ML20126E527

Amend 120 to License DPR-49,revising Tech Specs to Permit Implementation of Aprm,Rod Block Monitor & Tech Spec Program Improvements & Extended Load Line Limits for Uprating Energy Ctr

ML20126E527
Person / Time
Site:	Duane Arnold
Issue date:	05/28/1985
From:	Vassallo D Office of Nuclear Reactor Regulation
To:
Shared Package
ML112371149	List: ML20126E527
References
NUDOCS 8506170114
Download: ML20126E527 (62)

Text

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UNITED STATES 3,,

NUCLEAR REGULATORY COMMISSION o

h WASHINGTON, D. C. 20555 g

/a IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 120 License No. DPR-49 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Iowa Electric Light and Power Company, et al, dated August 17, 1984, January 11 and March 15, 1985, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (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; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; a'nd

~~

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

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 amended to read as follows:

8506170114 850528 yDR ADOCK 05000331 PDR 1

. (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.120, 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 NUC AR 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 D e l

l

ATTACHMENT TO LICENSE AMENDMENT N0. 120 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 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 vii 3.1-3 3.3-20*

3.12-8 3.12-9 3.12-10 viii*

3.2-2 3.6-7 3.12-11 1.0-2 3.2-2a*

3.6-7a 3.12-12 1.0-5 3.2-16 3.6-7b 3.12-13 1.1-1 3.2-17 3.6-31 3.12-14*

3.12-15 3.12-16 1.1-2 3.2-18 3.6-34 3.12-17 1.1-3 3.2-41 3.12-1 3.12-18 1.1-5 3.2-42 3.12-2 3.12-19 1.1-10 3.3-5 3.12-3 3.12-20*

3.12-3a 1.1-12 3.3-7a 3.12-4 3.12-21*

1.1-14 3.3-16 3.12-5 3.12-22*

1.1-15 1.1-19 3.3-17 3.12-5a 3.12-23*

1.1-20 3.3-18 3.12-6 3.1-1 3.3-19 3.12-7

• new page

DAEC-1 TECHNICAL SPECIFICATIONS LIST OF FIGURES Figure Number Title 1.1-1 Power / Flow Map 1.1-2 Deleted.

2.1-1 APRM Flow Biased Scram and Rod Blocks 2.1-2 Deleted 4.1-1 Instrument Test Interval Determination Curves 4.2-2 Probability of System Unavailability Vs. Test Interval 3.4-1 Sodium Pentaborate Solution Volume Concentration Requirements 3.4-2 Saturation Temperature of Sodium Pentaborate Solution 3.6-1 DAEC Operating Limits 4.8.C-1 DAEC Emergency Service Water Flow Requirement 3.12-1 Flow-Dependent Minimum Critical Power Ratio (MCPRp) l 3.12-2 Minimum Critical Power Ratio (MCPR) versus T (Fuel Types:

BP/P8X8R and ELTA) 3.12-3 Minimum Critical Power Ratio (MCPR) versus T (Fuel Type:

LTA 311) s

~ 3.12-4 Power-Dependent Minimum Critical Power Ratio Multiplier (K )

l p

3.12-5 Limiting Average Planar Linear Heat Generation Rate (Fuel Type:

LTA 311) 3.12-6 Limiting Average Planar Linear Heat Generation Rate (Fuel Type BP/P80RB301L) 3.12-7 Limiting Average Planar Linear Heat Generation Rate (Fuel Type P8DPB289) 3.12-8 Limiting Average Planar Linear Heat Generation Rate (Fuel Types:

BP/P80RB299 and ELTA) 3.12-9 Limiting Average Planar Linear Heat Generation Rate (Fuel Type P80RB284H)

RTS-182 vii 01/85 Amendment No. 120

DAEC-1 Figure Number T_i tl e 3.12-10 Flow-Dependent Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Multiplier (MAPFAC )

p 3.12-11 Power-Dependent Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Multiplier (MAPFAC )

p 3.12-12 Flow-Dependent Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Multiplier (MAPFAC ) for LTA-311 p

3.12-13 Flow-Dependent Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Multiplier (MAPFAC ) for SLO p

6.2-1 DAEC Nuclear Plant Staffing

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l RTS-182A viii 03/85 Amendment No.120

DAEC-1 5.*

OPERABLE-0PERABILITY A system, subsystem, train, ccmponent or device shall be OPERABLE or have OPERABILITY when it is capatie of performing its specified function (s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

6.

OPERATING Operating means that a system or component is performing its intended functions in its required manner.

1 7.

IMMEDIATE Immediate means that the required action will be initiated as soon as practical considering the safe operation of the unit and the importance of the required action.

8.

REACTOR POWER OPERATION Reactor power operation is any operation with the mode switch in the "Startup" or "Run" position with the reactor critical and above 1% rated power.

a) SINGLE LOOP OPERATION (SLO): REACTOR POWER OPERATION with only one of the two recirculation loops in operation.

9.

HOT STANDBY CONDITION Hot standby condition means operation with coolant, temperature greater than 212*F, reactor vessel pressure less than 1055 psig, and the mode switch in'the Startup/ Hot Standby position.

10. COLD CONDITION Reactor coolant temperature equal to or less than 212*F.

11. HOT SHUTDOWN The reactor is in the shutdown mode and the reactor coolant temperature greater than 212*F.

12. COLD SHUTDOWN The reactor is in the shutdown mode, the reactor coolant temperature equal to or less than 212*F, and the reactor vessel is vented to atmosphere.

RTS-182A 1.0-2 03/85 Amendment No.120

DAEC-1

19. ALTERATION OF THE REACTOR CORE (CORE ALTERATION)

The addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a componen~t to a safe conservative position.

20. REACTOR VESSEL PRESSURE Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.

21. THERMAL PARAMETERS a.

Minimum Critical Power Ratio (MCPR) - The value of critical power ratio (CPR) for that fuel bundle having the lowest CPR.

b.

Critical Power Ratio (CPR) - The ratio of that fuel bundle power which would produce boiling transition to the actual fuel bundle power.

c.

Transition Boiling - Transition boiling means the boiling regime between nucleate and film boiling.

Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.

d.

Limiting Control Rod Pattern - A limiting control rod pattern for rod withdrawal error (RWE) exists when a) core thermal power is greater than or equal to 30% of rated and less than 90% of rated (30% < P < 90%) and the MCPR is less than 1.70, or b) core thermal

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power is greater than or equal to 90% of rated (P > 90%) and the

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MCPR is less than 1.40.

e.

