is generated before the low flow trip setpoint is reached.
No credit was taken in the accident analyses for operation of these trips.
l Their functional capability at the open/close position settings is required to enhance the overall reliability of the Reactor Protection System.
Reactor Trio System Interlocka The Reactor Trip System interlocks perform the following functions:
P-6 Above the setpoint P-6 allows the manual block of the source Range reactor trip and de-energizing of the high voltage to the detectors.
Below the setpoint source range level trips are automatically reactivitated and high voltage restored.
I P-7 Above the setpoint P-7 automatically enables reactor trips on low flow or coolant pump breaker open in more than one_ primary coolant
- loop, reactor coolant pump bus undervoltage and urderfrequency, pressurizer low pressure and pressurizer high l
level.
Below the setpoint the above listed trips are l
automatically blocked.
BEAVER VALLEY - UNIT 2 B 2-7 Amendment No. 46
NPT'73 LIMITING SAFETY SYSTEM JBETTINGS BASES P-8 Above the setpoint P-8 automatically enables reactor trip on low flow in one or nore primary coolant loops.
Below the setpoint P-8 automatically blocks the above listed trip.
P-9 Above the setpoint P-9 automatically enables a reactor trip on turbine trip.
Below the setpoint P-9 automatically blocku a reactor trip on turbine trip.
P-10 Above the setpoint P-10 allows the manual block of the Intermediate Range reactor trip and the low setpoint Power Range reactor trip; and automatically blocks the Source Range reactor trip and Co-energizes the Source Range high voltage power.
Below the setpoint the Intermediate Range Teactor trip is automatically reactivated.
Provides input to P-7.
P-13 Provides input to P-7.
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1 9
'3EAVER VALLEY - UNIT 2 B 2-8 Amendment No. 46
~
l NPF-73 REACTIVITY CONTROL SYSTEMS ROD DROP TIME 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 $ 2.7 seconds *)from beginning of decay of stationary gripper coil voltage to dashpot entry with:
s.
T.,, 2 541'F, and j
b.
All reactor coolant pumps operating.
APPLICiBILITY:
MODE 3.
ACTIQ 1{:
a.
With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above i
limit prior to proceeding to MODE 1 cr 2.
SURVEILLANCE REQUIREMENTS 4.1.3.4 The ro.
drop time of full length rods shall be demonstrated through measur<: ment prior to reactor criticality, a.
For all rods following each removal of the reactor vessel head.
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.
.o
'vor cycle 4 operation only.
Approval pending for operation beyond cycle 4 with this rod drop time.
BEAVER VALLEY - UNIT 2 3/4 1-23 Amendment No. 46
NPF-73 REACTIVITY CONTROL SYSTEMS SHUTDOWN RCD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall Be within the insertion limits specified in the CORE OPERATING LIMITS REPORT.
APPLICABILITY:
MODES 1* and 2*#
ACTIOH:
With a maximum of one shutdown rod inserted beyond the insertion limit, l except for surveillance testing pursuant to Specification 4.1.3.1.1, within one hour either:
a.
Restore the rod to within the limit, or l
b.
Declare the rod to be inoperable and apply Specification 3.I.3.1.
SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion l limit:
a.
Within 15 minutes prior to withdrawal of any rods in control banks A, B, C or D during an approach to reactor criticality, and b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.
See Special Test exception 3.10.2 and 3.10.3
^^
f With Keff 1 1.0 BEAVER VALLEY - UNIT 2 3/4 1-24 Amendment No. 31 i
NPF-73 3/4.2
_ POWER DISTRIBUTION LIMITS BASES The specifications of thic section provide assurance of fuel integrity during condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:
(a) maintaining the minimum DNBR in the core ; the design DNBR limit during normal operation and in short term j transients, and (b) limiting the fission gas release, fuel pellet torperature 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 the LOCA analyses tre met and the ECCS acceptance criteria limit of 2200'F is not exceeded.
