ML19312C911
| ML19312C911 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/07/1976 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| NUDOCS 8001140574 | |
| Download: ML19312C911 (35) | |
Text
_ _ _ _ _
O ATTAC}{.'!ENT 1 l'ROPOSED TECilNICAL SPECIFICATION REVISIONS TO SUPPORT OCONEE 2, CYCI.E 2 OPERATION l
l l
t May 7, 1976 8001140f
Bqsey _1 Un i t, _2, 2 h.ive been generated using t
safetylinitspressutedforOconeQnit
"" U " ""
)
The BAW-2 r r it ical heat flux correlation and the React or6 "lbs/hr for of the design flow (131.21x10 flow rate of 107.6 percent flogate utilized is conservative compared to four-pump operation). The the actual measured flow rate.
To maintain the int"grity of the fuel cladding and to prevent fission overheating of the cladding product release, it is necessary to prevent This is accomplished by operating under normal operating conditions.
transfer, wherein the heat within the nucleate boiling regime of heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature.
The upper boundary" of the nucleate boiling regime is termed " departure from nucleate boiling (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures and the Although DNB is not an observable possibility of cladding failure.
the observable parameters of neutron parameter during reactor operatlon, and pressure can be related to reactor coolant flow, temperature, DNB through the use of the BAW-2 correlation (1).
The BAW-2 correlation
- power, has been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DSB ratio (DNBR), def ined as t he rat to of the heat flux that.could cause DNB at a flux, is indicative of the particular core location to the actual heat rargin to DNB.
The minimum value of the DNBR, during st eady-stat e and anticipated transients is operat ion, nomal operational t ransients, 1
i A DNBR of 1.30 corresponds to a 95 percent probabil i ty limited to 1.30.
at a 95 percent confidence level that DSB will not occur; this is considered The difference a conservative margin to DSB for all operating coaditions.
pressure and the indicated reactor coolant between the actual core out let system pressure has been considered in determining the core protection safety in these two pressures is nominally 45 psi; however, limits.
The difference only a 30 psi drop was assumed in reducing the pressure trip setpoints to d i
correspond to the elevated location where the pressure is actually measure.
The curve presented in Figure 2.1-1B represents the conditions at which 1.30 is predicted for the maximum possible thermal a minimum DNBR of (112 percent) when four regetorcoolant pumps are operating (minimum flow is 141.1x10 lbs/hr).
This curve is based on the power reactor coolant fuel densification following nuclear power peaking fattors with potential 1.50.
The l
N N
N J
F
= 2.67; F
= 1.78; F
=
i J
and fuel rod bowing effeets:
AH z
q j
design peaking combination results in a more conservative DNBR than any other power shape that exists during normal operation.
2.1-2B are based on the more restrictive of two The curves of Figure effects c.f potential fuel densification thermal limit s and include the and fuel rod howing:
2.1-3a
N 1.
The 1.30 DNBR limit produced by a nuclear peaking factor of F
= 2.67 or the combination of the radial peak, axial peak and positio8 of the axial peak that yields no less than a 1.30 DNBR.
2.
The combination of radial and axial peak that causes central fuel melting at the hot spot.
The limit is 19.8 kw/ft for Unit 2.
Power peaking is not a directly observable quantity, and, therefore, limits have been established on the bases of the reactor power imbalance produced by the power peaking.
The specif ied flow rat es f or Curves 1, 2, and 3 of Figure 2.1-2B correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.
The curve of Figure 2.1-1B is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-3B.
The maximum thermal power for t hree-pump operation is 86.4 percent due to a power level trip produced by the flux-flow ratio 74.7 percent flow x 1.07 = 79.9 percent power plus the maximum calibration and instrument error.
The maximum thermal power for other coolant pump conditions are produced in a similar manner.
For each curve of Figure 2.1-3B, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 or a local quality at the point of minimum DNBR less than 22 percent for that j
part icular reactor coolant pump situation. The 1.30 DNBR curve for four-j pump operat ion is more rest rict ive than any other reactor coolant pump j
situation because any pressure / temperature point above and to the left of the four-pump curve will he above and to the left of the other curves.
References (1)
Correlation of Critical lleat Flux in a Bundle Cooled by Pressurized Water, BAW-10000, March 1970.
(2) Oconee 2, Cycle 2 - Reload Report - BAW-1425 (Rev. 1), April 1976.
