ML20077D760
| ML20077D760 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 05/22/1991 |
| From: | Quay T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20077D765 | List: |
| References | |
| NUDOCS 9106050066 | |
| Download: ML20077D760 (20) | |
Text
. _ -.
p ete
_)
n\\
g l ' \\* '
i UNITED STATES
+*
- q\\d%l89e D.
i NUCLEAR REGULATORY COMMISSION
'c4 ff wAsmaton. o c. mss
' <,.....f NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendnent No.142 License No. DPR-46 1.
The h. cler Regulatory Comission (the Commission) has found that:
A.
The applicstion for amendment by Nebraska Public Power District (the licensee) dated July, 2,1990, as supplemented by letters dated March 8, and April 19, 1991, complies with the standards and require-j nents of the Atumic Energy Act of 1954, as amended (the Act), and the Cummission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Cummission; C.
There is reasonable assurance:
(1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations;
-D.
The issuance of this license anendnent will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is-in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9 e
P
l
, 2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in iha attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. DPR-46 is hereby amended to read as follows:
2.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.142, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY C0KMISSION
^
D M< cebw 2 77 Theodore R. Quay, Director Project Directorate IV-1 Division of Reactor Projects III, IV, and V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 22,1991
u l
ATTACH'iENT TO LICENSE AMENDMENT NO. 142 FACILITY OPERATING LICENSE NO. DPR-46 DOCKET NO. 50-298 Feplace the followirig pages of the Appendix A Technical Specifications with the enclosed pages.
The revised pages ure identified by Amendnent number and contain vertical lines indicating the area of change.
REMOVE PAGES INSERT PAGES iv iv 1
1 Sc 7
7 8
8 18 18 62 62 102 102 210 210 211 211 211a 211b 212 212 212b 212c 212d 213 213 214 214 214a 214a 214b 214b 214c 232 232 234 234
TABLE CT CONTENTS (Cont'd.)
Pare No
- c. ep ". r.. t_'-
.-r_
s".
s
-r-" tru.e.
.e.
e.r. e. r_ e. 3... r..
fag
..u......,,.. w...
...,.e, 20 6.2 Review and Audi:
6.2.1,A Statien Operations Review Ccami::te (SCRC) 220 20 A.; Membership 2:0 A.: Mee:ing 7:equency 220 A.3 Quorum 20 A.a Responsibili:ies 221 A.5 Authori:y
- 1 A.6 Records 221 A.7
?:ocedures 6.0.1.3 S??D Safety Review and Audi: Board (SRA3) 022 222
3.1 Memoe
ship 22 3.2 Mee:ing Frequency 22 3.3 Quorum 20 3.-
Review 0:3 3.5 Au:nority 23
!.6 Records 22 3.0 Auci:s 25 6.3 ?:ocecures and ?:ograms 225 6.3.1 Int:ccu:: ion 205 6.2.
?:ocecures 6.3.2 Maintenance and Tes: ?:ocedures 225 205 6.2.a Racistien Control ?:ocedures 2:6
.A Mi;h Radia:icn Areas 227 6,3.5 Temporary Changee 227 6.2.6 I:<er:ise of ?:ocedures 207 6.3.7
?:ograms 007
.A Sys: ems Inte5:i:7 Meni:::ing ?:og:am 2:'
.3
- odine Monitoring ?:cgran
.: ?os:-Acciden: Samp'.;,5 System (? ASS) 227 223 6.a Rec::d Re:ention
- S 6.4.1 5 year re:en: ion e.
6...-
,,,e
.e.e.....
2:0 6.
.3 yea: retention 230 6.5 S:a::en ?.epor-ing Requiremen:s 220 6.5.;
Routine Reper:s 20
.A n roduction 20
.5 5:ar:up Repor:
- 20
.0 Annual Repor:s 231
.3 Men:hly Operating Repor:
.: Annual Radiologics! Environmen:41 Repor 231
? Semiannual Radioactive 'da:erisl Release Report 22 22 l
.0 Core Cpera:tng Lim.:s Report Amendment No. 105,142 iv-
, e 1.0 DEFINITIONS The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved, A.
