ML20126E609

Amend 119 to License DPR-49,revising Tech Specs to Incorporate Changes Permitting Reactor Operation W/One Recirculation Loop Out of Svc

ML20126E609
Person / Time
Site:	Duane Arnold
Issue date:	05/28/1985
From:	Vassallo D Office of Nuclear Reactor Regulation
To:
Shared Package
ML112371152	List: ML20126E609
References
TAC-59235, NUDOCS 8506170150
Download: ML20126E609 (49)

Text

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,'% UNITED STATES

, [ i NUCLEAR REGULATORY COMMISSION t :j WASHINGTON, D. C. 20655

\...../

IOWA ELECTRIC LIGHT AND POWER COMPANY ,

CENTRAL IOWA POWER COOPERATIVE l CORN BELT POWER COOPERATIVE j DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE 4

Amendment No. 119 License No. DPR-49

1. The Nuclear Regulatory Comission (the Commission) has found that:

A. The application for amendment by Iowa Electric Light & Power Company, et al dated December 7, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, i as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the

• Comission; C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health

and safety of the public, and (ii) that such activities will be

, , conducted in compliance with the Comission's regulations; i D. The issuance of this amendment will not be inimical to the comon l defense and security or to the health and safety of the public; and E. The issuance of this amendme ' *s in accordance with 10 CFR Part 51 of the Comission's regt.iations and all applicable requirements i have been satisfied.

l 2. Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby j amended to read as follows:

j j

AN $ 1

. . (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.119, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: May 28, 1985 i

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ATTACHMENT TO LICENSE APENDPENT NO.119 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. E0-331 Revise the Appendix A Technical Specifications by removing the current pages and inserting.the revised pages listed below. The revised areas are identified by vertical lines.

LIST OF AFFECTED PAGES 1.1-1 3.3-7 1.1-2 3.3-7a**

1.1-3 3.3-18 1.1-5 3.3-20**

1.1-6 3.6-7 1.1-7 -3.6-7a**

1.1-8 3.6-7b** )

1.1-9 3.6-31 1.1-10 3.6-34 {

1.1-11 3.6-35 1.1-12 3.12-1 1.1-13 3.12-2

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1.1-14 3.12-3 1.1-15 3.12-4 1.1-16 3.12-7 1.1-17 3.12-3 1.1-18 3.12-9 .

1.1-19 (Figure 1.1-1) 3.12-9a*

1.1-20 (Figure 2.1-1) 3.12-10 1.1-21* 3.12-11 1.1-22* 3.12-13 (blank page) 1.1-23* 3.12-15 1.1-24* 3.12-16 3.1-3 3.12-17 -

l 3.2-16 3.12-18 3.12-19*

i 3.12-20*

• These pages have been deleted.

• • These are new pages g . . _

DAEC-1

. . SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability: Applicability:

Applies to the inter-related Applies to trip settings of the variables associated with fuel instruments and devices which thermal behavior, are provided to prevent the reactor system safety limits from being exceeded.

Objective: Objective:

To establish limits which To define the level of the ensure the integrity of the process variables at which

. fuel cladding, automatic protective action is initiated to prevent the fuel cladding integrity safety limits from being exceeded.

Specifications: Specifications :

The limiting safety system settings shall be as specified below:

A. Reactor Pressure > 785'asig A. Neutron Flux Trips and Core Flow > 10% of Rated

1. APRM High Flux Scram When The existence of a minimisn In Run Mode.

critical power ratio (MCPR) .

less than 1.07 for two For operation with the

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recirculation loop operation fraction of rated power

< (1.10 for single loop (FRP) greater than or equal operation) shall constitute to the maximum fraction of violation of the fuel cladding limiting power density integrity safety limit. (MFLPD), the APRM scra:n l o

trip setpoint shall be as )

8. Core Thermal Power Limit shown on Figure 2.1-1 and (Reactor Pressure 1785 psig shall be:

or Core Flow i10% of Rated S 1 (0.66W + 54)

When the reactor pressure is with a maximum setpoint of 120% rated power at 100%

<785 psig or core flow is less rated recirculation flow or Than 10% of rated, the core thermal power shall not exceed greater.

25 percent of rated thermal -

l power.

Amendment No. J4T, 119 l'l-1

a ... SAFETY LIMIT LIMITING SAFETY SYSTEM. SETTING C. Power Transient Where: $ = Setting in percent of rated power (1,658 MWt)

To ensure that the Safety limits

. . established in Specification W = Recirculation loop flow in 1.1.A and 1.1.8 are not exceeded, percent of rated flow, each required scram shall be Rated recirculation loop i initiated by its primary source flow is that recirculation sional. A Safety Limit shall be loop flow which assumed to be exceeded when scram corresponds to 49x106~

is accceplished by a means other Ib/hr core- flow.

than the Primary $ource Signal.

For a MFLPD greater than FRP, the- -

1 APRM scram setpoint shall be: I FRP 1

S < (0.66 W + 54)

- for two MFLPD recirculation loop operation, and

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- D. With irradiated fuel in the reactor vessel, the water level shall not be less than 12 in. 5 < (0.66 W + 50.5)

MFLPD above the top of the normal active fuel zone. Top of the for one recirculation loop operation.

active fuel zone is defined to be 344.5 inches above vessel zero NOTE: These settings assume (see Bases 3.2). Operation within the basic therinal design criteria. These criteria are LHGR < 13.4 KW/f t (8x8 array) and MCPR 7 values as indicated in d

Table 3.17-2 times Kf, where 4

Kf is defined by Figure 3.12-

1. Therefore, at full power, operation is not allowed with .

MFLPD greater than unity even if the scram setting is reduced.* If it is determined that either of these design criteria is being violated during operation, action must be taken immediately to return to operation within these criteria.

2. APRM High Flux Scram When in the REFUEL or STARTUP and '

HOT STANDBY MODE the APRM scram shallbesetatiessthanorequal to 15 percent of rated power.

• With MFLPD creater than FRP durinc power ascension up to 90% of r.ated' rather than adjusting the APRM power, ts, the APRM gain .aay be. adjusted setpoin such that APRM readinas are greater than or equal to 100% of MFLPD, provided that the adjusted APRM reading does not exceed 100% of rated power and a notice of adjustment is posted on the reactor l control panel.

Amendment No. M , 119 1.1-2

DAEC-1

. . -_ SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING

3. APRM Rod Block when in Run Mode.

For operation with MFLPD less than or equal to FRP the APRM Control Rod Block setpoint shall be as shown on Fig. 2.1-1 and shall be:

S;[ (0.66 W + 42)

The definitions used above for the APRM scram trip apply.

