ML20127H217

Proposed Tech Specs to Authorize Use of Partial Drilled Model of Fuel Type for Plant Future Use

ML20127H217
Person / Time
Site:	Monticello
Issue date:	08/10/1978
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20127H193	List: ML20127H188 ML20127H200 ML20127H217 ML20127H244
References
NUDOCS 9211180445
Download: ML20127H217 (24)

Text

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6 EXHIBIT.B SUPPLDIENT NO. 1 TO LICENSE AMENDMENT REQUEST DATED MARCH 21, 1978 This exhibit consists of the following pages revised to incorpo rate all of the proposed Technical Specification changes:

vil viii 1x 2

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9 10 ' (DELETED) 11 (DELETED) 12 (DELETED) 13 14 19 20 21 -

189B 189C 189D 189E 189F 189G 189H 1891 (DELETED)

~ 189J (DELETED) 189K (DELETED) 189L (DELETED) 190 9211180445 790810 PDR P

ADOCK 05000263 PDR _

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'D LIST OF FIGURES Figure No Page No ^

l 4.1.1 'M" Factor - Graphical Aid in the Selection of an Adequate Interval Between Tests 46  ;

4.2.1 System Unavailability 74 l 3.4.1 Sodium Pentaborate Solution Volume - Concentration Requirements 92 3.4.2 Sodium Pentaborate Solution Temperature Requirements 93 3.6.1 Change in Charpy V Transition Temperature versus Neutron Exposure 122 ,

3.6.2 Minimum Temperature versus Pressure for Pressure Tests 122A 3.6 3 Minimum Temperature versus Pressure for Mechanical Heatup or Cooldown Following Nuclear Shutdown 122B j 3.6.4 Minimum Temperature versus Pressure for Core Operation 122C 4.6.1 Deleted -

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4.6.2 . Chloride Stress Corrosion Test Results @ 500 F 123  ;

4.8.1 . Sampling locations - Radiation Environmental Monitoring Program 173 .

r 3.11.1. Deleted 4

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LIST OF FIGURES

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" Figure No Page No l 3.11.2 Deleted 3 11.3 Kg Factor versus Percent of Rated Core Flow *89M 6.1.1 NSP Corporate Organizational Relationship to On-Site Operating Organization 193 6.1.2 Functional Organization for on-Site Operating Group 194 i

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LIST OF TABLES Table No Pare No__

3.1.1 Reactor Protection System (Sc ram) Instrument Requirements 30 '

4.1.1 Scram Instrument Functional Tests - Minimum Functional Test Frequencies for Safety Instrumentation and Control Circuits 34 m

4.1.2 Scram Instrument Calibration - Minimum Calibration Frequencies for Reactor-Protection Instrument Channels 36 3.2.1 Ins trumentation that Initiates Primary Containment Isolation Functions 50

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3.2.2 Ins trumentat ion that Initiates Emergency Core Cooling Systems 53 3.2.3 ' Instrumentation that Initiates Rod Block 57 3.2.4 Ins trumentation that Initiates Reactor Building Ventilation Isolation and Standby Cas Treatment System Initiation 60 3 2.5 Trip Functions and Deviations 69 4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation Ins trumentat ion 61 3.6 1 Safety Related flydraulic Snubbers ~

121B 4.6.1 In-Service Inspection Requirements for Monticello 124 3.7.1 Primary Containment Isolation 153 4.8.1 Sample Collection and Analysis Monticello Nuclear Plant -Environmental 169 3.11.1 Maximum Average Planar Linear IIeat Ceneration Rate

, 189E 6.1.1 Minimum Shif t Crew Composition 194A 6.5.1 Protection Factors for Respirators 206 ix REV  !

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D. Immediate - Immediate means that the required action will be initiated as soon as practicable -

considering the safe operation of the unit and the importance of the required action. -

E. Instrument Functional Test - An ins t ru ment functional test means the injection of a simulated signal into the primary sensor to verify the proper instrument channel response, alarm, and/or initiating action.

