ML20137S773
| ML20137S773 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 09/24/1985 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Nebraska Public Power District (NPPD) |
| Shared Package | |
| ML20137S777 | List: |
| References | |
| DPR-46-A-094, TAC 42418 NUDOCS 8509300503 | |
| Download: ML20137S773 (38) | |
Text
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fo, UNITED STATES 3
NUCLEAR REGULATORY COMMISSION o
g
.t WASHINGTON, D. C. 20555
%,...../
NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 94 License No. DPR-46 l.
The Nuclear Regulatory Comission (the Comission) has foend that:
A.
The application for amendment by Nebraska Public Power District dated September 20, 1985, as supplemented September 23, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission; C.
There is reasonable assurance (i) that the activities authorized
-by this amendment can be conducted without endangering the health i
and safety of the public, and (ii) that such activities will be i
conducted in compliance with the Comission's regulations; D.
The issuance of this amendment 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 Comission's regulations and all applicable requirements have been satisfied.
l 2.
Accordingly, the license is amended by changes to the Technical Spec-i ifications as indicated in the attachment to this license amendment l
and paragraph 2.C(2) of Facility Operating License No. DPR-46 is hereby i
i amended to read as follows:
i 8509300503 S W ADOCK 05000298 PDR PDR P
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j (2) Technical Specification The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 94, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NU LEAR REGULATORY COMMISSION i
Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing i
Attachment:
Changes to the Technical j
Specifications Date of Issuance: September 24, 1985 i
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s ATTACHMENT TO LICENSE AMENDMENT NO. 94 FACILITY OPERATING LICENSE NO. DPR-46 DOCKET NO. 50-298 l
Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
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105, 106 (Blank) l 137 i
151 210 212 214 l
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RADIOLOGICAL TECHNICAL SPECZFICATIONS 8
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I TABLE OF CONTENTS Page No.
', J UV s:. (.XI f ;
.0n 1.0 DEFINITIONS
"~"
"'323 1-5 LIMITING SAFETY SAFETY LIMITS SYSTEM SETTINGS 1.1 FUEL CLADDING INTEGRITY 2.1 6 - 22 1.2 REACTOR COOLANT SYSTEM INTEGRITY 2.2 23 - 26 I
SURVEILLANCE 6
LIMITING CONDITIONS FOR OPERATION REQUIREMENTS i
3.1 REACTOR PROTECTION SYSTEM 4.1 27 - 46 3.2 PROTECTIVE INSTRUMENTATION 4.2 47 - 92 A.
Primary Containment Isolation Functions-47 B.
Core and Containment Cooling Systems Initiation 47 and Control (CS, LPCI, HPCI, RCIC, ADS)
C.
Control Rod Block Actuation 47 i
D.
Radiation Monitoring Systems'- Isolation and 48 i'
Initiation Functions 1.
Steam Jet Air Ejector Off-Gas System 48 2.
Reactor Building Isolation and Standby Gas 48 Treatment Initiation 3.
Liquid Radvaste Discharge Isolation 48 4.
Main Control Room Ventilation 48 j
5.
Mechanical Vacuum Pump Isolation 49 l
1 E.
Drywell Leak Detection i
49 F.. Primary Containment Surveillance,Information 49 l
Readouts i
G.
Recirculation Pump Trip l
49 H.
Post-Accident Monitoring 49 3.3 REACTIVITY CONTROL 4.3 93 - 106 i
1 A.
Reactivity Limitations A
93 B.
B 94 l
C.
Scram Insertion Times.
C 97 i
D.
Reactivity Anomalies D
98 i
E.
Restrictions E
98
'~
l F.
Recirculation Pumps F
98 G.
98a I
3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 107 - 113
[
i A.
Normal Operation A
107-B.
Operation with Inoperable Components B
108 C.
Sodium Pentaborate Solution l
C 108 i
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! Amendment No. 94
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TABLE OF CONTENTS ~(cont'd)
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[
]
Page No.
I SURVEILLANCE LIMITING CONDITIONS FOR OPERATION
, REQUIREMENTS,ation IOCntiCal SPC0ificallOns I
3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 114 - 131 A.
Core Spray and LPC1 Subsystems A
114
{
B.
Containment Cooling Subsystem (RHR Service Water)
B 116 C.
HPCI Subsystem C
117 D.
RCIC Subsystem D
118 E.
Automatic Depressurization Eystem E
119 3
F.
Minimum Low Pressure Cooling System Diesel 120 l
Generator Availability F
G.
Maintenance of Filled Discharge Pipe G
122 j
H.
Engineered Safeguards Compartments Cooling H
123 i
4 3.6 PRIMARY SYSTEM BOUNDARY 4.6 132 - 158 j
A.
Thermal and Pressurization Limitations A
132 j
B.
Coolant Chemistry B
133a i
C.
Coolant Leakage C
135 D.
Safety and Relief Valves-D 136 E.
Jet Pumps E
137 F.
Jet Pump Flow Mismatch F
137 G.
Inservice Inspection C
137 H.
Shock Suppressors (Snubbers)
H 137a i
j 3.7 CONTAINMENT SYSTEMS 4.7 159 - 192 A.
159 B.
Standby Gas Treatment System B
165 f
j C.
165a i
D.
Primary Containment Isolation Valves D
166 3.8 MISCELLANEOUS RADIOACTIVE MATERIAL SOURCES 4.8 185 - 186 3.9 AUXILIARY ELECTRICAL SYSTEMS 4.9
.193 - 202 l
A.
Auxiliary Electrical Equipment A
193 B.
Operation with Inoperable Equipment B
195 3.10 CORE ALTERATIONS 4.10 203 - 209 i
j A.
Refueling Interlocks A
203 B.
Core Monitoring B
205
)
C.
Spent Fuel Pool Water Level C
205 D.
Time Limitation ~
D 206 E.
Spent Fuel Cask Handling E
206 3.11 FUEL RODS 4.11 210 - 214e l
l A.
Average Planar Linear Heat Generation Rate (APLHCR)
A 210 j
B.
Linear Heat Generation Rate (LHCR)
B 210 C.
