ML20034A691
| ML20034A691 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 04/12/1990 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20034A692 | List: |
| References | |
| NUDOCS 9004240133 | |
| Download: ML20034A691 (25) | |
Text
,
- f' N 4'
UNITED STATES 8
NUCLEAR REGULATORY COMMISSION S
WASHINGTON, D C. 20666
^
r NEERASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO FACILITY OPERATING LICENSF f
- Amendment No.133 License No. DPR-46 t
1.
The Nuclear Replatory Comission (the Commission) has'found that:
A.
The application for. amendment by Nebraska Public Fcher District.
(the licensee) dated February 12, 1990, complies with the standards and requirements of the Atoraic Energy Act of 1954, as amended (the 1
Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as i
amended, the provisions cf the Act, and the rules and regulations.
of the Comission; t
C.
There is reasonable assurarice:
(i) that the activities authorized by this aniendment 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 amendment will not be inimical to the comn;cn defense and security or to the health and safety of the public; and E.
The issuance of this amer.dn.ent is in accordance with 10 CFR Part 51 of the C w.ission's regulations and all applicable requirements havt l
been satisfied.
l k
e I
e
t, 2-2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attechment to this liter.se arrendment and Paragraph 2.C.(2) of Facility Operating License No. DPR-46 is hereby amended to rt.ad us folicws:
2.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.133, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specificaticr.s.
3.
The license amendment is effective as of its date of issuarce.
FOR THE NUCLEI.R REGULATORY C0tEISSION
?
J U
.l h -
Frederick J. Hebdon, Director Project Directorate IV Division of Reactor Projects - !!!,
IV, V and Special Projects Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 12, 1990 r
s
ATTACHMENT TO LICENSE AMENDF.ENT NO.133 4
FACILITY OPERATING LICENSE NO. DPR-46 DOCKET NO. 50-298 Replace the following anges of the Appendix A Technical Specifications with the enclosed pages. Tae revised pages are identified by Arendrnent number and centain vertical lines indicating the area of change.
Delete Pages Insert Pages 1
1 5
5 6
6 7
7 8
8 12 12 17 17 99 99 102 102 104 104 210 210 211 211 211a Ella 211b 211b 212 212 212b 212b 212c 212c 212d 212d 212e 212f 2120 213 213 214 214 214b 214b 217 217
1.0 DETINITIONS i
The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved.
A.
IbArmal Parameters The critical power ratio is tha ratio of 1.
Critical Power Ratio (CPP3 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 NRG approved critical power correlation.
2.
Ma~imum Traction of Limitine Power Densitv The Maximum Traction of Limiting Power Density (MFLPD) is the highest value existing in the core of the Traction of Limiting Power Density (TLPD).
3.
Minimum Critieni Power Ratio IMCPRi - The minimum critical power ratio corresponding to the most limiting fuel assembly in the core.
4 Traction of limitine Power Densitv - The ratio of the linear heat generation rate (LhGR) existing at a given location to the design LHCR for that bundle type.
Design LHGR's are 13.4 K'J/f t for BPSXSR and 1988 LTA bundles and 14.4 K'J/ft for GESXBNB and Gell LTA bundles.
5.
Transition Beiline Transi: ion boilin6 means the boiling regime between nucleate and film boiling.
Transition boiling is the regime.in which i
both nucleate and film boiling occur intermittently with neither type being completely stable.
B.
Alteration of the Reseter Cere The ac: of moving any component in the region above the core support plate, below the upper grid and within the shroud.
Normal control rod movement wi:h the control rod drive hydraulic system is not defined as a core alteration.
Normal movement of in core instrumentation is not defined as a core alteration.
C.
Celd Conditten Reactor coolant temperature equal to or less than 212*F.
D.
Desf rn Pewer Design power means a steady state power level of 2486 thermal j
megcwatts. This is 104.4% of Rated Power (105% of rated reeam flow).
l l
E.
Inrineered Faferuard An engineered safeguard is a safety system the actiens of which are essential to a safe:y action required to maintain the consequences of postulated accidents within accep:able limits.
