ML20148J692
| ML20148J692 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/09/1978 |
| From: | Cavanaugh W ARKANSAS POWER & LIGHT CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20148J700 | List: |
| References | |
| 1-118-4, NUDOCS 7811160062 | |
| Download: ML20148J692 (24) | |
Text
.
P i
l ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 LITTLE ROCK. ARKANSAS 72203 (501)371-4422 November 9, 1978 WILLIAM CAVANAUGH ll1 Execudve Director Generation & Construction 1-118-4 Director of Nuclear Reactor Regulation 2
ATTN:
Mr. R. W. Reid, Chief Operating Reactor Branch #4 i
U. S. Nucicar Regulatory Commission Washington, D. C, 20555 Subj ect -
Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Proposed Technical Specifications (File:
1511.1) i 2
Gentlemen:
Enclosed is a proposed change to the Arkansas Nuclear One - Unit 1 (ANO-1)
Technical. Specifications which will incorporate parameter changes necessary for cyc1c'4 operation.
Twenty copies of the ANO-1 Cycle Reload Report have been included as supplementary information, j
An increased cycle length (387 EFPD) has prompted us to allow additional time for your review. Our third refueling outage is scheduled to begin in March, 1970.
Therefore, we request that this proposal be approved and returned no later than the second week in April in order to allow time for implementation prior to cycle 4 startup.
Pursuant to 10CFR170.12, we have detennined that this proposed change involves a complex safety issue and is, therefore, a Class IV amendment requiring remittance of $12,300. A check for that amount is enclosed.
)
Very truly yours, 4
lw-William Cavafiaugh II Executive Director, Gelteration 3
and Construction
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Enclosure 78113 g ocsg,
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MEMf3ER MIDOLE SOUTH UTlWTIES SYSTEM
l Using a local quality limi t of 22 percent at the point of minimun DNBR as a \\
ba.s2s for curve 3 of F2 cure 2.1-3 is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR, The DNBr. as calculated by the BAW-2 cora elation continually increases from point of minimun. D'4BR, so that the exit DNBR is always higher and is a funct ion of the prer.sure.
l The maximum thermal power for three pump operation is 85.4 percent due to a oower level trip produced by the flux-flow ratio (74. 7 percent flow x 1.057 =
78.9 percent power) plus the ma>> mum calibration and inst rumentation error.
The maximam thermal power for other reactor coolant pump conditions is pro-duced in a similar manner.
For each curve of Figure 2.1-3, a pres sure-temperature point above and to the left of the curve would result in a DNBR great er than 1.3 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor cool ant pump situation.
Curves 16 2 of Figure 2.1-3 are the most restrictive because any pressure / temperature point above and to the lef t of this curve will be above and to the left of the ot her curve.
REFE RENCES (1) Correlation of Crit ical Heat Fl ux in a Bundle Cooled by Pressurized Water, BAW-10000A, May, 1976.
(2)
FS AR, Sect i on 3. 2. 3.1.1. c I
a g.
'i
1 120 UNACCEPTABLE 18,112 25.112 OPERATION 1
ACCEPTABLE 100 4 PUMP 45,95
-45,92 OPERATION 18,85.4 25,85.4 r
80 ACCEPTABLE 2
3 & 4 PUMP 45,68.4 45,65.4 OPERATION 60 25,58.5 18,58.5 ACCEPTABLE 3
41.3,41.3 2,3 & 4 PUMP 43.1,43.1 40 OPERATION 20 I
i e
a 1
60
-40 20 0
20 40 00 Reactor Power Imoalance, %
CURVE REACTOR COOLANT FLOW (GPM) 1 374,880 2
280,035 104,441 CORE PROTECTION SAFETY LIMITS Figure 2.1-2 9b
o d
I 2.3 LIMITING SAFm SYSTD4 Sm1NGS, PROTECTIVE INSTBLHENTATION Applicability Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flov, number of pumps in operation, and high reactor building pressure.