2.1-3h
1 I
Bases - Unit 3 fission product integrity of the fuel cladding and to preventoverheating of the cladding un To maintain the is necessary to preventThis is accomplished by operating within the nucleate release, it transfer coefficient is operating conditions.
transfer, wherein the heat boiling regime of heatthe clad surface temperature is only slightly greater The upper boundary of the nucleate boiling large enough so that At this point, than t he coolant temperature.from nucleate boiling" (DNB).
which would j
regime is t ermed " depart ure transfer coefficient, is a sharp reduction of the heat possibility of cladding failure.
there in high eladding temperatures and thean observable parameter during reacto the result flow, temperature, Although DNB is not observable parameters of neutron power, reactor coolant l tfon.(1) and pressure can be related to DNB through the use of the W f DNB flux distributions. The local DNB a
l f or axially unif orm and non-uniform heat flux that would cause DNB at (DNBR), defined as the ratio of the heatflux, is indicative of the margin ratio j
particular core location to the actual heatThe minimum value of the DNBR, du to DNB.
is limited to 1.3.
A DNBR operat ional transients, and ant icipated transientsprobability at a 99 percent confidence of 1.3 corresponds to a 94.3 percentoccur; this is considered a conservative margin to l
level that DNB will. not The difference between the actual core DNB f or all operating conditions.
outlet pressure and the indicated reactor coolant system pressure has been
'the dif f erence considered in determining the core protection saf ety limits.is nominally 45 in these two pressures t rip setpoints to correspond to the elevat ed assumed in reduc ing the pressure is actually measured.
location where the pressure 2.1-1C represents the conditions at which a The curve presented in Figure l power (1127.)
minimum DNBR of 1.3 is predicted f or the maximum possible therma flow is when four reactor coolant the following nuclear power 131.3 x 106 lbs/hr). This curve is based on peaking f actors (2) with potential f uel densification ef f ects:
N N
N 1.78 ;F
= 1.50 2.67; F
=
F
=
4 AH The design peaking combination results in a more conservative DNBR than any exists during normal operation.
other shape that l
The curves of Figure 2.1-2C are based on the more restrictive of two therma fuel densification:
limits and include the ef fects of potential N = 2.67 produced by a nuelcar power peaking factor of F$he 1.
The 1.1 UNBR limit radial peak, axial peak and posit ion of or the combina t ion of ahe
- a. ial peak t hat yields no less than 1.3 UNHR.
s causes central fuel melting The combination of radial and axial peak that i
at the hot spot.
The limit is 19.8 kw/ft tor Unit 3.
2.
2.1-3c
Powe-peaking is not a directly observable quantity and therefore limits have been established on the bases of the reactor power imbalance produced by the power peaking.
The spei if led flow rates for Curves 1, 2, 3, and 4 of Figure 2.1-2C correspond to the expect ed minimum flow rates with four pumps, three pumps, one pump in each loop and two pumps in one loop, respectively.
The curve of Figure 2.1-1C is the most restrictive of all possible reactor coolant pump-maximum thermal power combina tions shown in Figure 2.1-3C.
The curves of Figure 2.1-3C represent the conditions at which a minimum DNBR of 1.3 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNER is equal to IP.,,(3) whichever condition is more restrictive.
Using a local quality limit of 15 percent at the point of minimum DNBR as a basis f or Curves 2 and 4 of Figure 2.1-3C is a conservative criterion even though the quality of the exit is higher than the quality at the point of minimum D'iBR.
The DNBR as calculated by the W-3 correlat ion cont inually increases from point of minimum DNBit, so that the exit DNBR is 1.7 or higher, depending on the pressure.
Extrapolation of the W-3 correlation beyond its published quality range of +15 percent is justified on the basis of experimental data.(4)
The maximum thermal power for three pump operation is 86.4% - Unit 3 due to a power level trip produced by the flux-flow ratio 75% flow x 1.07 = 80%
power plus the maxi aum calibration and instrument error. The maximum thermal power for other coolant pump conditions are produced in a similar manner.
A flux-flow ratio of 0.961 is used for single loop conditions.
For each curve of Figure 2.1-3C a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 15 percent for that particular reactor coolant pump situation. The 1.3 DNBR curve for four-pump operation is more rest rict ive t han any other react or coolant pump situat ion because any pressure /
temper.iture point above and to the lef t of the four pump curve will he above and to the lef t of the other curves.
REFERENCES (1) FSAR, Section 3.2.3.1.1 i
(2) FSAR, Sect ion 3.2.3.1.1.c (3) FSAR, Section 3.2.3.1.1.k 4
2.1-3d
i
{
(4) The following papers which were presented at the Winter Annual Meeting, ASME, !Jovember 18, 1969, during the "Tuo-phase Flow and lient Transfer in i
Ren! Hund 1 es Symposium:"
(a) Wilson, et al.
j
" Critical lleat Flux in tion-Uniform lleater Rod P,undles" (b) Cellerstedt, et al.
