ML20132C457

Proposed Tech Specs Allowing Operation W/One Recirculation Loop Out of Svc Greater than 24 H

ML20132C457
Person / Time
Site:	Cooper
Issue date:	09/20/1985
From:	ARKANSAS POWER & LIGHT CO.
To:
Shared Package
ML20132C447	List: ML20132C457 NLS8500260, Application for Amend to License DPR-46,revising Tech Specs to Allow Operation W/One Recirculation Loop Out of Svc Greater than 24 H.Fee Paid
References
TAC-42418, NUDOCS 8509270056
Download: ML20132C457 (34)

Text

T

, RADIOLOGICAL TECHNICAL SPECIFICATIONS ~

TABLE OF CONTENTS Page No.

1.0 DEFINITIONS ' 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 SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 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 D. Radiation Monitoring Systems - Isolation and 48 Initiation Functions

1. Steam Jet Air Ejector Off-Gas System 48

2. Reactor. Building Isolation and Standby Gas 48 Treatment Initiation

3. Liquid Radwaste Discharge Isolation 48

, 4. Main Control Room' Ventilation 48

5. Mechanical Vacuum Pump Isolation 49 4 E. Drywell Leak Detection .

49 F. Primary Containment Surveillance Information 49 l Readouts l

G. Recirculation Pump Trip , 49-H. Post-Accident Monitoring 49.

3.3 REACTIVITY CONTROL 4.3 93 - 106 A. Reactivity Limitations A 93 B. Control Rods B 94

,, C. Scram Insertion Times C .97

!~ D. Reactivity. Anomalies D 98 l

E. . Restrictions 'E 98 F. Recirculation Pumps F 98 C. Scram Discharge Volume #

_G 98a 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 107 - 113 A.-' Normal Operation .

A 107 B. Operation with Inoperable Components B- '108 i C. Sodium Pentaborate Solution C 108.

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r TABLE OF CONTENTS (cont'd)

Page No.

SURVEILLANCE

' LIMITING CONDITIONS FOR OPERATION REQUIREMENTS m

%.: -ca c.:e:#, cat.cns

' 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 System E 119 F. Minimum Low Pressure Cooling System Diesel Generator Availability F 120 G. Maintenance of Fil' led Discharge Pipe G 122 H. -Engineered Safeguards Compartments Cooling -H 123 3.6 PRIMARY SYSTEM BOUNDARY 4.6 132 - 158 A. Thermal and Pressurization Limitations A 132 B. Coolant chemistry B 133a 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 G 137 H. Shock Suppressors (Snubbers) H 137a 3.7 CONTAINMENT SYSTEMS 4.7 159 - 192 A.. Primary Containment A 159 B. Standby Gas Treatment System C. Secondary Containment B 165 ,

C 165a 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 A. Auxiliary Electrical Equipment A 193 B. Operation with Inoperable Equipment B 195 3.10 . CORE ALTERATIONS 4.10 203 - 209 A. Refueling Interlocks A 203 B. Core Monitoring b 205 C. Spent Fuel Fool Water Level: C 205 D. Time Limitation D 206 E. Spent Fuel Cask Handling ,

T. 206 3.11 FUEL RODS 4.11 210 - 214e A. Average Planar Linear Heat Generation _ Rate ' (APLHGR) A 210 B. Linear Heat Generation Rate (LHGR) -B 210-C.. Minimum Critical Power Ratio (MCPR) C 212 I

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~ "~ l SAFETY? LIM 1TS "LDiITING SAFETY SYSTEM SETTIWGS

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.l.l' FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY l 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

'the fuel thermal behavior, to prevent the fuel cladding integ-rity Safety Limits from being exceeded.

Objective Objective

' The objective of the Limiting Safe-

~ The objective of the Safdty 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 ,

4 Action cladding integrity Safety Limits from being exceeded. '

If a Safety Limit is exceeded, the Specifications 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 systeni trip

. Specifications settings shall be as specified below: -

A. Reactor Pressure >800 psia-and -1. Neutron Flux Trip Settings Core Flow >10% of Rated

- a. APRM Flux Scram Trip Setting (Run Mode)

The existence of a minimum crit- When the Mode Switch is in 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 greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />),

shall constitute violation of the S<0.66 W + 54%

_ _ .66 AW l fuel cladding integrity safety. where:

B. Core Thermal Power Limit (Recctor S = Setting -in percent of-Pressure <800 psia and/or Core rated thermal' power .

Flow <10%) (2381 MWt)