Li.near Heat Generation Rate - The heat output per unit length of fuel pin.

f.

Fraction of Rated Power (FRP) - The fraction of rated power is the ratio of core thermal power to rated thermal power of 1658 MWth.

g.

Total Peaking Factor (TPF) - The ratio of local LHGR for any specific location on a fuel rod divided by the core average LHGR associateo with the fuel bundles of the same type operating at the core average bundle power.

h.

Maximum Total Peaking Factor (MTPF) - The largest TPF which exists in the core for a given class of fuel for a given operating condition.

01/85 RTS-182 1.0-5 Amendment No.120

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

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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

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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 psig A.

Neutron Flux Trips and Core Flow > 10% of Rated 1.

APRM High Flux Scram When The existence of a minimum In Run Mode, critical power ratio (MCPR) less than 1.07 for two recir-The APRM scram trip l

culation loop operation [1.10

, setpoint shall be as shown for SINGLE LOOP OPERATION on Figure 2.1-1 and shall (SLO)] shall constitute be:

violation of the fuel cladding integrity safety limit.

S < (0.58W + 62) l B.

Core Thermal Power Limit with a maximum setpoint of 120% rated power at 100%

(Reactor Pressure,< 785 psi 9 rated recirculation flow or tw or Core Flow < 10% of Rated Q

p U

When the reactor pressure is

< 785 psig or core flow is 5 < (0.58W + 58.5)

Tess than or equal to 10% of rated, the core thermal power for SLO.

shall not exceed 25 percent of rated thermal power.

RTS-182A 1 1-1 03/85 Amendment No.120

DAEC-1 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING C.

Power Transient Where:

S = Setting in percent of rated power (1,658 MWt)

To ensure that the Safety Limits established in Specification W = Recirculaticn loop flow 1.1.A and 1.1.8 are not exceeded, in percent cf rated flow.

each required scram shall be Rated recirculation loop initiated by its primary source flow is that signal. A Safety Limit shall be recirculaticn loop flow assumed to be exceeded when scram which corresponds to is accomplished by a means other 49x106 lb/hr core flow, than the Primary Source Signal.

NOTE:

This setting assumes D.

With irradiated fuel in the operation within the basic reactor vessel, the water level thermal design criteria. These i-shall not be less than 12 in, criteria are LHGR < values given above the top of the normal in Section 3.12.B and MCPR >

active fuel zone. Top of the values as given in Section 7.12.C.

active fuel zone is defined to be If it is determined that either of 344.5 inches above vessel zero these design criteria is being (see Bases 3.2).

violated during operation, IMMEDIATE action must be taken to return to operation within these criteria.

2.

APRM High Flux Scram When in the REFUEL er STARTUP and H0T STANDBY MODE, tr e APRM scram,-

shall be set at less than or equal to 15 percent of rated r

power.

RTS-182 1 1.2 01/85 Amendment No. 120

=. ---

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DAEC-1 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 3.

IRM - The IRM scram shall be,. set l

at less than or equal to 120/125 of full scale.

8.

Scram and

. > 514.5 incnes Isolation on above 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.

s

• O RTS-182 1.1-3 01/85 Amendment No.120

DAEC - 1

1.1 BASES

FUEL CLADDING INTEGRITY A.

Fuel Cladding Integrity Limit at Reactor Pressure > 785 psig and Core Flow > 10% of Rated The fuel cladding integrity safety limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate r

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 uncertainti in the value of the critical power. Therefore, the fuel cladding integrity safety limit is defined as the critical power rat,io in the limiting fuel assembly for which more than 99.9% of the fuel rods"in the cere are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is generically determined in Reference 1, for two recirculation loop operation. This Safety Limit MCPR is increased by 0.03 for SLO.

RTS-182A 1.1-5 03/85 Amendment No. 120

DAEC-1 For analyses of the thermal consequences of the transients the MCPRs stated in Section 3.12 as a limiting condition of operation bound those which are conservatively assumed to exist prior to initiation of the transients.

As discussed in Reference 2, the core-wide transient analyses for SLO l

1s conservatively bounded by two-loop operation analyses and the flow-dependent rod block and scrsn setpoint equations are adjusted for SLO.

l 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 summary:

1.

The abnormal operational transients have bben analyzed to a power level of 102P. of 1658 MWt.

11.

The licensed maximum power level is 1658 MWt.

iii.

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

RTS-132A 1.1-10 03/85 Amendment No.120

DAEC-1 scram trip provides additional margin. An increase in the APRM l

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 effect 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.

s RTS-182 1.1-12 01/85 Amendment No. 120

DAEC-1 APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressure is greater than 850 psig.

1 s

RTS-182 1.1 14 01/85

-Amendment No. 120

n DAEC-1 3.

IRM l

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 instrunent were on range 5, the scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accommodate the increase in power level, the scram trip setting is also ranged up. The most significant sources of reactivity change during the power 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 RTS-182 1.1-15 01/85 Amendment No. 120

1N o'

l RATED WER = 1844 Mm APRM SCRAM = 0.58W + S2

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RATED FLOW = 49 Meathe p

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APRM MOD BLOCK = 0.58W

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(100,87)

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80

=

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6 w

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3

.----__{

f.

g 100% LOAD LINE 5

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l CORE THERMAL POWER l

SHADED AREA INDICATES ao LIMIT WHEN REACTOR REGION EXPANDED SY mESSURE IS < 785 X ENDED LOAD LINE PSIG OR CORE FLCW f

< 105 OF RATED 25

/

NATlRAL CIRCULATION w

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o 0

20 to so go 100 t

CCRE FLCW t% OF MATE 03 e

DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT &_ POWER CO1PANY TECHNICAL SPECIFIC TIONS APRM FLOW BIAS SCRAM RELATIONSHIP TO NORMAL OPERATING CONDITIONS FIGURO 1.1-1 RTS-182 Amendment No. 120 1*1 19 01/85

120

/

/

NORMAL

• APRM SCRAM TRIPg /

$1NGLE LOOP

• APR $ CRAM TRIP

• g 90

>j 7

7

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f j

MAL

• APRM ROD OLCCK TRIP

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ct 60

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-eSING E LOOP APRM ROD 8 OCK TRIP b

50 IE

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3 40 z

y 30 o

U 20 10 0

i O.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0110.0120.0 W - RECIRCULATION LOOP FLOW (% RATEO)

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n Op DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT & POWER COMPANY TECHNICAL SPECIFICATIONS Core Power Vs Recirc Loop Flcw FIGURE 2.1-1 RTS-182 1.1-20 01/85 Amendment No,'120

4 OAEC-1 LIMITING CONDITIONS FOR OPERATIONSURVEILLANCE RE0VIREMENT 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability:

Acolicability:

Applies to the instrumentation and Applies to the surveillance of associated devices which the instrumentation and initiate a reactor scram, associated devices which initiate reactor scram.

