ML20246J781
| ML20246J781 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 05/08/1989 |
| From: | Colburn T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246J787 | List: |
| References | |
| NUDOCS 8905170172 | |
| Download: ML20246J781 (24) | |
Text
- _ _ _ -
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%,j, UNITED STATES NUCLEAR REGULATORY COMMISSION h
.g WASHINGTON, D, C. 20555
.....f WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-266
. POINT BEACH NUCLEAR PLANT, UNIT NO.1 AMENDMENTTOFACILITYOPkRATINGLICENSE Amendnent No.120 License No. DPR-24
'1.
The Nuclear Regulatory Comission (the Counission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company 1
(thelicensee)datedAugust 26, 1988, as supplemented October 28, November 30, and December 23, 1988; ano as modified January 17, 1988 (sic) cortplies 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 CFP 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 (1) that the activitit.s 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 ar.d 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.
8905170172 890508 i
PDR ADOCK 0500 6
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' 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No, i
DPR-27 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.123, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective. November 1, 1989.
FOR THE NUCLEAR REGULATORY COMMISSION-I f
N.
Timothy G. Colburn, Acting Director Project Directorate III-3 Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation Attachmint:
Charges to the Technical Specifications Date of Issuance: May 8, 1989
~
4a na
.g-p, NUCLEAR REGULATORY COMMISSION
,j WASHINGTON, D. C. 20555
\\...../
i WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.123 License No. DPR-27 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (thelicensee)datedAugust 26, 1988, as suppleiwnted October 28, llovember 30, and December 23, 1988; and as modified January 17, 1988 (sic) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules ano regulations set forth in 10 CFR Chapter I; B.
The facility will oprate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; -
C.
There is reasonable assurance (1) that the activities authorized by this amenement can be: conducted without endangering the health and safety of the publit, and (ii) that such activities will be conoucted 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 Port 51 of the Comission's regulations ano all applicable requirements have been satisfied.
l l
e l
-2 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No.
DPR-24 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appenoices A and B, as revised through Amenditent No.120, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective immediately upon issuance. The Technical Specifications are to be isnplemented within 20 oays from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/(. W Timothy G. Colburn, Acting Director Project Directorate III-3 Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 8, 1989
e 4
ATTACHMENT TO LICENSE AMENDMENT N05.120
- AND123 **
TO FACILITY OPERATING LICENSE N05. DPR-24 AND DPR-27 DOCKET NOS. 50-266 AND 50-301 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by anendment number and contain marginal lines indicating the area of change.
~
REMOVE 1NSERT 15.2.1-1 15.2.1-1 15.2.1-2 15.2.1-2 Figure 15.2.1-1 Figure 15.2.1-1 Figure 15.2.1-2 15.2.3-2 15.2.3-2 15.2.3-3 15.2.3-3 15.2.3-6 15.2.3-6 15.3.1-10 15.3.1-10 15.3.1-19 15.3.1-19 15.3.3-8 15.3.3-8 15.3.3-9 15.3.3-9 15.3.3-10 15.3.10-2 15.3.10-2 15.3.10-11 15.3.10-11 15.3.10-12 15.3.10-12 Figure 15.3.10-1 Figure 15.3.10-1 Figure 15.3.10-3 Figure 15.3.10-3 F1gure 15.3.10-4 Figure 15.3.10-4 15.5.3-1 15.5.3-1 15.5.3-2 15.5.3-2
- For Unit 1, the amendment is effective immediately, with the TS changes to be implenmnted within 20 days.
- For Unit 2, The amendment is effective on November 1,1989.
l 1
15.2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 15.2.1 SAFETY LIMIT, REACTOR CORE Applicability:
Applies to the limiting combinations of thermal power, reactor coolant system pressure, an'd coolant temperature during operation.
Objective:
To maintain the integrity of 'the fuel cladding.
Specification:
1.
The combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure 15.2.1-1.
The safety limit is exceeded if the l
point defined by the combination of reactor coolant system average temperature and power level is at any time above the appropriate pressure line.
Unit 1 - Amendment No. JA 22,EE.120 15.2.1-1 Unit 2 - Amendment No. 2J,29.90,123
j L
Basis The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission l.