"Correlat ion of a Critical llea t Flux in a Bundle Cooled by Pressurized
)
Water" i
l I
i i
I 4
)
l i
0 1
1 l
)
1 r
I l
j i
1 I
I I
i 4
a 2.1-3e i
4
~
7400 _
2300 ACCEPTABLE OPERAll0N 2200 _
T E
a 6
2100 2
UNACCEPTABLE 5
OPERATION r
2000 O
\\
1000 1800 4
I 1
i 500 580 600 620 040 660 Reactor Outlet Temperature F
COPE PROTECT 10tl SAFEN LlfilTS U
,# P h NIT 2
}
s%f; OCONEE NUCLEAR STATION c ultoats FIGURE 2.1-1B 2.1-5
THERMAL. P0nER LEVEL *,
__ 120
(-21 3.112)
(21.3.112)
-- 110 g
KWiFT LIMIT KAcFT llMll ACCEPTABLE (33.102) 100 4 N 4P OPERATION
( 21.3.86.4)
~"
(21.3.86.4)
G
( 50.80)
-. 80 ACCEPTABLE 354 PO4P
-- 10 (33*76.4)
OPERATION
{
(-21. 3. 58. 9 )
_. 60 (21.3.58.9)
(-50.54.4) 50 ACCEPTABLE 2.3.&4 PU P 33*48 9)
__ 40 OPERATION
(-50.26.9)
_. 30 20 l
10 i
e I
i e
i I
I i
e i
I e
I
-70
-60
-50
-40
-30
-20
-10 10 20 30 40 50 60 70 Reactor Power lmualance, %
CURVE REACTOR COOLANT FLOW (LB/HR)
I 141.3 x 106 2
105.6 x 106 3
69 3 x 106 CORE PROTECTION SAFETY LIMITS UtilT 2 hgi ninan OCONEE NUCLEAR STATION FIGURE 2.1-2B
2400 1
2 3
2300 2200 w
ACCEPIABLE OPERAll0N l
?,
i 2100
~
I
~
<?
{
2000 o
UNACCEPTABLE OPERATION 1000 -
1800 -
a e
1 g
560 580 600 620 640 660 Reactor Outlet Temperature, F RE ACIOR COOL ANT FL0n PUMPS OPERATING CURVE (LBS/HR)
PC9ER (TYPE OF LIMIT}
I 141.3 x 106 (1000)*
112%
FOUR PUMP (DNBR LIMITED) 2 105.6 x 106 (74.7%)
86.4$
THREE PUMP (DNBR LIMIT.0) 3 69.. n 106
( 4 9,,' )
58.9%
ONE PUMP IN EACH' LOOP (00AllIY LIMITED)
- 107.65 0F CYCLE I DESIGN FLOW CORE PROJECTION SAFETY LIMITS M Uf11T 2 b[N, OCONEE NUCLEAR STAT 1
ut reau Fictae 2.1-3B
\\
1 2.1-11
. - =.
- Luri, no rmal plant op stian with all reactor coolant mps aperating, re utar trip is initia,_d wnen the reactor power level.eaches 105.5?. of rat. A power.
Adding to this the possible variation in trip setpoints due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 1127., which is more conservative than the value used in the safety analysis. (4)
Overpower Trip Based on Flow 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 acco=modate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power.
Analysis has demonstrated 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 electrical malfunction.
The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases.
The power level trip set point produced by the power-to-flow ratio provides overpower DNB pro-tection for all modes of pump operation.
For every flow rate there is a maxi-
.I mum permissible power level, and for every power level there is a minimum permissible low flow rate.
Typical power level and low flow rate combinations for the pump situtations of Table 2.3-1A are as follows:
l 1.
Trip would occur when four reactor coolant pumps are cperating if power is 105.57, and reactor flow rate is 100%, or flow rate is 94.87, and power level is 100?..
2.
Trip would occur when three reactor coolant pumps are operating if power is 78. S *: and reactor flow rate is 74.7% or flow rate is 71.1% and power level is 75%.
3.
Trip would occur when two reactor coolant pumps are operating in a single loop if power is 51.7% and the operating loop flow rate is 54.5% or flow rate is 48.5% and power level is 46%.
4.
Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 51.7% and reactor flow rate is 49.0% or flow rate is 46.4% and the power level is 49%.
The flux-to-flow ratios for Units 1 and 2 account for the maximum variation j
from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.
For safety calculations the maximum calibration and instrumentation errors for the power level trip were used.
The power-imbalance boundaries are established in order to prevent reactor rhermal limits from being exceeded. These thermal limits are either power peiking kw/ft limits or DNBR limits.
