ML20077K924
| ML20077K924 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 07/29/1991 |
| From: | NEBRASKA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML19302E996 | List: |
| References | |
| NUDOCS 9108090194 | |
| Download: ML20077K924 (26) | |
Text
- _ - _ _ _ _._
__ to NLS9100187 l
Page 16 of 16 i
REVISED TECHNICAL SPECIFICATIONS PAGES 9108090194 910729 Dh ADOCK 0500029:3 PDR
$AFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 17, 1 ; FUEL CLADDING INTEGRITY 2.1 FUEL CIADDING INTEGRITY Aonlicability Aonlicability The Safety Limits -established to The-Limiting Safety System Settings preserve tho fuel cladding integrity apply to trip settings of the apply.- to _ those variables which instruments and devices which are monitor the fuel thermal behavior, provided to prevent the fuel cladding integrity Safety Limits 5
from-being exceeded.
-Qbieccive Obiective The obj ective of the-Safety Limits is to establish' limits below which The objective of the Limiting Safety the integrity of the fuel cladding System Settings is to define the
- is preserved.
level of the process variables at which automatic protective action is Action initiated to prevent the fuel cladding integrity Safety Limits If a Safety Limit is exceeded, the from being exceeded, reactor shall be in at least hot shutdown within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, s.pecifications Specifications
^'
-A.
Reactor Pressure >800 esia and Core Flow >10% of Rated The limiting safety system The existence of a _ minimum trip settings shall be as
. MCPR) specified below:
critical ' power -ratio
(
less than 1.06 for two recirculat. ion loop operation 1.
Neutron Flux Trio Setting.g (1.07 for single-loop operation) shall constitute a.
APPR Flux Scram Trio settinn violation of the fuel (Run Mode) cladding integrity safety.
(1)
Flow Re fe renc ed Scram Trb B.
fore Thermal Power Limi.;;
Settine (Reactor Pressure <800 nsia When the Reactor Mode Selector and/or ForAFlow f10%)
is in the RUN position, the i-When the-reactor pressure is APRM flow referenced flux.
(800 psia or core flow is scram trip setting shall be:
less than 10% of. rated, the Ss0.58 W + 62% -. 5 8 t.W core thermal power shall not exceed 25% of rated-thermal where:
P "# '
S - Setting in percent of l
C.
Power Transtent rated thermal power
(
1#}
To ensure that the Safety Limit established in W - Two-loop recirculation Specification 1.1.A and 1.1.3 flow rata in percent of is not
- exceeded, each rated (rated-loop required scram shall be ini-recirculation flow rate tiated by its expected scram is that recirculation flow signal.
The Safety Limit rate which provides 100t shall be assumed to be core flow at 100% power;
{
exceeded when scram is
, g
,g accomplished by a means other T
P "d '
b" ~,#
than the e$pected scram ettectrce dri'/e t ww at c Qnal, the name cote flow 6-
SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS
.: 1.1-(Cong'd)!
- 2,1.A.l.a.
(Cont'd)
. D.-
.Q_old shutdown AW - O for-two -
recirculation-loop Whenever the reactor is in the cold--
operation.
l shutdown condiuion with irradiated fuel in the reactor. vessel, the (2)
Fixed APRM Flux Scram Trio Settine water level shall not be less-than 18 in, above the top of the normal The fixed APRM flux scram trip active fuel zor:e (top of active fuel setting shall-not be allowed to.
.)
is defined in ?igure 2.1.1),
exceed-120% of rated. thermal power.
- )
b.
APRM Flux Scram Trio Setting (Refuel
)
or Startun and Hot Standby Mode)-
i When the Reactor Mode Selector Switch is in the REFUEL or STARTUP.
position, the APRM scram shall be set at less than or equal to 15% of rated power, c.
IRM The IRM flux scram setting shall be
$120/125 of scale.
i e
w l
s,
.. _...,..-. ~ _
SAFETY LIMITSI LIMITING SAFETY SYSTEM SETTING 9 2.1.A.1 (Contfd) d._
APRM Rod Block Trio Setting The APRM rod block trip setting shall be:
S s.0.58 W + 50% --.58 AW-g with a maximwn of. s 108% of rated
- power, where:
S
- Rod block setting in RB percent of-rated thermal power (2381 MVt).
W and
-AW are
-- de fined in Specification 2.1.A.1.a.
I 2.
Reactor Water Low Level Scram and Isolation Trio Setting (excent MSIV) 2 +12.5 in, on vessel-level law t rwnents.
Ib F
f i
l.
-d s
r,--,
e
.-n.
-2;l Eases: _.. (cont ' d)
An increase in the_APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached :The APRM scram trip setting was reasonable range for determined by; an analysis.of margins required to provide a
maneuvering-during operation.