F. Instrument Calibration - An instrument calibration means the adjustment of an instrument signal output so t ha t it corresponds, within acceptable range, accuracy, and response time to a known value (s) of the parameter which the Instrument monitors. Calibration shall encompass ~'

the entire instrument including actuation, alarm or trip. Response time is not part of the routine instrument calibration but will be checked once per cycle.

G. Ilmiting Condit ions for Operation (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the-facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safety controlled.

H. . Limiting Safety System Setting (LSSS) - The limiting safety system settings are settings on instrumentation which initiate the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represents margin with normal l operation lying below these settings. The margin has been established so that with proper operation of the instrumentation, the safety limits will never be exceeded.

I.

Maximum Fraction of Limiting Power Density (MFLPD) - The maximum f raction of limiting power density is the highest value in the core of the ratio of the existing to the design linear heat generation rate.

I J. Minimum Critical Power Ratio (MCPR) - The minimum critical power ratio is the value of critical power ratio associated with the most limiting assembly in the reactor core. Critical power

' ratio (CPR) is the ratio of that power in a fuel assembly which is calculated by the CEXL correlation to cause some point in the assembly to experience boiling transition to the actual assembly op,erating power.

, l K. Mode - The reactor mode is that which is established by the mode-selector switch.

l L.

Operable - A sys tem or component shall be considered operable when it. is capable of performing its intended function in its required manner.

! M. Operating - Operating means that a system or component is performing its required functions in its required manner.

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l N. Operating Cycle - Interval between the end of one refueling outage and the end of the next ,  ;

subsequent refueling outage.

O. Power Operation - Power Operation is any operation with the mode switch in the " Start-Up" or "Run" position with the reactor ciritical and above 1% rated thermal power. ,

P.- Primary Containment Integrity - Primary Containment Integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied.

1. All nunual containment isolation valves on lines connecting to the reactor coolant system -~

or containment which are at required to be open during accident conditions are closed.

2. At least one door in the airlock is closed and sealed.

3. All automatic containment isolaiton valves are operable or are deactivated in the closed position or at- least one valve in each line having an in-operable valve is closed

4. All blind flanges and manways are closed.

Q Protective Instrumentation Logic Definitions

1. Instrument Channel - An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip system, a single trip signal related to the plant parameter monitored by that ins trument channel.

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2. Trip System - A trip system means an arrangement of instrument channel trip signals and ,

auxiliary equipment required to initiate a protection action. A trip system may require one or more instrument channel trip signals related to one or more plant parameters to  :

initiate trip system action. Initiation of the protective function may require tripping of a single trip system Ic.g., HPCI system isolation, off gas system isolation, reactor building isolation and standby gas treatment initiation, and rod block), or the coincident

, , tripping of two trip systems (e.g. , initiation of sc ram, reactor isolation, and primary containment isola tion) .

3. Protective Action - An action initiated by the protection system when a limit is exceeded.

! A protective action can be at channel or systen level.

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a 2 0 SAFETY LIMITS LIMITING SAFETY SYSTDi SETTINGS j 2.1 FUEL CLADDING. INTEGRITY 2.3 FUEL CIADDING INTEGRITY Applicability: Applicability:

Appli es to the interrelated variables Applies to trip settings of the instruments and associated with fuel thermal behavior. devices which are provided to prevent the reactor system safety limits from being exceeded.

Objective: Objective:

To establish limits below which the To define the level of the process variables integrity of the fuel cladding is preserved. at which automatic protective action is initiated to prevent the safety limits from being exceeded.

Specification: Specification:

A. Core Thermal Powe.r' Limit (Reactor The Limiting safety system settings shall be as Pressure > 800 Psia and Core Flow is specified below:

> 10% of Rated) .