Objective To provide automatic protection action to prevent any combination of process variables from exceeding a safety limit.
Specification 231 The reactor protection system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and Figure 2 3-2.
Bases The reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which vill cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the degree that a safety limit may be reached.
The trip setting limits for protection system instrumentation are listed in
[
Table 2 3-1.
The safety analysis has been based upon these protection system instrumentation trip set points plus calibrLtion and instrumentation errors.
Nuclear Overpover A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be i
detected by pressure and temperature measurements.
l During normal plant operation with all reactor coolant pumps operating, i
reactor trip is initiated when the reactor power level reaches 105.5 percent I
of rated power. Adding to this the possible variation in trip set points l
due to calibration and instrument errors, the maximum actual pover at which I
a trip would be actuated could be [12%, which is the value used in the safety analysis.
A.
Overpower trip based on flov and imbalance The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to i,
accommodate the most severe thermal transient considered in the design, the loss-of-coolant flov accident from high power. Analysis has demon-I strated that the specified power to flow ratio is adequate to prevent a DNBR of less than 1.3 should a low flow condition exist due to any ele-ctrical malfunction.
Am dment No.
21
/.
i L
.a I
'The power level, t rip set poi nt produced by the power-to-flow ratio provides bnt h high powc r level and low flow prot ec tion in the event 1
the reactor power level ineresses or the react or coolant. flow rate
' decreases. The power level t rip set ' point produced by the power to flow rat io provides overpower DNii prot ect 2on for all modes of pump operation, For eve ry flow rate tA cre is a ma2 2mtn permi ssible power level, anJ for every power level there is a m2 nimum perm 2 ssible low flow rate. Typical power level and lov flow rat e combinat ions for the pump situations of Table 2,3-1 are as follows:
1.
Trip would occur when four react or coolant pi.mps are operat ing 2f power is 10$.7 percent and reactor flow rat e is 100 percent or flow r at e i s 9.4.0, p e rc e nt an d pow e r level is 100 percent.
2.
Trip would occur when three react or cool ant panps are operating if power is 78.9 percent and reactor flow rate is 74.7 percent or flow rate is 70.9 percent and power level is 75 percent.
3.
Trip would occur when' one reactor cool ant pump is operating in each loop (tot al of two pumps operat ing) if the power is 52.0 Percent and reactor flow rate as 49.2 percent or flow rate is 46.3 percent and t he pow e r level 2s 49.0 percent.
l The flux / flow ratios account for ths neuims. cal ibrat ion and ins t rment at ion e rrors and the maa imum va ri at n on f rom the aserage value of the RC flow signal in such a manner that the react or prot ect ive sy stem rec eives a conservative indication of the RC flow.
No penalty in reactor coolant flow through the core was taken for an open core vent valve because of the core vent valse surveillance program during each refueling outage.
For safety analysis calculations the maximum call-bration and inst rumentation errors for the power level were used.
The power-imbalance boundaries are established in order to prevent reactor thermal i bni t s from being ex ceeded.
Th e s e t he rma l limits are either power peaking kW/ft limit s or DNBR limits. The ' reactor power Unbalance (power in top half of core minus power in the bot tom hal f of core) reduces the power level trip produced by the power-to-flow ratso so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the power, level trip associated reactor power-to reactor power imbalance boundaries by 105i7 percent for a 1 percent flow reduction.
B.
Pump monitors In conjunction with the power imbalance / flow trip, the pump moni-itors prevent the minimum core DNBR from decreasing below 1.3 by trip-
' ping the reactor due to the loss of reactor coolant pump (s). The pump monitors also restrict the power level for the number of pumps in operation.
.s C.
RCS Pressure j.
During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip set point is reached 11 before the nuclear overpower' trip set point. The trip setting limit l
' Amendment'No. p( 31 e-,
- ---c.-
-w i.