Obiective:

Ob.iective:

To assure the operability of the reactor protection To specify the type and frecuency system.

of surveillance to be applied to the protection instrumentation.

Specification:

Specification:

A.1 A.

The setpoints, minimum number Instrumentation systems shall be of trip systems, and minimum functionally tested and number of instrument channels calibrated as indicated in Tables that must be operable for 4.1-1 and 4.1-2 respectively.

each position of the reactor

.2 mode switch shall be as oiven Response time measurements (from in Table 3.1-1.

The designed actuation of sensor contacts or system response times from trip point to de-energization of the openino of the sensor scram solenoid relay) are not contact up to and includino part of the normal instrument calibration.

The reactor trip the opening of the trip actuator contacts shall not system response time of each exceed 50 milliseconds, reactor trip function shall be demonstrated to be within its As a minimum, the reactor limit at least once per 18

months. Each test shall include protection system instrumentation channels of at least one logic train such

^

Table 3.1-1 shall be operable that both locic trains are tested with response times as shown at least once per 36 months and in Table 3.1-2.

one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function.

.3 When it is determined that a channel has failed in.the unsafe condition, the other'RPS channels that monitor the same variable shall be functionally RTS-182 3.1 -1 01/85 Amendment No. 120

I TABLE 3.1-1 gg REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT gm

$5 "W

Minimum No.

Modes in Which Number of a

of Operable Function Must be Instrument Instrument Operable Channels Channels for Provided by Refuel Startup Run Design Action (1) g Trip System (1)

Trip Function Trip level Setting (6) o l.

Mode Switch in X

X X

1 Mode Switch A

Shutdown (4 sections) 1 Manual Scram X

X X

2 Instrument A

Channels 2

IRM High Flux

< 120/125 of Full X

X (5) 6 Instrument A

Tcale Channels 2

IRM Inoperative X

X (5) 6 Instrument A

L.

Channels i

2 APRM High Flu,x-Forltwo recirc loop X

6 Instrument A or B operation:

Channels

(.58W+62) (11) (12)

For SLO:

(.58W+58.5) (11) (12) 2 APRM Inoperative (10)

X X

X 6 Instrument A or 8 Channels 2

2 APRM Downscale

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

APRM High Flux in

< 15% Power X

X 6 Instrument A

i Startup S

Channels 3

2 High Reactor

< 1055 '.psig X(8)

X X

4 Instrument A

Pressure Channels

DAEC-1

' LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT C.

Control Rod Block Actuation C.

Control Rod Block Actuation 1.

SRM, IRM, APRM and Scram 1.

Instrumentation shall be

[

Discharge volume Roc Blocks functionally tested, calibrated i

and checked as indicated in The Limiting Conditions of Table 4.2-C.

Operation for the instrumenta-tion that initiates these System logic shall be control rod block are given functionally tested as in Table 3.2-C.

indicated in Table 4.2-C.

2.

Rod Block Monitor (RBM) 2.

When a Limiting Control Rod Pattern exists, an instrisnent (a)

Both RBM channels shall be functional test of the RBM

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demonstrated to be Operable shall be performed prior to prior to control rod with-ylthdrawal of the designated drawal when a Limiting Control rod (.s).,,

4_

Rod Pattern exists; otherwise, control rod withdrawal may take place with the RBM bypassed. A Limiting Control Rod Pattern exists when:

i) core thermal power is greater than or equal to 30% of rated and less than 90% of rated (30% < P <

90%)andtheMinimiim Critical Power Ratio (MCPR)islessthan1.70, or

11) core thermal power is greater than or equal to 90% of rated (P > 90%) and the MCPR is less~than 1.40.

~

When a Limiting Control Rod Pattern exists:

With one RBM channel inoper-able, control rod withdrawal shall be blocked within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, unless Operability is restored within this time period.

With both RBM channels inoper-able, control rod withdrawal shall be blocked until Operability of at least one channel is restored.

Amendment No. 120 3.2-2

DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT (b) The RBM control rod block set-points are given in Table 3.2-C.

The upscale High Power Trip Setpoint shall be applied when the core thermal power is greater than or equal to 85%

of rated (P > 85%). The up-scale Intermediate Power Trip Setpoint shall be applied when the core thermal power is greater than or equal to 65%

of rated and less than 85% of rated (65% < P < 85%). The upscale Low 7ower Trip Setpoint shall be applied when the core thermal power is greater than or equal to 30%

of rated and less than 65% of rated (30% < P < 65%).

The RBM can be Fypassed when core thermal power is less than 30%

of rated. The RBM bypass time delay (td2) shall be less than or equal to 2.0 seconds.

D.

Radiation Monitoring Systems -

D.

Radiation Monitorina Systems -

Isolation & Initiation Isolation & Initiation Functions Functions 1.

Steam Air Ejector Offcas 1.

Steam Air Ejector Offoas Syste;n System Instrumentation shall be (a) Except as specified in (b)

-functionally tesYed, calibrated below, both post treatment and checked as indicated in steam air ejector offgas Table 4.2.D.

system radiation monitors shall be operable during System logic shall be reactor power operation. The functionally tested as trip settings for the monitors indicated in Table 4.2-D.

shall be set at a value not to exceed the equivalent of the stack release limit specified in the Environmental Technical Specifications. The Steam Air Ejector isolation valves close immediately if the Steam Air Ejector Offgas Radiation Monitor output exceeds the trip setting.

(b)

From and after the date that one of the two steam air ejector radiation monitors is made or found to RTS-182 3.2-2a 01/85 l

Amendment No. 120

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TABLE 3.2-C i

INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS g5 Minimum No.

.m of Operable g.'.