' products to the reactor coolant. Overheating.of the fuel cladding is p.e-
. vented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface l
temperature is slightly above the coolant saturatica temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excess cladding temperature 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. This relation has been 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 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% probability at a 95% confidence level that DNB will not occur during steady state operation, normal operational transients, and anticipated transients and is an appropriate margin to DNB for all operating r.onditions.
The curves of Figure 15.2.1.1 are applicable for a core of 14 x 14 0FA. The curves also apply to the reinsertion of previously-depleted 14 x 14 standard fuel assemblies into an 0FA core.
The use of these assemblies is justified by a cycle-specific reload analysis.
The WRB-1 correlation is used to generate these curves.
Uncertainties in plant parameters and DNB correlation predictions are statistically convoluted to obtain a DNBR uncertainty factor.
This DNBR uncertainty factor establishes a value of design limit DNBR. This value of design limit DNBR is shown to be met in plant safety analyses, using values of input parameters considered at their nominal values.
Unit 1 - Amendment No. 86,120 15.2.1-2 Unit 2 - Amendment No. 21,90, 123
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Figure 15.2.1-1 REACTOR CORE SAFETY LIMITS I
. POINT BEACH UNITS 1 AND 2 660-658-2400 PSIA s
640-2250 PSIA 650-2000 PSIA W
,628-E' 6-
,t 1775 PSIA 6IB-I GOO-590-590 8.
.1
.2
.5 4
.6
.6
.7
.8
.9 1.
1.1 1.2 POVER treection or nominet Unit 1 - Amendment No. 86,120 Unit 2 - Amendment No. 2J.90.123
}
(3) Low pressurizer pressure'-
11865 psig for operation at 2250 psia l
primary system pressure 11790 psig for operation at 2000 psia primary system pressure (4) OvertemperatureaT(1*fb 3
1*I b 1
- K (T(1,7 3) - T') (1*Il)+K3 (P-P') - f(al))
<aTo (K 2
3 y
4 2
where oTo indicated AT at rated power, F
=
T average temperature, F
=
T' s
573.9 F
.l pressurizer pressure, psig P
=
2235 psig P'
=
K 1
1.30 y
K
=
0.0200 2
K
=
0.000791 3
25 sec T
=
y T
=
3 seC g
2 sec for Rosemont or equivalent RTD T
=
3 0 see for Sostman or equivalent RTD
=
2 sec for Rosemont or equivalent RTD T
=
4 0 sec for Sostman or equivalent RTD
=
and f(AI) is an even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests, where q and q are the percent power in the top and t
b
' bottom halves of the core respectively, and q
- 9 is total core t
b power in percent of rated power, such that:
(a) for q
~A within -17, +5 percent, f(al) = 0.
t b
(b) for each percent that the magnitude of q ~9b exceeds +5 percent, t
the AT trip setpoint shall be automatically reduced by an l
equivalent of 2.0 percent of rated power.
l l
Unit 1 - Amendment No. AA.EJ.EE.90,120 15.2.3-2 Unit 2 - Amendment No. A9,90,91,123
.+
4a s
$[
(c) for each percent that the magnitude of q ~9b exceeds -17 percent, t
the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of rated power.
_ (5) ' Overpower AT (h) 3 1ATg [K4 -KC
- 1) (1+ 45) T - K6N +T 5) -
~
5 S
1 4
where ATo' indicated AT at rated power, F
=
T average temperature, F-
=
T' 573.9 F-K 5
1.089 of rated power
.i 4
K
=
0.0262 for increasing T 5
~0.0 for decreasing T
=
6 0.00123 for T 1 T' K
=
0.0 for T < T'
=
T
=
10 sec 5
f (AI) as defined in (4) above, 2 sec for.Rosemont or equivalent RTD T
=
3.
0 sec for Sostman or equivalent RTD 2 sec for Rosemont or equivalent RTD T
=
4 0 sec for Sostman or equivalent RTD (6) Undervoltage - 175 percent of normal voltage (7) Indicated reactor coolant flow per loop -
190 percent of normal indicated loop flow (8) Reactor coolant pump motor. breaker open (a) Low frequency set point 155.0 HZ (b) Low voltage set point 175 percent of normal voltage.