THERMAL POWER LEVEL, 3
. 120 UNACCEPTABLE OPERATION 10,105.7 105.7 Mg = 0.747
" 100 g2 =
-0. 913 ACCEPTABLE 31,90 4 PUMP 30,92 OPERATION 78.95 10,78.95 80 15,78.95 ACCEPTABLE 3 & 4 PUMP 30,65.25
-31,63.25 OPERAIl0N 60 10,52.0 52 15,52.0 40 31,36.3 ACCEPTABLE 30,38.3 2,3 & 4 PUMP OPERATION 20 g
S S
E
+
+
1)
Il ll Il C
E E
E i
i I
I
-60 40 20 0
20 40 60 Power imoalance, 5 PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINTS Figure 2.3-2 14 9
. ~. -. - -
Table 2.3-1 Reactor Protection Svsten Tria Set:ing ti-its One Reactor Coolant Pu=p Four Reactor Coolant rumps Three neactor Coolant Pu ps Operating in Each toop Operating (Nortnal Ope r a t i n g (W-i n 21 (t -inal 0,'e ra:i ng shut.f m n Coerating Power - 1005)
_ Operating Power - T. )
Power - 4 9 *. )
E vpay s Nuclear power,1 of 105.5
- los.s 10s.5 s.0(3) rated, max Noc; ear power based on 1.057 times flow minus 1.057 t imes flow uinus 1.057 t 4.v s flow nimis
,,,g, flow (2) and.2m5alaare, redection due to reduction due to r eit e :, d:e *o
\\ of rated, ra t i.-b a l a nc e (s) i.-$a l a n c e ( s )
s-M l a w M s)
Nusi ca r powe r ba sed c.,n NA NA gypassed purp conttors) i of (4
rated, rax hi"h Teactor Coolant 2353 235$
2355 1720 si s tco pres st.re, psig, tu s G-Low reactor coolant sys-1300 1900 1800 P.ypu s e d tem pressure, psig, nin Variable low reactor (11.75Tou t-5103) ( 1)
(ll.75 Tout-5IUM(I)
- (11.75fout-5103}ll)
By, a s e. e d c o'a l a '. t systers pressure, psig, ain 2eac:nr coolant teay, 619 619 619 6 19 F, ram lii gh reactor buildin 4 (18.7 psia) 4(13.7 psia) 4 (13. 7 psia) 4(13. 7 psi pressure, pssg, rax (I) T is in degrees Fahrenheit (F).
(3) Autonatically set when other se;,ent s of the CPS (as specified) are bg:ssed out (2) Reactor coolant syste2 flow, %.
(4) The purp mon it or s a l so p roduc e a t r ip on: (a) loss of two reactor coolant pu.ps in one reactor coolant loop, and (b) loss of cne or tso reactor coolant pu.mps during two-purp operation.
,-w
- -- - - ~ ~ ~ ~
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6 Mininten volunes (including a 10*. sa fet y fac t or) ~ as speci fied by Figure 3. 2 1 for the horic acid addition tank or 35,659 gallons of 2270 ppm boron as boric acid solution in the horated water storage tank (3) will cach satisfy this requirement. The specification assures that adequate supplies are available whenever the reattor is heated ahose 200 F so that a singic failure will not prevent boration to a cold condition.
The minimum volumes of boric acid solu-tion given include the boron necessary to account for xenon decay.
The principal method of adding boron to the primary system is to pump the con-centrated boric acid solution (8700 ppm boron, minimum) into the makeup tank using the 25 gpm boric acid pumps.
The alternate method of addition is to inject boric acid from the boro-ated water storage tank using the makeup pumps.
Concentration of boron in the boric acid addition tank may be higher than the concentration which would crystalli:t at ambient conditions.
For this reason and to assure a flow of boric acid is availabic when needed this tank and its associated piping will be kept 10 F above the crystallization temperature for the concentration present. Once in the makeup system, the concentrate is suffi-ciently well mixed and diluted so that normal system temperatures assure borie acid solubility.
