ML20086K266
| ML20086K266 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 07/14/1995 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20086K263 | List: |
| References | |
| NUDOCS 9507200078 | |
| Download: ML20086K266 (17) | |
Text
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~
1ASES:
2.0 SAFETY LIMITS (Cont)
FUEL CLADDING Since the pressure drop in the bypass region is essentially all INTEGRITY (2.1.1) elevation head, the core pressure drop at low power and flows will (Cont) always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.
Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 102 lbs/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly' critical power at this flow is approximately 3.35 MWt.
With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.
Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis, GETAB (1), which is a statistical model that combines all of the uncertaintics in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) - Boiling Length (L),-GEXL, correlation.
The GEXL correlation is valid over the range of conditions used in the tests of the data used to develop the correlation.
These conditions are:
Pressure:
800 to 1300 psia Max Flux:
0.1 to 1.4 Mlb/hr-ft2 Inlet Subcooling:
0 to 70 Btu /lb 1
Rod Array 9x9 GE 11 array 1
)
i MINIMUM CRITICAL The fuel cladding integrity Safety Limit is set such that no fuel POWER RATIO damage is calculated to occur if the limit is not violated.
Since (2.1.2) the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not result in damage to BWR fuel rods, the critical power at (Cont)
Revision 15 -421-72 -195 -129 -133 B2-2 Amendment No.
7 7
7 1
9507200078 950714 PDR ADOCK 05000293 P
i I
IIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT I
3.3.
REACTIVITY CONTROL (Cont) 4.3 REACTIVITY CONTROL (Cont)
B.
Control Rods (Cont)
B.
Control Rods (Cont)
- 4. Control rods shall not be
- 4. Prior to control rod withdrawal withdrawn for startup or for startup or during
^
refueling unless at least two refueling, verify that at least source range channels have an two source range channels have observed count rate equal to or an observed count rate of at greater than three counts per least three counts per second.
second.
- 5. The RBM shall be operable as required in Table 3.2.C-1, or control rod withdrawal shall be blocked.
C.
Scram Insertion Times C.
Scram Insertion Times
- 1. Average scram insertion time
- 1. Following each refueling for all operable control rods outage, or after a reactor from de-energization of the shutdown that is greater than scram pilot valve solenoids to 120 days, each operable control drop out (DO) of Notches 04, rod shall be subjected to scram 24, 34, and 44 shall be no time tests from the fully greater than:
withdrawn position.
If testing is not accomplished with the Notch Average Scram nuclear system pressure above Position Times (seconds) 950 psig, the measured scram insertion time shall be 44 DO 0.508 extrapolated to reactor 34 DO 1.252 pressures above 950 psig using 24 Do 2.016 previously determined 04 DO 3.578 correlations. Testing of all operable control rods shall be completed prior to exceeding 40% rated thermal power.
Revision Amendment No. 15;-68;-124;-138 3/4.3-4 l
1 m.
'LIMITkNGCONDITIONFOROPERATION SURVEILLANCE REQUIREMENT 3.3.
REACTIVITY CONTROL (Cont) 4.3 REACTIVITY CONTROL (Cont)
C.
Scram Insertion Time (Cont)
C.
Scram Insertion Time (Cont)
- 2. Average scram insertion time
- 2. Within each 120 days of for the three fastest operable operation, a minimum of 10% of control rods in each group of the control rod drives, on a four control rods in all two-rotating basis, shall be scram by-two arrays from de-tested as in 4.3.C.1.
An energization of the scram evaluation shall be completed pilot valve solenoids to every 120 days of operation to dropout (DO) of Notches 04, provide reasonable assurance 24, 34 and 44 shall be no that proper performance is greater than:
being maintained.
Notch Average Scram Position Time (Seconds) 44 DO 0.538 34 Do 1.327 24 Do 2.137 04 DO 3.793 D.
- 3. The maximum scram insertion time for any operable control Once a shift, check the status of rod from de-energization of the pressure and level alarms for the scram pilot valve each accumulator.
solenoids to dropout of Notch 04 shall not exceed 7.00 seconds.
D.
