Text

Proposed Tech Spec Change RTS-74,allowing Single Loop Operation W/Bypass Flow Holes Plugged
	ML20099K559
Person / Time
Site:	Duane Arnold, 05000000
Issue date:	01/12/1977
From:	; IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:	;
Shared Package
ML20099K024	List: ML20099K053; ML20099K111; ML20099K116; ML20099K155; ML20099K172; ML20099K232; ML20099K268; ML20099K308; ML20099K324; ML20099K349; ML20099K371; ML20099K494; ML20099K508; ML20099K512; ML20099K542; ML20099K556; ML20099K559;
References
	FOIA-84-105 NUDOCS 8411290479
	Download: ML20099K559 (20)
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e PROPOSED CHANGE RTS-74 TO DAEC TECHNICAL SPECIFICATIONS I. Af fected Technical Soecifications Appendix A of the Technical Specifications for the DAEC (DPR-49) provides as follows:

Specifications 2.1.A, 3.6.F and 3.12.A specify certain limits on reactor operation concerning APRM High Flux Scram, APRM Rod Block, MaxLnum Average Planar Linear Heat Generation Rate (MAPLEGR) and recirculation loop operability.

II. Prooosed Change in Techn! cal Soecifications The licensees of DPR-49 propose the following change in the Technical Specifications set forth in I above:

Delete portions of the Technical Specifications and replace with the attached sheets as appropriate.

III. .Tustification for Prooosed Change This change is proposed so that reactor operation can.be continued for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after one recirculation loop is made or found to be inoperable. The analysis and justification for single loop operation is contained in the attached document NEDo-21226, August 1976, "Duane Arnold Energy Center License Amendment Sub- ,!

mittal for Single-Loop operation with the Bypass Flow Holes Plugged".

IV. Review Procedures This proposed change has been reviewed by the DAEC Operations Committee and Safety Committee which have found that this proposed.

change does not involve a significant hazards consideration.

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SAFETY-LIMIT LIMITING SAFETY'SYSTE1! SETTING-1.1 IUEL CLADDING INTEGRITY 2.1 -FUEL CLADDING INT $GRITY l

~ Applicability: Applicability:

Applies to the inter-- Applies to trip settings of :)

the-instruments and. devices related variables associated with-fuel' . which are provided to_ pre-

. thermal behavior vent the reactor system

, safety limits from,being i

' exceeded.

Objective: Objective:

To establish limits To define the. level of'the which ensure the inte- process variables at which grity of the fuel automatic protective . action i cladding. . is initiated to prevent the fuel cladding integrity safety-limits from being

! exceeded. .

Specifications: Specifications:

The limiting safs:y system

(_ , settings shall be as speci-i fied below:

A. Reactor Pressure.> 785 A. Neutron Flux ~Trius psig and Core Flow

> 10% of Rated. 1.a. APRi! High Flux Scrac When In Run Mede.

'The existence of a mini- '

mum critical power For operation with a peaking .

ratio.(MCPR) less than -factor less then 2.61 (7 x 7 array) '

1.07 shall constitute or 2.43 (8 x 8. array), the APri violation of the fuel scram trip setpoint shall be as -

cladding integrity shown on Fig. 2.1-1 and shall be:

safety limit. -

(0.66W'+ 54)

B. Core Thermal Power S _4 Limit (Reactor Pressure ,

1785 psig ~or Core Flow wi'th a maximum setpoint of

'110% of Rated) -~

120% rated. power at 1005 rated recirculation flow or '

~

When the reactor pres- greater. -

.sure is -y 785 psig or .