A. Neutron Flux Scram Mien the reactor pressure is > 800 Psia and core flow is > 10% of rated, the .l. APRM - The APRM flux scram trip setting existence of a minimum critical power' shall be:

ra tio (MCPR) less than 1.07 for 8x8 fuel and less than 1.07 for' 8x8R fuel shall S5 0.65 W + 55%

constitute violation of the fuel cladding where,-

integrity safety limit S= Setting of percent of rated thermal power, rated power beim 1670 tur U= recirculation drive flow in percent l 2.i/2.3

2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS B. Core Thermal Power Limit (Reactor except in the event of operation with a Pressure s 800 Psia or Core maximum f raction of limiting power density Flow $; 10% of Rated) for any fuel type in the core greater than the fraction of rated power, when the setting shall be modified as follows:

When the reactor pressure is ;E 800 psia or core flow is :$ 10% of rated, the cor thermal power shall not exceed 257. of S < (0.65 W + 55%) gFRPiEED rated thermal power.

where, C. Power Transients FRP = fraction of rated thermal power, rated power being 1670 MWt To insure that the safety limit established MFLPD = maximum fraction of limiting in Specification 2.1.A is not exceeded, each power density for any fuel type required scram shall be initiated by its in the core.

primary source signal as indicated by the plant process computer _-

2. IRM - Flux Scran setting shall be $1 20% of rated neutron flux B. APRM Rod Block - The APRM rod block setting shall be:

S 0 0.65 W + 43%

where, S= Setting of percent of rated thermal power, rated power being 1670 MWT

. U= recirculation drive flow in percent 4

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2.0 SAFETY LIMITS LINITING SAFETY SYSTEM SETTINGS except in the event of operation with a

.D. Reactor Water Level (Shutdown Condition) enximum f raction of limiting power density for any fuel type in the core greater than Whenever the reactor is in the shutdown the fraction of rated power, when the setting condition with irradiated fuel in the shall be modified as follows:

reactor vessel, the water level shall not 2

be less than that corresponding'to 12 FRP

' inches above the top of the active fuel S $ (0.65 U + -

when it is seated in the core. This where, 43%)MFLPD level shall be continuously monitored FRP = f raction of rated thermal power, whenever the recirculation pumps are not rated power being 1670 MWt ope ra t ing . MFLPD = maximum f raction of limiting power density for any fuel type in the core.

C. Reactor Iow Uater Level Scr4m setting shall be 2 -

10'6"above the top of the activ e fuel.

D. Reactor low Inw Water level ELCS initiation shall be 2 6'6" 1 6' 10" above the top of the active fuel.

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2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS r's E. Turbine Control Valve Fast Closure Scram shall initiate upon loss of pressure at the acceleration relay with turbine first stage pressure > 30%.

F. Turbine Stop Valve Scram shall be 5 10% valve closure from full open with turbine first stage pressure 2: 30%.

G. Main Steamline isolation Valve Closure Scram shall be 5 10% valve closure from full open.

II . Main Steamline Pressure initiation of main steam-line isolation valve closure shall be 2: 825 psig.

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f Bases:

2.1 The fuel cladding integrity limit is set such that no calculated f uel damage would occur as a result of ' an abnormal operational transient. Because fuel damage is not directly observable, a step-back l approach is used to establish a Safety Limit such that the MCPR is no less than 1.07. This limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate radioactive materials trom the environs. The integrity of this cladding barrier is related to its relative freedom from g~. -

perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously

, measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the protection system safety settings.

While fission product migration from cladding perforation is just as measurable as tha t from use related cracking, the thermally caused cladding perforations signal a threshold, beyond which still greater- thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with margin to the conditions which would produce onset of trans ition boiling. (MCPR of 1.0). These conditions represent a_significant departure from the condition intended by design for planned operation. The concept of MCPR, as used in the GETAB/GEXL critical power analysis, is discussed in Reference 1.

l A. ' Core Thermal Power Limit (Reactor Pressere > 800 psia and Core Flow > 10% of Rated.) Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad f ailure. However, the existence of critical power, i or boiling t ransition, is not' a directly observable parameter in an operating reactor. Therefore, I the margin to boiling transition is calculated from plant operating paramet,ers such as core

' power, core flow, feedwater temperature, and core power distribution. The margin for each fuel assembly is characterized by the critical power ratio (CPR) which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power. The ,

minimum value of this ratio for any bundle in the core is the minimum critical power ratio I (MCPR). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variables. The Safety Limit (T.S.2.1.A) has suf ficient conservatism to assure that in the event of an abnormal operational transient initiated from the Operating MCPR Limit (T.S.3.11.C) more than 99.9% of the fuel rods in the core are expected to avoid boiling transition. The margin between MCPR of 1.0 (onset of transition boiling) and the Safety Limit 13 2.1 Bases REV

Bases Continued:

t is derived from a detailed ' statistical analysis considering all of the uncertainties in monitoring the core. operating state ihcluding uncertainty in the boiling transition correlation as described in Reference 1. The uncertainties employed in deriving the Safety Limit are provided at the

'beginning of each fuel cycle.