The reactor power imbalance (power in the top half of core minus power in the bottom half of core) reduces the power leve: trip produced by the power-to-flow ratio such that the boandaries of Figure 2.3-2A
- Unit 1 are produced.
The power-to-flow ratio reduces the power l 2.3-2B
- Unit 2 2.3-2C - Unit 3
{
2.3-2
level t rip and associated reactor power / reactor power-imbal by 1. O '; '; ? - Un i t I for a 1% flow reduction.
ance boundaries 1.07% - Unit 2 1.07% - Unit 3 Fo r Un i t 1, the power-to-flow reduction ratio the power-to-flow reduction factor is 0 961 durinis 0.949, and for Units 2 and 3, Pupp Monitors g single loop operation.
The pump monitors prevent tripping the reactor due to the losa of reacter cthe minimum core DNBR from de monitoring pump operational status provides redundant oolant pump (s).
The circuitry by tripping the reactor on a signal diverse trip protection for DNB from that of the power-to-flow ratio.
The pump monitors also restrict pumps in operation.
the power level for the number of Reactor Coolant System pressure During a startup a eident from low power or a slow rod wit hdrawal fr power, the system high pressure set point is reached before t he nuclear over-power t rip set point.
The t rip set ting limit shown in Figure 2.3-1A - Unit 1 2.3-1B - Unit 2 for high reactor coolant naintain the systen pressure 2.3-lC - Unit 3 des ign t ransient. system pressure below the safety limit (2355 psig) has been es (1)
(2750 psig) for any The low pressure (1800) psig and variable low pressure (11.14 T
-4706) trip (1800) psig (1800) psig (10.79 T " -4539)
I setpoints shown in Figure 2.3-1A have been established t(16.25 T " -7756) 2.3-1B o maint In the DNB 2.3-lC ratio greater than or equal to 1.3 for those design accidents that a pressure reduction. (2,3) result in Due to the calibration and instrumentation errors the s variable low reactor coolant afety analysis used a system pressure trip value of (11.14 T I
-4746) out (10.79 T
-4579)
Coolant Outlet Temperature (16.25 T "' -7796) out The high reactor coolant outlet in Figure 2.3-1A has been established to preventtemperature trip setting limit (619 F) shown 2.3-1B excessive core toolant 2.3-1C temperatures in the operating range.
the safety analysis used a trip setDue to calibration and instrumentation
- errors, point of 620 F.
Reactor Building Pressure l
The high reactor building pressure trip setting limit positive assurance that a (4 psig) provides less-of-coolant accident, reactor trip will occur in the unlikely event a
of syst em pressure t rip.
even in the absence of a low reactor coolant 2.3-3 t
2400 P = 2355 psig T = 619 f 2300 2200 a
.i s
m 2100 ACCEPTABLE I
OPERAll0N E
R 9
2000 M
A E
8 D
UNACCEPTABLE 1900 Y
OPERATION
//
P = 1800 psig 1800 i
(5B7.5) l I
I I
540 560 580 600 620 640 Reactor Outlet Temperature, F PROTECTIVE SYSTelI'AXliul ALLO \\MBLE SETPOINIS M
(cuiroat%g UtilT 2 OCONEE NUCLEAR STATION FIGURE 2.3-1B 2.3-6
Ill E Dt ? P0 ef R L EVEL. '
- 120 UNACCEPTABLE OPERAll0N
__110
<107)
+
__100 7
ss h
ACCEPTABLE 4 PUMP
__ 90 4s OPERATION 80 (70 0)
ACCEPTABLE
_. 70 3 & 4 PUMP OPERATION
__ 60 (S24; 50 ACCEPTABLE.- 40 2.3.&4 Pll1P 30 OPERATION m
__ 20 o
9 10"'
2 ii g
nr ti in i
i i
i i
10 60 50 40 30 20 10 10 20 30 40 50 60 70 Poner Inma l anc e.
PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETP0lllTS UNIT 2
{ovat roai OCONEE NUCLEAR STATION w
FIGURE 2.3-2B 2 3-9 l
1.4 b l e 2.3-1H 1 nit 2 Rea< tor Pror oc t i ve, iv
- t e-r 1 r if S ;t t_i ng L i m i t_s No Reactor one Reactor Feur Reactor Three Reactor Coolant Pumps Coolant Pump Coolant Pumps Coolant Pumps Operating in A Operating in Ope ra t in,:
Operating Single Loop Each leep (Operating Power (Operating Po wee r (Operating Power (Operating Shutd wn RPS Segment
-100' Fated)
-75% Rated)
-467 Rated)
-49% Rated) ftvp3s 1.
Nuclear Power Max.
105.5 105.5 105.5 105.5 5.0'
(% Rated) 2.
