ML20248C395
| ML20248C395 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 09/28/1989 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Duquesne Light Co, Ohio Edison Co, Pennsylvania Power & Light Co |
| Shared Package | |
| ML20248C399 | List: |
| References | |
| DPR-66-A-144 NUDOCS 8910030445 | |
| Download: ML20248C395 (17) | |
Text
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION p
WASHINGTON, D. C. 20555
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DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY DOCKET NO. 50-334 BEAVER VALLEY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 144 License No. DPR 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applicatio-for amendment by Duquesne Light Company, et al.
(the licensee) ded May 9,1989 and supplement dated July 21, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized 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 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.
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b.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the atta:hment to this license amendment, and paragraph 2.C.(2) of Facility Operating License
(
No. OPR-66 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained.in Appendix A, as revised l
through Amendment No. 144, are hereby incorporated in the license.
The licensee shall opercte the facility in accordance with the Technical Specifications.
l 3.
This license amendment is effective on issuance.
FOR THE NUCLEAR REGULATORY COMMISSION W
Jo
. Stolz, Director Project Directorate I-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:Septeraber 28, 1989
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ATTACHMENT TO LICENSE AMENDMENT NO. 144 FACILITY OPERATING LICENSE NO. DPR-66 DOCKET NO. 50-334 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 2-4 B 2-4 B 2-6 B 2-6 3/4 1-22 3/4 1-22 3/4 2-5 3/4 2-5 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 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 8 3/4 2-6 B 3/4 2-6 B 3/4 4-1 B 3/4 4-1 B 3/4 9-4 B 3/4 9-4 l
l 2.1 SAFETY LIMITS I
BASES 2.1.1 REACTOR C, ORE The restrictions of this safety limit prevent overheating of the fuel l
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 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 WRB-1 correlation.
The WRB-1 DNB correlation has been l
developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat flux
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 established 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).
In meeting this design basis, uncertainties in nuclear and thermal parameters, and fuel fabrication parameters were combined statistically with the DNB correlation uncertainties to determine the plant DNBR uncertainty and establish the design DNBR limit such that there is at least a 95% probability witn 95% confidence level that
{
the minimum DNBR for the limiting fuel rod is greater than or equal to the DNBR limit.
For this application, the design DNBR limit is 1.21.
This DNBR value must be met in plant safety analyses using nominal values of the input parameters that were included in the DNBR uncertainty evaluation.
In addition, margin has been maintained in the design by meeting a
safety analysis DNBR limit of 1.33 in performing safety analyses.
The curves of Figure 2.1-1, show the loci of points of THERMAL POWER, l
Reactor Coolant System pressure and average temperature for which the 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.
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BEAVER VALLEY - UNIT 1 B 2-1 Amendment No. W,144
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SAFETY LIMITS BASES N
The curves are-based on an enthalpy hot channel factor, F H' Of 1.62 and a
reference cosine with a peak of 1.55 for. axial power i
p shape.
An allowance is included for an increase in F
at H
reduced power based on the expression:
A F[H $ 1.62 (1 + 0.3 (1-P)]
l 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 f(aI) function of the Overtemperature trip.
When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature aT trip will reduce the setpoint to provide protection consistent with core safety limits.
2.1.2 REACTOR COOLA.NT SYSTEM PRESSURE The restriction of this safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110%
(2735 psig) of design pressure.
The Reactor Coolant System piping and fittings are designed to ANSI O 31.1 and the valves are designed to ASA 16.5 which permit a maximum transient' pressure of 120% (2985) psig of component 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.
BEAVER VALLEY - UNIT 1 B 2-2 Amendment No.%144
J LTMITING SAFETY SYSTEM SETTINGS L
- BASES The ' Power Range Negative Rate trip provides protection to ensure that the minimum DNBR is maintained above the design DNBR limit for l
control rod drop accidents.
At high power a single or multiple rod drop accident could cause flux peaking which, when in conjunction with nuclear power beine 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.
For those transients on which reactor trip on power range negative rate trip is not postulated, it is shown that the minimum DNBR is greater than the design DNBR limit.
Intermediate and Source Ranos. Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor start-up.
These trips provide redundant protection to the low setpoint trip of the Power Ran'ge, Neutron Flux channels.
10+g Source Range Channels will initiate a Th reactor trip at about 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.
No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses;
- however, their functiona; capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.
Overtemperature LT The Overtemperature L T 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 delays from the core to the temperature detectors (about 4 seconds), 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 piping delays from the core to the loop temperature detectors.
With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown on Figures 2.1-1, 2.1-2 and 2.1-3.
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 notations in Table 2.2-1.
l BEAVER VALLEY - UNIT 1 B 2-4 Miendment No. lS.144 f
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' LIMITING SAFETY SYSTEM SETTINGS
, BASES through the pressurizer safety valves.
No credit was taken for operation of this trip in the accident analyses;
- however, its functional capability at the this specification specified trip setting is required by
(
Protection System.
to enhance the overall reliability of the Reactor Loss of Flow The loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.
Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 90% of nominal full loop flow.
Above 31% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90%
of nominal full loop flow.