For a MFLPD greater than FRP, the

. APRM Control Rod Block setpoint shall be:

S < (0.66 W + 42) for two MFLPD recirculation loop operation, .and S < (0.66 W + 38.5)

MFLPD .

for one recirculation loop operation.

4. IRM - The IRM scram shall be set at less than or equal to 120/125 of full scale.

B. Scram and > 514.5 inches Isolation on ihove vessel reactor low zero (+170"

water level indicated level)

C. Scram - turbine < 10 percent stop valve valve closure closure D. Turbine control valve fast closure shall occur within 30 milliseconds of the start of turbine control valve fast closure.

Amendment No. 345', 119 1.1-3

DAEC - 1

1.1 BASES

FUEL CLADDING INTEGRITY A.

Fuel Cladding Integrity Limit at Reactor Pressure > 785 psig and . Core Flow > 10% cf 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, l 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 l integrity safety limit is defined as the critical power ratio in the limiti,ng fuel assembly for which more than 99.9% of the fuel rods in the core are s

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 0.03 for single-loop operation.

Amendment No. M . 119 1.1-5

, , DAEC-1 B. Core Thermal Power Limit (Reactor Pressure 1785 psig or Core Flow 110%

of Rated)

At pressures below 785 psig, the core evaluation pressure drop (0 power, 0 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 always be greater than 4.56 psi.

Analyses show that with a flow of 28 x 101bs/hr 3

bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi driving head will be greater than 28 x 10 3 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

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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 power limit of 25% for reac' tor pressures below 800 psia or core flow less than 10% is conservative.

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 1.1. A or 1.1.8 will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when

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a scram is accomplished other than by the expected scre signal (e.g., scram from neutron flux folowing close of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a Safety Limit violation will be asstaned 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.

Amendment No. 119 1.1-6 l 1

e DAEC-1

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t The computer provided with Duane Arnold 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 i 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

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not be available for any scram analysis, Speelfication 1.1.C will be relied on to determine if a Safety Limit has been violated.

D. Reactor Water Level (Shutdown Condition)

During pertods when the reactor is shut down, consideration must also be given 4

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 th'e core height. Establishment of the Safety Limit at 12 inches above

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- the top of the fuel

• provides adequate margin. This level will be continuously monitored.

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• Top of the active fuel zone is defined to be 344.5 inches above vessel zero

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(See Bases 3.2).

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Amendment No. 119 g 1

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DAEC-1 i

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1.1 REFERENCES

I i -

1. " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A* l l

j 2. "Duane Arnold Energy Center Single-Loop Operation," NEDO-24272 July 1980. l l

i l

l

• Approved Revision at tim'e reload analyses are performed.

Amendment No. 119 1.1-8 w - , - ,, - - - -

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DAEC-1 t

2.1 BASES

LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL Lt. ADDING INTEGRITY The abnormal operational transients applicable to operation of the Duane Arnold Energy Center have been analyzed throughout the spectrum of planned operating conditions up to the thermal power condition of 102% of 1658 MWt in accordance with Regulatory Guide 1.49. The analyses were based upon plant operation in accordance with the operating map given in Figure 1.1-1 of the Technical Specifications.

In addition,1658 MWt is the licensed maximum power level of the Duane Arnold Energy Center, and this represents the maximum steady state power which shall not knowingly be exceeded.

Conservatism is incorporated in the transient analysis in estimating the controlling factors, such as void reactivity coefficient, control ,

rod scram worth, scram delay time, peaking factorsnand axial power shapes. These factors are selected conservatively with respect to their' effect on the applicable transient results as determined by the current analysis mode. Conservatism incorporated into the transient analysis is documented in Reference 1.

This choice of using conservative values of controlling parameters and initiating transients at the design power level produces more conservative results than would be obtained by using expected' values of control parameters and analyzing at higher power levels.

I*I'9 l Amendment No. 119

DAEC-1 For analyses of the thermal consequences of the transients the MCPRs 1

stated in Section 3.12 as a limiting condition of operation bound those which are conservatively asstrned to exist prior to initiation of the transients.

As discussed in Reference 2, the core-wide transient analyses for one recirculation pump operation is conservatively bounded by two-loop operation analyses and the flow-dependent rod block and scram setpoint equationsarefdjustedforone-pumpoperation.

Steady-state operation without forced recirculation will not be permitted, except during special testing. The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps.

in suninary:

1. The abnormal operational transients have been analyzed to a

+

power level of 102% of 1658 MWt.

ii. The licensed maximum power level is 1658 MWt.

iii. Analyses of transients employ adequately conservative values of the controlling reactor parameters.

• i Amendment No. 119 1.1-10 l

DAEC-1 iv.

The analytical procedures now used result in a more logical answer than the alternative method of assuming a higher starting power in conjunction with the expected values for 1 i

the parameters.

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i

Trip Settings i

j The bases for individual trip settings are discussed in the following

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paragraphs.

A. Neutron Flux Trips

1. APRM High Flux Scram (Run Mode) 1 l The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1658 MWt).

Because fission chambers provide the basic input signals, the APRM-system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.

Analyses demonstrate that with a 120 percent scram trip setting, none of the abnormal operational transients analyzed violate the fuel Safety Limit and there is a substantial margin to the threshold for fuel damage. Therefore, the use of flow referenced i

Amendment No. 119 l 1-11 I l

i DAEC-1  !