Thermal Parameters 1,
Critical Power Ratio (CPR)
The critical power ratio is the ratio of that assembly power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition of interest as calculated by application of an NRC approved critical power correlation.
2.
Maximum Fraction of Limitine Power Density The Maximum Traction of Limiting Power Density (MFLPD) is the highest value existing in the core of the Fraction of Limiting Power Density (FLPD).
3, Minimum Critien1 Power Ratio (MCPR) - The minimum critical power ratio corresponding to the most limiting fuel assembly in the core.
4.
Fraction of Limitine Power Density The ratio of the linear heat generation rate (LHCR) existing at a given location-to the design IllGR for that bundle type, Design LHGR's for each type of fuel are specified in the Core Operating Limits Report.
5.
Transition Boiline - Transition boiling means the boiling regime between nucleate and film boiling.
Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.
B.
Alteration of the Peactor Core The act of moving any component in the region above the core support plate, below the upper grid and within the shroud.
Normal control rod movement with the control rod drive hydraulic system is not defined as a core alteration.
Normal movement of in c re instrumentation is r.ot defined as a core alteration.
C.
Cold condition - Reactor coolant temperature equal to or less than 212*F.
D.
Desien Power-- Design power means a steady state power level of 2486 thermal megawatts, This is 104.4% of Rated Power (105% of rated steam flow).
E.
Enrineered Saferuard - An engineered safeguard is a safety system the actions of which are essential to a safety action required to maintain the consequences of postulated accidents within acceptable limits, i
E.A Dose Ecuf valent 1-131 - The DOSE EQUIVALENT I-131-shall be that concentration l
of I 131 (microcurie / gram) which alone would. produce the same thyroid dose if-L inhaled by an adult as the quantity and isotopic mixture of I 131, 1 132, 1-133, 1-134, and 1-135 actually present.
The dose equivalent I 131 concentration is calculated by:
equiv. I 131 - (I-131) + 0.0096 (I-132) +
l 0.18 (I-133) + 0.0025 (1 134) + 0.037 (I-135).
l E.B Exhaust Ventilation Treatment System - An EXHAUST VENTILATION TREATMENT SYSTEM (EVTS) is a system intended to remove radioiodine or radioactive material in particulate form from gaseous effluent by passing exhaust ventilation air l
through charcoal absorbers and/or HEPA filters before exhausting the air to the environment.. An EVTS is not intended to affect noble Eas-in gaseous effluent.
Engineered Safety Feature (ESF) gaseous treatment systems are not considered I
to be EVTS. The Standby Cas Treatment System is an ESF and not an EVTS. EVTS l
-are specifically identified in ODAM Figure 3 1.
Amendment No, 46,80,89,123,142
_____-_..._.___.m.._...
.__-.._____m..-~
u
'i.
.AA.
Core Operatine Limits Reeort The' Core Operating' Limits Report.is the unit specific document that provides core operating limits for.the current' reload cycle. These cycle specific core operating -
limits-shall b e' ~ determined for each reload cycle ' in accordance with
- Specification 6.5.,1,G.
Plant operation within these core operating limits is addressed in individual specifications, i
t.
4 L
l 9
I I.
l' i
Amendment No.142
-5c-
-.. -. ~ -
'g.
SAFI'TY 1.1MITS Llfilllfic SAFETY SYSTEM SFTTit;GS 1.1 (Cont'd) 2.1.A.1 (Cont'd)
D.
Cold Shutdown tN - O for two recirculation loop Whenever the reactor is in the cold operation, shutdown condition with irradiated fuel in the reactor vessel, the a.