. core flow is less than

' l'0% of rated, the core.

y thermal power shal1~not

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exceed 25 percent of rated thermal power. -

DAEC-l'

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SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING ~

C. Pouer Transient ..

To ensure.that the Where: S=Settinginpercent'oI{'

., Safety Limits establishe'd rated power W in Specification 1.1.A '(1,593 MWt).

and 1.1.B are not ex-l

_ceeded,.-each required W = Recirculation loop scram shall be initiated- flowlin percent of by'its primary source rated flow. Rated

, signal. A Safety Limit recirculationcloop shall be assumed to be .

flow is'thit~recircu-exceeded when scram is .

lation. loop flow;which accomplished by a means corresponds to 49x108 othef than the Primary lb/br core flow.

a Source Signal. '

7 MTPF = Actual'Maximam Total D. With irradiated fuel in peaking factor. '

the reactor vessel,. Se water level shall nct For a peaking factor. greater than

~

i be less than 12 in. 2.61 (7 x 7 array) or 2.43 (8 x - 8 3~

-above the top of the array), the APRM scram setpoint normal active' fuel zone- shall be:

, (' S < (0.66-W'+ 54)' IMTPF *U

. NOTE: . Thes'e settings assume b.

j operation within the basic

!. thermal;. design criteria.

These criteria are LHGR.S.

L 18.5 m/ft '(7; x .7 array) or- 13.4 1

W/ft (8 x 8 airay) and MCPR E values as indicated in Table 3.1'222 times K , where Kg is defined i by Figure 3.12 1. Therefore, at!

j full power, operation is not al-

~1 owed with total peaking factor greater than

even if the scran setting Q reduced.'. If it-is <

1 determined'that either of these. .

design criteria is being violated during operation, action cust be taken immediate1" to return to operation within these criteria.

b. APRM High Flux Scram When In Run -

o 3

Mode and Single Loop Operation.'

For single loop operatica with a peaking factor less than or equal

't'o '2 61 (7 'x 7 array) or 2 43

2.61-(7 x 7 array) or 2.43 (8 x 8 array), the APRM scram trip (-

[ -

(8 x-8 array) ,.y setpoint shall be: ' '

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SAFETY LIMIT LIMITING SAFETY SYSTEM SETTINO S gi (0.66 W + 50.7) with a maxi =um setpoint of 116.7 rated pcuer at 100%

rated recirculation flow or greater.

Where: S = Setting in percent of rated power (1,593 MWt)

W = Recirculation loop flew in percent of rated flow. Rated recircu-lation loop flow is that recirculation loop flow which6e rresponds to 49 x 10 lb/hr core flow.

MTPF = Actual Maxi =um Total peaking fcctor.

For a peaking factor greater than 2.61 (7 x 7 array) or i

2.43 (8 x 8 array), the APRM scram setpoint shall be:

S g6 (0.66 W + 50.7) {

I hTPF NOTE: These settings assume operation within the basic ther=al design criteria.

These criteria are LHGR 18.5 Kg/ft (7 x 7 array) or 13.4 KR/ft (8 x 8 array) and MCPR E values as indicated in Table 3.12-2 times K ,

where K isdefinedbyhigure 3.12- 1. g Therefore, operation is not allowed with total peaking factor greater than

even if the scram setting is reduced. If it is deter-mined that either of these design criteria is being s violated during operation, action =ust be taken 4- edi-

2.61 (7 x 7 array) or 2.43 ately to return to operation (8 x 8 array) within these criteria.

1.1-2a

DAEC-1

. , i SAFETY LDf1T LTMITING SAFETY SYSTEM SETTING 2 APRM High Flux Scram When in the REFUEL or STARTUP and HOT STANDBY MODE. The APRM scram shall be set at less than or equal to 15 per-cent of rated power.

3.a. APRM Rod Block When in Run Mode For operation with a peaking factor less then or equal to 2.61 (7 x 7 array) or ' 2.43 (8 x 8 array), the APRM Control Rod Block setpoint shall be:

S j$ (0.66 W + 42)

The definitions used above for the APRM scram trip apply.

For a peaking factor greater than 2.61 (7 x 7 array) or 2.43 (8 x 8 array), the APRM Control Rod Block setpoint shall be:

S 6 (0.66 W + 42)MTPF {*)

3.b. APRM Rod Block When in Run Mode and Single Loop Operation For single loop operation with a peaking factor less than or equal to 2.61 (7 x 7 array) or 2.43 (8 x 8 array), the APRM Control Rod Block setpoint shall be:

S di (0.66 W + 38.7)-

The cefinitions used above for the APRM scram trip apply.