REFERENCES (1)
FSAR, Section 9.1; 9.2 (2)
FSAR, Figure 6-2 (3)
FSAR, Section 3.3 i
I
'i 6
i l
Amendment No. 33 _,
Figurc 3.2-1 BORIC ACIO ADDITION TANK VOLUME AND CONCENTRATION REQUIREMENTS VS RCS AVERAGE TEMPERATURE 7000 (579 F,6138 gal) 6000 5000
~
w" OPERATION AB0VE AND T0 (532 F,4427 gal)
(579 F,4392 gal) g THE LEFT OF THE CURVES I
ACCEMABLE 4000 (500 F,3826 gal)
~
(532 F,3166 gal) 5 E
3000 E
a j
8700 PPM BORIC ACIO u
h
~
(400 F,2103 gal) 12000 PPM BORIC ACIO 1000 (300 F,877 gal)
(300 F,628 gal)
/
I I
t
't 0
200 300 400 500 000 RCS Average Temperature (F) h.
35a
~
(3. If c control -rod in the regulat ing or amici power shaping groups i s dec la red inope rable per Specs fiu.on 4. 7. 3. 2. operat ion' above 60 percent ni t he thermal power allowable for the reactor coolant pump enabinat ion may cont inue provided the rods in the group are posit loned such that the rod that was declared inoperable is con-tained within allowable group average position limits of Specifica-tion 4.7.1.2 and the withdrawal limits of Specifi cat ion 3. 5.2.5. 3.
3.5.2.3 The worth of single inserted control rods during criticality are limited by the rest rictions of Specification 3.1.3.5 and the Control Rod Position Limit s defined in Specification 3.5.2.5.
3.5.2.4 Quadrant tilt:
1.
Except fnr physics tests. if quadrant tilt exceed' 4.92% Power shall be reduced immediately to below the power level cutoff (see Figures
~
- 3. 5. 2. l A and 3. 5. 2-1 B ).
Mor eover, the power level tutof f value shall be reduced 2% for each 14 tilt in excess of 4.92% tilt.
For less than 4 pump operat ion. thermal power shall be reduced 2% of the thermal power al towable for the reactor coolant pump combin-ation for each 1% t ilt in excess of '4.92%.
2.
Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. the quadrant power tilt shall he reduced t.o l e s s t ha n 4.9'2% ex cept for physics test s, or the following adjust-ments in setpoints and limits shall he made:
a.
The protect ion system mnimum allowable set points (Figure 2.3 2) shall be reduced 2% in power forcaeh II( tilt.
b.
The cont rol rod group 'and APSR withdrpwal limits 'shall 'be f
reduced 2% in power for each 1% tilt in excess of 4.92%.
\\
e.
The operational imbalance limit s shall be reduce'd 2% in power for each 11 tilt in excess of 4.92%.
3.
If quadrant tilt is in excess of 25%, except for physics test s or diagnostic testing, the reactor will be placed in the hot shutdown condition.
Diagnostic testing during power operat ion with a quad-rant powe r t i l t is permitted provided the thermal power allowable for the reactor coolant pump combination is restricted as stated in 3.5.2.4.1 above.
4.
Quadrant tilt shall be monitored on a sinimum icequency of once every two hours during power operation above 15% of rated power.
3.5.2.5 Control rod positions:
l.
Technical Specification 3.1.3.5 (safety rod withdrawal) does not prohibit the exercising of individual safety rods as required by "lable 4.1-2 or apply to inoperable safety rod limits in Technical Spec i fi ca t i on 3. 5. 2. 2.
2.
Operat ing rod group overlap shall be 20% 15 between two sequential groups, except for physics tests.
Amendment No. J, 3i 47 1
e i
4 3.