Minimum Critical Power Ratio (MCPR)
C 212 1
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l Amendment No. 94 4(15x-xxxxxxxxj
S FETi~L'IMITS 7 IMITING SAFETY ~STSTEM' SETTINGS 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability Applicability The Limiting Safety System Settings The Safety Limits established to pre-apply to trip settings of the instru-serve the fuel cladding integrity ap-ments and devices which'are provided ply to those variables which monitor to prevent the fuel cladding integ-the fuel thermal behavior.
rity Safety Limits from being exceeded.
Objective Objective The objective of the Limiting Safe-The objective of the Safety Limits is ty System Settings is to define the to establish limits below which the level of the process variables at integrity of the fuel cladding is which automatic protective action preserved.
is initiated to prevent the fuel cladding integrity Safety Limits Action from being exceeded.
pec M cations If a Safety Limit is exceeded, the reactor shall be in at least hot A.
Trip Settings shutdown within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The limiting safety system trip Specifications settings shall be as specified below:
A.
Reactor Pressure 1800 psia and 1.
Neutron Flux Trip Settings Core Flow 310% of Rated a.
APRM Flux Scram Trip Setting (Run Mode)
The existence of a minimum crit-When the Mode Switch is fn ical power ratio (MCPR) less than the RUN position, the APRM 1.07 for two recirculation loop flux scram trip setting operation (1.08 for single-loop shall be:
operation) shall constitute S<0.66 W + 54%
.66 AW l
violation of the fuel cladding integrity safety.
where:
B.
Core Thermal Power Limit (Reactor S " Settin8 in Percent of Pressure <800 psia and/or Core rated thermal power Flow <10%)-
(2301 Mw')
W -Two-loop recirculation l When the reactor pressure is <800 flow rate in percent psia or core flow is less than of rated (rated loop 10% of rated, the core thermal recirculation flow power shall not exceed 25% of-rate is that recircu-rated thermal power.
lation flow. rate which provides 100% coreflow C.
Fower Transient at 100% power)
AW = Difference between To ensure that the Safety Limit tv -1 P and single-established in Specification 1.1.A and 1.1.B is not exceeded, 1 P effective drive each required scram shall be ini-fl w at the.same core tiated by its expected scram sig-fl "*
i nal. The Safety Limit shall be assumed to be exceeded when scram is accomplished by a means other j
than the expected scram signal.
Amendment No. jff, p2', JW, 94
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SAFETY f1MITS
~ L'lMITING' SAFETY' SYSTEM SETTINGS l
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i 1.1 (Cont'd) 2.1.A.1 (Cont'd) l D.
Cold Shutdown AW = 0 for two recirculation loop operation.
Whenever the reactor is in 2^
<r-n'-
non the cold shutdown condition a.
In the event of operation with a
~
with irradiated fuel in the maximum fraction of limiting power reactor vessel, the water density (MFLPD) greater than the level shall not be less than fraction of rated power (FRP),
18 in. above the top of the the setting shall be modified as normal active fuel zone (top follows:
of active fuel is defined in r
Figure 2.1.1).
S < (0.66 W + 54% - 0.66 AW)
- where, FRP = fraction of rated thermal power (2381 MWt)
MFLPD = maximum fraction of limiting power density where the limiting power density is 18.5 KW/ft for 7x7 fuel and 13.4 KW/ft for 8x8 fuel.
The ratio of FRP to MFLPD shall be 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.
I For no combination of loop I
recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.
i b.
APRM Flux Scram Trip Setting (Refuel or Start and Hot Standby Mode)
When the reactor mode switch is in the REFUEL or STARTUP posi-tion, 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.
Amendment No. )(, )9', g, (#, g, 94
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_ ~ _
ShFETY'L1MITS
- f1 METING" SAFETY SYSTEM SETTINGS
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2.1.A.1 (Cont'd) d.
APRM Rod Block Trip Setting -
The APRM rod block trip' setting shall be: :;
's SRB < 0.66 W + 42%
.66 AW where:
d block setting in S
=
RB percent of rated thermal power (2381 MWt)
W and AW are defined in Specifi-cation 2.1.A.I.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:
l RB -(0.66 W + 42% - 0.66 AW)
FRP S
- where, FRP = fraction of rated thermal power (2381 MWt)
MFLPD - maximum fraction of limiting power density where the limiting power density is 18.5 KW/ft for 7x7 fuel and 13.4 KW/ft for 8x8 fuel.
The ratio of FRP to MFLPD shall be 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.
2.
Reactor Water Low Level Scram and Isolation Trip Setting (except MSIV)
> +12.5 in, on vessel level instruments.
l Amendment No. K, K, )6 ff, f6, g, 94,xg8-x-xxxxxxxx
1.5 Bectat Fuel Cladding Integrity I
r A.
Fuel Cladding Integrity Limit at Reactor Pressure 2800 psia and l
Core Flow 210% of Rated l
- n 1
The fuel cladding integrity safety limit is set such that no "
fuel damage is calculated to occur if the limit is not violated.
Since the parameters which result in fuel damage are not directly 1
observable during reactor operation the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark 1
the beginning of the region where fuel damage could occur. Although I
it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at I
which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring I
the core operating state and in the procedure used to calculate j
the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected i
to avoid boiling transition considering the power distribution within the core and all uncertainties.
~
1 The Safety Limit MCPR is generically determined in Reference 1 for two i
recirculation loop operation. This safety limit MCPR is increased by O.01 for single-loop operation as discussed in Reference 2.
B.
Core Thermal Power Limit (Reactor Pressure < 800 psia and/or Core Flow < 10% of Rated)
At pressures below 800 psia, the core elevation pressure drop (0 power, O flow) is greater than 4.56 psi. At low power and all flows this pressure differential is maintained in the bypass region of the core.
Since the pressure drop in the bypass region is essentially all elevation 4
head, the core pressure drop at low power and all flows will alvags be 1
l greater than 4.56 psi. Analyses show that with a flow of 28 x 10 Ibs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.- Thus,thegundle.flowwith l
a 4.-56 psi driving head will be greater than 28 x 10 lbs/hr irrespective of total core flow and independent of bundle power for the range of l
bundle powers of concern. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors this corresponds to a core thermal power of more than 50%. Thus.
l a core thermal power limit of 25% for reactor pressures below 800 psi or core flow less than 10% is conservative.
i l
I Amendment No. 94
.rMr.