Control Rod Accumulators At all reactor operating pressures, a rod accumulator may be inoperable provided that no other control rod in the nine-rod square array around this rod has a:
- 2. Directional control valve electrically disarmed while in a non-fully inserted position.
)
- 3. Scram insertion time greater than the maximum permissible l
insertion time.
I If a control rod with an inoperable accumulator is i
inserted " full-in" and its directional control valves are j
electrically disarmed, it shall I
not be considered to have an inoperable accumulator.
Revision Amendment No. 65 -124 -146 3/4.3-5 1
5
BASES:
3/4.3 REACTIVITY CONTROL (Cont)
C.
Scram Insertion Times The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than the Safety Limit MCPR. Analysis of the limiting power transient shows that the negative reactivity rates resulting from the scram with the i
average response of all the drives as given in the above Specification, provide the required protection, and MCPR remains greater than the Safety Limit MCPR.
The scram times for all control rods will be determined at the time of each refueling outage. A representative sample of control rods will be scram tested during each cycle as a periodic check against deterioration of the control rod performance.
The limits on scram insertion time presented in Technical Specification 3.3.C include an allowance for the uncertainty in the location of the position indicator probes as well as an allowance for the uncertainty in the control rod positions themselves when dropout of the reed switches occur.
D.
Control Rod Accumulators Requiring no more than one inoperable accumulator in any nine-rod square array is based on a series of XY PDQ-4 quarter core calculations of a cold, clean core.
The worst case in a nine-rod withdrawal sequence resulted in a k gg e
<1.0 - other repeating rod sequences with more rods withdrawn resulted in k gg e
>1.0.
At reactor pressures in excess of 800 psig, even those control rods with inoperable accumulators will be able to meet required scram insertion times due to the action of reactor pressure.
In addition, they may be normally inserted using the control-rod-drive hydraulic system.
Procedural control will assure that control rods with inoperable accumulators will be spaced in a one-in-nine array rather than grouped together.
l l
i Revision Amendment No. 15r-42 -138 B3/4.3-6 l
5
9 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.1$'REACTORFUELASSEMBLY(Cont) 4.11 REACTOR FUEL ASSEMBLY (Cont)
B.
Linear Heat Generation Rate (LHGR)
B.
Linear Heat Generation Rate (LHGR)
During reactor power operation.
The LHGR as a function of core the LHGR shall not exceed the height shall be checked daily limits specified in the CORE during reactor operation at 2 25%
OPERATING LIMITS REPORT.
rated thermal power.
If at any time during operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the LHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillance and corresponding action shall continue until reactor operation is within the
-Minimum Critical Power Ratio C.
(MCPR) prescribed limits.
- 1. MCPR shall be determined daily C.
Minimum Critical Power Ratio during reactor power operation (MCPR) at > 25% rated thermal power "U ""Y # ""E"
- 1. During power operation MCPR Power level or distribution shall be a the MCPR operating limit specified in the Core that would cause operation with a limiting control rod pattern Operating Limits Report.
If at I" O*
any time during operation it is S ecification 3.3.B.S.
P determined by normal surveillance that the limiting
- 2. The value off in Specification value for MCPR is being 3.11.C.2 shall be equal to 1.0 exceeded, action shall be unless determined from the initiated within 15 minutes t restore operation to within the result of surveillance testing of Specification 4.3.C as prescribed limits.
If the g,17,y steady state MCPR is not returned to within the a) r is defined as prescribed limits within two (2) hours, the reactor shall be fave - f B brought to the Cold Shutdown
'~
condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />' Surveillance and corresponding 1.252 -rB action shall continue until reactor operation is within the prescribed limits.
l l
Revision Amendment No.
15,-27 42,-547-105;-133 3/4.11-2 2
1
'LIMITkNGCONDITIONSFOROPERATION SURVEILLANCE REQUIREMENTS 3.11 REACTOR FUEL ASSEMBLY (Cont) 4.11 REACTOR FUEL ASSEMBLY (Cont)
C.
Minimum Critical Power Ration MCPR C.