Instrument g >R Channels Per Number of Instrument Channels g

Trip System (1)

Trip Function Trip Level Setting Provided by Design Actlon 2

APRM Upscale (Flow Biased)

For two recirc loop operation:

6 inst. Channels (1) 1 3

< (0.58 W + 50) (2)

For SLE:

1 (0.58W + 46.5) (2) i 2

APRM Upscale (Not in Run Mode) i12indicatedonscale 6 Inst. Channels (1) 2 APRM Downscale l

> 5 indicated on scale 6 Inst. Channels (1) 1 (7)

Rod Block Monitor

)

a) Upscale (Power Referenced) 2 Inst. Channels (1)

1) Low Power Trip Setpoint

< 115/12T of full scale

+

i

2) Intermediate Power Trip Setpoint T 109/12.t of full scale w

3) lingh Power Trip Setpoint T 105/125. of full scale co i

b) Downscale

> 94/125 of full scale c) RBM Bypass Time Delay (td2) (10) 1 2.0 seconds j

2 IRM Downscale (3)

> 5/125 full scale 6 Inst. Channels (1) 1 2

IRM Detector not in (8) l Startup Position 6 Inst. Channels (1) 2

.IRM Upscale 1108/125 6 Inst. Channels (1) 2 (5)

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

Startup Position l

2 (5)(6)

SRM Upscale i105 counts /sec.

4 Inst. Channels (1) o R

g 1

Scram Discharge Volume 124. gallons 1 Inst. Channel (9)

Water level-High 9

DAEC-1 NOTES FOR TABLE 3.2-C 1.

For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function.

The SRM and IRM blocks need not be operable in "Run" mode, and the APRM

[except for APRM Upscale (Not in Run Mode)] and RBM rod blocks need not be operable in "Startup" mode.

If the first column cannot be met for

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one of the two trip systems, this condition may exist for up to seven 1

days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longen than seven days, the system shall be tripped.

If the first column cannot be met for both trip systems, the systems shall be tripped.

2.

W is the recirculation loop flow in percent of design. Trip level setting is in percent of rated power (1658 MWt).

s 3.

IRM downscale is bypassed when it is on its lowest range.

4.

This function is bypassed when the count rate is > 100 cps.

RTS-182 3.2-17 01/85 Amendment No. 120 b

DAEC-1 5.

One of the four SRM input's may be bypassed.

6.

This SRM function is bypassed when in the IRM rance switches are on a range 8 or above.

7.

There are three upscale trip levels. Only one is applied over a G

specified range of core thermal power. All RBM trips are bypassed below 30%

~

of rated power.

An RBM channel will be considered to be inocerable if less than one-half of the required number of. LPRM inputs are available.

8.

This function is bypassed when the mode switch is placed in Run.

9.

If the number of operable channels is less than required by the minimum number of. operable instrument channels per trip system requirement, place the inoperable channel in the tripped condition within one hour.

w 10.

RBM bypass time delay (td2) is set low enough to assure minimum rod movemefit while upscale trips are bypassed.

RTS-182-3.2-18 01/85 Amendment No. 120

DAEC-1 t

i The instrumentation which initiates CSCS action is arranced in a dual bus system. As for other vital instrumentation arranced in this fashion, the Specification preserves the effectiveness of the system even during periods when maintenance or testing is being performed.

An exception to this is when logic functional testing is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawal so that the MCPR does not decrease below the Safety Limit. The trip logic for this fune:fon is 1 out of n:

e.g., any trip on one of six APRM's, six IRM's, or four SRM's will result in a rod block.

The minimum instrument channel recuirements assure sufficient instrumentation to assure the single failure criterion is met.

The minimum instrument channel requirement; for the RBM may,be reduced by one for maintenance, testing, or calibration.

This time period is only 3% of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.

The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially during operation RTS-182 3.2-41 01/85 Amendment No. 120

DAEC-1 at reduced flow. The APRM provides gross core protection; i.e.,

limits the gross core power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than safety limit.

The RBM rod block function provides local protection of the core; i.e., the prevention of boiling transition in a local region of the core, for a single rod withdrawal error from a Limiting Control Rod Pattern.

The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level.

A downstale indication on an APRM or IRM is an indication the I

instrument has failed or the instrument is not sensitive enough.

In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented. The downscale trips are set at 5 indicated on scale for APRM's and 5/125 full scale for IRM's.

4 The flow comparator and scram discharge volume high level components have only one logic channel and are not required for,

safety. The flow comparator must be bypassed when in SLO.

RTS-182 3.2-42 01/85 Amendment No. 120

DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT e.

If Specification 3.3.8.3a 1)

The correctness of the control through d cannot be met, the rod withdrawal sequence input reactor shall not be started, to the RWM computer shall be or if the reactor is in the verified.

run or startup modes at less than 30% rated power, it shall 2)

The RWM computer on line be brought to a shutdown diagnostic test shall be condition immediately.

successfully performed.

f.

The sequence restraints 3)

Proper annunciation of the imposed on the control rods selection error of at least one may be removed by the use of out-of-sequence control rod in the individual rod position each fully inserted group shall bypass switches for scram be verified.

testing only those rods which are fully withdrawn in the 4)

The rod block function of the 100% to 50% rod density range.

RWM shall be verified by withdrawing the first rod as an out-of-sequence control rod no more than to the block point.

c.

When required, the presence of a second licensed operator to verify the following of the correct rod program shall be verified.

4.

Control rods shall not be 4.

Prior to control rod withdrawal withdrawn for startup or for startup or during

~

refueling unless at least two refueling, verify that at least source range channels have an two source range channels have observed count rate equal to an observed count rate of at or greater than three counts least three counts per second.

per second.

5.

During. operation with Limiting 5.

When a Limiting Control Rod Control Rod Patterns, either:

Pattern exists, an instrument functional test of the RBM a.

Both RBM channels shall be shall be performed prior to operable, or withdrawal of the designated rod (s).

b.

With one RBM channel inoper-able, control rod withdrawal shall be blocked within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, unless Operability is restored within this time period, or c.

With both RBM channels inoper-able, control rod withdrawal shall be blocked until Operability of at least one channel is restored, l

RTS-182 3.3-5 01/85 Amendment No. 120

o DAEC-1 LIMlTlNG CONDITIONS FOR OPERATION l

SURVEILLANCE REOUIREMENT levels are less than or equal to three times their established baseline levels, 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 4

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, 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 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing core flow, and/or by initiating an orderly reduction of core thermal power by inserting control rods, o

e See Soecifications 3.6.F.2 for SLO.