Reducing this operating cargin would increase the frequency of spurious scrams which have an adverse effect on reaccer safety because of the resulting thermal stresses, Thus,'the APRM scram trip settinb was selacted because it-provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that_ reduces the possibility of unnecessary scrams, I
b,
_APRM Flux Scram Trin Setting (Refuel or Start & Hot Standby Model
-For operation-in the_startup mode while the _ reactor-is et low pressure, the APRM
- scram setting of 15 percent of rated power provides adequate thermal margin between the setpoint_and the safety limit, 25 percent of rated, The margin is adequate to accommodate anticipated maneuvers associated wich power plant startup.
Effects of
~ increasing pressure at tero or low void concent are minor, cold water from sources available during startup _ is not much colder than that already in the system; temperature coefficients are small, and control rod patterns are constrained to be
+
uniform-by operating procedure backed up bi the-rod worth minimiter, and the rod sequences control system.
Vorth of individual rods is very low in a uniform rod pattern.
Thus, _ of all-possible sources of reactivity input, uniform control rod withdrawalL s the most probable cause of significant power rise.
Because the flux i
distribation associated with Juniform rod withdr:wal? does not involve high local
. peaks, and because several rods must be moved to change power by a significant
- percentage of rated power,. the rate of powar rise is very slow. Generally, the heat flux is in near ' _ equilibrium with the fission rate, in an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than-5 percent; of rated power per minute, and the APRM system would be more than adequate to assure a scram before the_ power could exceed the safety limit, The-15 percent APRM scram remains active until the mode switch is placed in the RUN position. This change can_ occur when_ reactor pressure is greater than Specification 2.1.A,6, h
k k
13-
2,1 Eases (Cont' d) c.
IRM Flux Scram Trip Setting The IRM system consists of 8 chambers, 4 in each of the reactor protection system logic channels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered by tha IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one half of a decade in size.
The IRM scram trip setting of 120 divisions is active in each r mge of the IRM.
F% exampla, if the instrument were on range 1, the scr-a setting would be 120 divisions for that range; likewise, if the ins t rwnenu "are on rary 3,
the scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accommodate the increase in power level, the scram trip setting is also ranged up.
The most significant nurces of reactivity change during the power increase are due to control red withdrawal.
For in-sequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing co%rol rods, that heat flux is in equilibrium with the neutron flux and en IRM scram would reruit in a reau or shutdown s ell before any Safety Limit is exceeled.
In order to ensure that the IRM provided adequa.e protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels.
The most severe case involves an initial condition in which the reactor is just suberitical and the IRM system is not yet on scale. This condition exists at quarter rod density.
Additional conservatir.m was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is by passed. The results of this analysis show that the reactor is tcrammed and peak power limited to one percent of rated power, thus maintaining MCPR 'bove the MCPR fuel cladding integrity safety limit.
Based on the above analysis, the IPJi provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
d.
APM Rod Block Trin Settine Reactor power level may be varied by moving control rods or by varying the recirculation flow rate.
The APRM system provides a control rod block which is dependent on recirculation flow rate to limit rod withdrawal, thus protecting against a MCPR of less than the MCPR fuel cladding integrity safety limit.
The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the Safety Limit increases as the flo~
decreases l'or the specified trip setting versus flow relati w %. therefore the sorst case MCPR which could occur during steady-state operation is at 10EL of rated thermal power because of the APF21 rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPPJi system.
l
y IJMI?115 CONDITIONS FOR ' OPERATION SURVEILIANCE REOUIREMENTS
-3.1 REACTOR PROTECTION SYSTEM-4.1 TREACTOR PROTECTION SYSTEg
= Applicability Applicability:
Applies to the instrumentation and Applies-to the-surveillance of_the -
associated devices which initiate a instrumentation and associated
-reactor scram.
devices-which initiate reactor scram.
Obiective:
Obiective:
To ) assure the _ operability of the To specify the type and frequency of reactor protection system.
surveillance to be applied to the protection instrumentation.
-1pecification:
Soeci fic a ti.Qnl. -
The-setpoints, ' minimum number of A.
Instrumentation systems shall trip systems, and minimum number of be-functionally tcsted-and instrument - channels that must be calibrated as indicated in operable for each position of the Tables 4.1.1 and 4.1.2 reactor. - mode - switch shall be _ as respectively.
given;in Table 3.1.1; B.
Deleted.
C.
Deleted.
D.
When it is determined that a channel has failed in the unsafe condition, the-other RPS channels that monitor the same variable shall be fun'ctionally c_e s t e d immediately before.the - trip system containing the failure is tripped, The trip system -
containing the unsafe' fallute may-be placed in the untripped condition during the' period in which surveillance testing is.
being performed on the-other RPS channels.
' COOPER IRJCLEAR STATION
. TABLE 3.1.1-'
REACTOR. PROTECTION SYSTEM INSTRUMENTATION' REQUIREMENTS Minimum Number.
Action Required i
j
.When Equipment Applicability Conditions-of Operable Peactor Protection' Mode Switch Position Trip. Level Channels Per; Operability is" System Trio Function.