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1.1 Besoa
(Cont'd)
_a C.
Power Transient Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit,of Specification n
1.1A or 1.1B will not be exceeded.
Scram times are,checkedjperiodically 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 of the plant design. The concept of not approaching a Safety Limit provided 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, APRM trip initiation, pressure scram initiation, etc. occur. This program also indicates when 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 normally will be available for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specification 1.1.C will be relied on to determine if a Safety Limit has been violated.
D.
Reactor Water Level (Shutdown Condition)
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 cladding temperatures and clad perforation.
The core can be cooled sufficiently should the water level be reduced to two-thirds the core height.
Establishment of the safety limit at 18 inches above the top of the fuel provides adequate margin.
Refe'rences for 1.1 Bases 1.
" Generic Reload Fuel Application," NEDE-24011-P (most current approved submittal).
2.
" Cooper Nuclear Station Single-Loop Operation," NEDO-24258, May,1980.
t LAmendmentNo.94
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s Ceccar Nue: ear sis: son
' r.-n:.11 Spec:fication
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, Amendment No. 94
-13. 149:4.5, 16-xxxxxxxx(
2.1 Bases
}
The abnormal operational transients applicable to operation of the CNS Unit have been analyzed throughout the spectrum of planned operating con-ditions up to 105% of rated steam flow.
The analyseslwer"E[ base ('uhodI 3
plant operation in accordance with Reference 3.
In addition,'2381'Wt is f
M the licensed maximum power level of CNS, and this represents the maximum steady-state power which shall not knowingly be exceeded.
I t
The transient analyses performed each reload are given in Reference 1.
Models and
[
model conservatisms are also described in this reference. As discussed in Refer-t ence 2, the core wide transient analyses for one recirculation pump operation is I
conservatively bounded by two-loop operation analyses and the flow-dependent rod block and scram setpoint equations are adjusted for one-pump operation.
l i
A.
Trip Settings The bases for individual trip settings are discussed in the following
[
paragraphs.
i t
1.
Neutron Flux Trip Settings h
I a.
APRM Flux Scram Trip Setting (Run Mode) i The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power
(
(2381 MWt). Because fission chambers provide the basic input signals, the APRM system responds directly to
(
i average neutron flux. During transients, the instanta-
[
neous rate of heat 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 l
be less than that indicated by the neutron flux at the scram setting. Analyses demonstrate that with a 120% scram trip setting, none of the abnormal operational transients i
analyzed violate the fuel Safety Limit and there is a j
substantial margin from fuel damage. Therefore, the use of flow referenced scram trip provides even additional margin.
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- Amendment No. pf',,8d', 94
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2.1 Beens
(Cont'd) i 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 determined by an analysis of margins required to provide a reasonable range for maneuvering,durjng..opcration.
Reducing this operating margin would increase the frequency 3hfyspurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was se-lected because it provides adequate margin for the fuel cladding integ-rity Safety Limit yet allows operating margin that reduces the possi-bility of unnecessary scrams.
The scra7 trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of maximum fraction of limiting power density (MFLPD) and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1. A.1.a. when the NFLPD is greater than the fraction of rated power (FRP). This adjust-ment may be accomplished by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced APRM High Flux Scram Curve by the reciprocal of the APRM gain change.
Analyses of the limiting transients show that no scram adjustment is required to assure MCPR above the safety limit when the transient is initiated from the operating MCPR limit.
i b.
APRM Flux Scram Trip Setting (Refuel or Start & 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 avail-able during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are con-strained to be uniform by operating procedure backed up by the rod worth minimizer, and the rod sequences control system. Worth of indivi-dual 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. Gen-i erally, the heat flux is in near equilibrium with the fission rate.
In an assumed uniform rod withdrawal approach to the scram level, the rate i
of power rise is no more than 5 percent of rated power per minute, and r
the APRM system would be more than adequate to assure a scram before l
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 Specifi-i cation 2.1.A.6.
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j Amendment No. 94
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2.1 Brete (Cont'd) c.
IRM Flux Scram Trip Setting The IRM system consists of 8 chambers, 4 in each.of the reactor protec-tion system logic channels. The IRM is a 5-decade instrument.which cov-ers the range of power level between that covered'b'yQh(SRM ahd,the APRM. The 5 decades are covered by the 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 range of the IRM.
For example, if the instrument were on range 1, the scram setting would be 120 divisions for that range; l
likewise, if the instrument were on range 5, 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 sources of reactivity change during the power in-crease are due to control rod withdrawal. For in-sequence control rod withdrawal, the rate of change of power is slow enough due to the phys-ical limitation of withdrawing control rods, that heat flux is in equi-librium with the neutron flux and an IRM scram would result in a reac-ter shutdown well before any Safety Limit is exceeded.
In order to ensure that the IRM provided adequate 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 subcritical and the IRM system is not yet on scale. This condition exists at quarter rod density. Additional conserva-tism was taken in this analysis by assuming that the IRM channel clos-est to the withdrawn rod is by-passed. The results of this analysis show that the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above the MCPR fuel cladding integrity safety limit. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
?
d.
APRM Rod Block Trip Setting i
Reactor power level may be varied b'y 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 flow decreases for the specified trip setting versus flow relationship; therefore the worst case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting. The actual power distri-bution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system. As with the APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum fraction of limiting power density exceeds the fraction of rated power, thus preserving the APRM rod block safety mar-gin. As with the scram setting, this may be accomplished by adjusting the APRM gain.
1 I
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- Amendment No. 94 e19ec.
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2.1 Beses
(Cont'd) 2.
Reactor Water Low Level Scram and Isolation Trip Setting (except-MSIV)
The set point for low level scram is above the bottom of the separator skirt. This level has been used in transieng 3nalyses,7 dealing with coolant inventory decrease.
The results reportedrio.3SAR Subsection 14.5 show that scram at this level adequately protects the fuel and the pressure barrier, because MCPR remains well above the MCPR fuel cladding integrity limit in all cases, and system pressure'does not reach the safety valve settings. The scram setting is approximately 25 in. below the normal operating range and is thus adequate to avoid spurious scrams.
3.