Minimum Critical Power Ration MCPR (Cont'd)
(Cont'd)
- 2. The operating limit MCPR values
- b. The average scram time to as a function of the r are dropout of Notch 34 is given in Table 3.3.1 of the determined as follows:
Core Operating Limits Report where r is given by n
specification 4.11.C.2.
Z Ni ri rave =
1"1 n
Z Ni i=1 Where: an n - number of surveillance tests performed to date in the cycle.
Ni-number of active control rods measured in the ith surveillance test.
ri - average scram time to dropout of Notch 34 of all rods measured in the ith surveillance test.
- c. The adjusted analysis mean scram time (rB) 18 calculated as follows:
r-1 l
1 lM r3 = y + 1.65l l
a l
l n
l Z Ni l l i=1 l
L_
_J Where:
mean of the distribution for p
average scram insertion time to dropout of Notch 34, 0.937 see.
Ni-total number of active control rods at B00 during the first surveillance test.
standard deviation of the a -
distribution for average scram insertion time to the dropout of Notch 34, 0.021 seconds.
1 Revision Amendment No. 24;-42-54;-59 -133;-138 3/4.11-3 5
.i
=
F Attachment C: Marked-up Tech Spec Pages i
i 1
l i
i BASES:
2.0
. SAFETY LIMITS (Cont)
FUEL CLADDING Since the pressure drop in the bypass region is essentially all INTEGRITY (2.1.1) elevation head, the core pressure drop at low power and flows will I (Cont) always be greater than 4.5 psi.
Analyses show that with a bundle flow of 28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.
Thus, the l bundle flow with a 4.5 psi driving head will be greater than 28 x 102 lbs/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.
With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis, GETAB (1), which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power.
The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) - Boiling Length (L), GEXL, correlation.
The GEXL correlation is valid over the range of conditions used in the tests of the data used to develop the correlation.
These conditions are:
, yog p,y, L Pressure:
800 to({400 psihM Max Flux:
0.1 to (1. 2M103J1b/hr-f t 2
/*4 Inlet Subcooling:
0 to {0B tu/lb (LocalPeaking:
1.01 as a uutuer rod 1.47 at an interior rod Axial Peaking:
Shape Max / Avg.
Uniform 1.0 Outlet Peaked 1.60 Inlet Peaked 1.60 Double Peak 1.46 and 1.38 Cosine 1.39 Rod Array 6,64 Rods in an 8x8 arra] 4x9 (,( ll o.rr 49 Rnde in an 7x7 arr y MINIMUM CRITICAL The fuel cladding integrity Safety Limit is set such that no fuel POWER RATIO damage is calculated to occur if the limit is not violated.
Since (2.1.2) the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not result in damage to BWR fuel rods, the critical oc,wer at (Cont)
Revision +'H-Amendment No. 15;-42;-72;-195;-129 -133 B2-2 7
~
l' LIMITING CONDITION FOR OPERATION SDRVEILLANCE RE0tfIREMENT 3.3 REACTIVITY CONTROL (Cont) 4.3 REACTIVITY CONTROL (Cont)
B.
Control Rods (Cont)
B.
Control Rods (Cont)
- 4. Control rods shall not be 4 Prior to control rod withdrawal withdrawn for startup or for startup or during refueling unless at least two refueling, verify that at least source range channels have an two source range channals have observed count rate equal to or an observed count rate of at greater than three counts per least three counts per second.
second.
- 5. The RBM shall be operable as required in Table 3.2.C-1, or control rod withdrawal shall be blocked.
C.
Scram Insertion Times
,)E #
C.
Scram Insertion Times f'lITheaveragescraminsertion
- 1. Following each refueling time, based on the outage, or after a reactor deenergization of the scram shutdown that is greater than pilot valve solenoids as time 120 days, each operable control zero, of all operable control rod shall be subjected to scram g
ofl,c-e rods in the reactor power time tests from the fully operation condition shall be no withdrawn position.
If testing greater than:
is not accomplished with the j(~ st/}-
s nuclear system pressure above
% Inserted Average Scram 950 psig, the measured scram From Fully Insertion insertion time shall be Withdrawn Times (set) extrapolated to reactor pressures above 950 psig using 10
.55 previously determined 30 1.275 correlations.