Unit 1 - Amendment No. 3.28,86.99,94, 120 15.2.3-3 Unit 2 - Amendment No. 32,99,91.98. 123
}:.
power distribution,'the reactor trip limit, with allowance for errors (2) is always below the core safety limit as shown on Figure 15.2.1-1.
If axial l
peaks are greater than design, as indicated by the difference between top and l
bottom power' range nuclear detectors, the reactor trip limit is automatically reduced (6)(7),
~
The overpower, overtemperature and pressurizer pressure system setpoints include the effect of reduced system pressure operation-(including the effects of fuel densification). The setpoints will not exceed the core safety limits as shown' in ' Figure 15.2.1-1.
The overpower limit criteria is that core power be prevented from reaching a value at which fuel pellet centerline melting would occur. The reactor is prevented from reaching the overpower limit condition by action of the nuclear overporer and overpower AT trips.
The high and low pressyre reactor trips limit the pressure range-in which the reactor operation is permitted. The high pressurizer pressure reactor trip setting is lower than the set pressure for the safety valves (2485 psig) such
+
that the reactor is tripped before the safety valves actuate. The low pressurizer pressure reactor trip trips the reactor in the unlikely event of a loss-of-coolant accident.(4)
The low flow reactor trip protects the core against DNB in the event of either a decreasing actual measured flow in the loops or a sudden loss of power to one or.both reactor coolant pumps.
The set point specified is consistent with the value used in the accident analysis.(8) The low loop flow signal is caused by a condition of less than 90% flow as measured by the loop flow instrumentation. The loss of power signal is caused by Unit 1 - Amendment No. E2,E6,120 15*2*3-6 Unit 2 - Amendment No. EB,90,123
Basis:
+
The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the site boundary will not exceed an appropriately.
small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 500 gpd in either steam generator. The values for l
the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conserva-tive for Point Beach Nuclear Plant.
Continued power operation f.or limited time periods with the. reactor. coolant's.
specific activity greater than 1.0 microcurie / gram Dose Equivalent I-131, but within the allowable limit shown on Figure 15.3.1-5 accommodates possible iodine spiking phenomenon which may occur foitowing changes in thermal power.
Operation with specific activity levels exceeding 1.0 microcurie / gram Dose Equivalent I-131 but within the limits shown on Figure 15.3.1-5 increase the 2-hour thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture.
Reducing T to less than 500 F nonnally prevents the release of activity avg should a steam generator tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that l
excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
A reduction in frequency of isotopic analyses following power changes may be permissible if iustified by the data obtained.
Unit 1 - Amendment No. /J.J02,120 15.3.1-10 Unit 2 - Amendment No. 76 J05,123
G.
OPERATIONAL LIMITATIONS The following DNB related parameters shall be maintained within the limits l
shown during Rated Power operation:
1.
T,yg shall be maintained below 578 F.
l 2.
Reactor Coolant System (RCS) pressurizer pressure shall be maintained:
}
12205 psig during operation at 2250 psia, or 11955 psig during operation at 2000 psia.
3.
Reactor. Coolant System raw measured Total Flow Rate 1181,800 gpm.(See Basis).
Basis:
The reactor coolant system total flow rate of 181,800 gpm is based on an assumed measurement uncertainty of 2.1 percent over thermal design flow (178,000 gpm).
The raw measured flow is based upon the use of normalized elbow tap differential pressure which is calibrated against a precision flow calorimeter at the begin-ning of each cycle.
i I
I l
15.3.1-19 Unit 1 - Amendment No. AA,EJ.BS, 120 j
Unit 2 - Amendment No. A9,90,123
' Assuming the reactor has been operating at full rated power for at least 100 days, the magnitude of'the decay heat decreases as follows after initiating hot shutdown.*
Time After Shutdown Decay Heat % of Rated Power
'l min.
3.6 30 min.
1.55 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1.25 8, hours 0.7 l
48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />
' 0. 4
- Based on ANS 5.1-1979, " Decay Heat Power in Light-Water Reactors" Thus, the requirement for core cooling in case of a postulated loss-of-coolant accident while in the hot shutdown condition'is'significantly redu'ced below the
~
requirements for a postulated loss-of-coolant accident during power operation.