Nuclear Power Max. Based 1.07 times, flow 1.07 times flow 0.961 t ir.w flow I.07 tines flow ILvp m ed on Flow (2) and Imbalance, ninus reduction minus reduc t ion ninus redaction minus reduction
(% Rated) due to imbalance due to imbalance due te imbalance due to imbalance 3
Nuclear Power.' tax. Based NA NA W (5) (6) 55%
Bvrassed on Pump Monitors. (% Rated) 4.
High Reactor Coolant 2355 2355 2355 2355 1720 System Pressure. psig. Max.
5.
Low Reactor Coolant 1P00 1800 1800 1800 sypassed System Pressure, psig. Min.
6.
Variable Low Reactor (IC.74 T
-4539)
(10.79 T
-4539)' '
'I II (10.74 T
-4539)
(10.79 T
-4539)
Bypassed
""E
""I N
Coolant System Pressure
{
psig. Min.
w 7.
Reactor Coolant Temp.
614 619 610 (6) 619 619 F., Max.
8.
High Reactor Building e
4 4
4 4
Pressure, psig, tbx.
(1) T is in degrees Fahrenheit ( F).
(5) Reactor power level trip set point produced "E
by pump contact monitor reset te 55.0%.
(2) Reactor Coolant Systen Flow, t.
(6) Specification 3.1.8 applies. Trip one of the (3) Administratively controlled reduction set two protection channels receiving outlet only during reactor shutdown.
t ernpe r a t ur e information from sensors in the idle loop.
(4) Automatically set when other segments of the RPS are bypassed.
6 If within one (1) hour of determination of an inoperable rod, it is not determined that a 1%ak/k hot shutdown margin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to the hot standby condition until this margin is established.
Following the determination of an inoperable rod, all rods shall h.
be exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised we2kly until 'the rod problem is solved.
If a control rod in the regulating or safety rod groups is 1.
elet 1.i r ed inoperable, power shall be reduced to 60 percent of the thermal power allowable for the reactor coolant pump co=-
bination.
If a control rod in the regulating or axial power shaping groups j.
of rated is declared inoperable, operation above 60 percent power may continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained within allowable group average position limits of Specification 3.5.2.2.a and the withdrawal limits of Specification 3.5.2.5.c.
3.5.2.3 The worths of single inserted control rods during criticality are limited hv the restrictions of Specification 3.1.3.5 and the control rod position limits defined in Specification 3.5.2.5.
3.5.2.4 Quadrant Power Tilt a.
1:xc ept for physics tests. i f t he maximum posi t ive qu:idrant power tilt exceuls +3.417 Unit 1. either the quadrant power t ilt shall I
- 3. 41'; Unit 2 4.92% Unit 3 he reduced to less than +3.41% Unit I within two hours or the 3.41% Unit 2 l
4.92% Unit 3 following act ions shall be taken:
(1) If four reactor coolant pumps are in operation, the allowable thermal power shall be reduced below the power level cutoff (as identified in specification 3.5.2.5) and further reduced by 2% of full power for each 1% tilt in excess of 3.41% Unit 1.
3.41% Unit 2 i 4.92% Unit 3 (2) If less than four reactor coolant pumps are in operation, the allowable thermal power for the reactor coolant pump combination shall he reilureil by 22 of full power f or each 12 t ilt.
3.5-7 j
( ') ) Except as provided in specification 3.5.2.4.b the reactor shalI be b rou r,h t to the hot shutdown condition within four hours if the quadrant power tilt is not reduced to less than 3.41% Unit I within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
'5.417 Unit 2 4.92% Unit 3 I
b.
If the quadrant tilt exceeds +3.41% Unit I and there is simultaneous 3.412 Unit 2 4.92% Unit 3 l
indication of a misaligned control rod per Specification 3.5.2.2, reactor operation may continue provided power is redr ed to 60%
of the thermal power allowable combination, for the reactor coolant pump c.
Except for physics test, if quadrant tilt exceeds 9.44% Unit 1, 9.44% Unit 2 l
a controlled shutdown shall be 11.07% Unit 3 initiated immeciately, and the reactor shall be brought to the hot four hours.
shutdown condition within d.
Whenever the reactor is brought to hot shutdown pursuant to
'3. 5. 2. 4.a (3 ) o r 3. 5. 2. 4. c above subsequent is permitted for the purpose of measurement reactor operation testing, and corrective action provided the thermal power and the power range high flux setpoint allowable for the reactor coolant ce=bination are restricted by a reduction of 2 percent pump power for each I percent of full tilt prior to shutdown.
for the maximum tilt observed e.
Quadrant power tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15 percent of rated power.