Sluitdown ' ' S tartup - Refuel Run Settine Trio Systems (1) Not Assured (1) 11 ode Ewitch in Shutdown
'X(7)'
X X
'X 1
qA flanut 1 Scram '
X(7)
.X X
X 1
A 1R11
'17)
X(7)
X X
(5)
's 120/125 of.in-3-
A' dicated scale liigh Flux l
InopecatIve X-X (5)'
3
.A APit!! (17)
X's (0.58W+62%-0 58AW).
2 C.
Illgh Flux (Flow biased)
(14)(19)-
/,
with a maximum of 120%
'?
of rated power liigh Flux (fixed)
X s'120% of Rated Power 2
A 111gh Flux
.X(7)
X(9)
X(9)
(16) s 15% Rated Power-2 A
Inoperative X(9)
X(9)
X (13) 2 A
Downseale
-(12)
(12)
(12)
X(11) 2 2.5%'
2 A
liigh iteactor Pressure X(9)
X(10)
X
's.1045 psig 2
A UBI PS-55 A,B,C, & D liigh Drywell Ptessure
- X(9)(8) X(8)
X 5 2 psig 2.
A or D PC-PS-12 A,B,C, & D iteact or I nw Water Level X'
X X
2.+
12.5 in. indi-2 A or D Uhl-l.lS-101 A,B,C, 6D
-cated level Scram Discharge. Instrument Volume X
X(2)
X 5 92 inches 3 (18)
A liigh Water Level Cit D - 1.S - 2 3 l A 6 B CitD-LS-2 34 A & B '
Citb-1.T-231 C & D Cl4D-1.T-234 C & D
~-
A
- 11. The APRM downscale trip function is only active when the Reactor Mode Selector Switch '
-is in RUN.1 When in RUN -.- this fimetion __is automatically bypassed when the companion 11RM instrumentation is operable-and not upscale.
- 12. The APRM downscale trip---is - aut omatically bypassed when the Reactor Mode Selector
. Switch is not in RUN.
13-An APRM will'be' considered inoporable if there are less than 2 LPRM inputs per level or there is less than 11 operable-LPRM detectors to an APRM, 14 W is the two loop. recirculation flow in percent of rated flow,
- 15. This note deleted.
l
- 17. The APRM and 'IRM instrument channels function in both the Reactor Protection System and Reactor Manual Control System (Control Rod Withdraw B'.4 k, Section 3,2.C.),
A failure of one-- channel will affee t both of these systems.
- 18. The minimum number opurable associtted with the Scram ' 'scharge Instrument Volume.are three instruments per Scram Dischrrge Instrument Volutne and three level devices per RPS channel.
L19. AW is the difference between two-;.oop and single-loop effective drive flow.and is used for single. recirculation loop operation.
AW-0 for two recirculation loop operation.
1 4
M'
_,~
- -.,. ~......, - -
l 1
i L3MITING CONDITIONS F,QR OPERATION 311gyEILIANCE REOUIREMENTS
-3.1 BASES (Cont'd.)
4 '1 BASES _(Cont'd,)
l there-isl proper overlap in the For.the.APRM system, drift of neutron monitoring system functions electronic apparatus is not the only i
Eand thus,;that adequate coverage-is consideration in. determining a
provided for all ranges of reactor calibration frequency, Change in operation, power-distribution-and loss of chamber sensitivity-dictate' a
~
calibration every seven days.
Calibration on.
this frequency assures plant operation at or below thermal limits.
A comparison of Tables 4.1.1 and -
4.1,2 indicates that two instrument channels have not been included in the latter table. These are: mode switch in shutdown and manual scram.
All of the devices or sensors associated ~
with these scram-functions are simple on off switches
- and, hence, calibration during operation is not applicable.
B.
The sensitivity of LPRM detectors decreases.with exposute to neutron flux at a. slow and approximately-constant rate, 'This is compensated for in..the APRfi. system by calibrating once a week using a heat balance data and by calibrating individual LPPJt's every six weeks 'of power operation above 20% of rated power.
9kb
me nHIJF1AFALLt LEFT BLME" i
's : -
_i-
^
d i'
COOPER NUCLEAR STATION TABLE.3.2.C (page 1) fI
'GONTROL ROD WITHDRAWAL BIDCK INSTRUMENTATION l
i flinimum Number Of Trip Imvel Setting Operable Instrument Function Channels / Trio Systesf5) i 4
i.
'APRf1 Upscale (Flow Bias) s (0.58V + 50% - 0.58 IN) (2)(13) 2(1) with a maximum of $ 108% of rated power 2(1)
APRM Upscale (Startup) 5 12%
2(1) i APRM lbwnscale (9) 2 2.5%
2(1)
I' A14:M Inoperative (10b) 1(15)
RBn Upscale (power referenced):
(14) b lowest Rated 3
MCPR Lirait LTSP*
ITSP*
IITSP*
2 1.20
$ 118.0 s 112.0 s 108.0 1
2 1.25 5 121.0 s 116.0 s 111.0 7
2 1.30 5 124.0 s 119.0 5 114.0 2 1.35 s 127.0 s 122.0 s 117.0 i
Not Applicable (15)
RBIl Power Range:
(14) i i
law Power Setpoint (LPSP) s 30% of rated power i
Internaediate Power Setpoint (IPSP)'
s 65% of rated power l-Itigh Power Setpoint (IIPSP)
S 851'of rated power RBt1 Downscale 2 94*
1(15)
EBM Inoperative (10c)(15) 1(15)
- Values are relative to REM initialization reference of 100/125 of full scale upon control rod selection for movement.
j
- 1. TSP - low Trip Setpoint.