E.A Dere Eeuivalent 1 131 The DOSE EQUIVALENT I 131 shall be that concentration of I 131 (microcurie / gram) which alone would produce the same thyroid dose if inhaled by an adult as the quantity and isotopic mixture of I 131. 1 122.
I 133, I 134 and I.135 ac:ually present.
The dose equivalent I *11 concentration is calcula:ed by: equiv.1 131 - (I 131) + 0.0096 (I 132) + 0. *.S (I 133) + 0.0025 (I 134) + 0.037 (I 135).
E.3 Exhaust Ver-flatien Treat rent Svsten An Elu.AUST VENTILATION TREATMENT SYSTEM (EVTS) is a system intended to remove radioiodine_ or radioactive material in par:iculate form from gaseous effluen: by passing exhaus: ventilation air
- hrough charcoal absorbers and/or HE?A filters before exhausting the air to the environment. An E*1TS is no: in: ended :o affec: noble gas in gaseous ef f*.uen:.
Engineered Safety Feature (EST) ;aseous treatment systems are no: conside re d to be I'!TS, The 5:andby Gas Trea: men: System is an ESF and not an E*lTS. I'l!S are specifically identified in OEA*1 Tigure 31.
Amendment NOS. H,80,87,133 1
+
3 '.
All automatic containment isolation valves are operable or de activated in the isolated position.
4 All blind flanges and aanways are closed.
P.A Purre Purrine Purge or Purging is the controlled process of discharging air or gas from a confinement to establish temperature, pressure, humidity, concentration cr other operating condition. in such a manner that replacement air or gas is required to purify the confinenent.
P.B Process control Prerram The Procers Control Program outline e the solidification of radioactive vaste from liquid systems. It does nct substitute for station operating procedures, but provides a general description of equipment, controls, and practices to be considered during waste solidification to assure solid wastes.
Q.
Rated Power Rated power refers to operation at a reactor power of 2381 megawatts thermal.
This is also termed 100% power and is the maximum power level authorized by the operating license. Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these par. neters when the reactor is at rated power.
R.
Reacter Pever overatien Reactor power operation is any operation with the mode switch in the "Startup/ Hot Standby" or "Run" position with the reactor critical and cbove 1% rated power.
S.
Etteter Verrel Prerrure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reae:cr vessel steam space detectors.
T.
Refueline Outare Refueling outage is the period of time be:veen the shutdown of the uni: prior to a refueling and the s:ar:up of the plant af:er that refueling.
U.
Safer" Limits The safe:y limits are limits within which the reasonable maintenance of the fuel cladding integrity and the reactor coolant system integrity are assured.
Violation of such a limit is cause for uni shu:down and review by the Nuclear Re5ulatory Commission before resumption of unit operation.
Opera: ion beyond such a limi: may not in itself result in serious consequences but it indicates an opera:iona'. deficiency subject to regslatory review.
Secondary con:sinment integrity means that the V.
Secordarv Contain ent Interrit" reac or building is intae: and the following condi: ions are met:
1 A: least one door in each access opening is closed.
2.
The standby gas' eatment sys:em is operable.
3, All automa:ic ventilation system isola: ion valves are operable or secured in the isclated position.
- ?,
She-fev-The reactor is in a shutdoun condition when :he mode suitch is in the
" Shutdown" or " Refuel" position, 1.
Ho: Shutdoun means condi: ions as above vi:h reac:or coolan: :empera:ure greater
- han 210'F.
2.
Cold Shu:do.:n means c:nditions as above 11:h rese:or coelan: :empe rature e qua'.
- o or less than 2'.2*T and :he reac:or vessel vented.
Amendment No ff 80.89,133 3
< SAFETY Lf MITS LIMITINC SATETY SYSTEM SETT7NCS 1.1 TUEL CIADDINC INTECRITY 2.1 TUEL CLADDINC INTECRITY Aeolicability Aenlicability The Safety Limits established to The Limiting Safety System Settings preserve the fuel cladding integrity apply to trip settings of the apply to those variables which instruments and devices which are acnitor the fuel thermal behavior, provided to prevent the fuel cladding integrity Safety Limits from being exceeded.