A recirculation pump shall not be started while 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

' ~ **

SHUTDGWN condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• 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 i

shall be monitored.

RTS-182A 3.3-7a 03/85 l

Amendment No.120

DAEC-1 drives is enforced. This demonstration is made by performing the hardware functional test sequence. The Group Notch restraints are automatically removed above 30% power.

During reactor shutdown, similar surveillance checks shall be made with regard to rod group availability as soon as automatic initiation of the RSCS occurs and subsequently at appropriate stages of the control rod insertion.

d.

The Source Range Monitor (SRM) system performs no automatic safety system function; i.e., it has no scram function.

It does provide the operator with a visual indication of neutron level.

The consequences of reactivity accidents are functions of the initial neutron flux. The requirement of at least 3 counts per second assures that any transient, should it occur, begins at or

'above the initial value of 10-8 of rafed power used inthe an,alyses of transients cold conditions. One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A l

minimum of two operable SRM's are provided as an added conservatism.

The RBM provides local protection of the core; i.e., the,pr.even-e.

tion of boiling transition in a local region of the core, for a single rod withdrawal error from a Limiting Control Rod Pattern.

The trjp point is referenced to power.

This power signal is RTS-182 3.3-16 01/85 Amendment No. 120

i e.,

DAEC-1 provided by the APRMs. A statistical analysis of many single control rod withdrawal errors has been performed and at the 95/95 level the results show that with the specified trip settings, rod withdrawal is blocked at MCPRs greater than the Safety Limit, thus allowing adequate margin.

This analysis assumes a steady state MCPR of 1.20 prior to the postulated rod withdrawal error.

The RBM functions are required when core thermal power is greater than 30% and a Limiting Control Rod Pattern exists. When both 5

RBM channels are operating either channel will assure required withdrawal blocks occur even assuming a single failure of one channel. When a Limiting Control Rod Pattern exists, with one RBM channel inoperable for no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, testing of the RBM prior to withdrawal of control rods assures that improper control rod withdrawal will be blocked (Reference 2). Requiring at least half of the normal LPRM inputs to be operable assures

~

that the RBM response will be adequate to protect against rod r

withdrawal errors, as shown by a statistical failure analysis.

The RBM bypass time delay is set low enough to assure minimun rod movement while upscale trips are bypassed.

A Limiting Control Rod Pattern for rod withdrawal error (RWE) exists when (a) core thermal power is greater than or equal to 30% of rated and less than 90% of rated (30% f, P < 90%) and the MCPR is less than 1.70, or (b) core thermal power is greater than or equal to 90% of rated (P ) 90%) and the MCPR is less than 1.40.

RTS-182 3.3-17 01/85 Amendement No. 120

OAEC-1 During the use of such patterns, it is judged that testing of the i

RBM system prior to withdrawal of such rods to assure its operability will assure that improper withdrawal dces not occur.

It is the responsibility of the Reactor Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns. Other personnel qualified to perform this function may l

be designated by the Plant Superintendent, Nuclear.

3.

Scram Insertion Times The control rod system is designed to bring the reactor subcritical at i

a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than the safety limit.

i

" 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.

I i

i l RTS-182 3.3-18 01/85 Amendment No. 120 l

r

t.

DAE C-1 4

4 Reactivity Anomalies During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poision in supplementary contaol is burned.

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

Requiring a reactivity comoarison at the specified frequency assures that a comparison will be made before the core reactivity chance exceeds 1%AX.

Deviations in core reactivity areater than 4% 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 desian conditions of the 4,

reactor system.

I l

RTS-182 3.3-19 01/85 Amendment No.120

l DAEC-1 3.3 and

4.3 REFERENCES

1)

NEDO 24087-3, 78NE0265, Class 1. June 1978 " General Electric Boilino Water Reactor Reload 3 (Cycle 4) Licensing Amendment for Duane Arnold Energy Center, Supplement 3: Application of Measured Scram Times".

2)

" Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program fcr the Duane Arnold Energy Center," NEDC-30813-P, December,1984 RTS-182 3.3-20 01/85 Amendment No. 120

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENT

[.

b. The indicated value of care flow rate varies from the value derived form loep flow measurements by more than 10%.

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

2. Whenever there is rectreulation flow from the reactor in the Startup or Run mode, and one l

recirculation pump 15 operating, the diffuser to lower plenum oifferential pressure shall be checked daily and the differential l

pressure of an individual jet pump in a loop shall not vary from the mean of all jet pump differential pressures in that loop by more than 10%.

F.

Jet Pumo Flow Mis' match F.

Jet Pumo Flow Mismatch

1. When both recirculation 1.

Recirculation pump speeds

' pumps are in steady state shall be checked and logged at operation, the speed of the least once per day.

faster pump may not exceed 122% of the speed of the 2.

a. Prior to SLO and core slower pump when core power thermal power greater than is 80% or more of rated the limit specified in power or 135% of the speed Figure 3.3-1, establish of the slower pump when baseline APRM and LPRM*

core power is below 80% of neutron flux noise levels, rated power.

provided that baseline values have not

2. If Specification 3.6.F.1 cannot be met, one

• Detector levels A and C of one recirculation pump shall be LPRM string per core octant plus tripped. The reactor may detector levels A and C of one LPRM be started and operated, or string in the center of the core operation may continue in shall be monitored.

SLO provided that:

RTS-182A 3.6-7 03/85 Amendment No. 120

DAEC-1 LIMITING CONDITIONS FOR OPERATION l

SURVEILLANCE REOUIREMENT

a. MAPLHGR multipliers as been previously established indicated in section since the last core 3.12.A are applied.

refueling. Baseline values shall be established during

b. During SLO and core SLO and core thermal power thermal power greater less than or equal to the than the limit specified limit specified in Figure in Figure 3.3-1, core 3.3-1.

flow must be greater than or equal to 39f. of rated,

b. Prior to SLO and core flow and greater than 45% of rated, establish baseline core (i) the Surveillance plate AP noise levels with Requirements of core flow less,than or equal 4.6.F.2.a have not been to 45% of rated, provided satisfied, immediately that baseline values have initiate action to not been previously reduce core thermal established during SLO since power to less than or the last core refueling, equal to the limit specified in Figure 3.3-1 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or (ii) the Surveillance Requirements of 4.6.F.2.a have been satified, continue to determine the APRM and LPRM neutron flux levels at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and also within 30 minutes after the completion of a core thermal power increase 3

of at least 5% of rated core thermal power while operating in this region of the power / flow map.