For a peaking f actor greater '

than 2.61 (7 x 7 array) or 2.43 (8 x 8 array), the APRM Control Rod Block setpoint -

shall be:

i

-* 2.61 (7 x 7 array) or 2.43 (8 x 8 array), S 'di (0.66 W + 38.7)MTPF {*I '

DAEC-1 B g SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 4 IRM - The IRM scra= shall be set at less than or equal to 120/125 of full scale.

B. Scram and Iso- 2 514.5 lation on re- inches above actor low vessel zero water level (+12" on .

level instru- I ments)

C. Scram - turbine 6 10 per-stop valve cent valve closure closure l

D. Turbine control valve fast l closure scram shall occur within 30 milliseconds of the start of turbine control valve fast closure.

E. Scram - main i 10 per-steam line cent valve isolation closure valve F. Main steam D. 880 psig isolation valve closure nuclear system low pressure G. Core spray and 2; 363 inches LPCI actuation- above vessel reactor low zero (-139.5 water level inches indi-cated level)

H. HPCI and RCIC 21 464 inches actuation - above vessel reactor low zero (-38.5 water level inches indi-cated level)

I. Main steam 1l 464 inches isolation above vessel ,

valve closure- zero (-38.5 reactor low inches indi-water level cated level)

J- Main steam (E 10 inches isolation Eg vacuum valve closure-loss of main condenser vacuum

... DAEC-1 during operation. Reducing this' operating margin would in-crease the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting~ther=al stresses. Thus, the APRM scram trip setting was selected be-cause it provides adequate margin for the fuel cladding in-tegrity Safety' Limit yet allows operating margin that reduces the possibility of unnecessary scrams.

The scram trip setting must be adjusted-to ensure.that the LHGR transient peak is not' increased for any combination of MTPF and reactor core thermal power. The scram setting is adjusted in accordance with the fonsula in Specification 2.1.A.1, when the maxLnus total peaking factor is greater.

than the design value. This adjustment may be accomplished

, by increasing the APRM gain and thus reducing the slope inter-cept point of the flow referenced APR}t High Flux Scram Curve by the reciprocal of the APRM gain change.

Analyses of the ILniting transients show that no scram adjust-ment is required to assure MCPR ?. 1.07 when the transient is i

initiated from MCPR' g values as indicated in Table 3.12.2 An evaluation of operation with one recirculation loop out of-service demonstrates that the scram trip setpoint must be modi-fied to assure Safety Limits are not violated. 'This evaluatice is presented in reference (2).

2 APRM High Flux Scram (Refuel or Startuo and Hot Standbv Model .

For operation in these modes the APRM scram setting of 15 per--

cent'of rated power and the IRM-High Flux Scram provide ade-

-quate thermal margin.between the setpoint and the Safety. Limit,. l

DAEC-1 25 percent of rated. The ' margin is adequate to accommodate anticipated maneuvers associated with power plant 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, tempera-ture. coefficients are small, and control rod patterns are

(. constrained to be uniform by operating proceduras backed up by the rod worth minimizer and the Rod Sequence Control System. Worths of individual rods are very low in a uniform rod pattern. Thus, of all possible sources of reactivity in-put, uniform control rod withdrawal is the most probable cause of significant power rise.

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because ,

several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram icvel, the rate of power rise is no more than 5 percent of rated power per minute, and the AFRM system would be more than adequate to assure a scram before the power could exceed the safety linit. The 15 percent APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressure is greater s

than 880 psig.

3. APRM Rod Block (Run Modei Reactor pcwer level may be. varied by =oving control rods or i

1.1-18 I

DAEC-l'

>= .

by varying the recirculation flow rate. The.APRM system pro-vides a control rod block to prevent rod withdrawal beyond a given power level at constant recirculation f1'ow rate, and thus prevents.a MCPR less than 1.07.- This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents excessive reactor power level increase resulting from control rod withdrawal. The flow variable trip setting provides substantial margin fran fuel damage, assuming a steady-state operation at the trip setting, over the entire y recirculation flow range. The margin to the Safety Limit in-creases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM red block trip ' setting. The

. l actual power distribution in the core is established by speci- 'l fied control rod sequences and is monitored continuously by the in-core LPRM system. As with the APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum total peaking factor exceeds the design value, thus preserving the APRM rod block safety margin. As with the scram setting, this may be accomplished by adjusting the APRM gain.