Obiective Obiective The objective of the Safety Limits is to establish limits belov vhich The objective of the Limiting Safe.
the integrity of the fuel cladding ty 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, reac:or 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 />.
Steeifications Seceificatiens
^*
A.
Reneter Prersure >POO esis and Core Tinw > nt of Fated The limiting safety sys:em : rip The existence of a mini:r.um settings shall be 'as'specified critical power ratio (MCPR) below:
l 1ess than 1.06 for two recirculation loop opera: ion 1.
hurren Fluv Tri, c --i n e
j (1.07 for single loop operation) shall constitute a.
APP 3! FluN Scra'- Tria' Se tti-e violation of the fuel cladding (Run Medes integrity safe:y.
k* hen the Mode Switch is in the 3.
Core Ther-a' Pe"er ti., i t RUN position. the A?P.4 flu::
Wnter Pressure eson esta scram trip se::ing shall be:
arm 'er cere Flou <'o o
- 'nen :ne reac:or pressure is 550.66 W + 3a; -. 6 6 ?.J.?
.s
<S00 psia or core f'.ow is less where:
than 10% of rated. the core thermal power shall no: exceed S - Se::ing in percent of 25'; of ra:ed thermal power.
rated thermal power (2381 'r.?:s C.
Pever Trnes'en-
'J
- Two loop recirculation o ensure tha: the S a:.e:v
..ow ra:e in percent of
,, m., :
e s t a b,. shed in ratec, (rated loop Spec,.,ication,. 1.A and 1.,.B recirculation flow ra:e
.s not exceeded. eac,n required
.s tu..a: recirculation.,,.cw scram shal,. be ini:ia:ed by ra:e which p rov,.de s
,w,.C.,,.
,:s expec:ed scram signal.
""******...,u.' ' " '* I '# '.
The Safe:y Limi: shall be assurec to be e::ceeded when 1.'.? - M f e r e nc e b e :ve e r l
s: ram is ac c omp '. is he d b,"
a
- wo '. cop att sin;1e '.r:;
scram signal.
effec:i"e dri.e f'e. a:
- .e me :: * :. m.
Amendment No. 48,82,88,9#,133 6-3
'.SATEtY LIMITS LIM 7TINC SAFETY SYSTEM SETT?NCS 1.1- (Cont'd) 2.1.A.1 (Cont'd)
D.
Cold Shutdevn AW - 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 than maximum fraction of limiting power 18 in, above the top of the normal density (KrLPD) greater than the active fuel tone (top of active fuel fraction of rated power (TRP), t.he is defined in Figure 2.1.1).
setting shall be modified as follows:
S $ (0.66 W + $4%
0.66 AW) A MFLPD 4
- where, FRP = fraction of rated thermal power (2381 MWt)
MPLPD - maximum fraction of limiting power density where the limiting power density is 13.4 )G/f t for BPSXSR and 1988 LTA fuel, and 14.4 1N/f: for CEBXSNS and Gell LTA fuel.
The ratio of TRP :o MTLPD shall be se: 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 wil'. be used.
For no combination of leo; recirculation flow ra:e and cere thermal power shall :he APP.'4 i'un scram trip setting be allowed :o
. exceed 100t of rated thermal pover.
b.
AFF_'4 Fluv. Serna Trie S*--inc me f se" er Star and Het S andke N d e '.
b Yhen the reac:or mode sui:ch is in the REFUEL or STARn'P pcsition. :he AFRM scram shall be se: a: less than or equal to 15: of ra:ed power.
c.
E Th. :Pei flun s.: ram se::in; incJ.1 :e 5'.M/125 of sca'.e.
Amendment No. 72,39,4,80,88,9#,
d' 133
1 1
=CAFETI LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.1.A.1 (Cont'd) d.
APRM Rod Block Trin Settine The APRM rod block trip setting.
shall be:
S s 0.66 V + 421 -.66 AV gg where:
SRB - Rod block setting in percent of rated i
thermal power (2381 MWt)
W and AW are defined in Specification 2.1.A.1.a.