If the APRM and/or LPRM*

neutron flux noise levels are greater than three times their established baseline values.

immediately initiate corrective action and restore the noise levels to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by

• Detector levels A and C of one LPRM string per core octant plus detector levels A and C of one LPRM string in the center of the core shall be monitored.

RTS 182A 3.5-7a 03/85 Amendment No. 120

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0t*REMENT increasing core flow and/or initiating an orderly reduction of core thermal power by inserting control rods,

c. During SLO and core flow greater than 45% of rated, and (i) the Surveillance Requirements of 4.6.F.2.b have not been satisfied, immediately initiate action to reduce core flow to less than or equal to 45% of rated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or

-(11) the Surveillance Requirements of 4.6.F.2.b have been satisfied, continue to determine core plate AP noise at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and also i

within 30 minutes after the completion of a core thermal power increase of at least 5% of rated thermal power.

If the i

core plate AP roise level is greater than 1.0 pst and 2 times its establish?1 baseline value, immediately 'jntt t ate corrective action and restore the noise levels to within the required limits within 2 hoars by decreasing core flow and/or initiating an orderly reduction,

of core thermal power by s

inserting control rods.

d. The' idle loop is isolated electrically by disconnecting the breaker to the recirculation pump motor generator (M/G) set drive motor prior to startup, or if disabled during reactor operattor., 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 startup of the idle rectrculation loop.

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

RTS-182A 3.6-7b 03/85 Amendment No. 120 i

e

DAEC-1 The jet pump flow deviation pattern derived from the diffuser c.

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.

Agreement of indicated core flow with established power-core flow relationships provides the most assurance that recirculation flow is i

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 summation of the flow indications for the sixteen individual jet pumps. The total core flow measuring instrumentation sums reverse i

j jet pump flow as though it were forward flow in the case of a failed i

jet pump. Thus the indicated flow is higher than actual core flow by at least twice the normal flow through,any backflowing jet pump.*

I Reactivity inventory is known to a high degree of confidence so that even if a jet pump f ailure occurred during a shutdewn period, t

subsequent power ascension would promptly demonstrate abnormal control rod withdrawal for any power-flow operating map point.

1 A nozzle-riser system f ailure could also generate the coincident f ail ~ re of a jet pump body; however, the converse is not true.

u l

• Note:

In the case of SLO, when the recirculatton pump is tripped, the flow l

through the inactive jet pumps is subtracted from the total jet pump flow, yielding the correct value for the total core flow,

)

RTS 182A 3.6 31 i

03/85 Amendment No. 120 l

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 SLO (Sec. 3.12, l

Ref. 11). Therefore, continuous operation under such conditions is appropriate. The reactor may also be operated in SLO up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without l

electrically isolating the idle loop. This short period of time permits corrective action to be taken to re-activate the idle loo'p or to implement the changes for continuous SLO.

l During periods of SLO, the idle recirculation loop is isolated by electrically l disarming the recirculation pump. This is done to prevent a cold water injection transient caused by an inadvertent pump start-up.

It is permissible l

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.

I 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 i

core flow to greater than or equal to 39% of rated, which corresponds to the core flow'et the 80% rodline with 2 recirculation pumps running at minimum speed, the region of the power / flow map where these oscillations are most s

likely to occur is avoided.

Individual APRM and/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 j

appropriate remedial action if such a condition exists.

RTS-182A 3.6-34 03/85 Amendment No. 120 1

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIDEMENTS 3.12 CORE THERMAL LIMITS 4.12 CORE THERMAL LIMITS Aeolicability Acolicability 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 conditions.

fuel rod operating Conditions.

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

fuel rods.

Specifications

_Soecific at ions A.

Maximum Averaae Planar Linear A.

Maximum Averaae Planar Linear Heat Generation Rate (MAPLNGR)

Heat 6eneration Rate (MAPLNGR) 1.

During reactor power The MAPLHGR for each type of fuel operation, the actual MAPLHGR as a function of average planar for each type of fuel as a exposure shall be determined function of average planar daily during reactor operation at exposure shall not exceed the

> 25% rated thermal power.

limiting value shown in Figs.

3.12-6, -7, -8 and -9 multiplied by the smaller of the two MAPFAC factors determined from Figs. 3.12-10 and 3.12-11.

For the central Lead Test I

Assembly (LTA 311), the actual MAPLHGR, as a function of average planar exposure, shall

+'

not exceed the limiting value shown in Fig. 3.12-5 multi-plied by the smaller of the two MAPFAC factors from Figs.

3.12-11 and 3.12-12.

2.

During SLO, the actual MAPLHGR for each type of fuel as a function of average planar exposure shall not exceed the limiting value shown in Figs.

3.12-5, 6,

7, 8 and 9 multiplied by the smaller of the two MAPFAC factors determined from Figs.

3.12 11 and 3.12 13.

RTS 182A 3 12 1 03/85 Amendment No. 120

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENT 3.

If at any time during reactor power operation (one or two loop) at >25% rated thermal power, it is determined by normal surveillance that the limiting value for 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 i

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 />.

4.

If the reactor is being operated in SLO 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 SHUTCOWN condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

4 5.

For either the one or two loop operating condition surveillance and I

corresponding action shall continue until the prescribed l

action is met.

B.

Linear Heat Generation Rate B.

Linear Heat Generation Rate (LH6s )

(_LNGR) 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 > 257.

rod in any BP/P8XSR or ELTA thermal power.

~

fuel assembly shall not exceed 13.4 KW/ft, while the LHGR of any rod in an LTA 311 fuel i

assembly shall not exceed 14.4 KW/ft.

8 RTS-182A 3.12-2 03/85 Amendment No. 120

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREME'lT 2.

If at any time during reactor power operation at >25% rated thermal power it is 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 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 />.

Surveillance and corresponding action shall continue until the prescribed limits are again being met.

C.

• Minimum Critical Power Ratio C.

Minimum Critical Power Ratio (MCFR)

(MCPR) 1.

During reactor power opera-MCPR shall be determined daily tion, the MCPR shall be equal during reactor power operation to or greater than the Operat-at > 25% rated thermal power and ing Limit MCPR, whicM is a folTowing any change in power function of core thermal level or distribution that would power, core flpw, fuel type cause operation with a Limiting and scram time r.