ITSP - Intermediate Trip Setpoint liTSP - High Trip Setpoint s
i Inct Upscale (8)
$ 108/125 of Full Scale 3(1) i i
IRn Downscale (3)(8) 2 2.5/125'of Full Scale 3(1)
\\.
3(1)
I IRM Detector Not Full In (8) 7 i
k..
~
~'
.--_.__m,
]
j r
COOPER NUCLEAR STATION TAP,LE 3.2.C (page 2).
[
CONTROL ROD WITl!LRAVAL BLOCK INSTRLMENTATION 1
i i
I Minimum Number Of.
Function Trip level Setting Operable Instru=ent r
l-Channels /Tric Svstemf5}'
[
l:
l IEn 1tmperative (8),
(loa) 3(1) i 5
SRM Upscale (8) s 1 x 10 counts /Second 1(1)(6) 1 i
SRf! Detector Uct Full In (4)-(8)
Q 100 cps) 1(1)(6).
l i
l SR!! luoperative (8)
(10a) 1(1)(6) i i
Flow Blas Comparator
- $ 10% Difference In Recire. Flows 1
flow Isias Upscale /Inop.
f 110I Recire. Flow I
S0:1 Ituunscale (8)(7)
> 3 Ccunts/Second (11) 1(1)(6) h I.
T SD7 L t e r 1. eve l lii gh 5 46 inches 1(12) e i
G D.' 31 E. 234E 1
1 7
1 1
i
[.
1 i
l 4
a
~
,.sw
--6q p
p--..
-~7,.
-er-'*
e---
WN--
- ww-- - -
rw-6T
--+m-wetr
- - - -.4-
NOTES FOR TABl.E 3.2.C 1.
For the startup and run positions of the Reactor Mode Selector Switch, the control 7
Rod Withdrawal Block Instrumentation trip system shall be operable for each function.
The SAM and IRM rod blocks need not be operable in the RUN mode, and the APRM (flow biased) rod blocks need not be operablo in the STARTUp mode.
The Control Rod Withdrawal Block Instrumentation trip system is a one oat of "n" trip system, and as such requires that only one instrument channel specified in the function column must exceed the Trip Level Setting to cause a rod block.
By utilizing the RPS bypass logic (see note S below and note 1 of Table 3.1.1) for the Control Rod Withdrawal Block Instrumentation, a sufficient number of instrument channels will always be operable to provide redundant rod withdrawal block protection.
- 2. W is the two loop recirculation flow rate in percent of rated.
Trip level setting is in percent of rated power (2381 MWt).
l 3.
IRM downscale is bypassed when it is on its lowest range.
4.
This function is bypassed when the count is 2100 cps and IRM above range 2.
For the APR4s and IRMs, the minimum number of channels specified is that minimum number required in each RPS channel and does not rafer to a minimum number required by the control rod block instrumentation trip function. By design only one of_two RBMs or one of four SRMs may be bypassed.
For the SRMs. the minimum number of channels specified is the minimum number required in each of the two circuit loops of the. control Rod Block Instrumentation Trip System.
l 6
IRM channels A,E,C,G all in range 8 or higher bypasses SRM channels A&C functions, IRM channels B,F,D,M all in range 8 or higher bypasses SRM channels B&D functions, 7.
This function is bypassed when IRM is above range 2.
8.
This function is bypassed when the Reactor Mode Selector Switch is placed in RUN.
t 9.
This function is-only active when the Reactor Mode Selector Switch is in RUN.
- 10. The inoperative trips are produced by the following functions:
a.
Mode switch not in operate (2)
Power supply voltage low (3)
Circuit boardo not in circuit l
(4)
Loss of negativo Supply voltage l-e,~.ree-ve--,--s,sm--.
, w. e - - -- ~w =
w,,-,
.. e -w m e,~
,n e w a r -. v,w a,
.,,v~ ~ w wr e ~ o
,w s,- w wm em, m -- w c.
HOTES FOR TABLE 3.2.0 (Continued) b.
APRM (1)
Mode switch not in operate (2)
Ltiss than 11 LPRM inputs S
(3)
Circuit' boards not in circuit c.-
RBM l
(1)
Mode switch not in operate l
(2)
Circuit boards not in. circuit (3)
RBM fails to null (4)
Less than required number of LPRM inputs for rod selected t
- 11. During spiral unloading / reloading, the SRM count rate will be below 3 cps for some period of time.