In the event of operation with a water level shall not-be less chan maximum fraction of limiting power 18 in, above the top of the normal density (MFLPD) greater than the-active fuel zone (top of active fuel fraction of rated power (FRP), the is defined in rigure 2.1.1),
setting shall be modified as follows:
S s (0.66 V 4 54.%
- where, FRP - fraction of rated thermal power (2381 MWt)
MFLPD - maximum fraction of limiting power density where the limiting power density for each type of fuel bundle is specified in the Core Operatin6 Limits Report.
The ratio of FRP to MFLPD shall be l
set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
[
For no combination of loop l;
recirculation flow rate and core I
thermal power shall the APRM flux scram trip setting be allowed to L
exceed 120% of rated thermal power.
[
b.
APRM Flux Scram Trin Set ting (Refuel pf' Start and Hot Standby Mode) l When the reactor mode switch is in the REFUEL or STARTUP pot.1i f on, the APRM scram shall be set at less than or equal to 15% of rated power.
c.
Ifd The IRM flux scram settlug shall be
$120/125 of scale.
. Amendment:No. 32,i;,46,80,88,94, 7
133, 142
.. hF Y L1MITS' L1HlTINC SAFETY SYSTEM SETTINGS 2.1.A.1 (Cont'd) d.
AEBM Pod P, lock Trin Setting The APRM rod block trip setting shall be:
S s 0.66 W 4 421 -. H AU RB where:
S
- Rod block setting in g3 percent of rated thermal power (2381 MWt)
W and AW are de fitie d in Specification 2.1.A.1.a.
In the event of operation with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
RB (0.66 V + 42%
0.66 AW)
FPP S
s MFLPD
- where, FRP - fraction of rated therinal power (2381 MWt)
MFLPD - maximum fraction of 4
limiting power density where the limiting power density for each type of fuel bundle is specified in the Core Operating Limits j
Report.
t The ratio of FRP to MFLPD shall be-set equal to 1.0 unless the actual operating value is Icss than the design value of 1.0, in which case the actual operating value will.be used.
I l
2.
Reactor Water Low level Scram and Jsolation Trio Settinr (except MSIV) l 2 +12.5 in, on vessel level inst ruments.
I f
Amendment No. 16.22,39,41,46,80,7#,
.g.
1 Ill,142
'4 2.1 hw (Cont'd)
An. increase in the APRM scram trip setting would decrease the margin present before the f uel cladding integrity Safety Limit.'s reached, The APPM scram trip setting was determined by an analysis of margins required to provice a reasonable range for maneuvering during operation.
Reducing this operating margin would increase the f requency cf spurious scrams which have an adverse ef fect on reactor saf ety because of the resulting thermal stresses. Thus, the APPA scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Liut yet allows operating margin that reduces the~ possibility of unnecessary scrams.
The scram trip setting must be adjusted to ensure that the LHCR transient peak is not increased for any combination of maximum f raction of limiting power density (MFi.PD) and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1. A.1,a, when the MFLPD is greater than the fraction of rated power (TRP). This adjustment may be accomplished by increasing the APPd gain and thus reducing the slope and intercept point of the flow referenced APRM High Flux Scram Curve by the reciprocal of the APP 4 gain change.
Analyses of the limiting transients show that no scram adjustment is required to assure MCPR remains above the safety limit when the transient is initiated from the operating MCPR limit specified in the Core Operating Limits Report, b.
APPM Flux Scram Trio Settine (Refuel or Start 6 Hot Standby Mode)
For operation in the startup mode while the reactor is at 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 accomodate anticipated maneuvers associated with power plant startup.
Effects of increasing pressure at zero or low void content 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 by the rod worth minimizer, and the rod sequences control system, Worth of individual rods is very low in a uniform rod pattern.
Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise.
Because the flux distribution associated with uniform rod withdrawals 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 power 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 APPM 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.
l r
Amen tent f.'o. N,142 l l
NOTES FOR TABLE 3.2.C 1.