Testing of all 50 2.00 operable control rods shall be 90 3.50 completed prior to exceeding 40% rated thermal power.
l Revision 477-Amendment No. 15;-68; 124;-138 3/4.3-4
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, LIMITING CONDITION FOR OPERATION SURVEILIANCE REQUIREMENT 3.3 REACTIVITY CONTROL (Cont) 4.3 REACTIVITY CONTROL (Cont)
~
' cram Insertion Time (Cont)
/
S C.
Scram Insertion Time (Cont)
C.*
{2.Theaverageofthescram
- 2. Within each 120 days of insertion times for the three operation, a minimum of 10% of fastest control rods of all the control rod drives, on a g
k k g6 groups of four control rods in rotating basis, shall be scram a two by two array shall be no tested as in 4.3.C.l.
An e
I5 h$d/[
greater than:
evaluation shall be completed every 120 days of operation to
% Inserted Avg. Scram provide reasonable assurance From Fully Insertion that proper performance is Withdrawn Time Sec.
being maintained.
10
.58 30 1.35 50 2.12 90 3.71
- 3. The maximum scram insertion h b
4 time for 90% insertion of any i rg ',-
operable control rod shall not D.
,,, [
exceed 7.00 seconds.
Once a shift, check the status of D.
Control Rod Accumulators the pressure and level alarms for ea accumulator, At all reactor operating pressures, a rod accumulator may be inoperable provided that no other control rod in the nine-rod square array around this rod has a:
i
- 2. Directional control valve electrically disarmed while in a non-fully inserted position.
- 3. Scram insertion time greater than the maximum permissible insertion time.
If a control rod with an inoperable accumulator is inserted " full-in" and its directional control valves are electrically disarmed, it shall not be considered to have an inoperable accumulator.
Revision 177 65 -124r-146 3/4.3-5 Amendment No.
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EASES:
3/4.3 REACTIVITY CONTROL (Cont)
C.
', Scram Insertion Times The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; i.e.; to prevent the MCPR from becoming less than the Safety Limit MCPR. Analysis of the limiting power transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above Specification, provide the required protection, and MCPR remains greater than the Safety Limit MCPR.
The scram times for all control rods will be determined at the time of each refueling outage. A representative sample of control rods will be scram tested during each cycle as a periodic check against deterioration of the d
, [2 control rod performance.
er[-
)w
(,y 7
In the analytical treatment of the transients, 290 milliseconds are allowed between a neutron sensor reaching scram point and the start of negative reactivity insertion.
This is adequate and conservative when compared to the typical time delay of about 210 milliseconds estimated from scram test results. Approximately 120 milliseconds later, the control rod motion is estimated to actually begin. However, 200 milliseconds is conservatively assumed for this time interval in the transient analyses and this is also cluded in the allowable scram insertion times of Specification 3.3.C.
D.
Control Rod A'ecumulators Requiring no more than one inoperable accumulator in any nine-rod square array is based on a series of XY PDQ-4 quarter core calculations of a cold, clean core.
The worst case in a nine-rod withdrawal sequence resulted in a keff
<l.0 - other repeating rod sequences with more rods withdrawn resulted in keff
>1.0.
At reactor pressures in excess of 800 psig, even those control rods with inoperable accumulators will be able to meet required scram insertion times due to the action of reactor pressure.
In addition, they may be normally inserted using the control-rod-drive hydraulic system.
Procedural control will assure that control rods with inoperable accumulators will be spaced in a one-in-nine array rather than grouped together.
i i
1 Revision 177 Amendment No. 152-425-138 B3/4.3 6
+
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LIMITING CONDITIONS FOR OPEPATION SURVEILIANCE REQUIREMENTS 3.11, REACTOR FUEL ASSEMBLY (Cont) 4.11 REACTOR FUEL ASSEMBLY (Cont)
B.
Linear Heat Generation Rate (LHCR)
B.
Linear Heat Generation Rate (LHGR)
During reactor power operation, The LHGR as a function of core the LHCR shall not exceed the height shall be checked daily limits specified in the CORE during reactor operation at t 25%
OPERATING LIMITS REPORT.
rated thermal power.