Because' the boiling transition correlation is based on a large quantity of full scale data, there is a very high confindence that operation of a fuel assembly at the MCPR Safety Limit would :not produce boiling . transition. Thus, although it is not required to establish the Safety.

Limit, additional margin exists between the Safety Limit and ' the actual occurrence of loss of cladding integrity.

- However, if boiling transition were to occur, clad perforation would not be expected. Cladding temperatures would increasef to approximatley 1100"F which is below the perforation temperature .

of the - cladding material. This has been verified by tests in the General Electric Test Reactor

. (CETR) where fuel siedlar in design to Monticello. operated above the boiling transition for:a

. significant period of time . (30 minutes) without clad perforation.

L If reactor' pressure should ever exceed 1400 psia during normal power-operating (the. limit of appalicability of the boiling transition correlation) it'would be assumed that the fuel cladding integrity Safety Limit ' has been violated.

In addition to the MCPR Safety . Limit, operation is constrained to a maximum design linear heat generation ' rate for any fuel type in the core. . -

.B.. Core Thermal' Power Ilmit (Reactor Pressurc 5 800 psia or Core Flow 5 10% of Rated) At pressure below 800 psia, the core elevation. pressure drop (0 power, O flow) is greater than 4.56 psi. At low powers ~ and all core flows, this pressure differential is maintained in the bypass region of the core.

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- t Bases Continued:

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 analyzed violate the fuel Safety Limit and there is a substantial margin from fuel damage. Therefore, the use of flow ref erenced scram trip provides even additional margin.

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 during operation. Reducing this operating margin would increase the frequency of spurious scrams which have an adverse ef fect on reactor safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams. Therefore, it is intended to ultimately replace (with prior NRC approval) the automatic flow referenced scram with a fixed 120 percent scram setting.

The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of maximum f raction of limiting power density and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.3.A.1, when the maximum fraction of limiting power density is greater than the fraction of rated power. If the APRM scram setting should require a change due to an abnormal peaking condition, it will be done by increasing the APRM gain and thus reducing the slope. and intercept point of the flow referenced scram curve by the reciprocal of the APRM gain change. Analyses of the limiting transients show that no scram adjustment is required to assure that the MCPR Safety LLeit-(T.S.2.1.A) is not exceeded when the -

transient is initiated from the Operating MCPR Limit (T.S.3.ll.C).

For operation in the startup mode while the reactor is at low pressure, the IRM scram setting of 20%

of ' rated power provides adequate thermal msrgin between the setpoint and the safety limit, 25% of rated. The margin is adequate to accommodate anticipated maneuvers associated with power plant s ta rtup. Ef fects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not .much colder than that already in the system, temperature coef ficients are small, and control rod patterns are constrained to be uniform by operating procedures 19 2.3 BASES REV

Bases Continued:-

backed up by the rod worth minimizer. Worth of individual rods is very low in a uniform rod  ;

pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the. rate of power rise is ve ry ~

l slow. ' Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5% of rated power per. minute, and the IRM system would be more' than . adequate to assure a scram before the power could exceed the safety limit. The IRM scram remains' active until the mode switch is placed in the run position. This switch occurs when reactor pressure is greater than 850 psig.

The = analysis to support operation at various power and flow relationships has considered opera-

-tion with either one or two recirculation pumps. During steady-state operation with one recircula-tion pump operating the equalizer line shall be open. Analysis of transients' from this operating condition are less severe than the same transients from the two pump operation.

The operator will set the APRM neutron flux trip setting no greater than that stated in Specifica-tion 2.3.A.1. However, the actual setpoint can be as much as 3% greater than that stated in Specification 2 3.A.1 for recirculation driving . flows less than 50% of design and 2% greater than that shown for recirculation driving flows greater than 50% of design due to the deviations discussed on page 18.