An evaluation of operation with one recirculation loop out of service demonstrates that the APRM Rod Block trip setpoint must be modified to assare that a MCPR.less-than the Limiting Condi-tion for Operation does not occur. This evaluation is presented in reference (2).

1.1- 18a m...... .. . .. . .

4. IRM The IRM system consists of 6 chambers, 3 in each of o

the reactor pr'tection system logic channels. The IRM is a 5-decade instrument which covers the range g- of power level between that covered by the SRM and h, 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 dec,ade in size. The IRM scram trip setting of 120 divisions is active in each range of the IRM. Fcr example, if the instrument were on range 1, the scrma setting would be 120 divisions for that range; likewise, if the instrument were on range 5, the scrnm 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 reaccivity 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 an

%s IRM scram would result in a reactor shutdown well before any Safety Limit is exceeded.

_ _ - - - - _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - - - - 1

l a; '

DAEC-1.

that occur during normal or inadvertent isolation valve h closure. With the scrams set at.10 percent of valve closure, neutron flux does not increase. To protect.the main condanser against overpressure, a-loss of. condenser vacuum i.litiates 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 Pumps 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 conditions result in adequate safety margins'for both the fuel and the system pressure.

2.1 REFERENCES

1. Linford, R. B., " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor,"-NEDO-10802, Feb., 1973.

2. Duane Arnold Energy Center License Amendment Submittal for Single-loop Operation with the Bypass Flow Holes Plugged, NEDO-21226, August, 1976.

a u-n . .

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Y TABLE 3.1-1 REACTOR PROTECTION SYSTD1 (SCRAM) INSTRUMENTATION REQUIRD'ENT Minimum No.

of Operable !! odes in Which Number of Instrument Punction Must be Instrument Channels Operable Channels l for Trip Trip Level Provided j Systa . (1) Trip Function Setting Refuel Startup Run by Design Action (1)

(6) 1 Mode Switch in X X X 1 $1 ode Switch A Shutdown (4 Sections) ,

1 Manual Scram X X X 2 Instrument A Channels 2 IleI !!igh Plux :$ 120/125 of rull X X (5) 6 Instrument A Scale Channels g w >

. 2 IRM Inoperative X X (5) 6 Instrument H

A M-I Chanre o lr. O I

2 APRM liigh Plux See Specification X 6 Instrument A or B 2.1.A.1.a or 2.1.A.1.b Channels 2 APIN Inoperati*e (10) X X X 6 Instrument A or D Channels 2 APRM Downscale D 5 Indicated on Scale (9) 6 Instrument A or D Channels 2 APRM liigh Plux in 4 151 Power X X 6 Instrument A Startup Channels 2 Ifigh Reactor 41035 puig X(8) X X 4 Instrument A Pressure Channels

/

O

, DAEC-1

( 7. Not required to be operable when prinary containment integrity is not required.

8. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.

9. The APm4 downscale trip is automatically bypassed when the Im4 instrumentation is operable and not high.

10. To be considered operable, APRM's A, B, C and D must

'have at least 9 LPRM inputs while APRM's E and F must have at least 13 LPRM inputs. Additionally each APRM must have g at least 2 LPRM inputs per level.

A.

11. Deleted

12. Deleted

13. The design permits closure of any two lines without a scram being initiated.

14.

The trip setting and alarm setting for the Main Steam Line High Itudiation Monitor shall be :$ 6 X ande_ _l X, respectively, "o m i Rated Power Background during the period prior to acnieving 50 per cent rated power for the first time.