When the reactor precrure is <800 W = Two-loop _ recirculation l psia or core' flow is less than. flow rate in percent 10% of rated,'the core thermal - of rated (rated loop power shall not exceed 25% of recirculation flow rated thermal power. rate is that recircu-lation flow. rate.which

• provides 100% coreflow C. . Power' Transient at .100% power)

• To ensure that the_ Safety Limit AW =. Difference between

established in Specification two-loop and single-1.1.A and 1.1.B is not exceeded, looPfeffective drive each required scram shall-be ini- flow at the same core

<tiated by its expected scram sig- flow. This factor is~

nal. ThefSafety. Limit shall b~e applied for single.

assumed to be exceeded when' scram 'lo p. operation' greater '

is accomplished by a means other: .than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ,

than the: expected scram signal.

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SAFETY' LIMITS ~~~C1MITING' SAFETY'STSTEM SETTINGS 1.1 (Cont'd) 2.1.A.1 (Cont'd)

D. Cold Shutdown AW = 0 for two recirculatioo loop operation and Whenever the reactor is in . single loop operation the cold shutdown condition less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

.with irradiated fuel in the reactor vessel, the water a. In the event of operation with a level shall not be less than maximum fraction of limiting power 18 in, above the top.of the density (MFLPD) greater than the normal active fuel zone (top fraction of rated power (FRP),

of active fuel is defined in the setting shall be modified as Figure 2.1.1). follows:

~

S 3 (0.66 W + 54% - 0.66 AW) FRP' MFLPD

. a where, FRP = fraction of rated thermal power (2381 MWt)

MFLPD = maximum fraction of limiting power density where the limiting power censity 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.

.For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.

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 3120/125 of scale.

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LIMITS ~~

~~ LIMITING SAFETY SYSTDI SETTINGS 2.1.A.1 (Cont'd)

d. APRM Rod Block Trip Setting The APRM rod block trip setting shall be:

SRB1 0.66 W + 42% .66 AW where:

S =

RB d block setting in percent of rated thermal power (2381 MWt)

W and AW are defined in Specifi-cation 2.1.A.1.a.

In the event of operatien with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:

SRB -( . 6 + -

. AW) FRP MFLPD where, FRP = fraction of rated thermal power (2381 MWe)

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) i

> +12.5.in. on vessel level instruments.

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' Fuel Cladding Integrity A. Fuel Cladding Integrity Limit at Reactor Pressure 2800 psia and Core Flow 210% of Rated 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 observable during reactor operation the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedure used to calculate 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 to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is generically determined in Reference 1 for two recirculation loop operation. This safety limit MCPR is increased by 0.01 for single-loop operation for a-period greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as discussed in Reference 2. .

B. Core Thermal Power Limit (Reactor Pressure < 800 psia and/or Core Flow < 10% of Rated) -

.l 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 head, the core pressure drop at low power and all flows will alwags be -

greater.than 4.56 psi. Analyses show that with a flow of 28 x 10 lbs/hr bun'dle flow, bundle pressure drop is nearly independent of bundle power and has aLvalue of 3.5 psi. Thus,'thegundleflowwith 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 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, a core thermal p'ower limit of 25% for reactor pressures below 800 psi or core flow less than 10% is conservative.

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1.1 Beens (Cont'd)

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 1.lA or 1.13 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 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. ,

References for 1.1 Bases

1. " Generic Reload Fuel Application," NEDE-24011-P (most current approved submittal).

2. ~ " Cooper Nuclear Station Single-Loop Operation," NED0-24258, May,1980.

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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 analyses were'basedlupon, plant operation in accordance with-Reference 3. In addition, 2381 MWe'is i

the licensed maximum power level of CNS, and this represents the maximum steady-state power which shall not knowingly be exceeded.

i 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-ence 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. Trip Settings

' The bases for individual trip settings are discussed in the following paragraphs.

1. Neutron Flux Trip Settings

a. APRM Flux Scram Trip Setting (Run Mode)

The average power range monitoring (APRM) system, 'which is -

calibrated using heat balance data taken during steady-1 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 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 vill be less than that indicated'by the neutron flux.ac.the

~

scram setting. Analyses demonstrate that with a 120% scram 4

trip setting, none of the abnormal operational transients analyzed violate the fuel Safety-Limit-and4t here is a substantial margin from fuel damage. . Therefore, the use

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of flow referenced scram trip.provides even additional margin.