B. APRM Control Rod Block Trips Reactor power'1evel may be varied by moving control rods or by 1 varying .the recirculation flow rate. The APRM system provides.a control rod block to prevent  ;

rod withdrawal beyond a given point at. constant recirculation flow rate, and thus to protect  ;

against' the condition of a' MCPR less than the Safety Lhmit (T.S.2.1.A). 'This rod block trip-setting, which 'is automatically varied with recirculation loop flow rate, prevents an increase in the- reactor power level to excessive . values due to control rod withdrawal. The flow variable trip setting provides substantial margin' from fuel- damage, asuming a steady-state operation at the. trip setting, over the . entire recirculation flow range. The margin to the Safety Limit N

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Bases Continued

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increases as the flow decreases for the specified trip setting versus flow relationship; the re fo re,

, the wors t 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 distribution in the core is established by the maximum f ract ion of limiting power density exceeds the fraction of rated >

thermal reactor power, the rod block setting is adjusted in accordance with the formula in Specifica tion 2.3.B. 'If the APRM rod block setting should require a change due to an abnormal #"*

peaking condition, it will be done by increasing the APRM gain and thus reducing the slope and ,

intercept point of the flow referenced rod block curve by the reciprocal of the APRM gain change.

The operator will set the APRM rod black trip settings no greater than that stated in Specification 2 3.B. However, the actual setpoint can be as much as 3% greater than that stated in Specification 2 3.B for recirculation driving flows less than 50% of design and 2% greater than that shown for recirculation driving flows greater than 50% of design due to the deviations discussed on Page 18.

C.- Reactor Iow Water Level Scram The reactor low water level scram is set at a point which will assure that the water level used in the bases for the safety limit is maintained.

, ' The _ operator will set the low wa'ter level trip setting no lower than 10'6" above the top of the

active fuel. However, the actual setpoint can be as much as 6 inches lower due to. the deviations discussed on page 18. -

D. Reactor Low Low Water Level ECCS Initiation Trip Point The emergency core cooling subsystems

, are designed to provide sufficient cooling to the core to dissipate the energy associated with the loss of coolant accident and to limit' fuel clad temperature to well below the clad melting temperature to assure. that core geometry remains intact and to limit any clad metal-water reaction to less than 1%.

The design of the ECCS components to meet the above criterion was dependent on three previously set parameters: the maximum break size, the low water. level scram setpoint, and the ECCS initiation i se t po int . ' To lower the setpoint for initiation of the ECCS could prevent the ECCS components from  ;

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t 3 0 LIMITING CONDITIONS EUR OPERATION 4.0 SUIiVEILIANCE REQUIRDIENTS 3.11 REACTOR FUEL ASSDiBLIES 4.11 REACTOR FUEL ASSEMBLIES Applicability Applicability i

The Limiting Conditions for Operation associated The Surveillance Requirements apply to with the fuel rods apply to those parameters the parameters which monitor the fuel which monitor the fuel rod operating conditions. rod operating conditions.

Objective j Objective The objective of the Limiting Conditions for Opera- The objective of the Surveillance Require-tion is to assure the performance of the fuel rods 4 ments is to specify the type and frequency of surveillance to be applied to the fuel rods.

Specifications Specifications A. Average Picnar Linear Heat Generatior, Rate (APLHGR) A. Average Planar Linear Heat Genera-s tion Rate (APLHGR)

During power operation, the APTT,GR for each type of fuel as a function of average planar The APLRGR for each type of fuel as exposure shall not exceed t'.e limiting value a function of average planar exposure -

given in Table 3.11.1 based on a straight shall be determined daily during line interpolation between data points. When reactor operation at 2 25% rated core flow is less than 90% of rated core flow, the rmal power.

the APLHGR shall not exceed 95% of the limit-l ing value given in Table 3.11.1. When core flow is less than 70% of rated core flow,

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the APLHGR shall not exceed 90% of the limit-ing value given in Table 3.11.1. If any l

tire during operation it is determined that the limit for APLHCR is being ex-ceeded, action shall be initiated within 15 189B 3.11/4.11 REV

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3.0 LIMITING -CONDITIONS POR OPERATION 4.0 SURVEILLANCE REQUIREMENTS minutes to restore operation to within the prescribed limits. Surveillance and corres-ponding action shall continue until reactor operation is within the prescribed limits.