Putting the reactor in the hot shutdown condition significantly reduces the poten-tial consequences of a loss-of-coolant accident, and also allows more free access to some of the engineered safety system components in order to effect repairs.
Failure to complete safety injection system repairs within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of going to the hot shutdown condition is considered indicative of a requirement for major maintenance and, therefore, in such a case, the reactor is to be put into the cold shutdown condition. When the failures involve the residual heat removal system, in order to insure redundant mear.s of decay heat removal, the reactor system.may remain in a condition with reactor coolant temperatures between 500 and 350 F so that the reactor coolant loops and associated steam generators may be utilized for redundant decay heat removal.
However, when the remaining RHR loop must be relied upon for redundant decay heat removal capability, reactor coolant temperatures shall be maintained between 350 F and 140 F.
With respect to the core cooling function, there is some functional redundancy for certain ranges cf break sizes.(2)
The operability of the Refueling Water Storage Tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection oy the ECCS in the event of either a LOCA or a steamline break.
The limits on RWST Unit 1 - Amendment No. 56,120 15*3*3-8 Unit 2 - Amendment No. 7J,123 I
minimum volume and boron concentration ensure that:
(1) sufficient water is available within containment to permit recirculation cooling flow to the core; (2) the reactor will remain subcritical in the cold condition (68 to 212 degrees-F) following a small break LOCA assuming complete mixing of the RWST, RCS, spray additive tank, containment spray system piping and ECCS water volumes with all control rods inserted except the most reactive control rod assembly (ARI-1);
i (3) the reactor will r,emain subcritical in the cold condition following a large break LOCA (break flow area greater than 3 ft2) assuming complete mixing of the RWST, RCS, ECCS water and other sources of water that may eventually reside in the sump post-LOCA with all control rods assumed to be out (AR0); and (4) long term subcriticality is maintained following a steamline break assuming ARI-1 and fuel failure is precluded.
(
The containment cooling function is provided by two independent systems:
(a) fan coolers and (b) containment spray which, with sodium hydroxide addition, provides the iodine removal function.
During normal power operation, only three of the four fan coolers are required to remove heat lost from equipment and piping within the containment.( ) In the event of a Design Basis Accident, any one of the following combinations will provide sufficient cooling to reduce containment pressure:
(1) four fan coolers, (2) two containment spray pumps, (3) two fan coolers plus one containment spray pump.(4) Sodium hydroxide addition via one spray pump reduces airborne iodine activity sufficiently to limit off-site doses to acceptable values.
One of the four fan coolers is permitted to be inoperable for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during power operation.
The component cooling system is different from the other systems discussed above in that the components are so located in the Auxiliary Building as to be acces-sible for repair after a loss-of coolant accident.
One component cooling water pump together with one component cooling heat exchanger can accommodate the heat removal load on one unit either following a loss-of-coolant accident, or during normal plant shutdown.
If during the post-accident phase the component cooling water supply is lost, core and containment cooling could be maintained until repairs were effected.(5)
Unit 1 - Amendment No. EE,76,120 15.3.3-9 Unit 2 - Amendment No. U,ED,123
o L
l l
A total of six service water pumps are installed, only three of which are required to operate during the injection and recirculation phases of a postulated loss of-coolant accident,(0) in one unit together with a hot shutdown condition in the other unit.
References (1) FSAR Section 3.2.1 (2) FSAR Section 6.2 (3) FSAR Section 6.3.2 (4) FSAR Section 6.3 (5) FSAR Section 9.3.2 (6) FSAR Section 9.6.2 Unit 1 - Amendment No.120 l
15.3.3-10 Unit 2 - Amendment No.123
1 B.
Power-Distribution Limits 1.
.a.
'Except during low power physics tests the hot channel factors defined.in the basis must meet the following limits:
F (Z) 5 (2.50)'x K(Z) for P i 0.5 9
P l
F(Z)1 5.00 x K(Z) for P 1 0.5 q
l N -
3H <
.70 x D 3 0.3 (1-P)]
l F
Where P is the fraction of full power at which the core is operating, it(Z) is the function in Figure 15.3.10-3 and Z is the~ core height location.of F.
q
- b. 'Following a refueling shutdown prior to exceeding 90% of rated power and at effective full power monthly intervals thereafter, power distribution maps using the moveable incore detector system shall be made to confirm that the hot channel factor limits are satisfied.