Except for physics tests or exercising control rods, a) the control rod withdrawal limits are specified on Figures 3.5.2-1A, 3.5.2-1B and 3.5.2-lC for four pump operation and, on Figures 3.5.2-2A, 3.5.2-2B and 3.5.2-2C for three or two pump operation and b) the axial power shaping control rod withdrawal limits are specified on Figures 3.5.2-4A, 3. 5. 2-4 B and 3.5. 2-4C. I f any of these control rod position limits are exceeded, corrective I
measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall be
's attained within four hours.
4 Except for physics tests, power shall not be increased above the power level cutoff (see Figures 3.5.21) unless the xenon reactivity is within 10 percent of the equilibrium value for operation at rated power and asymptotically approaching stability.
3.5.2.6 Reactor Power Imbalance shall be monitored on a frequency not 'to exceed two hours during power operation above 40 percent rated power.
Except for physics tests, imbalance shall be maintained within the envelopes defined by Figures 3.5.2-3A, 3.5.2-3B and 3.5.2-3C.
If the imbalance is not within the envelopes defined by i
Figures 3.5.2-3A, 3.5.2-3B and 3.5.2-3C corrective measures shall be taken to achieve an acceptable imbalance.
If an acceptable imbalance is not achieved within four hours, reactor power shall be reduced until imbalance limits are met.
3.5.2.7 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.
Bases The power-imbalance envelopes defined in Figures 3.5.2-3A, 3.5.2-38 and 3.5.2-3C are based on 1) 1.0CA analyses which have defined the maximum linear heat rate (See Fig. L5.2-4) such that the maximum clad temperature will not exceed the final Acceptance Criteria and 2) the Protective System Maximum Allowable Setyints (Figure 2.3-2).
Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified houndary.
Operation in a situation that would cause the final acceptance criteria to be approached should a LOCA occur is highly improbable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.* Conserva-tism is introduced by application of:
a.
Nuclear uncertainty factors b.
Thermal calibration
)
c.
Fuel densification effects j
d.
Hot rod manufacturing tolerance factors e.
Fuel rod bowing The 20 25 percent overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke.
Control rods are arranged in groups or banks defined as follows:
' Actual operating limits depend on whether or not incore or excore detectors are used and their respective instrument and calibration errors. The method used to define the operating limits is defined in plant operating procedures.
Amendment Nd.' //, [, 3 3 48 o
Group Function
[
l 1
Sa fe t y 2
Sa fe t y I
3 Sa fe t y i
p A
Safcty i
5 Regulating i
6 Regulating l
APSR (axial power shaping bank) 3
'the rod posi tion limi ts a re based on the mos t limi t ing of the following th roe cri teria: liCCS power peaking, shutdown margin, and potential ejected rod worth. As discussed above, compliance wi th the L:CCS power h
peaking critorion is ensured by the rod position limits.
1hc minimuu
. h arallable. rod worth, consistent wi th the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is withdrawn remains in the full out pori-tion (1).
1hc rod position limits also ensure that inserted rod groups i
w]Il not contain s ingic rod worths greater than 0.65". Ak/k at ra t ed pow r, l
1hese values have been shown to be safe by the safety analysis (2) of the hypothetical rod ejcetion accident.
A mximum single inserted centrol rod worth.of 1.04 nk/k is allowed by the rod positions iiuits at hot :ero l (
power.
A single inserted control rod worth of 1.0". nk/k at liegituting of li fc, hot, zero power would result in a lower transient peak therr.a1 power
- nid, the re fo re, less severe environmental consequences than a 0.05% 4h/k I
ejected rod worth at rated power.
Cont rol rod groups are wi thdrawn in sequence beginning with group 1.
G roup s y
5, 6, and 7 are overlappeil 20%.
1hc normal position at power is for groups l
6 and 7 to be partially inserted.