For the.startup and run positions of the Reactor Mode Selector Switch, the Control Red Withdrawal Block Instrumentation trip system shall be operable f or each lunction.
The SRM and IRM blocks need not be operable in "Run* mode, and the APRM (flow biased) and RBM rod blocks need not be operable in *Startup" mode.
The Control kod Vithdrawal Block Instrumentation trip system is a one out of an' 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 5 below and note 1 of Table 3.1.1) for the Control Rod Vithdrawal Block Instrument tion, a sufficient number of instrument channels will always be operable to provide redundant rod withdrawal block protterion.
'2. V is the two loop recirculation flow rate in percent of rated. Trip level setting is in percent of rated power (2381 MWt). N is the RBM setpoint selected (in percont) and is calculated in accordance with the methodology of the latest NRC approved version of NEDE-24011 P A.
The Core Operating Limits Report specifies the applicable value for N.
3.
IRM downscale is bypassed when it is on its lowest range.
4.
This function is bypassed when the count is 2 100 cps and IRM above range 2.
5.
By design one instrument channel; i.e., one APRM or IRM per RPS trip system may be bypassed. For the APRM's and IRM's, the minimum number of channels specified is that minimum number required in each RPS channel and does not refer to a minimum number required by the control rod block instrumentation trip function. By design only one of two RBM's or one of four SRM's may be bypassed. For the SRM's, 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.
For the RBM's, the minimum l
number of channels specified is the minimum number required by the Control Rod Block Instrumentation Trip System as a whole (except when a limiting control rod pattern j
exists and the requirements of Specification 3.3.B.5 apply).
f 6.
IRM channels A.E.C.C all in range 8 or higher bypasses SRM channels A&C functions.
l' IRM channels B F.D.H all in range 8 or higher bypasses SRM channels B6D functions.
I 7.
This function is bypassed when IRM is above range 2.
B.
This function is bypassed when the mode switch is placed in Run.
9.
This function is only active when the mode switch is in Run,
- 10. The inoperative trips are produced by the followi'ng functions:
a.
Mode switch not in operate (2)
Power supply voltage low (3)-
Circuit boards not in circuit (4)
Loss of negative supply volttge Anendment No. 77,92,94,10?,113.142 62-
s t
t 9
3.3 and 4.3 bkSES: (Cont'd) 5, The Fod Block Monitot (RBM) is designed to automatically prevent fuel damage in tt.e event of erroneous rod withdrawal from locations of high power density during high power level operation. Two channels are provided, and one of these may be bypassed from the console for maintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.
This system backs up the operator who withdraws control rods according to written sequences.
The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod althdrawal errots when this conditions exists.
A limiting control rod pattern is a pattern which results in the core being on a thermal hydraulic limit (i.e., MCPR equals the operating limit as specified in the Core Operating Limits Report, and LHCR - as defined in 1.0. A.4). During use of-such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods to assure its operability will assure that improper l
withdrawal does not occur.
It is the responsibility of the Reactor Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns.
Other personnel qualified to perform this function may be designated by the Division Manager of Nuclear Operations.
C.
Scram insertion Times The control rod system is designed to bring the reactor suberitical 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 l
this transient shows that the negative reactivity rates resulting f rom the scram j
provide the reqaired protection, and MCPR remains greater than the safety limit.
1 The surveillance requirement for scram testing of all the control rods af ter each j
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 scram times within the limits, but significantly longer than the average, should be viewed as an indication of a s stematic 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 co the ' typical time delay of about 210 milliseconds estimated from scram test results. Approximately the_first 90 milliseconds of each j
.of these time intervals result from the sensor and circuit delays; at this point. the pilot scram solenoid deenergized. Approximately 120 milliseconds later, l
- Amendment No. #4 s7/7,80,100.133,142 102
?.,'
M MITltiG CONDITIONS fop.QTERATION S!RVFILIMICE Rt 0111PFMDITS 3.11 FUF1 FODS 4.11 H FL PODS bpr.licability The Limiting Conditions for Applicability Operation associated;with the fuel-rods apply to those parameters which The Surveillance Requirements apply monitor the fuel rod operating to the parameters which monitor-the d tions, fuel rod opernin; conditions.