See Specification 3.10.B.
l i
- 12. With the number of OPERABLE channels less than required by the Minimum Number of Operable Instrument Char.nels/ Trip Syctem requirements, place the inoperable channel in the tripped condition within one hour.
13, AW is the difference between two. loop and single loop effective drive flow and is used for single recirculation loop. operation.
aW-0 for two recirculation loop operation.
14.-One set of power referenced RBM upscale trip settings (LTSP, ITSP and HTSP) is
. applied based on the lowest rated MCPR limit given in Specification 3.11.C.
The RBM power range satpoints control the enforcement of the appropriate upscale trips over the proper cere thermal power range as follows:
All RBM trips are automatically bypassed below the Low Power Setpoint (LPSP).
a.
-b.
The upscale Low Trip Setpoint (LTSP) is applied at the LPSP and up to the
- Inturmediate Power Setpoint-(IPSP).
The upscale Intermediate Trip Setpoint (ITSP) is applied at the IPSP up to the c.
High Power Setpoint (HPSP),
d.
The upscale High Trip Setpoint (HTSP) is applied at and above the HPSP.
- 15. The RBM.is only required when core thermal power is 2 30% of rated powec and a I
limiting control rod pattern (defined in Specification 3.3.B.5) exists. Requirements for operating-with-a limiting control rod-pattern are specified in Specification 3.3.B.S.b.
i
- - _. ~
l 3.2 DME (cont'd. )
C.
Control Rod Block Actuatirn The control rod block functions are provfded to prevent excessive control rod withdrawal so that MCPR does not decrease to the safety limit CPR.
The trip logic for this function is 1 out of n:
e.g., any trip on one of six APRMs. eight IRMs, or four SRMs will result in a rod block.
The minitnurn in.strument channel requirements assure sufficient instrumentation to i
assure the single failure criteria is met, The minitnurn instrument channel requirettents for the RBM may be reduced by one for maintenance, testing, or calibration. This time period is only 3% of the operating tirne in a month and does not significantly increase the rise of preventing an inadvertent control rod l
.'thdrawal.
The APRM rod block function il flow biased and prevents a significant reduction in MCPR, especially during opere' tion at reduced flow.
The APRM provides gross core protection; i.e., limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than the safety limit CPR.
The RBM rod block function provides local stection of the core; i.e.,
the prevention of the MCPR reaching the safety limit CPR in a local region of the corn, for a single rod withdrawal error frem a limiting control rod pattern.
Additional details are provided in Bases Section 3.3.B.S.
The IPJi rod block function provides local as well as gross coro protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the
-indicated level.
A downscale indication on an APRM or 1RM is an indication the instrument has failed or the instrument is r't sensitive enough.
In either case the instrument will not respond to changes in atrol rod motion and thus, control rod mation is prevented.
The downscale trips are set at 2.5 indicated on scale.
The flow comparator and scram discharge volume high level components have only one logic channel and are not required for safety.
The SDV high level rod block does provido adequate time to determine the - cause of the level increase and take corrective action prior to automatic scr m.
P The refueling interlocks also operate one logic channel, and are required for safety
-only when the modo switch is in the refueling position.
D.
Radiation Monitority Systens - Isolation and Initiation Fimedrng 1.-- Steam Jet Air Ejector Off Gas System Two air ejector off gas monitors are provided and when their trip point is reached, cause an isolation of the air ejector of f-gas line. Isolation is initiated when both instruments reach their high trip point or one has an upscale trip and the other a downscale trip.
There is a fif teen minute delay accountr-d for by the 30. minute holdup time of the off-gas before it reaches the stack.
l I
1 4
'n u.
LIMITING CONDITION FOR OPERATION SURVEILMNCE PEOUIREME!!I 3.3.B.3 (cont'd) 4.3.B.3.b (cont'd) e.
If Specifications 3.3.B.3a 1)
The correctness of the Banked through d cannot be met, the Position Vithdrawal Sequence reactor shali not be started, input to the RVM computer or if the reactor is in the shall be verified, run or startup modes at less than 20%
rated power, it 2)
The RWM computer on lins shall be brought to a
diagnostic test shall to ahutdown eondition successfully performed.
immediately.
3)
Proper annunciation of the f.
The sequence restraints selection error of at least imposed on the control rods one out of sequence control may be removed by the use of rod in each fully inserted the individual rod position group shall be verified, bypass switches for scram testing only those rods which 4)
The rod block function of the are fully withdrawn in the RWM shall be verified by 100%
to 50%
rod density withdrawing the first rod as an out of sequence control rod
- range, n
more than to the block 4.
Control rods shc11 not be P
1"
withdrawn for startup unless at least two source range When required, the presence of c.
chsnnels have an observed a second licensed operator or count rate equal to or other qualified employee to greater than three counts per verify the following of the
- second, correct rod program shall be verified.
Doeration with a
- Limiting, Control Rod Pattryn (for fad 4.
Prior to control rod W.11thdrawal Error. RWM withdrawal for startup, verify that at least two source range a.