In the event of operation with a maximum fraction of limiting power density (MTLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
g3 (0.66 W + 42: - 0.66 cW) _IEE_
S s
MTLPD
- where, TRP - fraction of rated thermal power (2381 MW:)
MTLPD maximum fraction of
' limiting power density where the limiting power density is 13.4 KW/ft for BPSXSR and 1968 LTA fuel, and 14.4 KW/f for CESXSSE and GEli LTA fuel.
l The ratio of TRP to MTLPD shall-be se: equal to 1.0 unless the actual operating value is less than the desi n value of 1.0, in which case 5
the actual opera:ing value will be' used.
2.
Pene-ar '?ater low 'e el Sett-snd Is31a-ter Trie Settinr re"e er - StS *"
t -12.5 in.
en vessel
'. e t *.
ins tru:re nts.
Amendment No. 16,32,37,41,46,80,94, S-133
'e 1.1 NLus:
(Cont'd)
C.
Power Transient Plant safety analyses have sheen that 'the scrats caused by exceeding any safety setting vill assure that the Safety Limit of Specification 1.1A or 1.1B will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate.
The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage.
However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature The concept of not approaching a Safety Limit provided of the plant design.
scram signals are operable is supported by the extensive plant safety analysis.
The computer provided with Cooper has a sequence annunciation program which will indicate the sequence in which events such as scram. AFRM trip initiation, pressure scram initiation, etc. occur.
This program also indicates uhen the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient. Thus, computer information normaly will be available for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specification l'.1.C will be relied on to determine if a Safety Limit has been violated.
D.
Peac or Water le"el /Shutdovn Conditieni During periods when the reactor is shutdown, consideration must also be given to water level requirements due to the effect of decay heat, If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladcing temperatures and clad perforation.
The core can be cooled sufficiently should the water level be reduced to two-thirds the core height. Establishment of He safety limit at 16 inches above the top of the fuel provides adequate margin.
Et erences for 1 1 Bases f
1.
" General Electric Standard Applics' ion for Reactor Tuel."
NEDE 24011 p A-(latest approved revision) 2.
" Cooper Suelear Station Single Loop Operation." 5E00 04:58, May. 1980.
I t
?
Amendment No. 94,133 12-
2.1 1ALti
The abnormal operational transients applicable to operation of the CNS Unit have been analyzed throughout the spectrum of planned operating conditions. The analyses were j
In addition, 2381 Mk'c is based upon plant operation in accordance with Reference 3.
the licensed maximum power level of CNS, and this represents the maximum steady.
state power which shall not knowingly be exceeded.
Models and The transient analyses performed each reload are given in Reference 1.
codel conservatisas are also described in this re ference.
As discussed in Reference 2, the core wide transient analyses for one recirculation pump operation is conservatively bounded by two. loop operation analyses and the flow dependent rod block and scram setpoint equations are adjusted for one. pump operation.
A.
Trir Settiner The bases for individual trip settings are discussed in the following paragraphs.
1.
Neutren riux Trie Settines a.
APRM T1ux Scram Trit. Settine (Run Medei The average power range meni:oring (AFPJ1) system, which is calibra:t:
using heat balance data taken during steady state conditions. reads in percent of rated thermal power (13613 :).
Because fission chambers provide the basic input signals, the AFPJ1 system responds directly to average neutron flux. During transients, the instantaneous rate of hes:
transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel.
Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram set:in5 Analyses demonstrate tha: with a 120% scram trip setting, none of the abnormal operational transients analy:ed violate the fuel Safety Limi: and there is a substantial margin - from fuel damage.
Therefore, the use of flow referenced scram trip provides even addi:iona; margin.
Amendment No. 67,f 0,N,133
3.3 and 4.3 BASES A.
Reactivity limitation 1.
The requirements for the control rod drive system have been identified by evaluating the need for reactivity control via control rod movement over the full spectrum of plant conditions and events. As discussed in subsection III.6 l
of the Updated Safety Analysis Report (USAR) the control rod system design is intended to provide sufficient control of core reactivity that the core could be made suberitical with the strongest rod fully withdrawn. This reactivity characteristic has been a basic assumption in the analysis of plant performance.