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2.1 Beres

(Cont'd)

An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached.

The APRM scram trip setting vas determined by an analysis of margins required to provide a reasonable range for maneuvering during operation.

Reducing this operating margin would increase the frequency ofispu~rious 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 scram 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 l the MFLPD 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,

b. APRM Flux Scram Trip Setting (Refuel or Start & Hot Standbv 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 mi-;in between the se'epoint and the safety limit, 25 percent of rated. The margin is adequate to accomodate anticipated maneuvers-assocf stor'. with power plant startup. Effects of increasing pressure at zere r low void content are minor, cold water from sources avail-able during startup is not much colder than that already in the system, temperature Mefficitats are small, and control rod patterns'are con-strained to be.tnifort by_ operating procedure backed up by the rod worth minimizer, ind t e rod sequences control system. Worth of indivi-

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dual rods is very 1.ow i t a uniform rod pattern. Thus, of all possible sources of reactiv ty it out, uniform control rod withdrawal is the most probable cause of s quif. cant power rise. Because the flux distribution associated with unii ra rcd withdrawals does not involve high local peaks, and because several r,ds m.tst be moved to change power by a significant percentage of rated pos tr, the -rate of power rise is very slow. Gen-erally, the heat flux it in near equilibrium with the fission rate. In an assumed uniform' rod wichdiswal' approach to the scram level, the rate of power rise is no more L ian 5 percent of rated power per minute, and the'APRM system would be mcee han adequate to assure a scram before the power could exceed the a sfe. v limit. The 15 percent APRM scram remains active until the modt swatch is placed in the RUN position.

This change can occur when recites pressure is greater than Specifi-cation 2.1.A.6.

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2.1---Bases (Cont'd)

c. IRM Flux Scram Trip Setting p ,

-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 by_the SRM and 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 e

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 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 ac'commodate 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-tor shutdown well before any Safety Limit is exceeded.

In order to ensure tha' the IIGi 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 i

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-i 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 l of rated power, thus maintaining MCPR ebove the MCPR fuel cladding integrity safety limit. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors.and continuous

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j withdrawal of control, rods in sequence and provides backup protection 1

1 for the APRM.

j d. APRM Rod Block Trip Setting i

Reactor power level may be varied by moving control rods or by varying i

the recirculation flow rate. The APRM system provides.a control' rod i block which is dependent on recirculation flow race 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 worstLease MCPR which could

. occur during steady-state operation is"at 103%'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 l-

' .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 i

fraction of rated power, thus ' preserving the APRM rod block safety mar -

( gin. As with the scram setting, this may be accomplished by adjusting L the APRM gain.

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.2.1 Basss: (Cont'd)

2. Reactor Water Lcw Level Scram and Isolation Trip Settine (except MSIV)

The set point for low level scram is above the bottom of the separator skirt. This level has b~een used in transient analyses. dealing with coolant inventory decrease. The results report'ed:.in SAR 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 the 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 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 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. D21s 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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(Cont'd)

5. Main Steam Line Isolation Valve Closure on Low Pressure  !

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- The low pressure isolation of the main steam 'linIes ,.(,

> n. Specif1-.

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).

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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).

4. Safety Analysis Report (Section XIV).

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l COOPER NUCLEAR STATION TABLE 3.1.I  ;

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REACTOR PROTECTION SYSTEM INSTRllHENTATION REQUIREMENTS l' ,

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l Minimum Number Action Required

{ Applicability Conditions

.of Operable When Equipment Reactor Protection Mode Switch Position . Trip Level Channels.Per Operability.is System Trip Function Shutdown Startup Refuel Run Setting Trip ~ Systems (1) Not Assured.(1)

Mode Switch in' Shutdown X(7) X X X 1 AI {'

Manual Scram ' X(7) X X X 1 A' l

IRM (17) X(7) X X (5)- < 120/125 of in- 3 A.