The-- Objective of the Limiting Obiective conditions for Operation is to assure the performance of the fuel The Objeevive of the Surveillance ds Requirements is to specify the type jf S
Y A.
Averare Planar Linear llent
- EE I "'
Generation Rate (APuiGR)
During steady state power operation' Snecifications the APuiGR for each type of fuel as a
function of average planar A.
Averare Planar Linear tient exposure shall not exceed the Generation Rate (APutGR) limiting value specified in the core Operating Limits Report for tw The APulGR for each type-of recirculation loop operation.
For fuel as a function of average single-loop operation these values planar exposure shall be are reduced for each fuel type as determined ' daily duiing specified in the Core Operating reactor operation at 2/5%
Limits Report.
If at any time rated thermal power.
during steady state operation it is determined by normal surveillance that the limiting value for APuiGR is being exceeded action shall be initiated within 15 minutes to
!~
restore operation to within the prescribed limits. -1f the APUlGR is not returned to within the prescribed limits within, two (2) hours, the reactor shall be brought to the Cold Shutdown condition I
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 L
until the prescribed limits are again being met.
B.
Linear Heat Generation Rate (DIGR)
During steady state power operation' B.
Linear llear Generation Rate the linear-heat generation rate
-(DiCR)
(UiGR) of any rod. in any fuel assembly at any axial location shall The UlGR as. a function of core l
l-not exceed the maximum allowable height shall be checked daily during uiGR as specified in the Core reactor operation at 2 25t rated Operating Limits Report
- thermal power.
i Amensnent No. 55,H,106,133.142
-210-
I
..,..,_.-.-.1,..,. _ :- 2..._
1 Amendment No. /7#,733,142
4 u
MMITINGCOMb1T10NSFOROPERATION St1RVEII.1MCE REQUIPEMENTS 3f at any time during steady state operation it is determined by normal surveillance that the limiting value for IJICR is being exceeded action shall then-be initiated to restore operation _to within the prescribed
- limits, Surveillance-and corresponding action shall continue until the -prescribed limits are again being met.
C.
Minimum Critieni Power Ratio (MCPR)
C.
Minimum Critical Pover Ratio (MCPR)
During steady state power operation MCPR shall be determined daily the MCPR for each type of fuel at during reactor power. operation at rated power and flow shall not be
) 25%
rated _thermel power and lower than the limiting value following any change in power level specified in the Core Operating or distribution that would cause 1.imits Reports for two recirculation operation with a limiting control loop operation.
If, at-any time rod pattern as described in the during steady state operation it is bases for Specification 3.3.B.S.
determined by normal surveillance that the limiting value for MCPR is r
being exceeded, action anall then be initiated within 15 minutes to restore operation to within the prescribed limits.
If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be
-brought to the Cold Shutdown condition within 36 hourn.
Surveillance and corresponding action shall continue until the prescribed limits are aEain being met.
For core flows other than rated the MCPR shall be the operating limit at rated flow times K, where K is specified in the bore Opera' ting Limits Report.
For one recirculation loop operation the MCPR limits at rated flow are O.01 higher than the comparable two loop values.
l
' Amendment No. /74,94,133,142 212-l L
-. -, - ~.-
{ q l'
\\;
- g.
l p,
4 l -
t 1
i I
t l
l t '.
I I
l'
"":TE!: !c': Atty it su :e l
('
i.
r
)
+
t-I' l
L l
l t
i I -
l 1
I,
l' A<hendment No. /pJ,fy3,142
,n),
_m._
,'o 1
-4 3.11 1% S f'S A.