A Limiting Control Rod channels have an obse rved Pattern for RWE exists when e unt rate of at least three either:
counts per second.
is 1)
Core thernal (power 90%
of 5.
Operation with a
Lirritint 2 30%
and rated power and the Control Rod Partern (for Rod MCPR is less than 1.70 Withdrwal Error. RWE) or 2)
Core thermal power is During operation when a
Limiting Control Rod Pattorn b 90% of rated power for RWE exists and only one and the MCPR is less RBM channel is operable, an than 1.40.
f""
'I ""1 C**'
f ferformedf"*"*"'friorto b.
During operation with a
I ichdraw E control Rod Pattern f
p
,t ot the control rod (s).
A 1)
Both rod block monitor Limitine, Control Rod Pattern for RWE is defined by (RBM) channels r. hall be Specification 3.3.B.S.
~
operable, or 2)
With one RBM chanal t
inoperable, control rod uitndrawal shall be blocked witlin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or 3)
With both RBM channels inoperable, control rod l
withdrawal shall be blochad
,unt11 operabilit" ot at least one ch'annel is I
restored.
.n.
3.3 and 4.3 BASES:
(Cont' d) 5.
The RBM provides local protection of the core; i.e., the prevention of boiling transition in-a local region of the core, for a single rod withdrawal error from a Limiting Control Rod Pattern as defined in Specification 3.3.B 5.
The trip point is referenced to a power signal provided by the APRMs.
A=
statistical analysis (Reference 5) of many single control rod withdrawal errors has been performed.
At the 95/95 level the results show that with the specified trip settings, rod withdrawai is blocked at MCPRs which are greater than Safety Limit 1.1. A, thus allowing adequate margin. This analysis assumed operation at steady state operating limit MCPRa'(Specification 3.11.C) prior to the postulated rod withdrawal error.
The RBM functions are required when core thermal-power is greater than 30% and a Limiting Control Rod Pattern (Specification 3.3.B.5) exists.
When both RBM channels are operating either channel will assure required withdrawal blocks occur even assuming a single failure of one channel. When a Limiting Control Rod Pattern exists, with one RBM channel inoperable for no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, testing of the RBM before withdrawal of control rods assures that improper control rod withdrawal will be blocked.
A Limiting Control Rod Pattern for rod withdrawal error (RWE) exists when (a) core thermal power is greater than or equal to 30% of rated and less than 90% of rated (30% $ P ( 90%) and the MCPR is less than 1.70, or (b) core thermal power is greater than or equal to 90% of rated (P 2 90%) and MCPR is less than 1.40.
Under these conditions, complete withdrawal of a control rod could result in-MCPR violating the safety limits.
Therefore. RBM operation is required to block control rod withdrawal. RBM setpoints have been selected such that the required control rod blocks shall occur even if one of the redundant RBM channels fails.
Should one RBM channel become inopetable during the use of such patterns, it is judged that testing of the RBM system before withdrawal of such rods assures the remaining channel operability.
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. The limiting power transient is defined in Reference 2.
Analysis of this transient shows that the negative reactivity rates resulting from the scram provide-the required protection, and MCPR remains greater than the safety limit.
The surveillance requirement for scram testing of all the control rods after each refueling outage and 10% of the control rods at 16 week intervals is adequate for determining the operability of-the control rod system yet is not so frequent as to cause. excessive wear on the control rod system components.
The numerical values assigned to the predicted scram performance are based on the analysis of data from other BWR's with control rod drives the same as those on Cooper Nuclear Station.
The ~ occurrence of scrat-times within the limits, but significantly longer than the average, should be viewed as an indication of a systematic problem with control rod
-drives.
In the analytical treatment of the transients which - are assumed to scram on high neutron flux 290 milliseconds are allowed between a neutron sensor reaching the scram - point and start of motion of the control rods.
This is adequate and conservative when compared to the typical time delay of about 210 milliseconds L
estimated from scram test results. Approximately the first 90 milliseconds of each
[
of these time intervals result from the sensor and circuit delays; at this point. the l
pilot scram solenoid deenergizes. Approximately 120 milliseconds later.
i 102 i
l
3.3 and 4,3 l}ASES:
(Cont'd)
G.
scram Discharge Volume To ensure the Scram Discharge Volume (SDV) does not fill with water, the vent and drain valves shall be verified open at least once every 31 days. This is to preclude establishing a water inventory, which if sufficiently large, could result in slow scram times or only a partial control rod insertion.
The vent and drain valves shut on a scram signal thus providing a contained volume (SDV) capable of receiving the full volume of water discharged by the control rod drives at any reactor vessel pressure. Following a scram the SDV is discharged into the reacw building drain system.
REFERENCES 1,
" General Electric Standard Application for Reactor Fuel," NEDE 24011-P A-(latest approved revision).
2.
" Supplemental Reload Licensing Submittal for Cooper Nuclear Station," (applicable reload document).
3.
General-Electric Service Information Letter No. 380, Revision 1, dated Fobruary 10, 1984.