High Flux dicated scale

. Inoperative )

X X (5) 3 A APRM (17). X < (0.66U+54%-0.66AW) ' FRP

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2 A or C High Flux (Flow biased) 5

- (14) (19) HFLPD i l ,

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l High. Flux X(7) X(9) ' X(9) ('16) < 15% Rated Power ~2 A.or C l s l

Inoperative X(9) X(9) X' l

(13) ~ 2 A or,C l .Downscale (11) ' X(12) 3 2.5% of indi- .2 A or'C i cated scale ,

t High Reactor Pressure 'X(9) X(10) X < 1045 psig 2 I NBI-PS-55 A.B.C. &~D, A')

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-High Drywell Pressure 'X(9)(8) X(8) .X~ < 2-psig 2 PC-PS-12 A,B,C, & D A or D j -Reactor Low Water Level X X L NBI-LIS-101~A.B.C, & D X 3 + 12.5 in. Indi- 2 A or D cated level ,

Scram Discharge Instrument Volume X X(2) X < 92 inches-High Water Level 3 (18). A

x. CRD-LS-231 A & B s CRD-LS-234 A & B y CRD-LT-231 C & D -

y CRD-LT-234 C & D X

. . . - _ _. .- .~ _ - . - , ~ . . . - .. .

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.11. The APRM downscale[tYip functio EiT 551y ic~tiva'whin~tha reactor moda 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 chadnels* 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

l three level devices per RPS channel.

1

19. AW is the difference between two-loop and single-loop effective drive flow and is

, -used for single recirculation loop _ operation for periods greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4 AW=0 for two recirculation loop operation and single loop operation less than-24 hours.

9 1

)

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~

-\~ XXXXXXXX L

~i

~

~LIMITINGCdEDIildNSFdR05ER'UbN SURVEIf5.ANCE REOUIREMENTS 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 apparatos is not thus, that' adequate coverage is-pro- the only consideration in deter-vided for all ranges of reactor oper- mining a calibration-frequency.

ation. Change in power distribution and loss of chamber sensitivity dictate a calibration every seven days. Cal-ibration on this frequency assures plant operation at or.below thermal limits.

A comparison of Tables 4.1.1 and 4.1.2 indicates that two instrument channels have not been included in the latter

" table. These are: mode switch in shut-

~

-down and manual scram. All of the de-vices or sensors associated with these scram functions are simple.on-off switches and, hence, calibration.during operation is not applicable.

B. The MFLPD is checked once per day 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 significantly and thus a daily check of-the MFLPD is adequate.

, The sensitivity of LPRM detectors de-creases with exposure to neutron flux-at a slow and approximately constant race. This is compensated for in the APRM system by calibrating once a week using a heat balance data and by cali-brating individual LPRM's every six weeks of power operation above 20%

of. rated power.

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 region where the initial heat 4

-42 1 xyxxxx =,-

!. COOPER NUCLEAR STATI'ON l TABLE 3.2.C CONTROL. ROD WITilDRAWAL BLOCK INSTRUMENTATION Minimum Number Of Function Trip Level Setting . Operable Instrument Channels / Trip System (5)

APRM Upscale (Flow Bias) 1 (0.66W + 42% - 0.66 AW) 'FRP ' (2) (13) 2(1)

A*"U Upscale (Startup) l 3 12%. MFLPD 2(1)

APRM Downscale (9) , $ 2.5%

2(1)-

APRM Inoperative (10b) 2(1)

[

i RBM Upscale (Flow Bias) $ 0.66W + (N - 66) (2) I jl RBH Downscale (9) > 2.5% 1 RBM Inoperative (10c) 1

IRM Upscale (8) $ 108/125 of Full Scale 3(1)

,I re q -*

e IRM Downscale (3)(8) 3 2.5% 3(1)  ! ,

IRM Detector Not Full In (8) 3(1) l IRM Inoperative (8) (10a) '3(1)

'SRM Upscale (8) $ 1 x 10 Counts /Second 1(1)(6)

SRM Detector Not Full In (4)(8) (> 100 cps) 1(1)(6)

SRM Inoperative (8) (10a) 1(1)(6)

Flow Bias Comparator <l10% Difference In Recirc. Flows 1 Flow Bias Upscale /Inop. 3 110% Recire. Flow.

1

, ' SRM Downscale (8)(7) . > 3 Counts /Second (11). l(1) (6)' i x

y SDV Water Level High $ 46 inches 1(12) y CRD-231E, 234E n

x

~ ~ ~ ~ ~ ~ ~ ~ ~ '

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 red 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-24011-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 blxk 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 t

-.n. xxxxxxXXt 1

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• NOTES'FOR TABLE 3.2.C (Continued)

, . . _ _ . . . , . .___ -= - .

b. APRM

, (1) ' Mode switch not in operate ,

(2) Less than 11 LPRM inputs (3) Circuit boards not in circuit

c. RBM i

3 (1).~ Mode switch not in operate (2) Circuit boards not in circuit (3) RBM fails to null (4) Less than required. number of LPRM it. puts 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.

c ,

12. With the number of OPERABLE channels less than required by the Minimum 2

Number of Operable Instrt. ment Channels / Trip System requirements, place j the inoperable channel in the tripped condition within one hour.

i 13. AW is the difference between two-loop and single-loop effective drive I

flow and is used for single recirculation loop operation for periods greater.than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. AW-0 for two recirculation loop operation and single loop operation less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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.D. - 'ex tivity 5 v slies

.the crit'c=1 rod 't l';prr t Iac -ill At a specific st u dy state icse condi- be cmpared to tSe exp . t--d c nfigura- . .

tion of the reactor actual ccetrol red tions at set <ct d vp-re ' g e aidit' i. e.