This latter trip will prevent the minimum value of the DNBR from going below the design DNBR limit I
during normal operational transients and anticipated. transients when 2
loops are in operation and the overtemperatureaT trip setpoint is adjusted to the value specified for all loops in operation.
With the overtemperature a T trip setpoint adjusted to the value specified for 2
loop operation, the P-8 trip at 66% RATED THERMAL POWER with loop stop valves open and at 71%
RATED THERMAL POWER with a loop stop valve closed will prevent the minimum value of the DNBR from going below the design DNBR limit during normal operational transients and l
anticipated transients with 2 loops in operation.
Steam 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 there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system.
Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam /Feedwater Flow Mismatch in coincidence with a
Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the
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functional capability of the specified trip settings and thereby anhence the overall BEAVER VALLEY - UNIT 1 B 2-6 Amendment No.144
1
- REACTIVITY CONTROL SYSTEMS ROD' DItOP TIME.
e LIMITING' CONDITION FOR OPERATION q
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3.1.3.4 The individual full length (shutdown and-control) rod drop time from the fully withdrawn position shall be 5 2.7 seconds from l
beginning. of-decay of stationary gripper coil voltage to dashpot entry with:.
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T 2 541*F, and 1
avg b.
All reactor coolant pumps eperating.
l APPLICABILITY:
MODE 3.
ACTION:
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 limit prior to proceeding to MODE 1 or 2.
b.
With the rod drop times within limits but determined with 2 reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to:
1.
5 61%
of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are open, or 2.
S 66%
of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are closed.
SURVEILLANCE REQUIREMENTS n
4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement 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.
BEAVER VALLEY - UNIT 1 3/4 1-22 Amendment No.'11, 144
POWER DISTRIBUTION LIMITS
- HEAT FLUX HOT CHANNEL FACTOR-Fg (Z)
LIMITING CONDITIONS FOR OPERATION i
3.2.2 Fg(Z) shall be limited by the following relationships:
Fg(Z) 1 2,40 [K(Z)) for P > 0.5 l
P Fg(Z) s.[4.80) [K(Z)] for P s 0.5 l
where P = THERMAL POWER RATED THERMAL POWER and-K(Z) is the function obtained from Figure 3.2-2 for a given core height location.
APPLICABILITY:
Mode 1 ACTION:
With Fg(Z) exceeding its limit:
a.
Reduce THERMAL POWER at least 1% for each 1% Fo(Z) exceeds the limit 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 a T Trip setpoints have been reduced at least 1%
for each 1%
Fn(Z) exceeds the limit.
The overpower aT Trip Setpoint Peduction shall be performed with the reactor suberitical.
b.
Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; THERMAL POWER may then be increased provided Fo(Z) is demonstrated through incore mapping to be within its limit.
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I BEAVER VALLEY - UNIT 1 3/4 2-5 Amendment No. S. 144
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o l.2 (o.o.l.0)
(6.o.l.0)
N l.O O
(no.s.o.s4) g C
O.8 LU N
_J 0.6 g
(iz.o.o.s4) moZ O.4 I
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O.2 x
0.0 0
2 4
6 8
10 12 14 CORE HEIGHT (PT)
Figure 3.2 2. K(Z)- Normalized F (Z) as a Function Q
of Core Height Amendment No. M.
BEAVER VALLEY - UNIT 1 3/4 2-7
W [PdWER DISTRIBUTION LIMITS g.' ',
EUCY MR EFTHALPY HOT CHANNEL FACTOR - FN LIMITING CONDITION FOR OPERATION
-"g
~ 3.2'.3' gHshallbelimitedbythefollowingrelationship:
F[H $ 1.62 (1 + 0.3 (1-P))
l where Pl= THERMAL POWER RATED THERMAL POWER APPLICABILITY:
MODE 1-ACTION:
With FN exceeding its limit:
3 a.
Reduce THERMAL POWER to less than 50% of RAi'ED THERMAL POWER within 2. hours and reduce the Power Range Neutron Flux-High Trip Setpoints to 5 55% of RATED THERMAL POWER within the
.next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, b.
Demonstrate thru in-core mapping that F[H is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit or reduce THERMAL 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 />, and c.
Identify and correct the cause of the out of limit condition prior' to increasing THERMAL POWER,[ subsequent POWER proceed provided
.that F
is demonstrated OPERATION may.
in-core mappingtobe'withinitsYimitatanominal through ~ RATED THERMAL POWER prior to exceeding this THERMAL 50%
of
- power, at a
nominal-75%
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.
BEAVER VALLEY - UNIT 1 3/4 2-8 Amendment No. V,144
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T/4.2' POWER DISTRIBUTION LIMITS
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. BASES
.The-specifications of this 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 1 the design DNBR limit during normal operation and.in short l
term transients, and (b) 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 the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.
The definitions of hot channel factors as used in these specifications are as follows:
F (Z)
Heat Flux Hot. Channel
- Factor, is defined as the maximum g
local heat flux on the surface of a
fuel rod at core e3evation 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.
3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)
.The limits on AXIAL FLUX DIFFERENCE assure that the F (Z) upper 9
bound envelope of 2.40 times the normalized axial peaking factor is l
not exceeded 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 inserti9n 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 i
associated core burnup conditions.