Averare Planar Linear Heat Generation Pete ( API)lCR )
This specification assures that the peak cladding temperature following the postulated design basis loss-of coolant accident will not exceed the limit specified in 10CFR50.46.
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 usembly 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 ussembly affect the calculated peak clad tempereture by less than i 20*T 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 temperatures are within the 10CFR$0.46 limit. The limiting value for APulGR for each fuel type is specified in the Core Operating Limits Report.
The APUICR values are reduced for single loop operation per Reference 10.
B.
Linear llent Generation Rate (UIGR)
This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densificacion is postulated.
The UiGR as a function of core height shall be checked daily during reactor operation at 2 251 power to determine if fuel burnup, or control rod movement has caused changes in power distribution.
For DiCR to be a limiting value below 251 rated thermal power, the MTFF would have to be greater a considerable margin when employing any then - 10 which is precluded by permissible control rod pattern. Pellet densification power spiking in CE fuel has been accounted for it, the safety analysis presented in References 1 and 2; thus no adjustment to the UICR limit for densification effects is required.
1 i
l knendment No. 62,80,94,133,142 214-l
3.11 Engn:
(Cont'd)
C.
til.ginum Critical Ppwer Ratio MCPR1 1
l f
The required cperating limit MCPR's at steady state operating conditions as specified in Specification 3.11C are derived f rom the established fuel cladding integrity Safety Limit and an analysis of abnormal operational transients (Reference 2).
For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any tin.e during the transient assuming instrument trip setting given in Specification 2.1.
To assure that the fuel cladding integrity Safety Limit is not exceeded during 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.
f actor is to define operating limits at other than rated The purpose of the K g flow conditions. At less than 100% flow, the required MCPR is the product of the operating limit MCPR and the K factor.
Specifically, the K factor flow kncrease provides the required thermal marg [n to protect against a transient.
The most limiting transient initiated from less than rated flow conditions is the recirculation pump speed up caused by a motor-generator speed control failure.
factors assure that For operation in the automatic flow control mode, the Kg the operating limit MCPR will not be violated should the most limiting transient occur at less than rated flow. In the manual flow control mode, the K factors assure that the Safety Limit MCPR will not be violated for the same g
postulated transient event.
The K factors as provided in the Core Operating Limits Report were developed g
The generically which are. applicable to all BWR/2, BVR/3, and BWR/4 reactott.
K factors were derived using the flow control line corresponding ti rated thermalpoweratratedcoreflowasdescribedinReference1.
The K factors are conservative for Cooper operation because the operating 11mithCPR's are greater than the original 1.20 operating limit MCPR used for the generic derivation of K.
g Egferences 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).
3-0.
Deleted 9.
Letter (with attachment),
R. H. Buckholz (CE) to
- p. S, Check (NRC),
" Response to NDC Request for Information on ODYN Computer Model,"
September 5, 1980.
10.
" Cooper Nuclear Station Single Loop Operation," NEDO 24258.
Amendment No. 62,80,94,142
-214a-
4
,r e.'11! ares:
A6.3, Ave r s e. and tocal LuCR The 1.HCR shall be checked daily to determine if fuel burnup, or control red movement has caused changes in power distribution, Since changes due to burnup are slov, and only a few control rods are moved daily, a daily check of power distribution is
- adequate, C.
Enfmum Critical Power parte IMcPRT. / Surveillance Reeutrementi At core thermal power levels less than or equal to 25% the reactor vill be operating at less than or equal to minieurs recirculation puttp speed and the moderator void content vill be very small.
For all desigaated control rod patterns which may be employed at this point, operating planc experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin, With this low void content, any inadvertent core flow increase vould only place operation in a more censervative mode relative to MCPR.
During inicial 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 unnecessarv, The daily requirement for calculating MCPR above 25% rated thermal power is sufficient since power distribution been si nificant power or control rod shifts are very slow when there have not 5
chan5es.