I l scram trip provides additional margin. An increase in the APRM l {

t scram trip setting would decreas! the margin present before the  ;

i

fuel cladding integrity Safety Limit is reached. The APRM scram i

trip setting was determined by an analysis of margins required to 1 l provide a reasonable range for maneuvering during operation. f

t Reducing this operating margin would increase the frequency of j spurious scrams which have an adverse effect on reactor safety i j because of the resulting thermal stresses. Thus, the APRM scram

] trip setting was selected because it provides adequate margin for i

i the fuel cladding integrity Safety Limit yet allows operating i j margin that reduces the possibility of unnecessary scrams, j The scram trip setting must be adjusted to ensure that the LHGR ,.

i transient peak is not increased for any combination of MFLPD and 1

j (

reactor core thermal power. The scram setting is adjusted in

) accordance with the formula in Specification 2.1.A.1, when the -

l I

{ maximum fraction of limiting power density is. greater than the l

l fraction of rated power. This adjustment may be accomplisheo by  !

increasing the APRM gain and thus reducing the slope and intercept i i -

p point of the flow-referenced APRM High Flux Scram curve by the 4

l reciprocal of the APRM gain change. l 1

Analyses of the limiting transients show that no scram adjustment is required to assure MCPR greater than or equal to the Safety l Limit when the transient is in'.tiated from MCPR > values as indicated 'in Figures 3.12-2 and 3.12-3. l i

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j ' Amendment No. 119 1.1-12 l

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DAEC-1

2. APRM High Flux Scram (Refuel or Startup & Hot Standby Mode)

T L For operation in these modes the APRM scram setting of 15 percent of rated power and the IRM High Flux Scram provide adequate thermal

( margin between the setpoint and the Safety Limit 25 percent of rated. The margin is adequate to accomodate anticipated maneuvers associated with power p1 M t startup. Effects of increasing pressure at zero or low void content are minor, cold water from

sources available during startup is not much colder than that

! already in the system, temperature coefficients are small, and i

j control rod patterns are constrained to be unifonn by operating j procedures backed up by the rod worth minimizer and the Rod i

Sequence Control System. Worths of individual rods are very low in

! a uniform rod pattern. Thus, of all possible sources of reactivity  !

l l input, uniform control rod withdrawal is the most probable cause of l significant power rise.

'Because the flux distribution associated with uniform rod

. withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of ,

rated power, the rate of power rise is very slow. Generally, the heat flux is near equilibritan with the fission rate. In an assumed uniform rod withdrawal approach to the scran level, the rate of I

power rise is no more than 5 percent of rated power per minute, and l the APRM system would be more than adequate to assure a scram l before the power could exceed the Safety Limit. The 15 percent f

I Amendment No. 119 g,g.g3 l i

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i l DAEC-1 1

APRM scram remains active until the mode switch is placed in the RUN i position. This switch occurs when reactor pressure is greater than 850 psig.

e  !

3. APRM Rod Block (Run Mode) l Reactor power level may be varied by moving control rods or by l

varying the recirculation flow rate. The APRM system provides a j control rod block to prevent rod withdrawal beyond a given power l l

level at constant recirculation flow rate, and thus prevents a MCPR less than the Safety Limit. This rod block trip setting, which is j automatically varied with recirculation loop flow rate, prevents 1

j excessive reactor power level increase resulting from control rod j withdrawal. 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 specif,ied trip setting versus flow relatio1 ship; therefore the worst case MCPR l

a which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting. The actual l .

4

,i power distribution in the core is established by specified control '

i rod sequences and is monitored continuously by the in-core LPRM system. As with the APRM scram trip setting, the APRM rod block trip l setting is adjusted downward if the maximum fraction of limiting i

j power density exceeds the fraction of rated power, thus preserving  !

j the APRM rod block safety margin. As with the scram sett'ing, this

{ may be accomplished by adjusting the APRM gain.

i  !

i Amendment No. 119 i

DAEC-1

4. IRM The IRM system consists of 6 chambers, 3 in each of the reactor protection system logic channels. The IRM is a 5-decade instrument which covers 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 active in each range of the IRM. For example, if the instrianent were on range 5, the scran would be 120 divisions on that range. Thus, as the IRM is ranged up to acconnodate the increase In power level, the scram trip set 61ng is also ranged up. The most significant sources of reactivity change during the power increase are due to control rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods that the heat flux is in equilibrium with the neutron flux, and the IRM scram would result in a reactor shutdown well before any Safety Limit is exceeded.

In order to ensure that the IRM provides adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents has been analyzed. This analysis included starting the accident at various power levels. Ths 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 conservatism was taken in this analysis by assuming that 1.1-15 l Amendment Nn. 119

DAEC-1 the IRM channel closest to the withdrawn rod is by-passed. The results of this analysis show that the reactor is scramed and peak power limited to one percent of rated power, thus maintaining MCPR above the Safety Limit. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

B. Scre and Isolation on Reactor Low Water Level The setpoint for the low level scrm is above the bottom of the separator skirt. This level has been used in transient analyses dealing with coolant inventory decrease. Analyses show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR.

is greater than the Safety Limit in all cases, and system pressure l does not reach the safety valve settings. The scr m setting is approximately 21 inches below the normal operating range and is thus adequate to avoid spurious scr es.

4 C. Scr e - Turbine Stop Valve Closure The turbine stop-valve closure scram anticipates the pressure, neutron flux, and heat flux increase that could result from rapid closure of the turbine stop valves. With a ser m setting at 10 percent of valve closure, the resultant increase in surface heat flux i

Amendment No. )(, g, lig 1.1-16 l i . .

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DAEC-1 is such that MCPR remains above the Safety Limit even during the l worst case transient that assumes the turbine bypass is closed. This i

j scram is by-passed when core thermal power is below 30 percent of l

)

rated, as measured by the turbine first stage pressure.

l O. Turbine Control Valve Fast Closure (Loss of Control Oil Pressure Scram) l i

i The control valve fast closure scram is provided to limit the rapid increase in pressure and neutron flux resulting from fast closure of 5 t i the turbine control valves due to a load rejection. It prevents MCPR from becoming less than the Safety Limit for this transient. l 1 l

l r

! E. F. and J. Main Steam Line Isolation on Low Pressure. Low Condenser

{ Vacuum. and Main Steam Line Isolation Scram

\ -

, The low pressure isolation of the main steam lines at 850 psig has 1

j been provided to protect against capid reactor depressurization and

)4 the resulting rapid cooldown of the vessel. Advantage is taken of

[*

i the scram feature that occurs when the main steam line isolation l valves are closed, to provide for reactor shutdown so that high power l

i operation at low reactor pressure does not occur, thus providing j protection for the fuel cladding integrity. Operation of the reactor i at pressures lower than 850 psig rwaires that the reactor mode j switch be in the STARTUP position, where protection of'thi fuel i

cladding integrity safety limit is provided by the IRM and APRM high l.

i AmendmentNo.Jd,119 1.1-17 l

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DAEC-1 neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation

' valve closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase. To protect the main condenser against overpressure, a loss of condenser vacum initiates automatic closure of the main steam isolation valves.