Compliance with this requirement can be demonstrated conveniently only at the time of initial fuel loading or refueling.
Therefore, the demonstration mus: be such that it will apply to. the entire subsequent fuel cycle. The demonstration shall be performed with the reactor core ' in the cold, xenon free condition and will show that the reactor is suberitical by at least R+ 0.38% Ak/h with the analytically determined strongest con:rol rod fully withdrawn.
The value of "R",
in units of %Ak/k, is the amount by which the core reactivity, in the most reactive condition at any time in the subsequent operating cycle, is calculated to be greater than at the time of the demonstration.
"R", therefore. is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning of life core reactivity. The value of "R" must be positive or :ero and mus: be determined for each fuel cycle.
The demonstration is performed with a control rod which is calculated to be the strongest rod. In determining this " analytically stronges:" rod, it is assumed that every fuel assembly of the same type has identical material properties.
In the actual core, however, the con:rol cell material proper:ies vary within allowed manufacturing tolerances, and the strongest rod O determined by a combination of the control cell geome:ry and local k=.
Therefore. an additional margin is included in the shutdown margin test to account for the fae: that the rod used for :he demonstration (the "analy:ically stronges:") is no: necessarily the strongest rod in :he core.
Studies have been made which compara experimen:a1 criticals with calculated criticals. These studies have shown tha: ac:ual criticals can be predicted within a given tolerance band.
For gadolinia cores the additional margin required due to con:rol cell :taterial manuf acturing tolerances and calcula:ional uncertainties has experimen: ally been determined to be 0.3S: ak/k.
k* hen this additional margin is demonstrated, it assures tha: the reactivity control requirement is me:.
2.
Reactivity :targin - inoperable con:rol rods.
Specification 3.3.A.: requires tha: a rod be taken ou: of service if 1:
Amendment No. 7F/,133
?9-
l 3.3 and 4.31631Ji:
(cont'd) 5.
The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density durin5 high power level operation. Two channels are provided, and one of these may be bypassed from the console for maintenance and/or testir.g. 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 withdrawal errors when this condition 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 defined on Figure 3.11, and LHGR - as defined in 1.0.A.4).
During use of such pat: erns, i; is judged tha: testing of the RBM system prior to withdrawal.of such rods f
to assure its operability will assure that improper withdrawal does not occur.
t 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 ofinoperable 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 Irrertien Times The control rod system is designed to bring the reac:or suberitical at a ra:e fas enough to prevent fuel damage ; i.e., to prevent the MCPR from becoming less than the safety lici, The limiting power transient is defined in Reference 2.
Analysis of l
this ::ansient shows that the negative reactivity rates resulting from the scram
[
provide the required protection, and MCFR remains greater than the safety limit.
The surveillance requiremen for scram testing of all the control rods af:er each refueling ou: age and 10% of the control rods at 16 week intervals is adequate for determining the operability of the control rod system yet is no: so frequent as to cause encessive wear on the control rod system components.
The nueerical values assi ned to the predic:ed scram performance are based on the 5
analysis of data from c:her BWR's with control rod drives the same as those' on Cooper helear 5:ation.
The occurrence of scram times within the limi:s, but significantly longer than the ave 35e. should be viewed as an indication of a systematic problem with control rod drives.
In the analytical treatment of the ::ansien:s which are assumed to scram on high neutron flux. 290 milliseconds are allowed between a neutron sensor reachin5 the scram point and star: of motion of the control rods.
This is adequate and conservative when compared to the typical time delay of about 210 milliseconds estimated from scram test results. Appronimately the firs: 90 milliseconds of each of these time in:ervals result from the sensor and :ircuit delays: at this poin:.
- he pile: scram solenoid deenergi:es. Approximately 120 milliseconds later, Amendment No. //5 A7,80,100,133 100-
t
^3.3 and 4.3 AA111:
(Cont'd) i G.
Scram Discharte 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 reactor building drain system.
I REFERENCES
-1.
" General Electric Standard Application for Reactor Tuel," NEDE 24011 P.A.(latest approved revision).
2.
" Supplemental Releac Licensing Submittal for Ceoper Nuclear Station," (applicable l
reload document).
l r
3.