4.
General Electric Service Information Letter No, 316. Reduced Notch Worth Proceduta, November, 1979.
5.
" Extended Load Line Limit and ARTS Improvement Program Analysis for Cooper Nuclear Station Cycle 14," NEDC 31892P, Revision 1. May 1991.
. z o:,.
2.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.11 FUEL RODS 4.11 BIEL RODS Apolicability Applicability The Limiting Conditior.
for operation associated with the fuel The Surveillance Requirements apply reds apply to those parameters which to the parameters which monitor the monitor' the fuel rod operating fuel rod operating conditions, conditions.
Obiective Objective The Obj ective of the-Limiting The Objective of the Surveillance Conditions for Operation is-to Requirements is to specify the type assure the performance of the fuel and frequency of surveillance to be
- rods, applied to-the fuel rods.
Epecificationg Specifications A.
Averare Planar Linear Heat A.
Averare Planar Linear Heat Generation Rate (APLHGR)
Generation Rate (APUiGB.)c During steady state power operation, with :both recirculation loops The APLHGR for each type of operating, the APUICR for each type fuel-as a function of average of fuel as a. function of average planar exposure shall - be planar exposure shall not exceed the determined daily dt: ing limiting value specified in the Core reactor operation at 25%
Operating Limits Report for two rated thermal power.
recirculation loop operation.
For sin 6 8-100P Operation these values 1
are reduced for each fuel type as specified in the Core Operating Limits. Report.
If at any time during steady state operation it is determined by normal surveillance that. the limiting value for APUICR is being exceeded action shall be -
initiated within 15' minutes to-restore operation' to within the prescribed limits. If the APuiGR is not returned to within the prescribed limits within two (2) hours,-reduce reactor power to 5; 25% -
of rated ~ power within~ the next-four (4)-hours.
Surveillance and
- corresponding action shall continue until -the prescribed limits are again being met.
210-
i-LIMITING CONDITIONS FOR OPERAllQ1{ '
SURVEILIANCE REOUIREMENTS I
3.11,B Linear Heat Generation Rate 4.11.8 Linear llent Cencration Rate fulGIL).
(IJ1GR)
Durins steady stato power The UiGR as a function of core operation,.the linear heat height shall be checked daily generation rate (UlGR) of any during reactor operation at rod-in any fuel assembly at 2 25% rated thermal power.
any axial location shall not exceed the roaximum allowable i
UlGR as specified in the Core operating Limits Report.
If at any time during steady state operation it is i
determined by normal surveillance that the limiting ' value for UlGR is being exceeded action shall then be initiated to restore operation to
-within the l
p r e s _c r i b e d l i ra i t s.
l Surveillance and corresponding action shall continue until the prescribed limits are again being met.
L l
1 111-l
.--.____._,_._., -._,._ _..-._. _..--.___2._._
. ~.... -...
.L__IMITING CONDITIONS FOR OPIMT.19H SfhY. EILIANCE REQUIBf.11ERIS 3.11.C.
Minimum Critieni Power Reig 4.11.C.
liinlitum Critical Power Ratio DiCPRI (MCPR)
During steady state power MCPR shall be determined daily operation the MCI R fc,r each during reactor power operatio'n type of fuel at rated power at ) 25% rated thermal power l
and flow shall not be lower and following any change in than the limiting value power level or distribution specified in the Core that could cause operation on Operating Limits Report for the operating limit MCPR.
two recirculation loop operation.
If, at any time during steady state operation it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall then be initiated within 15 minutes to restore operation to within the prescribed litnits.
If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, reduce reactoc power to s 25% of rated power within the next four (4) hours.
Surveillance and corresponding action shall continue until the prescribed limits are again being met.
For one recirculation loop I
operation the MCPR limits are 0.01 higher than the comparable two loop values,
-:L -
.--.--__________,_,,_m__,
I 3.11 BASEE A.
Av.tutelmar. Linear litat Generation _ Rate ( APulGR)
This specification assures that the peak cladding temperature following the postulated design bawit loss of coolant accident will not exceed the limit specified in 10CPR50.46.
The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assembly.
Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than i 20'F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated ternperatures are within the 10CFR50.46 litni t. The litniting value for APulGR is shown in Figure 3.11-1.
The flow cependent correction factor MAPFACr specified in the Core Operating Limits Report is applied to the MAPLGilR limits at rated conditions to assure that (1) the 10CFR50.46 limit would not be exceeded during a LOCA initiated from less than rated core flow conditions and (2) the fuel thermal mechanical design criteria would be met during abnormal operating transients initiated from less than rated core flow conditions (Reference 11).
The power dependent correction factor MAPFAC, specified in the Core operating Liraits Report is applied to the MAPulGR limits at rated conditions to assure that the fuel thermal mechanical derign criteria would be met during abnorrnal operating transienta initiated from lesa than rated core power conditions (Reference 11).
The APulGR valves are reduced for single loop operation per Reference 10.
B.