G. H. and I. Reactor low Water Level Setpoint for Initiation of HPCI and RCIC Closing Main Steam Isolation Valves, and Starting LPCI and Core Spray P eps These systems maintain adequate coolant inventory and provide core ,

cooling with the objective of preventing excessive clad temperatures. <

The design of these systems to adequately perform the intended function is based on the specified low level scram setpoint and

' initiation setpoints. Transient analyses demonstrate that these l conditions result in adequate safety margins for both the fuel and the system pressure.

2.1 REFERENCES

1. " General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A* '

I i

2. "Duane Arnold Energy Center Single-Loop Operation," NEDO-24272, July 1980.  !

• Approved revision n eber at time analyses are performed.

AmendmentNo.)Hf,119 1.1-18 l l

. ItATED Tite %\L POtfER = 1658 Mtf 6 RATED CO.E FLOW =

49 x 10 lb'/h'r 120 -

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I DUAND ARNOLD ENERGY CENTER IOWA ELECTRIC LIGitT & POWER CO!!PANY TECilNICAL GPECIFICATIONS APRM FLOU DIAS SCRMt RELATIONSilIP TO NORMAL OPF.RATI!!G CONDITIONS FIGURC 1.1-1 Amendment No. M . 119 1,1,19 ,

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WW Amendment No. 119 1.1-20

TA8tE 3.1-1 ' -

t

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- REACTOR PROTECTION SYSTEM (SCRAM) INSTRIMENTATION REQUIREMENT g Minimen l 3 No. of Modes in Which Number of 3 Operable Function N st be Instrument z Instrument Operable ChanneIs '

o Channels for Trip Provided by Refuel Startup Run Design Action (1) l System (1) Trip Function Trip Level Setting (6) i .

5

• 1 Mode Switch in X X X 1 Mode Switch A Shutdown (4 Sections) 1 Manual Scran X X X 2 Instrument A Channels 2 IRM High Flux < 120/125 of Full Scale X X (5) 6 Instrument A l Channels

';* 2 IRM Inoperative X X (5) 6 Instrument A Channels 2 APRM High Flux for two recirc loop operation X 6 Instrument A or B  !

<(.66W+54) Channels Tor one rec irc (FRP M LPO) (11) (12) loop operation

<(.66W+50.5)(FRP/MFLPD) (11) (12)

I 2 APRM Inoperative (10) 6 Instrument

, X X X A or 8  :

, Channels  !

t 2 APRM Downscale > 5 Indicated on Scale (9) 6 Instrument A or B Channels 5

2 APRM High Flux ~

< ISE Power X X 6 Instrument A  !

in Startup Channels 2 High Reactor -

< 1055 psig X(8) X X 4 Instrument A i Pressure .

Channels i

{

l

TABLE 3.2-C I;strumentation Th t laitictes Control Rod Blocis Minimus No. Number of of Operable Instrument -

2 Instrunent Channels .

2 Channels Per Provided by a Trip Systen Instrument Trip Level Setting Design Action 1 5 5 2 APRM Upscale (Flow Blased) 'for 2 recirc loop operation 5

1(0.66 W + 42) ( " * """'

• 0 i

! for 1 recirc loop operation 3

w I

1(0.66 W + 38.5) ( 0 j 2 APRM Upscale (Not in Run Mode) i 12 indicated on scale 6 Inst. Channels (1) 2 APRM Downscale > 5 indicated on scale 6 Inst. Channels (1) w 1 (7) Rod Block Monitor for 2 recirc loop operation

'3

'; (Flow Biased) FRP 2 Inst. Channels (1) 5 -<(0.66 W + 39)( WLPD)(2) for 1 recirc loop operation N

1(0.66 W + 35.5) ( -LPD '

1 (7) Rod Block Monitor Downscale

> 5 indicated on scale 2 Inst. Channels (1)

2 IRM Downscale (3) > 5/125 full scale 6 Inst. Channels (1) 2 IRM Detector not in '(8) 6 Inst. Channels (1)

Startup Position 2 IRM Upscale i 108/125 6 Inst. Channels (1) i 2 (5) -

SRM Detector not in (4) 4 Inst. Channels (1)  !

Startup Position 5

2 (5)(6) SRM Upscale i 10 counts /sec. 4 Inst. Channels (1) i 1 Scram Discharge Volume . < 24 gallons 1 Inst. Channel (9)

Water Level-High

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.D Reactivity Anomalies 4.3.0 Reactivity Anomalies The reactivity equivalent of During the startup test program i

the difference between the and startup following refueling actual critical rod outages, the critical rod configuration and the configurations will be compared expected configuration to the expected configurations during power operation shall at selected operating not exceed 1% a k. If this conditions. These comparisons limit is exceeded, the will be used as base data for reactor will be shut down reactivity monitoring during until the cause has been subsequent power operation determined and corrective throughout the fuel cycle. At actions have been taken as specific power operating appropriate. conditions, the critical rod

- configuration will be compared to the configuration expected based upon appropriately corrected past data. This comparison will be made at

< 1 east every full power month.

E. Recirculation Pumps E. Recirculation Pumps When the reactor mode switch With two recirculation pumps in is in startup or run operation and with core thermal position, the reactor shall power greater than the limit not be operated in the specified in Figure 3.3-1 and natural circulation flow total core flow less that 45%

mode. ,

of rated, establish baseline APRM and LPRM* neutron flux .

With two recirculation pinps noise levels within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in operation and with core provided that baseline values thermal power greater than have not been previously the limit specified in established since the last core Figure 3.3-1 and total core refueling.

flow less than 45% of rated.

the APRM and LPRM* neutron flux noise levels shall be detennined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and:

a) if the APRM and LPRM*

neutron flux noise

• Detector levels A and C of one .

LPRM string per core octant plus detector levels A and C of one LPRM i string in the center of the core

shall be monitored.

3.3-7 Amendment No.)6 )( 119

r

. DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT levels are less than or equal to three times their established baseline levels, 4 continue to determine the noise levels at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and also within 30 minutes after the completion of a core thermal power increase of at least 5% of rated

' core thermal power while operating in this region of the power / flow map, 1 or b) if the APRM and/or

- LPRM* neutron flux noise levels are greater than three times their established baseline levels, immediately initiate corrective action and restore the noise levels to within the required limits 1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by l increasing core flow and/or by initiating,an orderly reduction of core thermal power by inserting control rods.

,See Specifications 3.6.F.2 for operation with one ,

recirculation loop not in operation.

A recirculation pwnp shall not be started tile the reactor is in natural circulation flow and reactor '

power is greater than 1% of rated thermal power.