Turbine Stop Valve Closure Scram Trip Setting The turbine stop valve closure scram trip anticipates th'e pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of
<10 percent of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the MCPR j
fuel cladding integrity limit even during the worst case transient that assumes the turbine bypass is closed. This scram is bypassed when turbine steam flow is below 30% of rated, as measured by turbine first stage pressure.
4.
Turbine Control Valve Fast Closure Scram Trip Setting I
The turbine control valve fast closure scram anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection exceeding the capability of the bypass valves. The reactor protection system initiates a scram when fast closure of the control valves is initiated by the loss of turbine control oil pressure as sensed by pressure switches. This setting and the fact that control valve closure time is approximately twice as long as that for the stop valves means that resulting transients, while similar, are less severe than for stop valve closure. No significant change in MCPR occurs. Relevant transient analyses are presented in Paragraph 14.5.1.1 of the Safety Analysis l
, Report.
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b e Amendment No. 94
-204 xxxxxxxxj 9
.,,.e.
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2.1 Bases
(Cont'd) 5.
Main Steam Line Isolation Valve closure on Low Pressure a+cer Nuewar Swoon The low pressure isolation of the main steam lines,(Specifin; cation 2.1.A.6) was provided to protect against rapid reactor depressurization.
B.
Reactor Water Level Trip Settings Which Initiate Core Standby Cooling System (CSCS)
The core standby cooling subsystems are designed to provide suf-ficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident and to limit fuel clad temperature, to assure that core geometry remains intact and to limit any clad metal-water reaction to less than 1%.
To accomplish their intended function, the capacity of each Core Standby Cooling System component was established based on the reactor low water level scram set point. To lower the set point of the low water level scram would increase the capacity requirement for each of the CSCS components.
Thus, the reactor vessel low water level scram was set low enough to permit margin for operation, yet will not be set lower because of CSCS capacity requirements.
The design for the CSCS components to meet the above guidelines was dependent upon three previously set parameters: The maximum break size, low water level scram set point and the CSCS initiation set point. To lower the set point for initiation of the CSCS may lead to a decrease in effective core cooling. To raise the CSCS initia-tion set point would be in a safe direction, but it would reduce the margin established to prevent actuation of the CSCS during normal operation or during normally expected transients.
Transient'and accident analyses reported in Section 14 of the Safety l
Analyses Report demonstrate that these conditions result in adequate safety margins for the fuel.
C.
References for 2.1 Bases 1.
" Generic Reload Fuel Application," NEDE-24011-P, (most current approved submittal).
2.
" Cooper Nuclear Station Single-Loop Operation," NEDO-24258, May 1980.
3.-
" Supplemental Reload Licensing Submittal for Cooper Nuclear Station Unit 1,"
(applicable reload document).
l 4.
Safety Analysis Report (Section XIV).
l 2
Amendment No. 94
[x x'x.
xxxxxxxx
(
4 i
s 1
OCDCOf Nuclear C*at,an t'OhniCal SDeciflO3:lons "LEFT BLANK INTENTIONALLY" l
f I
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Amendment No. 94
-22n.
XXXXXXXX
COOPER NUCLEAR STATION k
TABLE 3.1.1 i
REACTOR PROTECTION SYSTEM INSTRUMENTATION REQUIREMENTS 2
A l
2 Minimum Number Action Required
,0 Applicability Conditions of Operable When Equipment Reactor Protection Mode Switch Position Trip Level Channels Per Operability ;is e
System Trip Function Shutdown Startup Refuel Run Setting Trip Systems (1) Not Assured 1(1)
Mode Switch in Shutdown X(7)
X X
X 1
A Manual Scram X(7)
X X
X 1
A IRM (17)
X(7)
X X
(5) f 120/125 of.in-3 A
High Flux dicated scale Inoperative X
X (5) 3 A
APRM (17)
X j (0.66W+54%-0.66AW) ' FRP 2
A or C
~
High Flux (Flow biased)
(14)(19)
MFLPD High Flux X(7) "
X(9)
X(9)
"(16) j 15% Rated Power 2
A or C Inoperative X(9)
X(9)
X (13) 2 A or C Downscale (11)
X(12) 3 2.5% of indi-2 A or C cated scale High Reactor. Pressure X(9)
X(10)
X j 1045 psig 2
A NBI-PS-55 A,B,C, & D High Drywell Pressure X(9)(8) X(8)
X j 2 psig 2
A orlD PC-PS-12 A,B,C, & D 6
- t l
Reactor Low Water Level X
X X
3 + 12.5 in. Indi-2 A or,D l NBI-LIS-101 A,B,C, & D cated level i
Scram Discharge Instrument Volume X
X(2)
X j 92 inches 3 (18)
A High Water Level CRD-LS-231 A & B xy CRD-LS-234 A & B y
CRD-LT-231 C & D y
CRD-LT-234 C & D X
W
- 11. -Th TAPRM dod ic~ ale trip function is only active when thn' reactor mode
~
, - switch is-in-RUN.
12.
The APRM downscale trip is automatically bypassed when the mode switch is not in RUN.
13.
An APRM will be considered inoperable 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.
16.
The 15% APRM scram is bypassed in the RUN mode.
17.
The APRM and IRM instrument channels function in both the Reactor Protection System and Reactor Manual Control System (Control Rod Withdraw Block, Section 3.2.C.).
A failure of one channel will affect both of these systems.
18.
The minimum number operable associated with the Scram Discharge Instrument Volume are three instruments per Scram Discharge Instrument Volume and three level devices per RPS channel.
19.
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.
t AmendmentNo.JW',f6',g,g,k,94
,-3),
XXXXXXXX
'$iMiiisildEi)1EidN53dRlPERAiiO SURVEiEUiCEItEOUIREMENU 3.1 BASES (Cont'd.)
4.1 BASES (Cont'd.)
there is proper overlap in the neu-For the APRM system, drift of tron monitoring system functions and electronic apparatus is not t
thus, that adequate coverage is pro-the only consideration in deter-
~
vided for all ranges of reactor oper-mining a~c'alibration frequency, ation.