3.1-7 April 1974

DAEC-1 (i NOTES FOR TA3LE 3.2-B~

1. Whenever any CSCS subsystem is required by subsection 3.5 to be operable, there shall be two operable trip . systems. If the first column cannot be met for one of the trip systems, that trip system shall be placed in the tripped condition or the reactor shall be placed in the Cold shutdown Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. Close isolation valves in RCIC subsystem. l l

3. Close isolation valves in HPCI subsystem.

4. Inst.r raent setpoint corresponds to 18.5" above the top of active fuel.

{

5. HPCI has only one trip system for these sensors.

I

6. The relay drop-out voltage will be measured once per operating 15 cycle and the data examined for evidence of relay deterioration, s

3.2-15

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TABLE 3.2-C INSTRUMENTATION TilAT INITIATES CONTROL ROD BLOCKS l

.1 Minimum No.

of Operable Instrument Number'of Channels Per Trip System Instrument Channels Instrument Trip Level Setting Provided by. Design Action 2 APRM Upscale (Flow See Specification Biased) 6 Inst. Channels (1) 2.1.A.3.a or 2.1.A.3.b

.)

i 2 APRM Upscale (Not in Run 6 12 indicated on scale Mode) 6 Inst. Channels (1) 2 APRM Downscale D5 indicated on scale 6 Inst. Channels (1) .O 1 (7) Rod Block Monitor 2 I'nst. Channels (Flow Biased) 6 (0.66W +41) (TPP* )I I (1) b 1 (7) Rod Block Monitor D 5 indicated on scale 2 Inst. Channels (1)

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

Startup Position 6 Inst. Channels (1) 2- IRM Upscale 4108/125 6 Inst. Channels (1) 2 (5) SRM Detector not in (4) 4 Inst. Channels (1).

Startup Position 2 (5) (6) SRM Upscale 6 105 counts /sec. 4 Inst. Channels (1) 1 (7) Rod Block Monitor (Flev 2 Inst. Channels Biased) (Single Loop (2) (1)

Operction)

$ (0.66W +37.7) ( ,*pp) ,

{

2.61 (7:x 7. array) or 2.43 (8 x 8 array)

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS j(

b. The indicated value of core flow rate varies from the value derived frcm loop flow measurements by more than 10%.

c. The diffuser to lower plenum differential pressure read-ing on an individual jet

! pump varies from the mean of j all jet pump differential pressures by more than 10%.

2. Whenever there is recircu-lation flow with the reactor in the Startup or Run mode,.

and one recirculation pump is operating, the diffuser to lower plenum differential pressure shall be checked daily and the differential pressure of an individual jet pump in a loop shall not vary from the mean of all jet pump differential pres-sures 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 pumps are in steady state shall be checked and logged operation, the speed of at lear ' once per day.

the faster pump may not exceed 122% of the speed of'the slower pump when core power is 80% or more of rated power or 135% of the speed of the slower pump when core power is below 80% of rated power.

2. From and after the date that one recirculation loop is made or found to be inoperable for any reason, continued reactor opera-tion is permissible only during -

the succeeding 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months unless the modifications to APMI High Flux Scram, APDi Rod Block Monitor and M\PLHGR's as speci-(~ fled in Specifications 2.1.A.1.b, L- 3.1.A.3.b and 3.12.A are made.

1 LIMITING CONDITIONS FOR OPERATION SURVEILIANCE REOUIREMENTS If these requirements cannot be met, an orderly shutdown shall G. Structural Integrity

(

be initiated and the reactor 1.a. Nuclear Class I Components -

shall be in a cold shutdown Components within the reactor condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . coolant pressure boundary (as

. defined in Article IS-120 of 3 Following 1-pump operation, the the ASME Boiler and Pressure discharge valve of the lower Vessel Code) will be pressure speed pump may not be opened tested.pricr to startup follow-unless the speed of the faster '

ing each reactor refueling out-pump is less than 507. of its age. During the pressure test, rated speed and the requirements components will be inspected of Specification 3.6.A.4 and for leakage without removal of 3.6.A.5 are met. insulation. The test pressure and temperature will be main-G. Structural Integrity tained for at least four hours

-prior to the final leakage in-The structural integrity of the spection. The test pressure primary system boundary shall will not be less than the be maintained at the level re- system nominal operating pres-quired.by the original accept- sure at 1007. rated reactor ance standard throughout the power. The pressure test will life of the plant, be conducted at a vessel tem-perature above the nil ductility

. temperature of the vessel.