F. If Specifications 3.3.A through D above cannot be met an orderly shutdown '

shall be initiated and the reactor shall be in the COLD SHUTDOWN condition within 24 ,

hours.

• Detector levels A and C of one LPRM string per core octant plus -

detector levels A and C of one LPRM string in the center of the core shall be monitored.

l Amendment No. 119 3.3-7a

DAEC-1 After initial fuel loading and subsequent refuelings when operating above 950 psig, all control rods shall be scram tested within the constraints imposed by the Technical Specifications and before the 40% power level is reached. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

4. Reactivity Anomalies During each fuel cycle excess operative reactivity varies as fuel depletes and

, as any burnable poison in supplementary control is burned. The magnitude of l this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state. Power operating base conditions provide the most sensitive and directly interpretable data relative to core reactivity. Furthermore, using power operating base conditions permits frequent reactivity comparisons.

~

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 design conditions of the reactor system.

5. Recirculation Pumps APRM and/or LPRM oscillations in excess of those specified in section 3.3.E could be an indication that a condition of thermal hydraulic instabil,ity exists and that appropriate remedial action should be taken. These specifications are based upon the guidance of GE SIL #380, Rev. 1, 2/10/84 Amendment No. ) W M , 119 3.3-18 l _ _

f (

70 .

{

I i

arveillance magion 60

• per 3.3.E or 3.6.F.2.b  :-

t 50  ;

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- I I

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i Un 1 I t i  ;

20 30 40 . 50 60 70 80 s

Total Core Flow (5 of Rated) ,,

DAEC Iowa Electric Light *& Power Cogany j TechnicalSpecklications Thermal Power vs C) pre Flow Limits for Thermal Hydraylic stability Survel115nce Fievre 3.3-1 j

Amentent No.119 3.3 20

DAEC-1 e

. LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENT

b. The indicated value of core flow rate varies from the value derived form loop flow measurements by more than 101

c. The diffuser to lower plenum differential pressure readino on an individual .iet pump varies from the mean of all ,iet opp differential pressures by more than 10%.

2. Whenever there is recirculation flow from the

• reactor in the Startup or Run mode, and one recirculation ptro is operating, the diffuser to lower plente differential cressure shall be checked daily and the differential pressure of an individual

,iet pe o in a loop shall not vary from the mean of all

,iet otep differential pressures in that loop by more than 10%.

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

1. When both recirculation 1. Recirculation pump speeds pinos are in steady state shall be checked and locoed at operation, the speed of the least once per day, f aster ptmp may not exceed 122% of the speed of the 2. a. Prior to noeration with one o slower ptsp when core power recirculation pump not in is 80% or rore of rated operation and core thermal power or la5% of the speed power creater than the Ifmit of the slower pino when specified in Floure 3.3-1, core power is below 80% of establish baseline APRM and rated power. LPR W neutron flux noise levels, provided that

2. If Specification 3.6.F.1 baseline values have not

• cannot be met, one recirculation pino shall be tripped. The reactor may *Datector Icvels A and C of one be start./ and operated, or LPRM Strino per core octant plus operation may continue with detector levels A and C of one LPRM one recirculatton loop not string in the center of the core in operation provided shall be monitored.

t thatt f

AmendmentNo.ff,119 3.6 7

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT

a. MAPLHGR multipliers as been previously established indicated in section since the last core 3.12.A are applied, refuelino. Baseline values shall be established with

b. With one recirculation one recirculation pump not pump not in oneration in operation and core and core thermal power thermal power less than or greater than the limit enual to the limit specified specified in Floure in Fioure 3.3-1.

3.3-1, core flow must be creater than or ecual to b. Prior to operation with one 39% of rated, and recirculation pump not in operation and core flow (1) the Surveillance creater than a5% of rated, Requirements of establish baseline core

• 4.6 F.2.a have not been plate AP noise levels with satisfied,immediately core flow less than or eoual initiate action to to 45% of rated, provided reduce core thermal that baseline values have power to less than or not heen previously ecual to the Ifmit estabitshed with one specified in Floure 3.3-1 recirculation pump not in within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or operation since the last core refueling.

(11) the Surveillance Recuirements of 4.6 F 2.4 have been satlfled,continueto determine the APRM and LPRM neutron flux levels at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and also within 30 minutes after the completion of a core

• thermal power increase of at least 5% of rated core thermal power while operatino in this reoion of the power / flow map.

If the APRM and/or LPRM*

. neutron flux noise levels are creater than three times their established baseline values immediately intItate corrective action and restore the noise levels to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by

• Detector levels A and C of one LPRM string per core netant plus detector levels A and C of one LPRM ,

string in the center of the core shall be monitored.

Amendment No. 119 3.6-7a l 9

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENT increasino core flow and/or initiatino an orderly reduction of core thermal power by insertina control rods.

c. With one recirculation ounp not in operation and core flow areater than 45% of rated, and (i) the Surveillance Reauirements of 4.6.F.2.b have not been satisfied, immediately initiate action to reduce core flow to less than or eaual to 45% of rated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or

~

(ii) the Surveillance Requirements of 4.6 F.2.b have been satisfied, continue to determine core plate AP noise at least once oer 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and also within 30 minutes after the completion of a core thermal power increase of at least 5% of rated thermal power. If the core plate AP noise level is greater than 1.0 osi and 2 times its established baseline value, immediately initiate corrective action and restore the noise levels to within the reauired .

limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by decreasina core flow and/c-initiatina an orderly reduction of core thermal power by insertina control rods.

d. The idle loop is isolated electrically by disconnectino the breaker to the recirculation pump motor Generator (M/G) set drive motor prior to startup, or if disabled during reactor operation, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Refer to Specification 3.6.A for startuo of the idle recirculation 1o00.

e. The recirculation system controls will be placed in the manual flow control mode.

Amendment No. 119 3.6-7b l

DAEC-1

c. The jet punp flow deviation pattern derived from the diffuser to lower plenum differential pressure readings will be used to further evaluate jet pump operability in the event that the jet pumps fail the tests in Section 4.6.E.1 and 2. l Agreement of indicated core flow with established power-core flow relationships provides the most assurance that recirculation flow is not bypassing the core through inactive jet pumps. This bypass flow is reverse with respect to normal jet pump flow. The indicated total core flow is a sunmation of the flow indications for the sixteen individual jet pumps. The total core flow measuring instrumentation sums reverse jet pump flow as though it were forward flow in the case of a failed jet pump. Thus the indicated flow is higher than actual core flow by at least twice the normal flow through any'backflowing jet pump.*

Reactivity inventory is known to a high degree of confidence so that .