Change in power distribution and i
loss of chamber sensitivity dictate i
j.
a calibration every seven days. Cal-i
.ibration on this frequency assures plant operation at or below thermal limits.
l 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 shut-i down and manual scram.- All of the de-vices or sensors associated with these scram functions are simple on-off l
j switches and, hence, calibration curing ';
operation is not applicable.
i B.
The MFLPD is checked once per day l
to determine if the APRM scram requires adjustment. This will nor-mally be done by checking the LPRM readings. Only a small number of control rods are moved daily and thus the MFLPD is not expected to change f
significantly and thus a daily check i
of the MFLPD is adequate, j
The sensitivity of LPRM detectors de-creases with exposure to neutron flux 4
at a slow.and approximately constant rate. This is compensated for in the APRM system by calibrating once a week using a heat balance data and by cali-t brating individual LPRM's every six weeks of power operation above 20%
of rated power.
4 It is highly improbable that in
[
actual operation with MFLPD < FRP that MCPR will be as low as the MCPR fuel cladding integrity safety
[
limit. Usually with power densities of this magnitude the peak occurs low in the core in a low quality l
region where the initial heat l
i i
i I
Amendment No. Jef, yl, g, g 94 :.
uxn:.-
1 i
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._ ~
COOPER NUCLEAR STATION f.
TABLE 3.2.C CONTROL ROD WITHDRAWAL BLOCK' INSTRUMENTATION l
k I
i g
i i
z Minimum Number Of :
Function Trip Level Setting Operable Instrumenti Channels / Trip System (5)
~
~
APRM Upscale (Flow Bias)
< (0.66W + 42% - 0.66 AW)
FRP (2)(13) 2(1) l APRM Upscale (Startup) 512%
MFLPD 2(1) l APRM Downscale (9) 1 2.5%
2(1)
.APRM Inoperative (10b) 2(1)-
RBM Upscale (Flow Bias)
< 0.66W + (N - 66) (2) 1 l
RBM Downscale (9) 1 2.5%
1 RBM Inoperative (10c) 1 4
s IRM Upscale (8)
< 108/125 of Full Scale 3(1)
D' IRM Downscale. (3)(8) 3 2.5%
3(1)
IRM Detector Not Full In (8) 3(1)
IRM Inoperative.(8)
(10a) 3(1)
I i
SRM Upscale (8)
< 1 x 10 Counts /Second 1(1)(6)
SRM Detector Not Full In-(4)(8)
(3 100 cps) 1(1)(6) i i
SRM Inoperative.(8)
(10a) 1(1)(6) l l
Flow Bias Comparator
< 10% Difference In Recire. Flows 1
i Flow Bias Upscale /Inop.
< 110% Recire. Flow I
l l
SRM Downscale (8)(7) 3 3 Counts /Second (11) 1(1)(6) xy SDV Water Level High
-< 46 inches 1(12) y CRD-231E, 234E J
N i'
X g
.,p w.-
NOTES FOR TABLE 3.2.C 1.
For the startup and run positions of the Reactor Mode Selector Switch, the Control Rod Withdrawal Block Instrumentation trip system shall be operable for'each function. 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 Rod Withdrawal Block Instrumentation trip system is a one'out 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 5 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 l
setting is in percent of rated power (2381 MWt). N is the RBM setpoint selected (in percent) and is calculated in accordance with the methodology of the latest NRC approved version of NEDE-240ll-P-A.
3.
IRM downscale is bypassed when it is on its lowest range.
4.
This function is bypassed when the count is > 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 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 exists and the requirements of Specification 3.3.B.5 apply).
6.
IRM channels A.E.C.G all in range 8 or higher bypasses SRM channels A&C functions.
IRM channels B,F,D H 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 mode switch is placed in Run.
9.
This function is only active when the mode switch is in Run.
This function is automatically bypassed when the IRM instrumentation is operable and not high.
- 10. The inoperative trips are produced by the following functions:
a.
SRM and IRM (1) Mode switch not in operate (2) Power supply voltage low (3) Circuit boards not in circuit i
t AmendmentNo.)J',S5',94
,;ffx~.
xxxxxxxx
NOTES FOR TABLE 3.2.C (Continusd)
~ ~ ~ ~ - " ' - ~ ~ ~
'~
b.
APRM (1) Mode switch not in operate (2) Less than 11 LPRM inputs (3) Circuit boards not in circuit c '.
RBM (1) Mode switch not in operate (2) Circuit boards not in circuit-
~
1 (3) RBM fails to null (4) Less than required number of LPRM inputs for rod selected 11.
During spiral unloading / reloading, the SRM count rate will be below 3 cps-for some period of time. See Specification 3.10.B.
12.
With the number of OPERABLE channels less than required by the Minimum
~ Number of Operable Instrument Channels / Trip System 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.
1 f
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i Amendment No, f4', )4', 94
.-62a-xxxxxxxx
LIMITING'CORDIT 5NS"FOR OPEfATTOM SURVEILLANCE RECUIREMENTS -
~
3.3.C (Cont'd.) ~
~~
4.3.C (Cont'd.)
3.
The maximum scram insertion time for D.
Reactivity Anomalics 90% insertion of any operable control rod shall not exceed 7.00 seconds.
During the startup test program and startup following refueling outages, D.
Reactivity Anomalies the critical rod configurations will At a specific steady state base condi-be compared to the expected configura-tion of the reactor actual control rod tions at selected operating conditions.
inventory will be periodically com-These comparisons will be used as base pared to a normalized computer pre-data for reactivity monitoring during diction of the inventory.
If the subsequent power operation through-difference between observed and pre-out the fuel cycle. At specific power dicted rod inventory reaches the operating conditions, the critical rod equivalent of 1% Ak reactivity, the configuration will be compared to the reactor will be shut down until the configuration expected based upon ap-cause has been determined and correc-propriately crtrected past data. This tive actions have been taken as comparison will be made at least every appropriate.
full power month.
E.
Restrictions F.
Recirculation Pumps If Specifications 3.3.A through D 1.
With two recirculation pumps in above cannot be met, an orderly operation and with core thermal power shutdown shall be initiated and the greater than the limit specified in reactor shall be in the Shutdown Figure 3.3.1 and total core flow less condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
than 45% of rated, establish baseline l
F.
Recirculation Pumps levels within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, provided that l
1.