'lhe quadrant powe r ti lt limits set forth in Specif.ical ion 3.5.2..I have been I
est ablished within the thermal analysis design base using the dc rinition o f iptadrant powe r til t gi sea in Technical Speci ficativis, Section 1.0
'ihese limi ts in con.innet ion with the control rod pos i t ion Iin.i t s in Speeif-ic.ition 3.5.?.5.3 ensure that des inn peak heat r,te eciim ia are not e w eeded thiri ng norma l oper.it ion when i ncludi ng the e f fect :. of pot en t ia l fuel denii-j iication.
d
. and a x i a l i mha lance mon i tori ng i n Sprei fica l ions.i. 5..'. 4.6 l
'the spiad rant tilt a n d.i. 5..'. S. 4, respect ively, will normally be performed in the plant com-i puter.
- lhe two hour frequency for monitoring these quantitics will provido
(
adequate surveillance when the computer is out of service.
During t he physics tes t inn program, the high flux trip setpoints are ad:.dnis-t rai i vely net an follows to ensure that an additional safety margin is pro-vided:
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(280,102) 100 OPERATION IN THIS (280,92) go REGION IS NOT ALLOWED 80 RESTRICTED (270,80)
REGION 70 S
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50
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40 t
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PERMISSIBLE G?ERATING REGION 20 - /
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O 20 40 60 60 100 120 140 160 180 200 220 240 260 280 300 Rod index.
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20 40 00 80 100 Group 7 1
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20 40 60 80 100 Group 6 OPERATION AFTER 250 1 10 EFPD AH 0-1 CYCLE 4 1
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20 40 60 80 100 Figure 3.5.2-1C Group 5
u f
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EESTRICTED FOR 2 & 3 PUMP 110 (24,102)
(175,102)
(215,102) 100 RESTRICTED RESTRICTED FOR 3 PUMP FOR 3 PUMP 90 80 i
70 x
60 (24,64)
E 0
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20 40 60 30 100 Group 7 I
ROD POSITION LIMITS FOR TWO AND THREE PUMP O
20 40 60 80 100 Group 6 OPERATION FROM 0 TO 100 t 10 EFPD ANO-1, CYCLE 4 i
I t
't t
1 Figure.3.5.Z-2A 0
20 40 60 80 100
)
Group 5
^
II0 (132.102)
(174,102)
(230,102) 0 RESTRICTED 00 -
4 FOR 3 PUMP OPERATION IN THIS g
50 REGICN IS NOT 8
ALLOWED 80 70 (79,64)
(300,64) '
q x
60 3
RESTRICTED FOR (60 50)
N 50
-2 & 3 PUMP 1
r ESTRICTED FOR 3 PUMP 40 (22.36) f 1
30 PERMISSIBLE OPERATING REGION y
20 10
/
(0,0) 0/
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index. % WD t
i I
I I
t 0
20 40 60 80 100 Group 7 I
I l
I I
I O
20 40 60 80 100 ROD POSITION LIMITS FOR TWO AND THREE PUMI Gr up 6 OPERATION FROM 100 1 i0 TO 250 t i0 EFPD A N O -1, CYCLE 'I t
t 1
I f
f 0
20 40 60 80 100 Figure 3.5.2-2B Group 5 7
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, ~ ~
I10 (15',102)
(270,102) 4 100 RESTRICTED 90 CPERATION IN THIS REGION FOR 3 PUMP 15 NOT ALLOWED 20 70 (100,70)
E