3.5.2.5 Control Rod Positions Technical Specification 3.1.3.5 does not a.
prohibit of individual safety rods as required by Table 4.1-2 or apply tothe exercis inoperable safety rod limits in Technical Specification 3 5 2 2 b.
Operating rod group overlap shall be 25% + 5% between two sequential groups, except for physics tests.
c.
Except for physics tests or exercising control Il, rods, the control rod withdrawal limits are specified on Figures 3.5.2-1A1 and lI 1.5.2-IA2, (Unit 1),
i 3.5.2-1B1, 3.5.2-1B2 and 3.5.2-1B3 (Unit 2),
l3 and 3. 5. 2-101, 3. 5. 2-lC2. and 3. 5. 2-IC3 (Unit 3) for four pump operat ion and on Figures 3.5.2-2A1, 3.5.2-2A2 (Unit 1), 3.5.2-2B1, l
3.s.2-2B2 3.5.2-2B3 (Unit 2), and 3.5.2-2C (Unit 3) for three or 3.5-8 h
(3) Except as provided in specification 3.5.2.4.b. the reactor shall be brought to the hot shutdown condition within four hours if the quadrant power tilt is not reduced tc less than 3.41% Unit I within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.41% Unit 2 I
4.92% Unit 3 h.
If the quadrant tilt exceeds +3.41% Unit I and there is simultaneous 3.41% Unit 2 l
4.92% Unit 3 indication of a misaligned control rod per Specification 3.5.2.2, reactor operation may continue provided power is reduced to 60%
of the thermal power allowable for the reactor coolant p ur,ip combination.
c.
Except for physics test, if quadrant tilt e>:ceeds 9.44% Uni t 1, 9.447 Unit 2 l
11.077. Unit 3 a cont rolled shutdown shall be initiated immediately, and the reactor shall be brought to the hot shutdown condition within four hours.
d.
Whenever the reactor is brought to hot shutdown pursuant to 3.5.2.4.a(3) or 3.5.2.4.c above, subsequent reactor operation is permitted for the purpose of measurement, testing, and corrective action provided the thermal power and the power range high flux setpoint allowable for the reactor coolant pump combination are restricted by a redu' tion of 2 percent of full power for each I percent tilt for the maximum tilt obse rved prior to shutdown.
Quadrant power tilt shall be monitored on a minimum frequency e.
of once every two hours during power operation above 15 percent of rated power.
3.5.2.5 Control Rod Positions Technical Specification 3.1.3.5 does not prohibit the exercising a.
of individual safety rods as required by Table 4.1-2 or apply to inoperable safety rod limits in Technical Specification 3.5.2.2.
b.
Operating rod group overlap shall be 25% + 5% between two sequential groups, except for physics tests.
e.
Except for physics tests or exercising control rods, the control r oil w It h Ir.iw.ii I imit s are sper if ied on Figures 3.5.2-l Al and 1.'. 2-IA2 (Itn i t 1).
- 1. 5. 2-I tt i.
- 1. 5. 2-182 and 3. 5.2-1 B3 (Unit 2),
, i n.1 t.N.'
lel, t. S. '
10.'..in.1
- 1. 5. 2-1C 1 (Uni t 3) for four pump
..i ci.it t..n.i n.1
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4. 'i. ' ' A l. 4. %. 2.' A.'
(Unit 1), 3.5.2-2H1, j
t. *.
/ I t.'. 4. '.. t.'It t t t:n ii.).
.in 1
- 1. 5..'- 2C (l'n i t 1) for t bree or 3.5-8
two pump operation.
If the control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position.
Acceptable control rod position shall then be attained within two hours. The minimum shutdown margin required by Specification 3.5.2.1 shall be maintained at all times.
i d.
Except for physics tests, power shall not be increased above the power level cutoff as shown on Figures 3.5.2-1A1, 3.5.2-1A2 (Unit 1), 3.5.2-1B1, 3.5.2-1B2, and 3.5.2-1B3 (Unit 2), and 3.5.2-1C1, 3.5.2-lC2, 3.5.2-lC3 (Unit 3), unless the following requirements are met.
(1)
The xenon reactivity shall be within 10 percent of the value for operation at steady-state rated power.
(2) The xenon reactivity shall be asymptotically approaching the value for operation at the power level cutoff.
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 envelope defined by Figures 3.5.2-3Al, 3.5.2-3A2, 3.5.2-3Bl. 3.5.2-3B2,l 3.5.2-3B3, and 3.5.2-3C.
If the imbalance is not within the envelope defined by Figure 3.5.2-3A1, 3.5.2-3A2, 3.5.2-3B1, 3.5.2-3B2, 3.5.2-3B3,l and 3.5.2-3C, correct ive measures shall be taken to achieve an acceptable imbalance.