The definitions of hot
<-hannel factors as used in those specifications are as follows:
Fo(Z)
Heat Flux Hot Channel Factor, is defined as the maximum local 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.
F" Nuclear Enthalpy Rise Hot Channel Factor, is defined as the d
integral of linear power along the rod with the highest integrated power to the average rod power.
1/4.2.1 AXIAL FIyX DIFFERENCE (AFQ1 The limits on AXIAL FLUX DIFFERENCE assure that the Fn(Z) upper bound envelope times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.
Target flux difference is determined at equilibrium xenon inditions.
The full length rods may be positioned within the core in
-ccordance with their respective insertion limits and should be inserted ear 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 FOWER 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.
Although it is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE withir. the target band about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels.
This deviation will not affect the xenon redistribu+ 1on suf ficiently to change the envelope of peaking factors which may be raached on a subsequent return to RATED THERMAL POWER (with the AFD within tha target band) provided the time BEAVER VALLEY - UNIT 2 B 3/4 2-1 Amendment No. 46
NPF-73 POWFR DTSTRIBUTION LTMITS BASES
&XIAL FLUX DIFFERENCE (AFD)(Continued) duration limit of the deviation is limited.
Accordingly, a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided for operation outside of the target band but within the limits j
specified in the CORE OPERATING LIMITS REPORT for TilERMAL POWER levels l between 50% and 90% of RATED THERMAL POWER.
For THERMAL POWER levels between 15% and 50% of RATED THERMAL PCWAR, d.:viations of the AFD outside of the target band are less signincant.
The penalty of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> actual time reflects this reduced significance.
Prcvisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm.
The computer de "ninos the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for
)
at least 2 of_4 or 2 of 3 OPERABLE excore channels are outside the target i
band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER.
During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, respectively.
Figure B 3/4 2-1 shows a typical monthly target band near the beginning of core life.
3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL i
FACTORS Fn(Z) and F"as The limits on heat flux and nuc. ear enthalpy hot channel factors ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the ECCS acceptance criterie. limit of 2200*F.
Each of these hot channel factors are measurable but will normally only be determined periodically as specified in Specification 4.2.2 and 4.2.3.
This periodic surveillance is sufficient to insure that the hot channel :' actor limits are maintained provided:
a.
Control rods in a single group move together with no individual rod insertion differing by more than 12 steps from the group demand position.
~
b.
Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6.
BEAVER VALLEY - UNIT 2 B 3/4 2-2 Amendment No. 31 l
b l
l
. NPFo73 7%
7%
95 i
90 i
l 85 i
1 I
- 80 b
75 i
l7 65 60 g
y i
TARGET
_ g _55
[
FLUX -
H DIFFERENCE O 50 45.
h40
$ 35 b _30 I.
2 i
w 1 25 i
20 l -
i 15 j
s 10-1[
5 20
-15
-10
-5 0
+5
+10
+15
+20 L
INDICATED AXIAL FLUX DIFFERENCE (%)
FIGURE B 3/42-1 TYPICAL INDICATED AXIAL FLUXDIFFERENCE (AFD)
VERSUS THERhfAL POWER ATBOL BEAVER VALLEY - UNIT 2 83/42-3
=__
NPF-73 POWER DISTRIBUTION LIMITS BASES 1/_L.2_.2 and 3/4.2.3 HEAT FLUX AND HUCLEAR ENTHALPY ff0T CHANNEL FACTORS Fo(Zi and F" (Continued) an i
c.
The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.
j d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE is maintained within the limits.
The relaxation in F" ac a function of THERMAL POWER allows 48 changes in the radial power shape for all permissible rod insertion limits.
F" will be maintained within its limits provided au conditions a throu a d aLeve, are maintained.
v When an Fo measurement de taken, both experimental error and manufacturing tolerance must be allowed for.
5% is the appropriate experimental error allowance for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance.