Inventory will' be periodically cc,m- TFe se et 1pa cIn,r s v f 11 ',e or s d u h.u a -

cred to a netcalized cc;puter pre- data for reactivity c..iwcing du. ing diction of the inventery. If the- 5"hsequent pever cpa r r f 3n .hrcush-differetc, N tween observed and pre- .

out the fuel cycle, e.t trecific ,v-r dict:d rod i.xentery reaches' the garnt'ng cc.dit! m , de critical red er;uivalent of 1% !k reactit ity, the configuration will Le cs .pn ed 'to the -

c nfiguration expected b wed i. pen ap-

- reactor will be shut dcvn unti'l the cause has been determined and ccrrec- prepriately corrected p u t data. His "tive actions have been taken as co: pari >on will be -ade at Ierst ev'i:y appropricte. f t'll rever nonth.

.E. Festrictions F. h eirculation runps

~

if Specifications-3.3.A through D .

1. *it ch two recirculation pu: ps in aa ve car.not be =et, an orderly _

operation and with core therral pcwer 4 r .m shall he initiated and the. greater than'the limit specified in r- < er.aS a11 Se 'n *,e Shu t dc un- Figure 3.3.1-and total core flew less

'c: 11 - . 24 * .r s , than !.5% of rated, estaM ish baseline

, MFX and T. IMP n.autren f'ux noise

1. F- T , '

• r ,4 * -

L.S levels within 2 hcurs, previded that l 1. 'A . - ce 3 t ! .n y .; .all not be '5:meline values have not-been pre-s t .- id .*l'.e the Yrvtor is in v'cusly established since the last ut n.n'. ..tcculm itn v and ructor core refueling.

pc a i s t. . e v . ; * * :n % of rated .,

u. 3) .,rior c to oparntion with 'one rc-the cal *c.er. circulation purp not in opera-

2. 'Jith two m.'irculation pu .ps in crera- -tien and core thernal pcwer tion r d with core thermal pcwer greater than the limit specified gr e r

• ban the Ifmit specif"ed in in Figure.3.3.1 establish Fi nre 3.3.1 and total ccre flow less. baseline APM: and LPFd!* rentren tbn 45% of rated, the APM and h?Rn* flux noise levels, prov'dod thit s-uu.sa f'ux noise levels shall- be baseline values have not been dotcrrfnad within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and: previcusly established sL:ce t' e it.st core refueling. Baseline a) if LLe Ai 01 and LPMi* neutron values shall be esta*dist.ed with.

rin ,se 1crets are less than one recirculation pir:p not in or -

• al to-three tines their ~

operatten and ccre thertal pever est thi!shed I;aseline' icvels, con-

. lees ti.an or equal to the -liait tinue to d.iternine the noise speciffed 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) Frior to-cperat' ion with one re-the completion of a core therreal -circulation puep not in opera-power increase of at least 5% of

~

tion and cere flow greater than rated core thereal power while 45% of rated, establich baseline operating in this region of the core plate SP noise levels with pcuer/flev cap, or core flow less than or equal to 45% of rated, provided that

• Detector levels A and 'C of one 1.PRM baseline values have not been str'ng per core octant plus detector levels previcusly established with 6ac A and C of one LFRM string in the center of recirculation purp not in the core shall be cenitored. cperation since the last cote-refueling.

3

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• : at rd drain c2't s eb il be-t ive aet'en wd v.wtore tha~roise .c

.ycid and verified cpen at

. vels .te .-f tt :n the reluired.1 Git s i.w t'ence. v.*ry 31'J..ys lad,

> within 2 hcurs by 'increwing cure .

prjer to i cacterr s tyr t-up, i fIcv, and/or ty initiating-an orderly i ryduction ef core ther s1 pcver by T. The $37 vent ord drain valves

+

f r9erting c< ntrol rods. - shall he verified to cicse itMn 30 m:cccds af ter re ceipt.