Linant llent Ceut. ration Rate (QLGil).
This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densification is postulated, The UICR as.a function of cor9 height shall be checked daily during reactor operation at 2 25% power to determine if fuel burnup. or control rod movement has caused changes in power distribution.
For UlGR to be a limiting value below 25% rated thermal power, the MTPF would have to be greater a considerable margin when employing any than 10 which is precluded by permissible control rod pattern. Pellet densification power spihing in CE fuel has been accounted for in the safety analysis presented in Referenceu 1 and 2; thus no adjustment to the UICR limit for densification offects is required.
115-
l t il Bases:
(Cont'd) i C.
Minimum Critical Power Ratio (MCPR) 1 The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.11C are derived from the established fuel cladding integrity Safety Limit-and an analysis of abnormal operational transients 3
(References 2 and 11).
For any abnormal operating transient analysis l
evaluation with the initial cc,ndition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety 1.imit MCPR at any time during the transient assuming instrument trip settin6 given in Specification 2.1, j
To assure that the fuel cladding integrity Safety Limit is not exceeded during i
any anticipated abnormal operational transient, the more limiting transients have been analyzed to determine which result in the largest reduction in critical power-ratio (CPR).-
The models used in the transient analyses are discussed in Reference 1.
Flow dependent and power dependent MCPR limits (MCPRr and MCRPr) are used to define the required Operating Limit MCPR (OLCPR) such that-the above Safety e
= Limit MCPR requirement is met-for all power / flow conditions.
MCPRr provides the thermal margin required to protect the fuel from transients resulting from inadvertent core flow increases.
MCPR, protects the fuel from the other l
limiting abnormal operating transients, including localized events such as a rod withdrawal-error.
Direct scram on Turbine Stop Valve Closure or-Turbine Control Valve fast closure provides the fastest response to an abnormal operating transient such as load rejection, turbine trip, or feedwater controller failure. These direct scrams are bypassed at low power (P33,,, ), to reduce the frequency of scrams (30% of rated power),
during power ascension. For operation at or above P39,,,
the required OLMCPR is the larger of MCPR, or MCPR, at the existing core power / flow state; _where MCPRr and MCPRy are determined in the Core Operating Limits Report by multiplying the scram time dependent MCPR limit for rated power and flow MCPR(100) by the K, factor. Below 30% of rated power, when the direct scrams are bypassed, a slightly more severe transient response results.
To. compensate for the more severe transient response, two power dependent MCPR limits are_ established, one for high flow (>50% of rated) conditions and one for -low flow (550% of rated) conditions. -These limits are specified in the Core Operating Limits Report. Further information on the MCPR operating limits for off-rated conditions is presented in Reference 11.
References for Bases 3.11 1.
" General Electric Standard Application for Reactor Fuel," NEDE 24011 P-A.
(The approved revision at the time the reload analyses are performed.)
The approved revision number shall be identified in the Core Operating Limits Report.
2.
" Supplemental Reload Licensing-Submittal for Cooper Nuclear Station," (applicable reload document).
3 8.
Deleted 9.
Letter (with attachment), R. H. Buckhol: (CE) to P. S. Check (NRC).
" Response to NRC Request for-Information on ODYN Computer Model." September 5, 1980.
I 10.
" Cooper Nuclear Station Single-Loop Operation," NEDO 24258.
11.
" Extended Lo'ad Line Limit and ARTS Improvement Program Analysis for Coopet Nuclear Station Cycle 14," NEDC 31892P, Revision 1. May 1991.
214a-
_.__....,_,_-.__.-_,,.___._m.
i 0 11 Aaa#A:
A&B. Averare and Local LHGR The IllGR shall be checked daily to deteritilne if fuel burnup, or control rod movement I
has caused changes in power distribution.
Since changes due to burnup are slow, and only a few control rods are moved daily, a daily check of poser distribution is adequate.
(Surveillance Recuirement)
C.
Minimum Critical Power Ratio fliCPR)
At core thermal power. Icvels less than or equal to 25%. the reactor will be operating at less than or equal to minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excens of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR.
During initial start up testing of the plant, a MCPR evaluation was made at 25% thermal power level with minimum recirculation pump - speed.
The MCPR margin was thus demonstrated such that subsequent MCPR evaluation below this power level was shown to be unnecessary. The daily requirement for calculating MCPR above 25% ratH thermal power is-sufficient since power distribution shifts are very slow when t..e re have not been significant power or control rod changes.
The requirement for calculating MCPR when an operating limit MCPR is approached ensures that MCPR will be known following a
= change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.
214b-
1 to NLS0100187 l
General Electric Report NEDE 31892P, Revision 1 EXTENDED IAAD LINE LIMIT AND ARTS IMPROVEMENT l'ROGRAM ANALYSES FOR COOPER NUCLEAR STATION CYCLE 14 The following document has been classified as Proprietary Information and should be handled accordingly, In accordance with the provisions of 10 CFR 2,790, this document should be withheld form public disclosure.
i
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