If the APLJ1GR is not returned to within the prescribed limits within two (2) hours, the #'A 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 />.

B. Linear Heat Generation Rate ( LHGR) B. Linear Heat Generation Rate ( LHCR)

During power operation, the LHGR as a function The LHCR as a function of core height d

of core height shall be limited to: shall be checked daily during reactor operation at 2 25% of rated thermal LHGR 5 3.4(1 .022 X/L) . power.

where, X = Elevation from the bottom of the core L = Fuel Column length If at any time during operation it is de- s 4

termined that the limiting value for LHGR is --

being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed -limits. Surveillance

{ and corresponding action shall continue t until reactor operation is within the pre-j scribed limits. If the LHGR 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 />.

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i 3.0 LIMITING CONDITIONS FUR OPERATION 4.0. SURVEILLANCE REQUIRDiENTS i

C. Minimum Critical Power Ratio (MCPR) C. Minimum Critical Power Ratio (MCPR) I During power operation, the Operating MCPR MCPR shall be determined daily during Limit shall be > 132 for 8x8 fuel and reactor power operation at 2 25% rated 2 1.32 for 8x8R fuel at rated power and thermal power and f ollowing any change flow. If at any time during operation it in power level or distribution which.

is determined that the limiting value for has the potential of bringing the core MCPR is being exceeded, action shall be to its operating MCPR limit.

initiated within 15 minutes to restore operation to within the prescribed limits.

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits. .If the -

steady state MCP:t is not returned to within the prescribed limits within two (2) houre,, the reactor shall be breught to the Cold Shutdown condition i-within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. For core flows other than rated the Operating MCPR Limit shall-be the'above applicable MCPR value times Kg where Kg is as shown in Figure 3.11.3.

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Bases 3.11 A. Average Planar Linear IIcat Generation Rate -(APLilGR)

This specification ' assures that the peak cladding temperature following the postulated design basis los s-o f-c oo la nt accident will not exceed the limit specified in the 10CFR50, Appendix K.

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 af fect the calculated peak cladding temperature by less than +20 relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is suf ficient to assure that calculated temperatures are 'within the 10CFR50 Appendix K limit. The limiting value for APLilGR is given by this specification.

Reference 6 demonstrates that for lower initial core flow rates the potential exists for earlier DNB during postulated LOCA's. Therefore a more restrictive limit for APLIIGR is required during reduced flow conditions.

Those abnormal operational transients, analyzed in FSAR Section 14.5, which result in an automatic reactor scram are not considered a violation of the LCO. Exceeding APLHGR limits in such cases need not be reported.

B. LilGR This specification assures . that the linear heat' generation rate in any rod is less than'the design linear heat generation if fuel pellet densification is pos tulated. The power spike penalty specified is based on the analysis presented in .Section 3.2.1 of Reference 1 and in References 2 and 3, and assumes a linearly increasing variation and 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.

Those abnormal operational . transients, analyzed in FSAR Section 14.5, which result in an automatic reactor scram are .not considered a violation of the LCO. Exceeding LilGR limits in such cases need not be reported.

s I 189F 3.11 RASES REV J

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Bases Continued:

C. Mini, mum Critical Power Ratio (MCPR)

The ECCS evaluation presented in Reference 4 assured the steady state MCPR prior to the postulated los s-o f-c oola nt ar.cident to be 1.18 for all fuel types. In addition, the ECCS analysis presented in Reference 6 assumed an initial MCPR of 1.74 for reduced flow conditions. The Operating MCPR Llmit of l 1.32 for 8x9 fuel and 1 32 for 8x8R fr 'etermined f rom the analysis of transients discussed in t Bases Sections 2.1 and 2.3. By maint. . an operating MCPR above these limits, the Safety LLait ,_

(T.S. 2.1.A) is maintained in the event of the most limiting abnormal operational transient.  ;

For operation with less than rated core flow the Operating MCPR Limit is adjusted by multiply'ng the above limit by K . Reference 5 discusses how the transient analysis done at rated conditions encompasses the reduced flow situation when the proper Kg factor is applied.