3. The reactor rey be staited and epc-. of a signal for ecatrol rod rated, or operation r:ay centinue with

- scram once.per reS cline cve.l._e..

~

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one recirculation Icop not in opera-tion prcvided that;

3. SDV vent and drain va'.ve opera- ,

! bii t ty - shall be ve rified folicw-i a. with cne recirculatica pu .p not -ini; h y r.aintermee or .ndifica-in cperation and core thermal tion to ar.y portion (electrical l pcwer greater t' an the limit or recFanical) of the SDV shich j specified in Figure 3.3.1, core may affect the operation of the ficy nst be gr?ater than or vent and drain vavles.

! e. ual t o ' 45% of rat ?d , . rid 4

(i) the Se ceilla.ic: F.quireca ts of

4. 3. F. .a have ht ' h e e n c y t 's f f ed , .

f r edD tely hitinte. an .a to  !

r educ e c c,re t he rna l ;. ewe r t o ' u s than or er;ual tti the I'mit uprei-

, fied in F#sure 3.3.1 wid.!n 4 ,

hours, or (ii) th.e Surveillance Requirecents of '

4 4.3.F.2.a have been satisfied,

(

centinue to deterrine the APEM

, and L7;M neutren flux levels at icast 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 nicutes after the 4- cc:.pletion of a core therreal i pcuer inerense of at least 5% of

, - rated core tharr.a1 pcwer while

~

j- operating in this region of the

pwer/ flew rap. . If the APFM

, and/or I.M* neutron flux noi::e levels are greater than three tires their established =baceline .

values, itmediately initiate corrective action and' restore the noise levels to within the

[ required limits within 2 hcurs by

• Detector levels A and C of one LPRM

l. ' string per core octant 'plus detector levels

, A and C'of one h?RM string in the ccnter of the core shall be u onitcred..

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it, ing ccre r:c and/<>r .it t .t inu :.n 3 r e rly r.~'uc t ien o f cer 2 t'.erral re.e r by i: ." r tir.g e -trel rod: .

b. ';f th ene recirculaticn ; ep not in i,eraticn and core flow greater thcn -

45% of rated, and (i) t!.c F':: eeilicnce Eequir cr erts of 6.3.F.2.b have not been satisfied,

i. .:df O'ly initiite action to raduce u;re flow to less than or equal to 45" of rated v'ithin 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.2.5 have been satisfied, centinue to determine cere plate AP noise at least once per 8 hcurs and also within 30 minutes after the cc:;1etion of r. core the:=al pcwer increase nf .it l e. ;st .'. of rater' hernal pcwer. If the cere p!.te AP neise Iccel is grcater than 1.0 psi and 2 tines its esta-blished baseline value, inme-diately initiate corrective action and restore the e.o'se 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 decrces-ing core flew and/or iritiating an orderly reductien of core thernal pcwer by inserting centrol reds,

c. The id:e ic p is isol.ited electrically by dis.:c nne.:t ing the b ecake r to the t . cir celat ien , t;.r.p rot or sencra tor

("./G) S e t s' r ".v e ctor prior to ut:irt-e p , o r i f .' , :M e d i : r ' ., $ r e a c t o r e

p.-ratien, within 24 Peurs.

d. 'ihe recirculation yetem controls will be placed in the rauual flow control

r. ode.

-9Sc-

- - - . .. - - , -- - - --- - - . ~ . . . .-. . . ..

~ ~ ' ' ~ ~ ~~

i 3.'3 and 4.3 BASES: lCont'd)~

i

~ 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.

I D. Reactivity Anomalies During each fuel cycle excess operative reactivity varies as fuel depletes and as i

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, ancmalous behavior in the excess reactivity may be detected i 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 i sensitive and directly interpretable data relative to core reactivity. Furthermore, using power operating base conditions permits frequent reactivity comparisons.

Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% Ak.

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 j design conditions of the reactor system.