If an acceptable imbalance is not achieved within two hours, reactor power shall be reduced until imbalance limits are met.
3.5.2. / The control rod drive patch panels shall be locked at all times with limited access to be authorized by the manager.
i 4
3.5-9
U " "." 8 The power-inhalance envelope defined in Figures 3.5.2-3A1, 3.5.2-3A2, 3.5.2-3Bl. 3.5.2-3B2, 3.5.2-3B3, and 3.5.2-3C is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5-2-4) such that the maxirium clad temperature will not exceed the Final Acceptance Criteria.
Corrective measures will he taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundary.
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 t hei r limits while simul t aneously all other engineering and uncertainty at factors are also at their limits.**
Conservatism is introduced by application of:
a.
Nuclear uncertainty factors b.
Thermal calibration c.
Fuel densification effects d.
Ilot rod manufacturing tolerance factors The 25% + 5% 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:
Croup Function 1
Safety 2
Safety 3
Safety 4
Safety 5
Regulating 6
Regulating 7
APSR (axial power shaping bank) l The rod position limits are based on the most limiting of the following three criteria:
ECCS power peaking, shutdown margin, and potential ejected rod worth.
Therefore, compliance with the ECCS power peaking criterion is ensured by the rod position limits.
The minimum available rod worth, consis-tent with 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 position (l).
The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.5% Ak/k (Unit 1) or 0.65% Ak/k (Units 2 and 3) at rated power. These values have been shown to be safe by the safety analysis (2,3,4) of the hypot het ica l rod eject ion acc ident. A maximum single inserted control rod worth of 1.02 Ak/k is allowed by the rod posit lons limits at hot zero power.
A s i n;',l e insert e I cont rol rod wort b of I.02 Ak/k at beginning-of-life, hot w ro power would resol in a lower t ransient peak t hermal power and, I here-fore, less severe env i ronment a l consequences t han a 0.5Z Ak/k (Uni t 1) or 0.652 Ak/k (Units 2 and D eject ed rod worth at rated power.
- Actual operating limits lepend en 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 procedurer.
3.5-10
Control rod groups are withdrawn in sequence beginning with Group 1.
Groups 5, 6, and 7 are overlapped 25 percent. The normal position at power is for Groups 6 and 7 to be partially inserted.
Tin-yn.nl e an t power lit l i ne i t 6 nel l os t le in '; pee ll ie al ion ' t. 's. ?. 4 l ea v e In en est ablinhed wi t h runsides at ion of pot ent la l el l er t s 01 c e nt liowi ny. (lin i t r 1.ind 1 only).nul f uel densi f f rat ion to prevent the linear heat rate peaking increase associated with a positive quadrant power tilt during normal power operation from exceeding 5.10% for Unit 1.
The limits shown in Specification 3.5.2.4 5.10% for Unit 2 l
7.36% for Unit 3 are measurement system independent. The actual operating limits, with the appropriate allowance for observability and instrumentation errors, for each measurement system are defined in the at at ion operat ing procedures.
The quadrant tilt and axial imbalance monitoring in Specification 3.5.2.4
.od 3.5.2.6, respectively, normally will be performed in the process computer.
The two-hour frequency for monitoring these quantities will provide adequate surveillance when the computer is out of service.
Allowance is provided for withdrawal limits and reactor power imbalance limits to be exceeded for a period of two hours withour specification violation.
Acceptable rod positions and imbalance must be achieved within the two-hour time period or appropriate action such as a reduction of power taken.
i operating restrictions are included in Technical Specification 3.5.2.5d to prevent excessive power peaking by transient xenon. The xenon reactirity must be beyond the "undershoot" region and asymptotically approaching its equilibrium value at the power level cutof f.
REFERENCES 1FSAR, Section 3.2.2.1.2
'FSAR, Section 14.2.2.2 I
FSAR, SUPPLEMENT 9 B&W FUEL DENSIF1 CATION REPORT BAW-1409 (UNIT 1) j BAW-1396 (UNIT 2)
BAW-1400 (UNIT 3) i, 3.5-11 l
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-- ' FIGURE 3.5.2-1B2 3.5-14a
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2 MD 3 PUMP OPEPATI0f! FROM 150110 EFPD TO 2E7110 EFPD Mg UfJIT 2 F,( Y FIGURE 3 5 2 2B2 i r:*e si OCONEE NUCLEAR STATION 3 5-19a Ic
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10 12 Axial Location of Peak Power
= rom Battcm of Core, ft i
LOCA LIMITED t%XIiUi ALLQ9RE g LIIEAR HEAT PATE pu rcere, OCONEE NUCLEAR STATION W
FIGURE 3.5.2-4 Y
l 3.5-24 1
3.Ii
?GXIML'M POW'l.R RESTRICTIO!'