General Electric Service Infor.v.ation Letter No. 380, Revision 1, dated February 10 1984 4
General Electric Service Information Letter No. 316. Reduced Netch 's'erth Proc
- dure.
November, 1979.
10--
Amendment No. 117/, 133 i
L' i.1MITN:G DONDITIONS FOR OPERAT70N St'RVEILLANGE RE017TREMENTS 3.11 MtEL RODS 4.11 RTEL RODS Aeolicability Arelicability The Limiting Conditions for Operation associated with the fuel The Surveillance Requirements apply rods apply to those parameters which to the parameters which monitor the monitor the fuel rod perating fuel rod operating conditions.
conditions.
Ob4eetive Objective The Objective 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.
j applied to the fuel rods, i
irecificatiens Srecifications i
A.
Averare Placar Linear Heat A.
Averare Planar linear Heat Generation Rate (APluGR)
Generation Rate (APLHCR)
During s:endy state power operation, the APMGR for each type of fuel as The APMGR for each type of a
function of average planar fuel as a function of average expecure shall not exceed the planar exposure shall be limiting value shown in determined daily during reactor Figure 3.11 1 for two recirculation operation at 225: rated therr.e.1 loop operation.
For single loop power.
operation, the limits are reduced to 0.77 of the curves' value for the BPSXSR and 1986 LTA fuel and to 0.75 l
of the curves' value for the GESXSNB and CE11LTA fuel.
If a: any time during steady state operation it is determined by normal surveillance that the limiting value for APMGR is being exceeded action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APMGR is not returned to vi hin the prescribed limits within two (2) hours, the reactor shall be brough 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 un:11 the prescribed limits are again being met.
S.
Linear Feat Generation Rate (LMGRT 3.
Linear Heat Gereratien Rate During steady state power operation.
(LHGR4 the linear heat generation rate l
The LHGR as a function of c:re (1.HGR) of any red in any fuel assembly at any axial location shall height shall be checked daily I
no: exceed 13.4 K'J/f: for SPSXSR and during reac:or opers: ion at i
1985 LTA fuel; or 14.4 K'J/f: for i
2 25% ra:ed thermal power.
l GEli LTA and GESXSNS fue..
i i
Amendment No. 55,H,106,133 210
e 13 3
- e 12
,2 s
s n-10 g
$4 2d s
9 gg 5E 2=
1 6
e 3
20 d0 50 PLANAR AVERAGE EXPCSURE (GWd/ST)
DATA COORDINATES GWD/ST kW/ft 0.2 11.07 1.0 11.25 2.0 11.49 3.0 11.73 4.0 11.90 6.0 12.23 7.0 12.41 8.0 12.61 9.0 12.80 10.0 12.93 12.5 12.93 15.0 12.59 20.0 12.02 25.0 11.35 45.0 7.12 50.5 5.92 51.3 5.30 Figure 3.11 1.1 Maxi =u= Average Planar !.inear Hea: Generation Rate versus Exposure vi h !.?C Modift:a: ion and Bypass Holes Plugged. GI3X3N3 Tuel Amendment No. //8,133
-21'-
'o i
12 k
33 I N
Y l
6 e
I d
k 10 I
5 WY t
l yh 8
s:
I i
~.
z N
g I
6 O
20 40 50 PLANAR AVERAGE EXFCSURE (GWd/ST)
DATA COORDINATES GWD/ST kWft 0.2 11.3 20.0 11.3 25.0 11.1-30.0 10.6 35.0 10.1 40.0 9.7 45.0 9.0 50.0 8.0 55.0 7.2 61.5 6.1 Figure 3,11 1.2 Maximum Average Planar Linear Hea: Generation Ra:e versus Exposure with L?CI Modification and Sypass Holes ? lugged. GEli LTA Tuel Amendment No. #8,133 211a-
l j
J l
J e
d 13-j
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o 2400 tocco :5000 20000 tuco 3 Coco 38400 40000 48@00 50D00 PLANAR AVERAGE EXPOSURE (Wd/t)
Tigure 3.11-1.3 Maximu:n Average Plana: Linear Beat Generation Rate versus Exposure with LPCI Modification and Bypass Flow Ecles Plugged, P8tR3 65L and BPCDR306$L Tuel and 1988 LTA.