For core Control Rod Pattern as define #

thermal power gr(ea)ter than or in Section 3.2.C.2(a). During equal to 25% of rated and less operation with a Limiting Control than 30% of rated (25% < P <

Rod Pattern, the MCPR shall be 30%), the Operating LimTt MCPR

' determined at,,least once per 12 15 given by Fig. 3.12-4 For

hours, core thermal power greater than or equal to 30% of rated (P > 30%), the Operating Limit MCPIT is the greater of either:

a) The applicable flow-dependent MCPR (MCPR )

p determined from Figure 3.12-1, or b) The appropriate rated power MCPR from Figures 3.12 2 and 3.12-3

[MCPR(100))multipliedby the applicable power.

dependent MCPR multiplier RTS 182A 3.12-3 03/85 l

Amendment No. 120 i

DAEC-1 e

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENT (K ) determined from n

Fi@ure 3.12-4.

2.

During SLO with core thermal power greater than or equal to 25% of rated, the Operating Limit MCPR is increased by adding 0.03 to the above determined Operating Limit MCPR.

3.

If at any time during reactor power operation (one or two recire, 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 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 4.

If the reactor is being operated in SLO, 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 SHUT 00WN condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

5.

For either the one or two recirc.

loop operating condition survetilance and corresponding action shall continue until the l

prescribed action 15 met.

I RTS-182A 3.12-3a 03/85 l

Amendment No.120 9

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 1

analyses are performed using General Electric calculational models l

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 j

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 j

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 1

temperatures are within the 10CFR50.46 limit.

The flow-dependent correction factor (Figure 3.12-10) applied to the MAPLHGR limits at rated conditions assures that (1) the 10CFR50.46 limit would not be exceeded during a LOCA initiated from less than rated core flow conditions and (2) the fuel thermal-mechanical design criteria would be met during abnormal operating transients initiated from less than rated core flow conditions (Reference 14).

i RTS-182 3.12-4 01/85 Amendment No. 120

DAEC-1 The power-dependent correction factor (Figure 3.12-11) applied to

i the MAPLHGR limits at rated conditions assures that the fuel thermal-mechar.ical design criteria would be met during abnormal operating transients initiated from less than rated power conditions (Reference 14).

1 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 SLO were derived in Reference 13.

i i

8.

Linear Heat Generation Rate (LHGR) a This specification assures that the linear heat generation rate in any rod is less than the design linear heat eneration rate and that j

the fuel cladding 1% plastic diametral strain linear heat generation rate is not exceeded during any abnormal operating

[

l transient if fuel pellet densification is postulated.

The power 1

~

spike penalty specified is based on the analysis presented in j

Section 3.2.1 of Reference 3 and in References 4 and 5, and assumes a linearly increasing variation in axial gaps between core bottom l

and top, and assures with a 95% confidence, that no more than one fuel rod exceeds the design linear heat generation rate due to f

L power spiking.

The LHGR as a function of core height shall be checkeddailyduringreactoroperationat125%powertodetermine l

if fuel burnup, or control rod movement has caused changes in power i

RTS-182A 3.12-5 03/85 l

Amendment No. 120 l

.i DAEC-1 distribution. For LHGR to be a limitino value below 25% rated thermal power, the Maximum Total Peaking Factor (MTPF) would have to be greater than 10 which is precluded by a considerable margin when employing any permissible control rod pattern.

C.

Minimum Critical Power Ratio (MCPRI 1.

Operating Limit MCPR The required operatino limit MCPR's at steady state operating conditions as specified in Specification 3.12.C are derived from the established fuel cladding integrity Safety Limit MCPR value, and an analysis of abnormal operational transients (2). For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit it is required that the resul.t.ing MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrtnent trip settings given in Specification 2.1.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in critical power ratio (CPR).

The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

RTS-182 3.12 5a 01/85 l

l Amendment No. 120

DAEC-1 i

The limiting transient, which determines the required steady state MCPR limit, is the transient which yields the largest ACPR.

The minimum Operating Limit MCPR of Specification 3.12.C bounds the sum of the Safety Limit MCPR and the largest ACPR.

l TherequiredMCPRsatratedpower[MCPR(100)]aredeterminedby the methods described in References 11 and 12.

These limits

'were derived by using the GE 678 scram times, given in Section 3.3.C which are based upon extensive operating plant data, as well as GE test data.

The ODYN Option B scram insertion times were statistically derived from the 678 data to ensure that the 7

resulting Operating Limit from the transient analysis would, I

with 95% probability at the 95% confidence level, result in the Safety Limit MCPR not being exceeded. The scram time parameter j

(t), as calculated by the following formula, is a measure of the conformance of the actual plant control rod drive performance to l

that used in the ODYN Option-B li, censing basis:

l t ave - tb t '-

Ta

- Tb where: tave = average scram insertion time to Notch 38, as measured by surveillance testing tb

= scram insertion time to Notch 3B used in the ODYN Option B Licensing Basis.

i i

ta

= 678 scram insertion time to Notch 38 i

As the average scram time measured by surveillance testing (tave),exceedstheODYNOptionBscramtime(tb),theMCPRsat j

rated power (MCPR(100)] must be adjusted using Figures 3.12 2 I

e and 3.

OU85 l

n nt No,120

I DAEC-1 2.

MCPR Limits for Other Than Rated Power and/or Rated Flow Conditions At less than 100% of rated power and/or flow the required Operating Limit MCPR is the larger value of the flow-dependent MCPR (MCPRp) or the power-dependent multiplier (K ) times the rated power MCPR [MCPR(100)]

p at the existing core power / flow state.

The required Operating Limit MCPR is a function of flow in order to protect the fuel from inadvertent core flow increases such that the Safety Limit MCPR requirement can be assured.

The MCPR s were calculated such that, for the maximum core flow rate y

and core thermal power along a conservative load line, the limiting bundle's relative power was adjusted untti the MCPR was slightly above the Safety Limit MCPR. Using this relative bundle power, the MCPRs were calculated at different points along this conservative load line

~

corresponding to different : ore flows.

The' resulting MCPRp'i' are

~

given in Figure 3.12-1.

For operation above 30% of rated thermal power, the core power-dependent MCPR operating limit is the rated power MCPR [MCPR(100)], multiplied by the f actor given in Figure 3.12-4, i.e., X. For operation below 30%

p of rated thermal power, where the direct scrams on turbine control valve fast closure and turbine stop valve closures are bypassed, absolute MCPR limits are established.