Near the end of each inspection interval, one system pressure test will be upgraded to a system hydrostatic pressure test. The hydrostatic test will be identical with the pressure test, except that the minimum test pressure will be higher and the test will be witnessed by an authorized in-spector. The test pressure

, will not be less than 1.08 times the system nominal oper-ating pressure as required by Subsection IS-522'of the Winter 1972 Addenda to Section XI of the ASME Boiler and Pressure Vessel Code.

b. Nuclear Class II Components - .,

Near the end of each inspection interval, the following systems (as defined in Subsection ISC-261 of the Winter 1972 Ad-dendum to Section XI of the .

ASME Boiler'and Pressure

(

3.6-8

P

~

DAEC-1 807 power cases, respectively. If the reactor is operating on one pu=p, the

. loop select logic trips that pu=p before making the loop selection.

An evaluation of ECCS performance and transient analyses has been made for single loop operation (Ref. 2). This evaluation shows that with modifica-tions to APRM High Flux Scram, APRM Rod Block and MAPLHGR's, continuous operation may be allowed. The short period of time allowed to operate with-out setpoint changes permits appropriate corrective action to be taken.

Requiring the discharge valve of the lower speed loop to remain closed until the speed of faster pump is below 507. of its rated speed provides assurance when going from one to two pump operation that excessive vibration of the jet pump risers will not occur, s

3.6-29

. un,~~-

s.

- LIMITING CONDITION FOR OpERATIO? SURVEILLANCE REQUIREMENT 3.12 CORE THERMAL LIMITS 4.12 CORE THERMAL LIMITS Applicability: Applicability:

The Limiting Conditions for The Surveillance Require-Operation associated with ments apply to the para-the fuel rods apply to those meters which monitor the parameters which monitor fuel ~ rod operating con-

. the fuel rod operating ditions.

conditions.

Objective -

Objective The Objective of the Limit- The Objective of the Sur-ing Conditions for Operation veillance Requirements is to assure the performance is to specify the type of the fuel rods. and frequency of surveil-lance to be applied to' the fuel rods.

Specifications Specifications A. Maximum Average Planar A. Maximum Average Planar Linear Heat Generation Linear Heat Generation Rate (MApLHGR) Rate (MAPLEGR) f ,

\.. During reactor power operation, The MAPLHGR for each type the actual MAPLEGR for each type '

of fuel as a function of of fuel as a function of average average planar exposure planar exposure shall not exceed shall be determined daily the li=iting value shown in Figs. during reactor' operation 3.12- 2, 3.12-3, 3.12-4 and 3.12-5. at > 25% rated thermal During periods of single loop , power.

operation the 'li:aiting value shall be no more than 0.86 times the ~

values shown in Figures 3.12-2, -

3.12-3, 3.12-4 and 3.12-5. If at any time during reactor power operation it is determined by nor-mal surveillance that the limiting value for >RPIRGR (IAPLEGR) is being exceeded, cction shall then.

be initiated within 15 minutes to restore operation to within the prescribed limits. If the >%PLHGR

.(IAPLEGR) is not returned to within -

the prescribed limits within two hours, the reactor shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Surveillance and -

corresponding action shall continue l until the prescribed limits are t

again being met.

, DAEC-1

'3*.12 BASES: CORE THERMAL LIMITS A. Maxi ,

Averaze Planar Linear Heat Generation Rate (MA?LEGR)

This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 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 the rods of a fuel assembly at any axial location and is only dependent secondarily on the

, rod to rod pcwer distribution within an assembly. Since expected local varia-tions 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 duel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 CFR Part 50, Appendix K limit. The limiting values for MAPLHGR's are shown in Figures

(

3.12-2, 3.12-3, 3.12-4 and 3.12-5.

The calculational procedure used to establish the MAPLHGR's shown on Figures 3.12-2, 3.12-3, 3.12-4 and 3.12-5 is based on a loss-of-coolant accident analysis.