The measured hot channel factors shall be increased in the following way:
(1) The measurement of total peaking factor, F eas, shall be increased by three percent to account for manufacturing tolerances and further increased by five percent to account for measurement error.
(2)Themeasurementofenthalpyrisehotchannelfactor,FfH shall be increased by four percent to account for measure-ment error.
c.
If a measured hot channel factor exceeds the full power limit of Specification 15.3.10.B.1.a. the reactor power and power range high setpoints shall be reduced until those limits are met.
If subsequent flux mapping cannot, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, demonstrate that the full power hot. channel factor limits are et, the overpower and overtemperature 6T trip setpoints shall be similarly reduced and reactor power limited such that Specification 15.3.10.B.1.a above is met.
Unit 1 - Amendment No. 25,f9,E6,120 15.3.10-2 Unit 2 - Amendment No. 30,55,90, 123 l
-l
l
' An upper bound envelope of 2.50 times the normalized peaking factor axial l
dependence of Figure 15.3.10-3 consistent with the Technical Specifications on power distribution control as given in Section 15.3.10 was used in the
)
large and small break LOCA analyses. The envelope was detemined based on f
allowable power density distributions at full power restricted to axial flux difference (AI) values consistent with those in Specification 15.3.10.B.2.
The results of the analyses based on this upper bound envelope indicate's peak clad temperatur"e of less than the 2200 F limit.
When an F measurement g
is taken, both experimental error and manufacturing tolerance must be allowed for.
Five percent is the appropriate allowance for a full core map taken with the moveable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance.
In the design limit of FaH, there is eight percent allowance for uncertainties which means that normal operation of the core is expected to result in a design F i1 OD.08.
H The logic behind the larger uncertainty in this case is that (a) normal perturbations in the radial power shape (i.e., rod misalignment) affect F g, in most cases without necessarily affecting F, (b) while the operator has a g
direct influence on F through movement of rods, and can limit it to the g
desired value, he has no direct control over FAH, and (c) an error in the predictions for radial, power shape which may be detected during startup physics tests can be compensated for in F by tighter axial control, but compensation g
N for F is less readily available.
When a measurement of F is taken, H
3H experimental error must be allowed for and four percent is the appropriate allowance for a full core map taken with the moveable incore detector flux mapping system.
Measurements of the hot channel factors are required as part of startup physics tests, at least each full power month operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based upon measured hot channel factors.
The incore map taken following initial loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns.
The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational anomalies which would, otherwise, affect these bases.
Unit 1 - Amendment No. 6,EE,120 15*3'10-11 Unit 2 - Amendment No. EE,90,123
i Axial Power Distribution The limits on axial flux difference (AFD) assure that the axial power distri-imit bution is maintained such that the F (Z) upper bound envelope of F times q
the normalized axial peaking factor [K(Z)] is not exceeded.during either 1
normal operation or in the event of xenon redistribution following pov:er changes.
This ensures that the power distributions assumed in the large and 1
small break LOCA ana' lyses will bound those that occur during plant operation.
Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer'through the AFD monitor alarm. The computer determines-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.AFD limits and the thermal power is greater than 50 percent of Rated Power.
Unit 1 - Amendment No. 26,49,86,120 15.3.10-12 Unit 2 - Amendment No. 3J,EE,90, 123
e l
FIGURE 15.3.10-1 CONTROL BANK INSERTION LIMITS POINT BEACH UNITS 1 AND 2 3
(2 1% '
( 68.5% )
3o3
/
l BAfK 8 INSERTION l
/
/
/
w
/
/
/
(81) so
/
??)
l70
/
/
G
/
/
/
/
I BR C If6ERTION l
/
so b
/
/
~
hso
/
/
j
/
/
E.