- L EO PERMISSIBLE OPERATING REGION
<o N
50 (68,50) u (50,45) o
$ 40
-(26.3Bi e
E 30 f 0.30)
I RESTRICTED FOR 3 PUMP l
20
[
RESTRICTED FOR 2 & 3 PUMP g3 s O d I
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20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index. % WD 1
I I
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20 40 60 60 100 Group 7 i
f I
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ROD POSITION LIMITS FOR TWO iND THREE O
20 40 60 80 100 Group 6 PUMP CPERATION AFTER 250 1 50 EFPD ANO i, CYCLE 4 I
t t
f f
I O
20 40 60 80 100 Figure 3.5.2-2C Group 5
i j
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Power, % of 2568 HWt i
-12.0,102
+8.2,iO2 i
100 l I
-11.6,92 L
+9.6,92 90
,+ii.2,80
-i9.2,80 80 70 l
RESTRICTED h
RESTRICTED REGION REGION mo 50 x i
o 4
1 w
a e
co w
i 40 cx ww E
30 ce w
L 20 e
10 t
i i
i i
t i
i t
i
-50
-40
-30
-20
-i0 0
iO 20 30 40 50 Axial Power imbalance, %
i OPERATIONAL POWER IHBALANCE ENVELOPE FOR
'i OPERATION FROH 0 T0 100 1 10 EFPD j
A H 0 -1, CYCLE 4 i
Figure 3.5.2-3A i
48d
{,
' 'l 1
a
4 Power, % of 2568 MWt 1
-15.3,102
+8.2,102 100
-13.4,92
~
+9.6,92 90
-18.0,80 80 h +11.2,80 i
70 e*
60 Q
ex 50 RESTRICTED RESTRICTED x
REGION d
LIO G RE010N U
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T.x 30 5
n.
20 l'0 t
i f
I i
t i
1 1
I
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-40 20
-10 0
i0 20 30 40 50 Axial Power imbalance, %
OPERATIONAL POWER lHBALANCE ENVELOPE FOR OPERATION FROM iOO 1 10 TO 250 1 10 EFP0 A N O -l, CYCLE I Figure 3.5.2-3B i
48dd s!
t
i Power, % of 2568 MWt ll0 t
-10.8,102 r
+14.3,102
<. +15.4,92
-10. 4,92 90
,+16.4,80
-21.5,80 80 i-70 60 r
4=
50 a.
x o
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RESTRICTED RESTRICTED w
40 a
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30 xa w
A 20 4
10 i
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1
-40
-30
. -20
-10 0
i0 20 30 40 50 Axial Power Imbalance, %
- l' OPERATIONAL POWER IHBALANCE ENVELOPE FOR OPERATION AFTER 250 t 10 EFPD AHO -l, CYCLE Li Figure 3.5.2-3C 48ddd
6,102 M,102 o
o 100 4'9 90 RESTRICTED REGION RESTRICTED 30,80 80 0,80 REGION 70 60 i
w$
50
~
100,50 o
Drt.
l 40 u
k PERMISSIBLE OPERATING REGION 30 20 10 i
0 I
1 I
I I
t) 10 20 30 40 50 60 70 80 90 100 APSR, /. WITHDRAWN I
l APSR POSITION LlHITS FOR OPERATION l FROH 0 TO 100
Figure 3.5.2-4A I
y 48f
- g
fl 17,102 o
100 RESTRICTED REGION 90 N'17,92
,80 80 70 u
60 w
g 50 100,50 a
f 40 5
a.
30 PERMISSIBLE OPERATlHG REG 10H 20 10 0
I I
I I
I I
I I
I 0
10 20 30 40 50 60 70 80 90 100 APSR,7. Withdrawn APSR POSITION LlHITS FOR OPERATION FROM 100 t 10 TO 250 t 10 EFP0 ANO-l, CYCLE il Figure 3.5.2-4B l
h i
Y.
r; i
48g l
4
20,102 100 i 20,92 RESTRICTED REGION 90 80 30,80 70 h
e 60 N
]
50 100,50 40
-~
E PERMISSIBLE OPERATING REGION 30 20
~
10 O
l I
I I
I I
I I
I a
0 10 20 30 40 50 60 70 80 90 iOO APSR, % Withdrawn I
i APSR POSITION LlHITS FOR OPERATION AFTER 250 t 10 EFPD ANO-i, CYCLE 4 Figure 3.5.2-4C I
l 48h
.o y