The analysis was perfor=ed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR Part 50 A ccmplete discuselon of each code employed in the analysis is presented in Reference 1. Differences in this analysis as cc= pared to previous analyses performed using the methods in Reference 1 are: (1) The analyses assu=es a 1

fuel assembly planar power consistent with 102". of the MAPLHCR shewn in Figures 3.12-2, 3.12-3, 3.12-4 and 3.12-5; (2) Fission prodect decay is computed as- %

suming an energy release rate of 200 MEV/ Fission; (3) Pool boiling is assu=ed

c. iter nucleate boiling is lost during the flow stagnation period; (4) The effects of core spray entrainment and counter-current flow limiting as de-scribed in Reference 2, are included in the reflooding calculations. The pro- (

cedures used to determine the MAPLEGR's for use during single loop operation are described in Reference (9).

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Iowa Electric Light and kwer Company June 24, 1983 NG-83-0671 Mr. Harold Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Duane Arnold Energy Center Docket No: 50-331 Op. License No: DPR-49

Subject:

Single Recirculation loop Operation

Dear Mr. Denton:

In accordance with the requirements of 10 CFR 50.59 and 50.90, we transmitted our proposed technical, specification change regarding single recirculation lcop operation on October 17, 1980, which was subsequently amended in our December 18, 1981 transmittal. We hereby amend that application and amendment with the enclosed technical specification page changes, which are intended to supersede those previously submitted.

This amendment has been reviewed by the Duane Arnold Energy Center Operations Committee and the Safety Committee. A check for $4,000 was submitted with our origina-1 application and, therefore, no further fee is required.

Three signed and 37 additional copies of this application are transmitted herewith. Pursuant to the requirements of 10 CFR 50.91, a copy of this application and analysis of no significant hazards considerations is being sent to our appointed state official. This application, consisting of the foregoing letter and enclosures, is true and accurate to the best of my knowledge and belief.

IOWA ELECTRIC L GHT AND POWER COMPANY (MI ~ BY u h tmd \ bbu Alv

" Tichara 4. McGaugripf Manager, Nuclear DivG ici RWM/RAB/dmh* Subscribed and sworn tABefore Me'on Attachment this d74 M day of Uc o / / - 1983.

-cc: R. Browning g S.' hill & b )

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F. Apicella Notary Public in and for tne State of Iowa T. Houvenagle NRC Resident Office

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, f Evaluation of Change With Respect to 10 CFR 50.92 The enclosed application is judged to involve no significant hazards based upon the following information:

i) The enclosed application is for permanent licensing of Single Loop-Operation (SLO), i.e. operation with one recirculation loop out of service. The NRC, by granting Iowa Electric temocrary license amendments in the past for SLO, has determined that such operation does not involve-a significant hazards consideration. As the present-application foliows the guidelines previously established by the NRC for SLO, the enclosed amendment is therefore judged to involve no significant hazards as well.

ii) 'The enclosed amendment request is judged to involve no significant-hazards based uoan the precedent set by the NRC'is granting similar requests for other Operating Reactors.

iii) In the April 6th 1983 Federal Register the NRC published a list of examples of amendments that are not likely to involve a sionificant hazards concern. Example number four of that list states:

"A relief granted upon demonstration of acceptable operation from an operating restriction that was imposed because acceptable operation was not yet demonstrated.

This assumes that the operating restriction cnd the criteria to be applied to a request for relief have been estab.ished in a prior review and that it is justified in the satisfactory way that the criteria have been met."

As the NRC has previously established the criteria for SLO and the enclosed application follows those criteria, the amendment request is judged to fall within the scope of the above example.