A recirculation pump shall not be baseline values have not been pre-started while the reactor is in viously established since the last natural circulation flow and reactor core rclueling.
power is greater than 1% of rated 2.
a)
Prior to operation with one re-thermal power.
circulation pump not in opera-2.
With two recirculation pu=ps in opera-tion and core thermal power tion and with core thermal power greater than the limit specified greater than the limit specified in in Figure 3.3.1 establish Figure 3.3.1 and total core flow less baseline APRM and LPRM* neutron than 45% of rated, the APRM and LPRM*
flux noise levels, provided that neutron flux noise levels shall be baseline values have not been determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and; previously established since the "E*
a) if the APRM and LPRM* neutron values shall be established with flux noise levels are less than or equal to three times their ne recircul8ti n pump not in peration and core thermal power established baseline levels, con-less than or equal to the limit tinue to determine the noise specified in Figure 3.3.1.
levels at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and also within 30 minutes after b)
Prior to operation with one re-the completion of a core thermal circulation pump not in opera-power increase of at least 5% of tion and core flow greater than rated core thermal power while 45% of rated, establish baseline operating in this region of the core plate AP noise levels with power /fIcw map, or core flow less than or equal to 45% of rated, provided that
- Detector levels A and C of one LPRM baseline values have not been string per core octant plus detector levels previously established with one A and C of one LPRM string in the center of recirculation pump not in the core shall be monitored.
operation since the last core
- Amendment No. g, g g, 94
, xxxxxxxx
~ LIMITI G dONDITI'ONS'FdR315R TidN'
' SURVEILLANCE" REQUIREMENTS
~
3.3.F (Cont'd.)
4.3 (Cont'd.)
b) if the APRM and/or LPRM* neutron flux G.
Scram Discharge Volume noise levels are greater than three times their established baseline 1.
The scram discharge volume (SDV) levels, immediately initiate corree-vent and drain valves shall be tive action and restore the noise cycled and verified open at levels to within the required limits least once every 31 days and within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing core prior to reactor start-up.
flow, and/or by initiating an orderly reduction of core thermal power by 2.
The SDV vent and drain valves inserting centrol rods.
shall be verified to close within 30 seconds after receipt 3.
The reactor may be started and ope-of a signal for control rod
^
rated, or operation may continue with
. scram once per refueling cycle.
One recirculation loop not in opera-tion provided that; 3.
SDV vent and drain valve opera-bility shall be verified follow-a.
with one recirculation pump not ing any maintenance or modifica-in operation and core thermal tion to any portion (electrical power greater than the limit or mechanical) of the SDV which specified in Figure 3.3.1, core may affect the operation of the flow must be greater than or vent and drain vavles.
equal to 45% of rated, and (i) the Surveillance Requirements of 4.3.F.2.a have not been satisfied.
immediately initiate action to reduce core thermal power to less than or equal to the limit speci-fled in Figure 3.3.1 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or (ii) the Surveillance Requirements of 4.3.F.2.a have been satisfied, continue to determine the APRM and LPRM neutron flux levels at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and also within 30 minutes after the completion of a core thermal power increase of at least 5% of rated core thermal power while operating in this region of the power / flow map.
If the APRM and/or LPRM* neutron flux noise.
levels are greater than three times their established baseline values, immediately initiate corrective action and restore the noise levels to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by
- Detector levels A and C of one LPRM string per core octant plus detector levels A and C of one LPRM string in the center of the core shall be monitored.
a Amendment No. 94
,,,98a-xxxxxxxx!
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Amendment No. 94
-98b-
LIMITING C05IIIT'I5NS EOR OPERATION
' SURVEILLANCE REQUIREMEWTS "
~
~
3.3.F (Cont'd.)
increasing core flow and/or initiating an orderly reduction of core thermal power by inserting control rods.
b.
With one recirculation pump not in operatien and core flow greater than 45% of rated, and (i) the Surveillance Requirements of 4.3.F.2.b have not been satisfied, immediately initiate action to reduce core flow to less than or equal to 45% of rated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or (ii) the Surveillance Requirements of 4.3.F.2.b have been satisfied, continue to determine core plate AP noise at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and also within 30 minutes after the completion of a core thermal power increase of at least 5% of rated thermal i
power.
If the core plate AP noise level is greater than 1.0 psi and 2 times its esta-blished baseline value, imme-i diately initiate corrective action and restore the noise levels to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by decreas-ing core flow and/or initiating an orderly reduction of core thermal power by inserting control rods.
c.
The idle" loop is isolated electrically by disconnecting the breaker to the recirculation pump motor generator (M/C) set drive motor prior to start-up, or if disabled during reactor operation, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d.
The recirculation system controls will j
be placed in the manual flow control mode.
I i
Amendment No. 94
-x~ 9 6c-xxxxxxxx w-
3.3 cnd 4.3 EASES (Cont'd)
~
~ ~ ' ~~
~~
the control rod motion is estimated to actually begin. However, 200 milliseconds is conservatively assumed for this time interval in the transient analyses and this is also included in the allowable scram insertion times of Specification 3.3.C.
The time to deenergize the pilot valve scram solenoid is measured during l
the calibration tests required by Specification 4.1.
D.
Reactivity Anomalies i
During each fuel cycle excess operative reactivity varies as fuel depletes and as l
any burnable poison in supplementary control is burned. The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state. Power operating base conditions provide the most sensitive and directly interpretable data relative to core reactivity.
Furthermore, using power operating base conditions permits frequent reactivity comparisons.
P L
Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% Ak.
l Deviations in core reactivity greater than 1% Ak are not expected and require thorough evaluation. One percent reactivity limit is considered safe since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the reactor system.
j F.