1 F. Recirculation Pumps a

Until analyses are submitted for review and approval by the NRC which prove that '

, 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 t

reactor is in natural circulation above 1" of rated thermal power. ' Specifications 3.3.F.2, 3, and 4 are based upon providing assurance that neutron flux limit cycle o oscillations, which have a small probability of occurring in the high power / low I

flow corner of the operating domain, are detected and suppressed. BWR cores j 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 l considered in the thermal / mechanical design of GE BWR fuel, occur in a stable mode, and are found to be of negligible consequence. However, under certain l l high power / low flow conditions that could occur during a recirculation pump

' trip and subsequent Single Loop Operation (SLO) where reverse flow occurs in inactive jet pumps, a hydraulic / reactor kinetic feedback mechanism can be j enhanced such that sustained limit cycle oscillations of flow noise with peak to peak levels several times normal values are exhibited. Although large 4 margins to safety limits are maintained when these limit cycle oscillations occur, they are to be monitored for, and suppressed when flux noise exceeds j? 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 2-probability of limit cycle oscillations occurring is negligible. The thermal i power, core flow, and neutron flux noise level limitations are prescribed in

,- accordance with Reference 3.

4 e

1 a

~

-II XXXXXXXX

+ " * ~ - - --- *r-+--t'tve-~-r----ew y re ---+-,w*ww.W w 1=-ww ==y+e-r e =m---- ' y - - , v -w g we + m w m-e y er ,re--v- +==v e ea wy-=-g--w , y c - , v-e emee teem'*- e--