If at any time during operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the LHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillance and corresponding 1 &, @g p action shall continue unt 1 C.
Minimum Crleical Power Ratio
%rW29, n <.. reactor operation is with n the
- Q3 43ew (MCPR)
. x..n - a 4 prescribed limits.
nailp g J.,.
- 1. MCPR shall be determined;dailf g %
- .2C yr @ inimum Critical Power Ratio during reactor power operation p f g.
hiisW WZYX ((MCPR)"
at > 25% rated thermal pw n gd M.;
n
~P and following any change >in u W m
in <
s a.
-fB During power operation MCPR
- *mW*
~
pawer level; gar distribuctortqw%p+A O ;;
un
%u. hall be a the MCPR operating n4
~
ten virh;"_ :s w
that would-cause operation yltha;.
t :;<
limit specified in_the Core t; a limiting' control rod pattern. % %.
ie
. i ut tcrn 4;P
~
w.,.J
. M.. m perating Limits Report.
If at E
J as desoribed_in the bases?fw g m %.
tt orm
} 3 **If1**ti n 3.3.B.S.
S m @. #p W ec.anf time,during operation,it is 1
dNh7dat,erminedby~ normal.
P -
- O surveillance
- that the limiting' M,. -
d Ciut.mm.. he for MCPR is being 2'. The value of,r. in Spe'cificath+NC ~CA
. o xa a
w n ~L mos 3.11. C. 2. shall be. equal;to31g~w e %+ w
- exceeded, action shall be m~
0
+
. s==
ra unless deteimined from the initiated within 15-eibutes to
-. = -
-r result of-surveillance testingiu h r
'o tiue
,3
. restore, operation to within the a
of Specification 4.3'.C as-c
.r-prescribed limits., If the c
follows' nc :
steady state MCPR is not y '- Qp 3 N
returned to within the
,M L
Y. prescribed limits within two a) r is defined as ggp I (2) hcurs, the reactor shall be kl
- 7
-f brought to the Cold Shutdown ave B
condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillance and corresponding
- fB action shall continue until reactor operation is within the prescribed limits.
/EN Revision 177 Amendment No. 15--2b -39;-42--s4--105--133 3/4.11-2
g-3 CN TTwITING CONDITIONS FOR OPERATION SURVEILIANCE REQUIREMENTS 3.11; REACTOR FUEL ASSEMBLY (. Cont) 4.11 REACTOR FUEL ASSEMBLY (Cont)
C.
' Minimum Critical Power Ration MCPR C.
Minimum Critical Power Ration MCPR' i
.(Cont'd)
(Cont'd) g,,, p g g41, gy
)
- 2. The operating limit MCPR values
- b. The average scram time to as a function of the r are ffh som insertion positio given in Table 3.3.1 of the
.is determined as follows:
Core Operating Limits Report o
.where r is given by n
specification 4.ll.C.2.
I Ni ri rave "
I"1 i
n I'
'Ni-i=1 Where: an n - number of surveillance tests performed to date in the cycle.
Ni-number of active control rods measured in the in surveillance g, q,,-[ ; of. A/o del s y test.
ri - average scram time to fhe 30%
/fnsertionpositiongfallrods
~ measured in the im surveillance i
l
-- te s t.
2
- c. The adjusted analysis mean scram time (rB) is calculated as follows:
F "o[ $
l l
\\
su rB = y + 1.65 l l
a i
l n
l i
l I
Ni l
}
l i=1 l
l L.
J Where:
i mean of the distribution for i
p de,ge./h e[ /,d/Jy
,w(the 304 position, 0.945 sec3 average scram insertion time to MO - total number of active control rods '
g, P soc cL.oroh the. -C Os V s n.ra.,f/ wet. +uf. :
a - standard deviation of the distribution for average _ scram insertion time g the 10% de (position, 0.064 sec.
d r op o.o l o f YO N~Ch 3 $ O'
- ~ NA Revision 177
. Amendment No. 24r-42-54r-59 -133r-138 3/4.11-3 t
~_ - -., -.-