Those aonormal operational tr - ~.'er 's analyzed in FSAR Section 14.5, which result in an automatic reactor scram are not consider wt a v:olation of the 100. Exceeding MCPR limits in such cases need not be repo rted. <  !

References

1. " Fuel Densifiestion Ef fects in General Electric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NEDM-10735, August, 1973.

2. Supplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 14, i 1974 (US AEC Regula tory Staf f) .

3. Communication: VA Moore to IS Mitchell, "Mocified CE Model for Fuel Densification," P. ket 50-321, March 27, 1974.

4. " Loss-of-coolant Accident Analysis Report for the Monticel's Nuclear Generating Plant," NEDO-24050, September 1977, L 0 Mayer (NSP) to V Stello (USNRC), September 15, 1977.

5. " General Electric BWR Generic Reload Application for 8x8 Fuel," NEDO-2036C, Revislam 1, November 4 1974.

6. " Revision of Low Core Flow Ef fects on LOCA Analysis for Operating BWR's," R L C 'iley (GE) to D G Eisenhut (USNRC), September 28, 1977.

I 189G 3 11 BASES REV

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Bases 4.11  !

I The APLIIGR, LHGR and MCPR shall be checked daily to determine if fuel burnup, or control rod movement have r caused enanges in power distribution. Since changes due to burnup are slow, and only a few cmtrol rods -

1 are removed daily, a daily check of power distribution is adequate. For a limiting value to occur below [

25% of rated themal power, an unreasonably large peaking factor would be required, which is not the case  ;

for operating control rod sequences. In addition, the MCPR is checked whenever changes In the core power n 1evel or distribution are made which have the potential of bringing the f uel rods to their thermal-hydraulic  ;

limits.

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189H 6011 BASES REV f NEXT PAGE IS 189M  !

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b 5.0 DESIGN FEATURES

• 5.1 Site A. i The reactor center line is located at approximately 850,810 feet North and 2,038,920 feet East as j

determined on the Minnesota State Grid, South Zone. The nearest site boundary is approximately l 1630 feet S 30 W of the reactor center line and the exclusion area is defined by the minimum O fenced area shown in FSAR Figure 2 2.2a. Due to the prevailing wind pattern, the direction of i

[ maximum integrated dosage is SSE. The southern property line follows the northern boundary of i

1 the right-mf-way for the Burlington Northern Railway.  :

5.2 -Reactor 1' i l  !

A. The reactor core shall consist of not more than 484 fuel assemblies.

B. The reactor core shall. contain 121 cruciform-shaped control rods. The control rod mterial shall i i

be boron carbide powder (B4 C) compacted to approximately 70% of theoretical density. i 4

5.2 Reactor Vessel A. The pressure vessel shall be designed for a pressure of 1250 psig and a temperature of 575 F.

The coolant recirculation system shall be designed for a pressure of 1148 psig on suction side of i l

. pump and 1248 psig at pump discharge. Both the pressure vessel and recirculation system shall be i

, designed .in accordance with the ASME Boiler and Pressure Vessel Code Sections III and II. -

i~, 5.4 containment j

A. The primary containment shall be of the pressure suppression type having a drywell and an 3bsorption t

chamber constructed of steel. The dryvell shall have a volume of approximately 134,200 ft and i-is designed to conform to ASME Boiler and Pressure Vessel Code Section III Class B for an internal t pressure of 56 psig at 281 F and an external pressure of 2 psgg at 281 F. The absorption chamber shall have a total volume of approximately 176,250 ft  ;

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R.W

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EXHIBIT C  :

SUPPLDIENT NO. 1 TO LICENSE AMENDMENT REQUEST DATED 11 ARCH 21,1978 This exhibi t consists of the report NED0-24133 entitled, " Supplemental- '

Reload Licensing Submittal for Monticello Nuclear Generating Plant

+

Reload 6".

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