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Figure 3.11-1.l. Maxi =un Average ? lana: Linear Eea: Generati n ? ate versus E:posure with LPC1 Modification and 3ypass Flow Ecles Plugged SDF3:S3, ?SDF3:52 ar.d 3?!!?l:52 Fuel.
l Amendment No. 93,133
- 115-
. LIMMINC CONDITIONS FOR OPERATION Sl'RVEILULNCE REDUIREME*.iS
. If at any time during steady state operation it is determined by normal surveillance that the limiting value for DIGR is being exceeded action shall then be initiated to restore operation to within the prescribed limits.
Surveillance and corresponding action shall ec.ntinue until the prescribed limits are again being met.
C.
Minimum Critical Power Ratie (MCPR)
C, Minimum Critical Power Ratio (MCFP3 During steady state power operation MCPR shall be determined daily the MCPR for each type of fuel at during reactor power operation a":
rated power and flow shall not be
> 2M ' rated thermel power and lower than the limi:ing value shown following any change in power level in Figure 3.11+2 for tw or distribution that would cause recirculation loop operation.
If.
operation with a limiting control-at any time during steady state r
rod pattern as described in the operation it is determined by normal surveillance that the limiting value bases for Specification 3.3.B.5.
i for MCPR is being exceeded. action shall then be initiated within 15 minutes to restore opera: ion to within the prescribed limits.
If the steady state MCPR is not returned co within the prescribed limits within two (2) hours. the reactor shall be brouch: 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 ac: ion shall continue un:11 :he prescribed limits are again being me:.
For core flows other than rated :he MC?R shall be the opera:ing limit a ra:ed flow times K,. where K, is as l
calcula:ed in Table 3.11.1. '
i For one recircula: ion loop operation
~
the MCPR limi:s at ra:ed flow are 0.01 higher than :he comparable two loop values.
Amendment No. M,N,133 212'
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1.10 0.0 C. 2 0.4 0.6 0.8 1.0 Y
(based on tested measured scram time as defined in Reference 9)
Fi;;ure 3.112a GE3X3N3 Fuel (SCC to ICC)
Amendment No. B0,133
-212b-l
_______._______-]
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Ti pre 3.11-2b GEli LTA Fuel (30C to ECC)
Amendment No. 50,80,133 212e-
1 l
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(based on tested measured scram time as denned in Reference 9)
Figure 3.11 2: BPBXSR and 1988 LTA Tuel (3CC to ECC) i Amendment No. //0,80,FJ,133
-212d-r
Table 3.11.1 BWR/2-4 TLOW DE?ENDENT MCPR MULTIPLIER (K,) WITH GEIL FLUS LOW FLOW ADJUSTMENT i
I Tor 40% < WT s 100%,
W - MAX [1.0, A - 0.00641*kT]
kT s 40%,
& - [A 0.00441*WT)*[1.0 + 0.0032*(40 WT))
i whare WT - Percent of Rated Core Flow, and A-constant which depends on the Flow Control mode and the Scoop Tube Setpoint as noted below.
t SCOOP TUBE E.OU CONTROL MODE SETPOINT A
MANUAL 102.5%
1.3308 MANUAL 107.0%
1.3528 MANUAL 112.0%
1.3793 MANUAL 117.0%
1.4035 e
AUT0".ATIC N/A 1.4410 Amendment No. 40,133 212
,3.11)hSI),
A.
Averare Planar Linear Heat Generation Rate (APMGR)
This specification assures that the peak cladding temperature following the postulated design basis loss of coolant accident will not excewd the limit l
specified in 10CFR$0.46.
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 20'T relative to the peak temperature for a typical fuel design, the limit on the averaSe linear heat generation rate is sufficient to assure that calculated temperatures are within the 10CTR50.46 l
limit. The limiting value for API.HGR is shown in Figure 3.11 1.
The AFLHGR valves are reduced for single loop operation per Reference 10.
l B.