Target flux differences for other THERMAL POWER levels are BEAVER VALLEY - UNIT 1 B 3/4 2-1 Amendment No. 1,7%144
_ _ = _ _ _ _ _ _ _ _ _ -
.. _ _ - _ - _ _ _ _ - - = _ _ _ - - - _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _ - _ _ -
. ' POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL ZAFTORS - Fg(Z) andF{Ji The limits on heat flux and nuclear enthalpy hot channel factors ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded 2) in the event of a LOCA the peak fuel clad temperature will not exceed the ECCS acceptance criteria limit of 2200*F.
Each of these hot channel factors are measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.
This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided:
a.
Control rod in a
single group move together with.no individual rod insertion differing by more than i 12 steps from the group demand position.
b.
Control rod groups are sequenced with overlapping groups as described in specification 3.1.3.5.
c.
The control rod insertion limits of Specifications 3.1.3.4 and 3.1.3.5 are maintained.
d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE is maintained within the limits.
The relaxation in F5 as a
function of THERMAL POWER allows H
changes in tge radial power shape for all permissible rod insertion limits.
F will be maintained within its limits provided AH conditions a thru d above, are maintained.
When an F
measurement is
- taken, both experimental error and g
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.
N The specified limit of F
contains an 8%
allowance for meags thak
- normal, uncertainties which full
- power, three loop H $ 1.62/1.08.
g operation will result in F i
BEAVER VALLEY - UNIT 1 B 3/4 2-4 41endment No.'14,144
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POWER DISTRIBUTION LIMITS
(%-
BASES
(
Fuel rod bowing reduces the value of DNB ratio.
Margin has been raintained between the DNBR value used in the safety analyses (1.33)
I and the design limit (1.21) to offset the rod bow penalty and other penalties which may apply.
The radial peaking factor F
(2) is measured periodically to provide assurance that the*Not channel
- factor, F (Z),
remaggg within its limits.
The F
o limit for Rated ThermaI Power F'Y as provided in the Rad!El Peaking Factor Limit Report per specification 6.9.1.14 was determined from expected power control maneuvers over the full range of burnup conditions in the core.
3/4.2.4 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 made during start-up 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.
The two-hour time allowance for operation with a tilt condition 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 uncertainty on F
is reinstated by reducing the maximum allowed g
power by 3 percent for each percent of tilt in excess of 1.0 1
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l BEAVER VALLEY - UNIT 1 B 3/4 2-5 Amendment No. M. W S.
l
- 144,
.IFOWER DISTRIBUTION LIMITS u,
BASES 3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the maintained within the normal steady state envelope of parameters are 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 DNBP 1 the design DNBR limit throughout each analyr.ed transient.
l 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 parameters 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 pr. ovide 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 1 B 3/4 2-6 Miendment No 'SJ 44
,3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACff0R COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the design DNBR limit during all l
normal operations and anticipated transients.
In Modes 1 and 2, with one reactor coolant loop not in operation, THERMAL POWER is restricted to 5 31 percent of RATED THERMAL POWER until the Overtemperature oT trip is reset.
Either action ensures that the DNBR will be maintained above the design DNBR limit.
A loss of flow in two loops will ~cause a reactor trip if operating above P-7 (11 percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a
reactor trip if operating above P-8 (31 percent of RATED THERMAL POWER).
In MODE 3,
a single reactor coolant loop provides sufficient heat removal capability for removing decay heat;
- however, due to the initial conditions assumed in the analysis for the control rod bank withdrawal from a suberitical condition, two operating coolant loops are required to meet the DNB design basis for.this Condition II event.
In MODES 4
and 5,
a single reactor coolant loop'or RHR 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 loops to be OPERABLE.
The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during 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 275'F are provided to prevent RCS pressure transients, 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.
BEAVER VALLEY - UNIT 1 B 3/4 4-1 Miendment No. 42,144
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' REFUELING OPERATIONS t,
BASES The results of the spent fuel pool criticality analysis (August 1986) for Westinghouse STD/ Vantage SH and OFA/ Vantage 5 fuel in three of l
four storage locations show that there is more than 0.3% margin to the k,ff limit of 0.95 with all uncertainties included.
Based on the sensitivity study completed with this analysis, an increase in the maximum allowed enrichment for fuel stored in the spent fuel storage racks from 4.00 to 4.05 W/o will increase the maximum rack k
f by less than 0.002.
Therefore, with Westinghouse 17 x 17 SkDfVantage 5H and OFA/ Vantage 5 fuel enriched at 4.05 W/o stored in l
the spent fuel racks in three of four storage locations and with all of the assumptions and conservatism presented in the criticality analysis, the maximum rack k,ff will be less than 0.95.
3/4.9.15 CONTROL ROOM EMERGENCY HABITABILITY SYSTEMS The OPERABILITY of the centrol room emergency habitability system ensures that the control room will camain habitable for operations l
personnel during and following all credible accident conditions.
The ambient air temperature is controlled to prevent exceeding the allowable equipment qualification temperature for the equipment and instrumentation in the control room.
The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent.
This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A",
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