APP.I.i c ab i l.1 t v Appiles to the nuclear steam supply syste m of Unit 3 reactor.
Objec t i ve To maintain c<,re life margin in reserve until the system has performed under <>iserat inn c onditions and design objectives for a significant period of time.
SPFf_I f I cji t_l o n The first reactor core in Unit 3 may not he operated beyond 10,944 effective full power hours until supporting analysis and data pertinent to fuel clad collapse under fuel densification conditions have been approved by the Directorate of 1.icensing.
Iti se s The licensing itaff has reviewed the ef fect s of fuel densification for the f ir t.t core in Oconee Unit 1 and conc 1utled that clad collapse wil1 not take place wit hin t he first fuel cycle (10,944 effective full power hours).
Detailed clad creep collapse analyses are yet to be performed to demonstrate t h.it clad collapse will not occur d: ring operation beyond the first fuel cycle.
w 3.11-1
Table 4.1-2 l
!;1NIMU:1 EQUIP;fFNT TEST FICEOUENCY Test Frequency item 1.
Control Rod Movement Movement of Each Rod Bi-Weekly 50% Annually 2.
Pressurizer Safety Valves Setpoint I
25% Annually 3.
Main Steam Safety Valves Setpoint 4.
Ref ueling System Inter loc'..s Functional Prior to Refueling 1
5.
Main Steam Stop Valves Movement of Each Stop Monthly Valve Daily Evaluate 6.
Re.ietor Coolant System 1,cakage Functional Annually 7.
Condenser Cooling Water System Gravity Flow Test Functional Monthly 8.
High Pressure Service Water Pumps and Power Supplies 9.
Spent Fuel Cooling System.
Functional Prior t,o -
Refueling 10.
11ydraulic Snubbers on Visual Inspection Annually Safety-Related Systems Vent Pump Casings Monthly and Prior 11.
liigh Pressure and Low to Testing Pressure injection System 12.
Reactor Coolant System Flow Validate Flow to be Once Per Fuel at liast:
Cycle 6
Unit 1 141.30 x 10 lb/hr b
Unit 2 141.30 x 10 lb/hr I
6 l'ni t 3 131.32 x 10 lb/hr (1)
Applicable only when the reactor is critical is above 200 F and at a steady-(2)
Applicable only when the reactor coolant state temperature and pressure.
( l) Operatiny, pumps exeIuded.
4.1-9
s 4.2.10 For Unit 1, rycle 3 operation, the surveillance capsules will be removed frem the reactor vessel and the provisiens of Specification 4.2.9 will be revised prior to Cycle 4 operation For I? nit 2, Cycle 2 operatica, the curveillance capsules will be removed from the tion 4.2.4 will be revised prior to Cycle 3 operationreactor vessel 3, Cycle 1 operation, For Unit from tl'e the surveillance capsules will be removed visiens of Precification 4.2.9 will be revised prior to Cy operation.
4.2.11 During t he first two refueling periods, two reactor coolant system pipia.e elbows shall be ultrasonically inspected alo
. longitudinal
, 'lds (4 ng their and for cracks inches beyone each side) for clad bonding in bath the clad and base metal.
be inspe:ted are identified in BEW Report The elbows to 1970.
1364 dated December Bases The surveillance program has been developed to comply with S ASME Boiler and Pressure Veasel Code, ection XI of the Inservice Inspection of Nuclear Reactor Coolant S stems, 1970, twledinc 3
1970 uinter adderda, places major erphasis on the area of highest edition.
The program areas where fast stress concentrations and on neutroa irradiation might be sufficient to change material properties.
The reactor vessel specimen surveillance program for Unit based on equivalent 1 and Unit 2 is emo sure tir.es of 1.8, 19.8, 30.6 and 39.6 years.
The contents of the different type of capsules are defined below.
A Type B Tvoe Weld ?!rcerial HAZ 5f st erial HAZ Material Baseline Mattrial Bas. ' n e-Material For Unit 3, the Reactor Vessel Surveillance Program is based exposure times of 1.8, 13.3, 26.7, and 30.0 years.
on equivalent selected and fabricat2d as specified in ASTM-E-185-72.The specimens have been Early inspection of Reactor Coolant System piping elbows is when explosively clad with sensitized stainless steeldesirable considered e
ase metal observed during the two annual inspections, surveillance requireIf no degradation revert to Section XI of the ASME Boiler and Pressure Vessel C d ments will o e.
J f
i 4.2-3 k
.