/
/
b
/
/
B R D INSERTION l 3 /
/
(22)
/
/
/
[
10
/
(28.5 %)/
0 S 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 power LEWL (% OF RATED POWER)
Unit 1 - Amendment No. 75,AS,EE,EE, 120 Unit 2 - Amendment No. 30,EE,90,93, 123
r '.;
1 FIGURE 15.3.10-3 POINT BEACH UNITS 1 AND 2 HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE 1.2 (0.0,1.0)l (6.0,1.0)l 1.1 (12.0. 92) l 3,
.9
~
.8
.7
.g.e
.5
.4
.3
.2
.1 00 1
2 3
4 5
6 7
8 9
10 11 12 I
coas attorr (FT)
Unit 1 - Amendrt.ent No. JA,22,EE,120 Unit 2 - Amendment No. JE,29,99,123
__m-__
________m.m.
+
FIGURE 15.3.10-4 FLUX DIFFERENCE OPERATING ENVELOPE POINT BEACH UNITS 1 AND 2 4
(-8.1001!
19,1001l'
?
95
!i i
(-8.90)l (9,901 l' I
85 I
\\
l,' lfl, l l' l
h i
i i
l l:
i i
r 1
i h
i
~
l 1
ii Ii l
l
'lj
!i l
l
. li li i!
il I
ll
!l l
Il i
l! l l. !
- l 1
3 j i
l (
i 75 7
,l l
i!l l
l l
l ili I'
l l
I L
I li li i
70 l,
j l
l
!i jl l
ll ll l
I I
I 65 l ',
i 4, !
i i
g
- L
[
!l l
/
i 55
/
l i
I I
i
(-29,50)l i
l i
I t27,50)l S 40 30 20 10
-s o
5 to 15 20 25 30 35 40 DELTA I (f.)
J l
Unit 1 - Amendment No. $5,120 j
Unit 2 - Amendment No. 90, 123 l
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- 'L 15~. 5. 3 ~
- REACTOR' p
Applicability Applies to the reactor core, Reactor Coolant System,~and Emergency Core' Cooling
' Systems.
Objective To deffne'those design features which are essential in providing for safe system operation.
4 Specifications A.
Reactor Core 1.
General The uranium fuel is in the form of slightly enriched uranium dioxide pellets.
The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. 'The reactor core is made up of 121 fuel assemblies.
Each fuel assembly nominally contains 179 fuel rods (1)
Where safety limits are not violated, limited substitutions of fuel rods.by filler rods consisting of Zircaloy 4 or stainless steel, or by vacancies, may
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be made to replace damaged fuel rods if justified by cycle specific reload analysis.
2.
Core A reactor core is a core loading pattern containing any combination of 14x14 0FA and 14x14 upgraded 0FA fuel assemblies.
The core may also contain previously depleted 14x14 standard fuel assemblies.
The use of previously depleted 14x14 standard fuel assemblies will be justified by a cycle specific reload analysis.
Unit 1 - Amendment No. 22,28,EE J05,120 15.5.3-1 Unit 2 - Amendment No. 32,98,JJJ,123
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3.
Burnable absorber and/or water displacer rods are incorporated for reactivity and/or power distribution control. ' The burnable absorber rods consist of borated pyrex glass clad with stainless steel (4)
The water displacer rods are empty burnable absorber rods containing no pyrex glass.
Another type of burnable absorber may consist of a thin coating of zirconium diboride on the radial surface of selected fuel rod pellets.
4.
There are 33 full-length RCC assemblies in the reactor core. The j
full-length RCC assemblies contain a 142-inch length of silver-indium-cadmium alloy clad with the stainless steel.
5.
Neutron source assemblies are used to prov.ide a requi. red minimum count rate during startup operations. The core contains at least l
two such assemblies, each containing four. source rodlets comprised of a uixture of. antimony and beryllium.
6.
Peripheral power suppression assemblies (PPSA) are used to reduce neutron fluence at the welds in the beltline region of the reactor vessel.
Peripheral fuel assemblies may contain PPSAs, which utilize part-length hafnium absorber rods in the assembly guide tubes.
'l B.
The design of the Reactor Coolant System complies with the code
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requirements.(0) 2.
All high pressure piping, components of the Reactor Coolant System and their supporting structures are designed to Class I requirements, and have been designed to withstand:
a.
The design seismic' ground acceleration, 0.06g, acting in i
the horizontal and 0.04g acting in the vertical planes simultaneously, with stresses maintained within code allowable working stresses.
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