The specified limit of F" contains an 8% allowance for 45 uncertainties which means that normal, full power, three loop operation will result in F" less than or equal to the design limit 48 specified in the CORE OPERATING LIMITS REPORT.
Puel rod bowing reduces the value of DNB ratio.
Margin has been maintained between the DNBR value used in the safety analyses'and the design limit to oftset the rod bow penalty and other penalties whicM may apply.
The radial peaking reactor Fy(Z) is measured periodically to provide assurance that the hot channel facter, Fo ( Z), remains within its limit.
The F,, Aimit for Rated Thermal Power (F"")
U
-o provided in the CORE OPERATING LIMITS REPORT was determined from expected power control maneuvers over the full range of burnup conditions in the Core.
BEAVER VALLEY - UNIT 2 B 3/4 2-4 Amendment No. 46
NPF-73 POWER DISTRIBUTION LIMITS BASES 1/4.2.4 OUADRANT POWER TILT RATIO The Quadrant Power Tilt Ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are mado during startup testing and periodically during power operation.
The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts.
tin condition The two-hour time allowance for operation with a greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod.
In the event such action does not correct the tilt, the margin for uncertainly on Fn is reinstated by reducing the maximum ellowed power by 3 percent for each percent of tilt in excess of 1.0.
3/4.2.5 DNB PARAMETER The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.
The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR greater than or equal to the design DNBR limit throughout each analyzed transient.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these paramotors through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.
BEAVER VALLEY - UNIT 2 B 3/4 2-5 Amendment Nc. 46
! I NPF-73 I
3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT _ LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in I
operation and maintain DNBR above the design DNBR limit during all normal operations and anticipated transients.
In MODES 1 and 2,
with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In MODE 3, a aagle reactor coolant loop provides sufficient heat removal capability f^s removing decay heat; however, due to the initial conditions assumed in the analysis for the control rod bank withdrawal from a suberitical condition, twc operating coolant loops are required to meet the DNB design basis for this condition II event when the rod control system is capable of control bank rod withdrawal.
In MODES 4 and 5, a single reactor coolant loop or RER subsystem provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.
Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR locps to be OPERABLE.
The cperation of one Reactor Coolant Pump or one RHR pump providou adaquate flow to ensure mixing, prevent stratification and produce gradual reactivity changes durino boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction
- will, therefore, be within the capability of operator recognition and control.
The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 350*F are provided to prevent RCS pressure transiento, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50.
The RCS will be protected against overpressure transients and will not exceed the limits of appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above nach of the RCS cold leg temperatures.
BEAVER VALLEY - UNIT 2 D 3/4 4-1 Amendment No. 46 l
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i NPF-73 BEASTOR COOLANT SYSTEM BASES 3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its safety Limit of 2735 psig. Each safety valve is designed to relieve 345,000 lbs. per hour of saturated steam at the valve set point.
The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.
During operation, all pressurize-code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig.
The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.
Demonstration of the safety valves' Alft settings will occut only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.
Safety valves similar to the pressurizer code safety valves were tested under an Electric Power Research Institute (EPRI) program to determine if the valves would operate stably under feedwater line break accident conditions.
The test results indicated the need for inspection and maintenance of the safety valves to determine the potential damage that may - have occurred after a safety valve has lifted and either discharged the loop seal or discharged water through the valvo.
Additional action statements require safety valve inspection to determine the extent of the corrective actions required to ensure the valves will be capable of performing their intended function in the future.
3/4.4.4 PRESSURIZER The requirement that 150 Kw of pressurizer heaters and their associated controls and emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.
3/4.4.5 STEAM GENERATORS One OPERABLE steam generator in a non-isolated reactor coolant loop provides sufficient heat removal capability to remove decay heat after a reactor shutdown.
The requirement for two OPERABLE steam generators, combined with other requirements of the Limiting Conditions for Operation ensures adequate BEAVER VALLEY - UNIT 2 B 3/4 4-2 Amendment No. 39
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