~~

~

3.3 and 4.3 BASES:

~~~

(Cont'd)' ~ ~ ~

G. Scram Discharge Volume To ensure the Scram Discharge Volume (SDV) does not fill with water, the vent and drain valves shall be verified open at least once every 31 days. This is to preclude establishing a water inventory, which if sufficiently large, could result in slow scram ti=es 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.

REFERENCES

1. Licensing Topical Report GE-BWR Generic Reload Fuel Application, NEDE-24011-P, (most current approved submittal _).

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.

a e

4

- -104_- xxxxxxxx '

, LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUZREMENTS 3.6.E Jet Pumps 4.6.E. Jet Pumps

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 be operable. If it is determined run modes, jet pump operability shall that a jet pump is inoperable, or be checked daily by verifying that the ff 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 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 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.

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 jet pump differential pressures by more than 10%.

F. Jet Pump Flow Mismatch F. Jet Pump Flow Mismatch

1. Deleted. 1. Deleted.

2. Following one-pump operation, the dis-charge valve of the low speed pump may not be opened unless the speed of ,

the faster pump is equal to or less

'than 50% of its rated speed.

G. Inservice Inspection G. Inservice Inspection To be considered operable, com- Inservice inspection shall be per-ponents shall satisfy the require- formed in accordance with the ments contained in Section XI of requirements for ASME Code Class 1, the ASME Boiler and Pressure Vessel 2, and 3 components contained in Code and applicable Addenda for Section XI of the ASME Boiler and continued service of ASME Code Pressure Vessel Code and applicable Class 1, 2, and 3 components except Addenda as required by 13 CFR 50, where relief has been granted by the Section 50.55a(g), excep: where Commission pursuant to 10 CFR 50, relief has been granted by the Section 50.55a(g)(6)(1). Commission pursuant to 10.CFR 50, Section 50.55a(g)(6)(1).

3 l

t 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 Requiring the discharge valve of the lower speed loop to remain closed until 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.

G. Inservice Inspection I 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 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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EIMITING CONDITIONS FOR' OPERATION SURVEILMICE 'REOUIREMENTS

~3.11'TUEL RODS -

4.11 FUEL RODS Applicability Aeolicability The Limiting Conditions for Operation The Surveillance Requirements apply associated with the fuel rods apply t to the parameters which. monitor the those parameters which monitor the fuel fuel rod operating conditions.

rod operating conditions.

Objective Objective The Objective of the Limiting Condi- The Objective of the Surveillance tions for Operation is to assure the performance of the fuel rods. Requirements is to specify the type and frequency of surveillance to be Syecifications applied to the fuel rods.

A. Average Planar Linear Heat Specifications Generation Rate (APLHGR)

During steady state power opera- A. Average Planar Linear Heat tion, the'APLHCR for each type of Generation Rate (APLHCR) fuel as a function of average planar exposure shall not exceed The APLHGR for each type of fuel the limiting value shown in Figure as a fcnction of average planar 3.11-1 for two recirculation loop. exposure shall be determined For single-loop operation greater daily during reactor operation than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the values in these at 125% rated thermal power.

curves 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 and corres-ponding action shall continue until the prescribed limits are again being met. B. Linear Heat Generation Rate (LHGR)

B. Linear Heat Generation Rate (LHCR) 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 1 25%

rate (LHCR) of any rod in any fuel rated thermal power.

assembly at any axial location shall not exceed the maximum allow-able LHCR as calculated by the following equation:

max $ LHGR ~

(

d max LHGRd

= Design LHCR = G KW/ft.

(AP/P) = Maximum power spiking penalty = N

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LIMITING CO fTIONS FOR OPERATION SURVEILLA[CEREOUIREMENTS LT = Total core length - 12 feet i

3 L = Axial position above bottom i of core G = 18.5 kW/ft for 7x7 fuel - ' ur :v, bundles ~ ' "

0'

~" "

= 13.4 kW/ft for 8x8 fuel j 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 2

prescribed limits. Surveillance and corresponding action shall .

continue until the prescribed lim-

its are again being met.

i C. Minimum Critical Power Ratio (MCPR) C. Minimum Critical Power Ratio (MCPR)

During steady state power opera-tion the MCPR for each type of fuel MCPR shall be determined daily at rated power and flow shall not be during reactor power operation lower than the limiting value shown at > 25% rated thermal power in Figure 3.11-2 for two recircula- and following any change in

j. power level or distribution that tion loop operation. If, at any time during steady stata oper- would cause operation with a

{ limiting control rod pattern as ation it is determined by normal

- surveillance that the limiting described in the bases for Spec-ification 3.3.B.5.

7 value for MCPR is being exceeded, action shall then be initiated within 15 minutes to restore oper-ation to within the prescribed i

limits. If the steady state MCPR is not returned to within the pre-

. scribed limits within two (2) j hours, the reactor shall be brought to the Cold Shutdown con-

dicion within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Su rveil-l lance and corresponding action shall continue until the pre-scribed limits are again being met.

For core flows other than rated the MCPR shall be the operating j limit at rated flow times K g, where Kg is as shown in Figure 3.11-3 i

! For one recirculation loop oper-f ation greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the MCPR limits at rated flow are

, 0.01 higher than the comparable two-loop values.

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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. _ ., , j : . .g. J ,

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 functioc of core height shall be checked daily during reactor operation i

at > 25% power to determine if fuel burnup, or control rod movement has caused changes in power distribution. For LHGR 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 8x8 fuels.

d i

j i

i

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__ _ -. . . _ _ . _ ~ . _ _ ~_._ _ _ _ _ _ _ _ _ . - _ ,

1 3.11 Bassa: (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.the established fuel cladding integrity Safety Limit and an analysis of-abnormal" opera' ~ t ional transients (Reference 2). For any abnormal operating transient analysis evaluation with the initial cohdition 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 the largest reduction in critical power ratio (CPR). The models used in the transient analyses are discussed in Reference 1.

The purpose of the gK factor is to define operating limits at other than 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 Kg factor provides the required thermal margin to protect against a flow 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.

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 g factors assure that the Safety Limit MCPR will not be vio-laced for ene same postulated transient event.

1

)

E 4

t t

-214aw. xxxxxxXXi

,_n. e -- --.--.-n--.--,. ,n,,, ---,,---.n.,y - , , , , , , .._,.,,,..n,.7, . ~, sm-,_n...,---s, n,g, w,,, --, - , - - , . , , , - - . - - - . , . . , . -

3.11 Basas: (Cont'd) 1 The Kg factor curves shown in Figure 3.11-3 were developed generically i which are applicable to all BWR/2, BWR/3, and BWR/4 reactors. The K factorswerederivedusingtheflowcontrollinecorrespondingtorafed thermal power at rated core flow as described in Reference,1.. , ,,

The K factors shown in Figure 3.11-3, are conservative for Cooper opera-

~

I tionbecausetheoperatinglimitMCPR'saregreaterthantheoriginal 1.20 operatirg limit MCPR used for the generic derivation of Kg .

References for Bases 3.11

1. Licensing Topical Report, General Electric Boiling Water Reactor, Generic Reload Fuel Application. (NEDE-24011-P), (most current approved submittal).

2. " Supplemental Reload Licensing Submittal for Cooper Nuclear Station

j. Unit 1," (applicable reload document).

J J

t l

1 1

4 b

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4.11 Bases:

_l A&B. Average and Local LHGR The LHGR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution. . .,,

Since changes due to burnup are slow, and only a few control rods. ,,

are moved daily, a daily check of 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 MCPR evaluation was made at 25% thermal power level with minimum recirculation pump speed.

The MCPR margin was thus demonstrated such that subsequent MCPR evaluation below this power level was shown to be unnecessary. The daily requirement for calculating MCPR above 25% 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.

t 1

t

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!