- even if a jet pump failure occurred during a shutdown period, subsequent power ascension would promptly demonstrate abnormal control rod ' withdrawal for any power-flow operating map point.

A nozzle-riser system failure could also generate the coincident failure of a jet pump body; however, the converse is not true.

• Note: In the case of single recirculation loop operation, when the recirculation pump is tripped, the flow through the inactive jet pumps is subtracted from the total jet pump flow, yielding the correct value for the total core flow.

3.6-31 Amendment No.,7f, 119

DAEC-1

, , 80% power cases, respectively. If the reactor is operating on one pump, the loop select logic trips that pump before making the loop selection.

An evaluation has been provided for ECCS performance during reactor operation with one recirculation loop not in operation (Sec. 3.12, Ref.11). Therefore, continuous operation under such conditions is appropriate. The reactor may also be operated up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with one recirculation loop not in operation without electrically isolating the idle loop. This short period of time permits corrective action to be taken to re-activate the idle loop or to implement the changes for continuous operation with one recirculation loop not in operation.

. During periods of Single Loop Operation (SLO), the idle recirculation loop is isolated by electrically disarming the recirculation pump. This is done to prevent a cold water injection transient caused by an inadvertent pump start-up. It is permissible to leave the suction and discharge valves open during SLO to allow flow through the loop in order to maintain the temperature.

However, if for some reason the discharge valve is inoperable it should be closed and electrically disarmed. This is done to prevent degradation of LPCI flow during a LOCA. With the discharge valve disarmed, the temperature in the loop can be maintained by opening the bypass valve, as the loop selection logic will close the bypass valve, isolating the loop, prior to opening the LPCI injection valve.

APRM and/or LPRM oscillations in excess of those specified in Section 3.6.F.2 could be an indication that a condition of thermal hydraulic instability exists and that appropriate remedial action should be taken. By restricting core flow to greater than or equal to 39% of rated, which corresponds to the core flow at the 80% rodline with 2 recirculation pumps running at minimum speed, the region of the power / flow map where these oscillations are most likely to occur is avoided. Individual APRM or LPRM channels exhibiting excessive flux noise may be discounted upon verification that a true condition of thermal hydraulic instability does not exist by observation of the remaining available APRM and/or LPRM channels. These specifications are based upon the guidance of GE SIL #380, Rev. 1, 2/10/84.

Above 45% of rated core flow in SLO there is the potential to set up high flow-induced noise in the core. Thus, surveillance of core plate AP noise is required in this region of the power / flow map to alert the operators to take appropriate remedial action if such a condition exists.

AmendmentNo.[,119 3.6-34 l

DAEC-1 Requiring the discharge valve of the lower speed loop to remain closed until the speed of faster punp is below 50% of its rated speed provides assurance when going from one to two pump operation that excessive vibration of the jet punp risers will not occur.

3.6.G & 4.6.G BASES:

REACTOR COOLANT SYSTEM Structural Integrity A pre-service inspection of Nuclear Class I Components was conducted to assure freedom from defects greater than code allowance; in addition, this served as a reference base for future inspections. Prior to operation, the reactor coolant system as described in Article IS-120 of Section XI of the ASME Boiler and Pressure Vessel Code was inspected to provide assurance that the system was free of gross defects. In addition, the facility was designed such that gross defects should not occur throughout plant life.

The pre-service inspection program was based on the 1970 Section XI of the ASME Code for in-service inspection. This inspection plan was designed to reveal problem areas (should they occur) before a leak in the coolant system could develop. The program was established to provide reasonable assurance, L that no LOCA would occur at the DAEC as a result of leakage or breach of pressure-containing components and piping of the reactor coolant system, portions of the ECCS, and portions of the reactor coolant associated

~

auxiliary systems.

A pre-service inspection was not performed on Nuclear Class II Components because it was not required at that stage of DAEC construction when it would have been used. For these components, shop and in-plant examination records i

of components and welds will be used as a basis for comparison with in-service inspection data.

AmendmentNo.j?T$119 3.6-35

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.12 CORE THERMAL LIMITS 4.12 CORE THERMAL LIMITS Applicability Applicability The Limiting Conditions for The Surveillance Requirements Operation associated with the apply to the parameters which fuel rods apply to those monitor the fuel rod operating parameters which monitor the fuel conditions.

rod operating conditions.

Objective Objective TheObjectiveoftheLimiting The Objective of the Surveillance Conditions for Operation is to Requirements is to specify the assure the performance of the type and frequency of fuel rods. surveillancento be applied to the fuel rods.

Specifications Specifications

~

A. Maximum Averaae Planar Linear A. Maximum Averaoe Planar Linear neat ueneration Kate (nartnbx) neat benerat16n Kate (MAVLnbK)

During reactor power operation, The MAPLHGR for each type of fuel the actual MAPLHGR for each type as a function of average, planar of fuel as a function of average exposure shall be determined planar exposure shall not exceed daily during reactor operation at the limiting value shown in Figs. > 25% rated thermal power and any 3.12-5 7 -8 and -9. For 31ange in power level or single, loop, ope, ration, the values distribution that would cause in these curves are reduced by operation with a limiting control multiplying by 0.87. If at any rod pattern as described in the time during reactor power bases for specification 3.3.2.

at During operation with a limiting operation

>25% rated(oneortwoloop) thermal power, i tis control rod pattern, the MAPLHGR tretermined by normal surveillance (LAPLHGR) shall be determined at that the limiting value for least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.. -

MAPLHGR (LAPLHGR) is being exceeded, action shall then be initiated within 15 minutes to ^

restore operation to within the prescribed limits. If the MAPLHGR'(LAPLHGR) is not returned to within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, reduce reactor power to -< 25% of rated thermal power, or to such a power level that the limits are again bein met, within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. g If the reactor is being operated with one recirculation loop not in operation and cannot be returned to within prescribed limits within this 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period, the reactor shall be brought to the COLD SHUTOOWN condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

For either the one or two loop 1 operating condition surveillance I and corresponding action shall  !

continue until the prescribed l l limits are again being met.

I 3.12-1 Amendment No.)Hf, JW, JIM 119

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT B. Linear Heat Generation Rate B. Linear Heat Generation Rate (LHGR) (LHUR)

1. During reactor power The LHGR as a function of core operation the linear heat height shall be checked daily generation rate (LHGR) of any during reactor operation at > 25%

rod in any P8X8R, BP8X8R or thermal power and following any ELTA fuel assembly shall not change in power level or

, exceed 13.4 KW/f t, while the distribution that would cause LHGR of any rod in an LTA-311 operation with a limiting control fuel assembly shall not rod pattern as described in the exceed 14.4 KW/ft. bases for Specification 3.3.2.