Recirculation Pumps Until analyses are submitted for review and approval by the NRC which prove that f
recirculation pump startup from natural circulation does not cause a reactivity insertion transient in excess of the most severe coolant flow increase currently analyzed, Specification 3.3.F.1 prevents starting recirculation pumps while the reactor is in natural circulation above 1% of rated thermal power. Specifications 3.3.F.2 and 3 are based upon providing assurance that neutron flux limit cycle oscillations, which have a small probability of occurring in the high power / low flow corner of the operating domain, are detected and suppressed. Bk'R cores typically operate with neutron flux noise levels of 1%-12% of rated power (peak to peak) due to random boiling and flow noise. These flux noise levels are i
considered in the thermal / mechanical design of GE Bk'R fuel, occur in a stable mode, and are found to be of negligible consequence. However, under certain high power / low flow conditions that could occur during a recirculation pump trip and subsequent Single Loop Operation (SLO) where reverse flow occurs in i
inactive jet pumps, a hydraulic / reactor kinetic feedback mechanism can be l
enhanced such that sustained limit cycle oscillations of flow noise with peak to peak levels several times normal values are exhibited. Although large
[
margins to safety limits are maintained when these limit cycle oscillations i
occur, they are to be monitored for, and suppressed when flux noise exceeds the three time baseline value by inserting rods and/or increasing coolant
[
flow. The line in Figure 3.3.1 is based on the 80% rod line below which the l
probability of limit cycle oscillations occurring is negligible. The thernal power, core flow, and neutron flux noise leve! limitations are prescribed in
[
4 accordance with Reference 3.
l
[
k t
i i
Amendment No. f J d y d 94
.n103-d xxxxxxxx
.__..m.
]
3.3 end 4'3 BA5ESi~~(C5itP'd)
~
~
" ~ ~ ~ ~ ~ ~
1 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 l
preclude establishing a water inventory, which if sufficiently large, could result 1
in slow scram times or only a partial control rod insertion.'
i
.i The vent and drain valves shut on a scram signal thus providing a contained volume j
(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.
1 REFERENCES 1.
Licensing Topical Report CE-BWR Ceneric Reload Fuel Application, NEDE-240ll-P, j
(most current approved submittal).
i 2.
" Supplemental Reload Licensing Submittal for Cooper Nuclear Station Unit 1,"
(applicable reload document).
]
3.
General Electric Service Information Letter No. 380, Revision 1, dated February 10, 1984.
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Amendment No. 94
+1,0,4-xxxxxxxx 4
- _ _. _. ~.
" INTENTIONALLY LEFT BLANK" l
t I
Amendment No. 94
-105,o 06-1 4.,
xxxxxxxx
LIMIT Ed'CONDI50 S FOR 0PERATIUk SUNEI'LLANCE REQUIREMENTS
~
~
~
1 T
j 3.6.E Jet Pumps 4.6.E.
Jet Pumps j
1.
Whenever the reactor is in the start-1.
Whenever there is recirculation flow up or run modes, all jet pumps shall with the reactor in the startup or i
be operable.
If it is determined run modes, jet pump operability shall i
that a jet pump is inoperable, or be checked daily by verifying that the if two or more jet pump flow in-following conditions do not occur sim-struments failures occur and cannot ultaneously:
be corrected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an j
orderly shutdown shall be initiated a.
The recirculation pump flow differs and the reactor shall be in a cold by more than 15% from the established j
Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
speed flow characteristics.
j b.
The indicated value of core flow rate varies from the value derived from loop flow measurements by more than 10%.
c.
The diffuser to lower plenum differen-tial pressure reading on an individual jet pump varies from the mean of all i
jet pump differential pressures by more than 10%.
F.
Jet Pump Flow Mismatch F.
Jet Pump Flow Mismatch i
1.
Deleted.
1.
Deleted.
l 2.
Following one-pump operation, the dis-1 i
charge valve of the low speed pump l
1 may not be opened unless the speed of the faster pump is equal to or less
]
than 50% of its rated speed.
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Inservice Inspection C.
Inservice Inspection To be considered operable, com-Inservice inspection-shall be per-ponents shall satisfy the require-formed in accordance with the i
ments contained in Section XI of requirements for ASME Code Class 1, l
the ASME Boiler and Pressure Vessel 2, and 3 components contained in i
Code and applicable Addenda for Seccion XI of the ASME Boiler and continued service of ASME Code Pressure Vessel Code and applicable 1
Class 1, 2, and 3 components except.
Addenda as required by 10 CFR 50, where relief has been granted by the Section 50.55a(g), except where Commission pursuant to 10 CFR 50, relief has been granted by the j
Section ~50.55a(g)(6)(1).
Commission pursuant to 10 CFR 50, j
Section 50.55a(g)(6)(1).
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Amendment No. J4$ 94
.y) 37.-
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1 3.6.E & 4.6.E BASES (Cont'd) jet pump body; however, the converse is not true. The lack of any substantial stress in the jet pump body makes failure impossible without an initial nozzle riser system failure.
F.
Jet Pump Flow Mismatch i,
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4 Requiring the discharge valve of the lower speed loop to remain closed until j
the speed of faster pump is equal to or less than 50% of its rated speed provides assurance when going from one to two pump operation that excessive vibration of the jet pump risers will not occur.
i G.
Inservice Inspection The inservice inspection program conforms to the requirements of 10 CFR 50 Section 50.55a(g). Where practical, the inspection of components conforms to the requirements of ASME Code Class 1, 2, and 3 components contained in Section XI of the ASME Boiler and Pressure Vessel Code.
If a Code required inspection is impractical, a request for a deviation from that requirement is s
submitted to the Commission in accordance with 10 CFR 50, Section 50.55a(g)(6)(1).
Deviations which are needed from the procedures prescribed in Section XI of the ASME Code' and applicable Addenda will be reported to the Commission prior to the beginning of each 10-year inspection period if they are known to be required at that time. Deviations which are identified during the course of inspection will be reported quarterly throughout the inspection period.
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Amendment No.,S(',' 94 352n.
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LIMITING CONDITIONS FOR OPERATION SURVEYLLANCE REQUIREMENTS 3*.11 FUEL RODS 4,71 FUEL RODS
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Applicability gg The Limiting Conditions for Operation associated with the fuel rods apply t The Surveillance Requirements apply those parameters which monitor the fuel to the parameters which monitor the rod operating conditions.
fuel rod operating conditions.
Objective Objective The Objective of the Limiting Condi-i tions for Operation is to assure the The Objective of the Surveillance performance of the fuel rods.
Requirements is to specify the type and frequency of surveillance to be Specifications applied to the fuel rods.
j A.
Average Planar Linear Heat g
f Generation Rate (APLHGR)
During steady state power opera-A.