Linear Heat Generation Rate (LHCR)
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 LHGR as a function of core height shall-be checked daily l
during reactor operation at 2: 25% power to determine if fuel burnup, or control rod movement has caused changes in power distribution.
For LHCR to be a limiting value below 25% rated thermal power, the MTPF would have to be greater than 10 which is precluded by a considerable margin when employing any per nissible control rod pattern. Pellet densification power spiking in GE fuel has been accounted for in the safety analysis presented in References 1 and 2:
thus no adjustment to the 1.HGR limit for densification effects is required.
Amendment No. 62.80,94.133-
-21e-
l, 3.111AA.ta:
(Cont'd)
The K factors as calculated by Table 3.11.1 were developed generically which are l
applikable to all Bn/2, BW/3, and B1;R/4 reactors.
The K,. factors were derived using the flow control line correspondin5 to rated thermal p6wer at rated core flow as described in Reference 1.
The K factors are conservative for Cooper operation because the operating limit l
g MCPR's are greater than the original 1.20 operating limit MCPR used for the generic derivation of K.
f References for Bases 3.11 1.
" General Electric Standard Application for Reactor Fuel,"~
NEDE 24011 P A*(1stest approved revision).
2.
" Supplemental Reload Licensing Submittal for Cooper Nuclear Station,"
(applicable reload document).
3-8.
Deleted 9.
Letter (with attachment), R. H. Buckholt (CE) to P. S. Check (NRC), " Response to NRC Request for Information on ODYN Computer Model," September 5, 1980.
10.
" Cooper Nuclear Station Single Loop Operation," NEDO 24258, i
a i
Amendment No. BO,80,N,106,133
- 1:,b -
4 5.0 NAJOR DESIGN FEATURES 5.1 Site Features The Cooper Nuclear Station site is located in Nemaha County, Nebraska, on the west bank of the Missouri River, at river mile 532.5. This part of the river is referred to by the Corps of Engineers as the Lower Brownville Bend.
Site coordinates are approximately 40' 21' north latitude and 95' 38' west longitude. The site consists of 1351 acres of land owned by Nebraska Public Power District.
About 205 acres of this property is located in Atchison County, Missouri, opposite the Nebraska portion of the station site. The land area upon which the station is constructed is crossed by the Missouri River on the east and is bounded by privately owned property on the north, south, and west.
At the west site boundary, a county road and Burlington Northern Railroad spur pass the site.
The reactor (center line) is located approximately 3600 feet from the neares:
property boundary. No part of the present property shall be sold or leased by the applicant which would reduce the minimum distance from the reactor to the nearest si:e boundary to less than 3600 feet without prior NRC approval.
The protected area is formed by a seven foot chain link fence which surrounds the site buildings.
5.2 Resetor A.
The reactor shall contain $48 fuel assemblies.
Each assembly shall consis:
of a matrix of Zircalloy clad fuel rods with an initial compcsition of slightly enriched uranium dioxide (UO ) as fuel material.
Tuel assemblies shall be 2
limited to those fuel designs approved by the NRC for use in BWRs.
B.
The core shall contain 137 cruciform shaped control rods. The control material shall be boren carbide powder (B C) compacted to approximately 70% theoretical 3
density, except for the Hybrid I control rods which contain approximately 15%
C.
Lead Tes: Assembly (LTA) control blades and fuel assemblies of different design than described above may be installed under the provisions of 10CFR50.59 in conjunction with vendor tes: programs. The LTAs shall have been analyzed using methods previously approved by the NRC. The licensee will provide the NRC with '
a report describinq the LTAs and analyses not less than 30 days prior to startup.
5.3 Resetor Vessel The reactor vessel shall be as described in Section IV-20 of the SAR. The applicable design shall be as described in this section of the SAR.
5,4 Containment A.
The principal design para =eters for the primary containment shall be as given in Table V 2 1 of the SAR.
The applicable design shall be as described in Section XII 2.3 of the SAR.
B.
The secondary containment shall be as described in Section V 3.0 of the SAR.
C.
Pene: rations :o the primary containmen and piping passing through such Amendment No. 118,133
-217-
....