During operation with a limiting If at any time during reactor control rod pattern the LHGR power operation at >25% rated shall be determined at least once

, thermal power it is- per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall then be initiated within 15 minutes to restore operation to within the prescribed limits. If the LHGR is not returned to within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, reduce reactor power to < 25% of Rated Thermal Powe'r, or to such a power level that the limits are again being met, within the ~

next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Surveillance s and corresponding action shall continue until the prescribed limits are again being-met.

l l

Amendment No.) 87 M 119 3.12-2

DAEC-1

. , LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

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

During reactor power operation, MCPR shall be determined daily MCPR for one or two during reactor power operati.on recirculation loop operation at > 25% rated thermal power shall be > values as indicated and Tollowing any change in in FigureT 3.12-2 and -3. power level or distribution These values are multiplied by that would cause operation with

! Kf which is shown in Figure a limiting control rod pattern 3.12-1. Note that for one as described in the bases for recirculation loop operation Specification 3.3.2. During the MCPR limits at rated flow operation with a limiting are 0.03 higher than the control rod pattern, the MCPR comparable two-loop values. If shall be determined at least at any time during reactor once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

power operation (one or two loop) at >25% rated thermal power, it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall then be initiated within 15 minutes to restore operation to within the prescribed limits.

If the operating MCPR is not returned to within the prescribed limits within two hours, reduce reactor power to

< 25% of rated thermal power, or to such a power level that .

the limits are again being

- met, within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. '

If the reactor is being operated with one recirculation loop not in operation, and

, cannot be returned to within

. prescribed limits within this 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period, the reactor shall be brought to a COLD SHUTDOWN condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

For either the one or two loop operating condition surveillance and corresponding action shall continue until the prescribed limits are again ,

being met.

i

~

AmendmentNo.[,,,MC119

DAEC-1 3.12 BASES: CORE THERMAL LIMITS A. Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)

This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident (LOCA) will not exceed the limit specified in 10CFR50.46. LOCA

_ analyses are performed using General Electric calculational models which conform to the requirements of 10 CFR Part 50, 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 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 10CFR50.46 limit.

For two recirculation loop operation, the calculational procedures used to establish the MAPLHGR's shown on Figures 3.12-5 thru 3.12-9 are documented in Reference 7. The reduction factors for one recirculation loop operation were derived in Reference 13.

3.12-4 AmendmentNo.[.119

DAEC-1 i . .

2. MCPR Limits for Core aws Other than Rated Flow The purpose of the Kg 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 Kf factor. Specifically, the Kf 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 Kf factors assure that the operating limit MCPR of values as indicated in Figures 3.12-2 and 3.12-3 will not be violated should the most limiting transient occur at less than rated flow. In the manual flow control mode, the K f factors assure that the Safety Limit MCPR will not be violated for the same postulated transient event.

The K f factor curves shown in Figure 3.12-1 were developed generically and are applicable to all BWR/2, BWR/3 and BWR/4 reactors. The K f factors were derived using the flow control line corresponding to rated thermal power at rated core flow, as described in Reference 2.

The Kf factors shown in Figure 3.12-1 are conservative for Duane Arnold operation because the operating limit MCPR of values as indicated in Figures 3.12-2 and 3.12-3 is greater than the original 1.20 operating limit MCPR used for the generic derivation of K .f l

Amendment No. f , % 119 3.12-7 i

. DAEC-1 4.12 BASES: CORE THERMAL LIMITS 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 minimum recirculation pump speed and the moderator void content will be very snall. 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 state. relative to MCPR. During initial start up testing of the plant, a MCPR evaluation will be made at 25%

thermal power level with minimum recirculation punp speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be 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 assures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.

Amendment No.119 3.12-8

' - DAEC-1 3.12 REFERENCES

1. Duane Arnold Energy Center Loss-of-Coolant Accident Analysis Report, NED0-21082-03, June 1984.

2. General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A**.

3. " Fuel Densification Effects on General Electric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NEDM-19735, August 1973.

4. Supplement 1 to Technical Reports on Densifications of General Electric -

Reactor Fuels, December 14,1973 (AEC Regulatory Staff).

5. Communication: V. A. Moore to I.S. Mitchell, " Modified GE Model for Fuel Densification," Docket 50-321, March 27,1974. -

6. R.B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NED0-10802).

7. General Electric. Company Analytical Model for loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K, NEDE 20566, August 1974.

S. Boiling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, NED0-24087, 77 NED 359, Class 1, December 1977.

9. Boiling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, Supplement 2: Revised Fuel Loading Accident Analysis, NED0-24087-2.

10. Boiling Water Reactor Reload-3 Licensing Anendment for Duane Arnold Energy Center, Supplement 5: Revised Operating Limits for Loss of -

_ Feedwater Heating, NED0-24987-5.

11. Letter, R. H. Buchholz (GE) to P. S. Check (NRC), " Response to NRC Request for Information on ODYN Computer Model," September 5,1980.

12. Letter, R. H. Buchholz (GE) to P. S. Check (NRC), "0DYN Adjustment

., Methods for Determination of Operating Limits," January 19, 1981.

13. Duane Arnold Energy Center Single Loop Operation, NED0-24272, July 1980. l

• • Approved revision number at time reload fuel analyses are performed.