This limit is taken directly from Figure 3.12-4 This limit protects the fuel from abnormal operating transients, RTS-182 3.12-7 01/85 Amendment No.120

DAEC-1 i

I including localized events, such as a rod withdrawal error, other than i

those resulting from inadvertent core flow increases, which are covered.

by the flow-dependent MCPR limits. This power-dependent MCPR limit was i

developed based upon bounding analyses for the most limiting transient c

i at the given core power level.

Further information on the MCPR f

operating limits for off-rated conditions is presented in Reference 14.

At thermal power levels less than or equal to 25% of rated thermal i

i l

power, operating plant experience indicates that the resulting MCPR I

value is in excess of the requirements by considerable margin, j

Therefore, monitoring of MCPRs below this power level is unnecessary.

The daily monitoring of MCPRs above 25% of rated thermal power is j

sufficient, since power distribution shifts are very slow, provided that no significant changes in core flow or control rod pattern have taken j

place.

l s

i l

During SLO, the Operating Limit MCPR must be increased by 0.03 to j

account for'the increased uncertainty in the core flow and Transversing L*

In-coreProbe(TIP)readingsusedinthestatisticalanalysestoderive l

i the Safety Limit MCPR (see Reference 13).

}

l I

i I

I i

t t

I l

1' RTS 182A 3.12 8 03/85 Amendment No. 120 i

- - = - - - -

DAEC-1 4

r

)

j 4.12 BASES: CORE THERMAL LIMITS 4

j C.

Minimum Critical Power Ratio (MCPR) - Surveillance Requirement i

j I

At core thermal power levels less than or equal to 25%, the reactor will be f

i operating at minimum recirculation pep speed and the moderator void content i

will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin.

e i

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 257.

thermal power level with minimum recirculation pep speed.

The MCPR margin j

will thus be demonstrated such that future MCPR evaluation beinw this power

]

level will be shown to be unnecessary. The daily requirenent for 1

3 calculating MCPR above 25% rated thermal power is sufficient since power i

distribution shifts are very slow when there have not been significant power i

s i

or control rod changes.

The requirement for calculating MCPR when a f

limiting control rod pattern is approached assures that MCPR will be known

]

following a change in power or power shape (regardless of magnitude) that l

could place operation at a thermal limit.

i i

RTS 182 3.12 9 01/85 Amendment No. 120 I

DAEC-1 3.12 REFERENCES 1.

Quane 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, NE0M-19735, August 1973.

4.

Supplement : to Technical Reports on Densifications of General Electric Reactor Fuels, December 14, 1973 (AEC Regulatory Staff).

5.

Communication:

V.A. Moore to 1.5. 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 (NE00-10802).

7.

General Electric Company Analytical Model for loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K, NEDE-20566, August 1974 8.

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

9.

Bolling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, Supplement 2: Revised Fuel Loading Accident Analysis, NE00-24087-2.

10. Boiling Water Reactor Reload 3 Licensing Anendment for Duane Arnold Energy Center, Supplement 5: Revised Operating Limits for Loss of Feedwater Heating, NE00 24987-5.

s

11. Letter,R.H.Buchholz(GE) top.S. Check (NRC),"ResponsetoNRC Request for Information on 00VN 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, NE00 24272, July 1980.

14. " Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement ( ARTS) Program for the Ouane Arnold Energy Center," NEDC 30813 P December, 1984.

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

RTS 182 3.12 10 01/85 Amendment No. 120

t e.

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DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS MINIMUM CRITICAL POWER RATIO AT RATED POWER [MCPR(100)]. VERSUS T FUEL TYPE: BP/P8X8R AND ELTA FIGURE 3.12-2 RTS-182 3.12-12 Amendment No. 120 i

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RTS-182 3,12 14 01/85 Amendment No. 120

MAPLHGR vs FUEL EXPOSURE LT A311 14.0 1 d.5 1 $.5

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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)*

• 1 GWd/t = 1000 mwd /t DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY 4

TECHNICAL SPECIFICATIONS LIMITING AVERAGE PLANAR LINEAR HEAT GENERATION RATE AS A FUNCTION OF PLANAR AVERAGE EXPOSURE..

FUEL TYPE: LTA-311 i

FIGURE 3.12-5 RTS-182A 3.12-15 03/85 Amendment No. 120

e DAEC-1 MAPLHGR vs FUEL EXPOSURE i

PSDR8301 L 13.0 g

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• 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: BP/P80RB301L FIGURE 3.12-6 RTS-182 3.12-16 01/85 Amendment No. 120

e r.

DAEC-1 MAPLHGR vs FUEL EXPOSURE PSDP8289 13.0 w"

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• 1 GWd/t = 1000 mwd /t s

DUANE ARNOLO ENERGY CENTER IOWA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS LIMITING AVERAGE PLANAR LINEAR HEAT i

GENERATION RATE AS A FUNCTION OF PLANAR AVERAGE EXPOStJRE FUEL TYPE: P80PB289 l

FIGURE 3.12-7 l

l RTS-182 3.12-17 01/85 Amendment No. 120 w

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MAPLHGR vs FUEL EXPOSURE P8DRB299/ ELTA O

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~

1 0

DUANE ARNOLD ENERGY CENTER 10WA ELECTRIC LIGHT AND POWER COMPANY TECHNICAL SPECIFICATIONS i

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

FUEL TYPE:

BP/P8DRB299 and' ELTA l

l FIGURE 3.12-8 RTS-182A 3.12-18 03/85 l

Amendment No. 120

t e

DAEC-1 MAPLHGR vs FUEL EXPOSURE PSDRB254H 7

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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: P80RS284H FIGURE 3.12-9 RTS-182 3.12-19 01/85 Amendment No. 120

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I DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT AND POWEll COMPANY TECHNICAL SPECIFICATIONS FLOW-OEPENDENT MAXIMtN AVERAGE PLANAR LINEAR HEAT GENERATIDN RATE (MAPLHGR)

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RTS-182 3.12-21 01/85 Amendment No. 120

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MULTIPLIER (MAPFAC l FOR LTA-311 i

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DIMNE ARNOLD ENERGY CENTER I

1(hfA ELECTRIC LIGHT AND POWER C0pFANY TECHNICAL SfECIFICATIONS FLON CERNOENT MAxlMlM AKRAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGil MillilPLIER (MAPFAC l FOR SLO FIGtRE 3.12p13

!