The requirement for calculating MCPR vhen a limiting control rod pattern 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.
Amendment No. 62,80,9#,142
. mb.
l
l l
'4
- . f.l.E (Cen:'d)
Sumarized and :abulated results in the forma: of Table 6.51 c.
of analyses of sanples required by the radiological environmental monitoring program, and taken during the report period. In the event tha: seme results are not available for inclusion with the report, the. report shall be submitted notin5 and explaining the reasons for the missing results.
The missing data shall be submit:ed as soon as possible in a supplementary report, d.
A summary description of the radiological environmen:al monitoring program including any changes: a map of all sampling locations keyed to a table giving dis:ances and directions from the reactor: and :he results of par:icipation in the Inter laboratory Comparison Program, required by Specification 3.21.C.
F Semiannual Radioac:ive Material Release Report 1.
A repor: of radioactive materials released from the Station du:.nS the preceding six months shall be submitted to the 2:RC wi:nin 60 days af:er January 1 and July 1 of each yearw.
2.
A Semiannual Radioactive Ma:erial Release Repor: shall include the following:
a.
A summarv by calendar quarter of the quantities of radioactive liquid and gaseous effluents released from :he S:ation.
The data should be reported in the format reco=. ended in Regulatory Guide 1.21, Appendix B. Tables 1 and 2 b.
A sum. ary of radioactive solid waste shipped from the S:stion, including informa:Lon named in Specification 4.21.I..
A sumary of meteorolo5 cal data collected during the year i
c.
shall be included in the Semiannual Report submitted wi:hin 60 days af:er January 1 of each year.
d.
A lis: and brief description of each unplanned release of gaseous or liquid radioac:ive effluent that causes a limi:
in Specification 3.21.B.l.a.
3.21.B.2.a.
3.21.C.l.a.
3.21.C.2.a. or 3.21.c.2,a :o be exceeded.
Calcula:ed offsi:e dose :o humans resui:in5 from the release e.
of affluents and : heir subsequent disre:sion in :ne a:mosphere reported in accordance wi:h Regula:ory Guide 1.21.
6.5.1.C
.Q.e r. 5 t a - i r ti-i s S m r -
Core opera:in5 limi:s shall be es::blisned nd documented in the Core Coerating
'imi:s Reper: prior to each reload cycle, or prio: :o any remaining pornen of a reload cycle, for the following; a.
The Average Plana:
Linear Hea Generation Rates
( API.HCR) for Spec:.fication 3.ll.A.
- !: should be no:ed tha: this da:a has not normally been available to the Discrie: within 60 days anc a ve:bal ex:ension has :ypically been required frem :he NRC CSS Projec:
Manager.
Amenoment No. 89,142 232-
~
Core Operatinr_ Limitt. Fenort (Continued)
The Linear Heat Ceneration Rate for Specification 3.11.E.
b.
c.
The K core flow MCPR adjustment factor for Specification 3.11.C.
g The minimum critical power ratio (MCPR) for Specification 3.11.C.
d.
e.
The rod block monitor upscale setpoint for Table 3.2.C of Specification 3.2.C.
The analytical methods used to determine the core operating limits shall be
- Ceneral those previously reviewed and approved by the NRC in NEDE 24011 P A, Electric Standard Application for Reactor Fuel." (The approved revision at the time the reload analyses are performed.) The approved revision number shall be identified in the Core Operating Limits Report.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits, and accicent analysis limits) of the safety analysis are met.
The Core Operating Limits Report, including uny mid cycle revisions or supplements thereto, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
6.5.2 Renortable Events A Reportable Event shall be any of those conditions specified in Section 50.73 to 10CFR Part 50.
The NRC shall be notified and a report submitted pursuant to the requirements of Sect ion 50.73. Each Reportable Event shall he reviewed by SORC and the results of this review shall be submitted to SRAB and the Nuclear Power Group Manager.
l Amendment No. 86,89,102,122,142 234-
.