Amendment Nog,,g,119 3.12-9 l i

. e -

1.4 i i , , , , ,

N -

@ K FACTOR .

a. f R

N 1.3 y

~

$ AUTOMATIC FLOW CONTROL l

1.2 _.

u x

MANUAL FLOW CONTROL w Scoop-Tube Set-Point Calibration g positioned such that -

a Flowmax = 102.5%

! "o = 107.0%

1.0 *

.= 112.0%

l = 117.0%

)

I I I I t l g 30 40 50 60 70 80 90 100 CORE FLOW, t DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT & POWER, COMPANY TECHNICAL SPECIFICATIONS K AS A FUNCTION OF f

CORE PLOW l PIGURE 3.12-1

Option B Option A

t 1.30 - - 1.30

- 1.28 1.26 - _

u, 1.25 - _ 1.25 t=0.75 _

. 3 -

m -

5 -

e 1.20 - _ 1.20 7 _

g -

3 -

1.15 - - 1.15 1.10 - - 1.10 I

0.0 O!2 I 0!4 I 0!6 I V!8 I 1.0 (based on tested measured scram time as defined in Reference 11)

DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS MINIMUM CRITICAL POWER RATIO (MCPR)

VERSUS T

FUEL TYPES

BP/P8X8R and ELTA l

! FIGURE 3.12-2 Amendment No.) #,) Y/, M 119 3.12-11 l

DAEC-1 Option B

~

Ootion A

1.30 - -

_ ,- 1.30

- 1.28 m - -

{s 1.25 -

_ 1.25 t u -

e 1.21 - -

- 1.20 - '

m - 1.20 m -

o _ -

l _

1.15 - -

- 1.15 i - -

t j 1.10 - _-

1

~ _ 1,3n 0.0 I ' I '

O!2 O!4 O!6 '

O.'8 1.0 1

' T (based on tested measured scram time as defined in Reference 4

DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER C0fiPANY

TECHNICAL SPECIFICATIONS MINIMUM CRITICAL POWER RATIO (MCPR)

VERSUS r .

FUEL TYPE: LTA 311 4

FIGURE 3.12-3'

- -) .

Amendment No.)(Jg.119 3'I2*l2 I

" j .. . -

. 3. . . v.y . -

y h . *

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

MAPLHGR vs FUEL EXPOSURE .

LTA311 14.0

-l 1 .5 1:.5 3

' 13.0

'2

' .o u

?

'Y ~

E 12.0 12 2 3 .o J -

(

a \

,8 e 4 3 K1 2

y  %

g 11.0 , \

en c N .s 5 8 f

~

10 0 g

a = 37 _

b

{

9.0 9 1

a.o 0.0 0.2 1.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 Planar Average Exposure (GWd/t)*

~/ When. core flow is equal to or less than 70% of rated, the MAPLHGR 1

shall not exceed 95% of the limiting values shown.

• 1 GWd/t = 1000 mwd /t DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS

' LIMITING AVERAGE PLANAR LINEAR HEAT GENERATION RATE AS A FUNCTION OF PLANAR AVERAGE EXPOSURE

}

FUEL TYPE: LTA 311' FIGURE 3.12-5

.)

Amendment No. )ff,M1I )tf, J W , 119 3.12-14 i

-..t ...

_ . . :V P . .'" .

MAPLHGR vs FUEL EXPOSURE P8DRB301L

-l 1:;.3 19 4 t -q .2 12.0 1E A 11 .5 11 .5 11 g 5 11.0 \

5 h$6 E $

E E 10.0 9

.  ; a- l s E m

3 3 9.0

$ N7 8.0 , ,

0.0 0.2 1.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 Planar Average Exposure (GWd/t)*

-1/ When core flow is equal to or less than 70% of rated, the MAPLHG.R shall not exceed 95% of the limiting values shown. Values shown are for two recirculation loops. Reduction factors for one recirculation loop operation are given in Section 3.12. A. s

• 1 GWd/t = 1000 mwd /t DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS l LIMITING AVERAGE PLANAR LINEAR HEAT GENERATION RATE AS A FUNCTION OF PLANAR AVERAGE EXPOSURE' FUEL TYPE: BP/P80RB301L FIGURE 3.1'2-6 Amendment No.)MI A , 119 3.12-15 l

v .

MAPLHGR vs FUEL EXPOSURE PSDPB289 13.o N

%I E 12 .1 12.o ,; ' l' 9 k - 11 .2 13 2 111p f 1 4 I

11.o q .3 Ih en E \

3 c 10.o y

_a 46 ,

"Z a :c g 9.0 8 9 2J 8.o 0.0 0.2 1.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 Planar Average Exposure (GWd/t)*

-1/ When core flow is equal to or less than 70% of rated, the MAPLHGR shall not exceed 95% of the limiting values shown. Values shown.are for two recirculation loops. Reduction factors for one recirculation loop operation are given in Section 3.12.A. ,

• 1 GWd/t = 1000 mwd /t DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS LIMITING AVERAGE PLANAR LINEAR HEAT GENERATION RATE AS A FUNCTION OF PLANAR AVERAGE EXPOS 0RE-FUEL TYPE: P80PB289 FIGURE 3.1'2-7 i

Amendment No. % 119 3.12-16 i L

MAPLHGR vs FUEL EXPOSURE 13.0 P8DRB299/ ELTA

-J -

2 3 u .2

.tn 12.0 j '

Nk 11.y 1 .5 t% /

.= 3 3 ,0, c .9 1c .9 ' 'S _o F8 g;

N N

ae N 3

$ 10.0 30 j7

~%

3:E

% 9.0 E

3 8.0 0.0 0.2 1.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 Planar Average Exposure (GWd/t)*

-1/ When core flow is equal to or less than 70% of rated, the MAPLHGR shall not exceed 95% of the limiting values shown. Values shown'are for two recirculation loops. Reduction factors for one recirculation loop operation are given in Section 3.12. A. ,-lexcept ELTA).

• 1 GWd/t = 1000 mwd /t DUANE ARNOLD ENERGY CENTER 104A ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS LIMITING AVERAGE PLANAR LINEAR HEAT GENERATION RATE AS A FUNCTION OF PLANAR AVERAGE EXPOSURE' FUEL TYPE: BP/P80RB299 and ELTA FIGURE 3.12-8 Amendment No.J W, M ) W ) M 119 3.12-17

MAPLHGR vs FUEL EXPOSURE 13.0 PSDRS284H bl t h.O 12 .0 gg 12.0

,,,y g x .m Eo

,q q t1 .2 11,.2 1 ;.y/' ,3 11.0 T8 I h] C .4

• b Ey 10.0 \q 5 e  %

3E 1 9.0 '

a s.O .

0.0 0.2 1.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 Planar Average Exposure (GWd/t)*

~1/ When core flow is equal to or less than 70% of rated, the MAPLHGR shall not exceed 95% of the limiting values shown. Values showrf are for two recirculation loops. Reduction factors for one recirculation loop operation are given in Section 3.12.A. s

• 1 GWd/t = 1000 mwd /t o

DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS LIMITING AVERAGE PLANAR LINEAR HEAT GENERATION RATE AS A FUNCTION OF PLANAR AVERAGE EXPOSURE' FUEL TYPE: P80RB284H FIGURE 3.1'2-9 Amendment No. M 119 3.12-18

- - _ _