Average Planar Linear Heat tion, the APLHGR for each type of Generation Rate (APLNGR) fuel as a function of average planar exposure shall not exceed The APLHCR for each type of fuel the limiting value shown in Figure as a function of average planar 3.11-1 for two recirculation loop exposure shall be determined operation. For single-loop oper-daily during reactor operation ation the values in these curves at > 25% rated thermal power.
~
are reduced by 0.84 for 7x7 fuel, 0.86 for 8x8 fuel. 0.77 for 8x8R fuel and 0.77 for P8x8R fuel.
If at any time during steady state operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the APLHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillance end corres-ponding action shall continue until the prescribed limits are again being met.
B.
Linear Heat Generation Rate (LHGR)
B.
Linear Heat Generation Rate (LHGR)
The LHGR as a function of core During steady state power opera-height shall be checked daily tion, the linear heat generation during reactor operation at > 25%
rate (LHGR) of any rod in any fuel rated thermal power.
assembly at any axial location shall not exceed the maximum allow-able LHGR as calculated by the following equation:
max 0 LHCRd [1 - ((AP/P)(L/LT))]
LHGR = Design LHCR =
G KW/ft.
d (AP/P),,g = Maximum power spiking penalty =
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l Amendment No. M 94 410-
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~SURVEILf'AN'CE REQUIREMENTS '~~
LIMITisi COND'I'I SS'FOR' PER TION T
-LT = Total core length - 12 feet
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L = Axial position above bottom of core G = 18.5 kW/ft for 7x7 fuel bundles
= 13.4 kW/ft for 8x8 fuel bundles N = 0.038 for 7x7 fuel bundles
= 0.0 for 8x8 fuel bundles If at any time during steady state operation it is determined by nor-mal surveillance that the limiting value for LHGR 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 lim-its are again being met.
C.
Minimum Critical Power Ratio (MCPR)
C.
Minimum Critical Power Ratio (MCPR)
During steady state power opera-MCPR shall be determined daily tion the MCPR for each type of fuel during reactor power operation at rated power and flow shall not be at > 25% rated thermal power lower than the limiting value shown and following any change in in Figure 3.11-2 for two recircula-power level or distribution that tion loop operation.
If, at any would cause operation with a time during steady state oper-limiting control rod pattern as ation it is determined by normal described in the bases for Spec-surveillance that the limiting ification 3.3.L 5.
value for MCPR is being exceeded, action shall then be initiated within 15 minutes to restore oper-ation to within the prescribed limits. If the steady state MCPR is not returned to within the pre-scribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown con-dition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveil-lance and corresponding action shall continue until the pre-s_cribed limits are again being met.
For core flows other than rated the MCPR shall be the operating limit at rated flow times K,
g where K is as shown in Figure g
3.11-3 For one recirculation loop oper-ation the MCPR limits at rated i
flow are 0.01 higher than the comparable two-loop values.
l AmendmentNo,J(,94 12-xxxxxxxx
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3.11 BASES A.
Average Planar Linear Heat Generation Rate (APLHGR)
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will. not, exceed the limit specified in the 10CFR50, Appendix K. G,,
"g' 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 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 temperatures are within the 10CFR50 Appendix K limit.
The limiting value for APLHGR is shown in Figure 3.11-1.
.B.
Linear Heat Generation Rate (LHGR)
This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel pellet densi-fication is postulated. The power spike penalty specified is based on the analysis presented in Section 5 of Reference 1 and assumes a linearly increasing variation in axial gaps between core bottom and top, and assures with a 95 confidence, that no more than one fuel rod exceeds the design linear heat generation rate.due to power spiking. The LHGR as a function of core height shall be checked daily during reactor operation at > 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 permissible control rod pattern. Pellet densification power spiking in 8x8 fuel has been accounted for in the safety analysis presented in Reference 2; thus l
no adjustment to the LHGR limit for densification effects is required for l
8x8 fuels.
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AmendmentNo.,5f'JOI94
-22 4 rx.
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3.11 Bmean:
(Cont'd)
C.
Minimum Critical Power Ratio (MCPR)
The required operating limit MCPR's at steady state operating conditions as specified in Specification 3.11C are derived from,$he es,tablished fuel t
cladding integrity Safety Limit and an analysis o'f.-abnormalyoperational 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 time 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 th,e largest reduction in critical power ratio (CPR). The models used in the transient analyses are discussed in Reference 1.
The purpose of the K factor is to define operating limits at other than f
rated flow conditions. At less than 100% flow, the required MCPR is the product of the operating limit MCPR and the K factor. Specifically, the g
Kg. factor provides the required thermal margin to protect against a flow j
increase 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.
l For operation in the automatic flow control mode, the K, factors assure that 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 vio-f lated for the same postulated transient event.
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j Amendment No.g,[ 94
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-3.11 Bases:
(Cont'd) t The K factor curves shown in Figure 3.11-3 were developed generically g
which are applicable to all BWR/2, BWR/3, and BWR/4 reactors. The K factorswerederivedusingtheflowcontrollinecorrespondingtora[ed thermal power at rated core flow as described in Reference,,1.
g,,,
y l
men, e cir cre ~
i The K factors shown in Figure 3.11-3, are conservative for Cooper opera-tionbecausetheoperatinglimitMCPR'saregreaterthantheoriginal l
{
1.20 operating limit MCPR used for the generic derivation of K.
g 4
i References for Bases 3.11 i
1.
Licensing Topical Report. General Electric Boiling Water Reactor,'
Generic Reload Fuel Application, (NEDE-24011-P), (most current approved submittal).
i 2.
" Supplemental Reload Licensing Submittal for Cooper Nuclear ~ Station i
Unit 1," (applicable reload document).
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l Amendment No.f?6I)6I 94
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4.11 B aan:
&B. Average and Local LHCR The LHCR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution.9,,g,,g,,
Since changes due to burnup are slow, and only a few contEol _r,oda:caar,
are moved daily, a daily check of power distribution is adequate.
C.
Minimum Critical Power Ratio (MCPR) - (Surveillance Requirement)
At core thermal power levels 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 excess 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 HCPR 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% rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when 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, j
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- AmendmentNo.,6fI,p6I94
-214c.<.
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