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.f 't Revision 2 to Proposed Change RTS-124 to the Duane Arnold Energy Center l

Technical Soecifications The purpose of this change to th'e previous revision, RTS-124A submitted December 18, 1981, is to update tne submittal for permanent licensing of Single Loop Operation (SLO) to be consistent with the Reload Licensing amendment for Cycla 7-(Amendment 88, issued April 25, 1983). The major changes deal'with the: increased Minimum Critical Power datio (MCPR) operating limits for 8 X 8 ari P8 X SR fuel and new Maximum Average Planar Linear Heat Generation Rate (MAPLHliR) curves for the fresh fuel for Cycle 7. Also this revision reflects the new NRC position which allows isolation of the out-of-service recirc. loop by electrically isolating the recirc. pump instead of closing the suction valve, as proposed in the previous submittal. This change allows the temperature in the idle loop to be maintained, reducing the thermal stress on the pipe. Several administrative changes are also included. dealing with updating references and deletion of blank pages. For clarity and conciseness, a list of the affected pages is included which supersedes those previously submitted.

The changes being made are as follows:

1) Change the List of Tables on page vi to show that Table 3.12-2 is moved to page 3.12-10.

2) Change 1.1.A to include the safety limit FCPR for SLO.

3) Add the APRM Flow-Biased Flux Scram and Rod Black Equations for SLO to Sections 2.1.A.1 and 2.1.A.3 respectively as well as Figure 2.1-1 and to Tables 3.1-1 and 3.2-C (pages 1.1-19, 3.1-3 and 3.2-16). Also add the SLO-modified Rod Block Monitor (R6M) equation to Taole 3.2-C.

4) Add Section 2.1.A.4 detailing APkM Flux noise surveillance requ'irements for SLO, and add supporting Bases to page 1.1-3.

5) Update the bases for Section 1.1.A (page 1.1-5) anc 2.1 (page 1.1-9) for SLO.

6) Update the Referencas for Section 1.1, 2.1 and 3.12 to include the SLO analysis for DAEC and to show the new title of the GE Reload Fuel Licensing Topical Report.

7) Paragraph 4.6.E.b has been moved from page 3.6-7 to page 3.6-6.

8) Change Sections 3.3.E and 3.6.F.2 to allow SLO for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and add the necessary Limiting Condition for Operation (LCO).

9) Add the core plate aP surveillance requirement to Section 4.6.F.2 and supporting cases to page 3.6-34.

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Revision 2 Proposed Changas - RTS-1248 Page Two

10) Change the bases for Sections 3.6 and 4.6 to allow SLO and to reference the appropriate supporting analysis. Also, clarify the bases on jet pumos (section 3.6.E) and out-of-service loop isolation (section 3.6.F) for SLO.

11) Add the core thermal limits LCO's for SLO to section 3.12.A and 3.12.C.

12) Update the bases for Section 3.12 to support SLO and to reference the SLU snalysis.

13) Consolidate the text on page 3.12-Sa onto 3.12-6.

14) Update the MCPR limits given in Table 3.12-2 to reflect the increased operating limits for Cycle 7 and to show the comparable limits.for SLO.

15) Add required reference for SLO to footnote on Figures 3.12-5 through 3.12-9.

16) The following pages are renumbered as a result of deleting blank pages:

1.1-6 1.1-18 1.1-7 1.1-19 1.1-8 3.12-6 1.1-9 3.12-8 1.1-10 3.12-9 1.1-11 3.12-10 1.1-12 3.12-12 1.1-13 3.12-13 1.1-14 3.12-14 1.1-15 3.12-15 1.1-16 3.12-1b 1.1-17 i

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Affected Pages-vi 3.2-16 vii 3.3-7 1.1-1 3.6-6 1.1-2 ,3.6-7 1.1-3 3.6-31 1.1-5 3. 6-34 --

1.1 3.12-1 1.1-7' 3.12-3 1.1-8 3.12-4' l.1-9 3.12-5a*

1.1-10 3.12-6 1.1-11 3.12-7 1.1-12 3.12-8 1.1-13 3.12-9 1.1-14 3.12-9a*

1.1-15 3.12-10 1.1-16 3.12-11 1.1-17 3.12-12 1.1-18 3.12-13 1.1-19 3.12-14 1.1-20* 3.12-15 1.1-21* 3.12-16 1.1-22* 3.12-17*

1.1-23* 3.12-18*

1.1-24* 3.12-19*

1.2-7 3.12-20*

3.1-3

These pages have been deleted.

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DAEC-1:

1 TABLE NO. TITLE PAGE N0'.

4.2-0 ~ Minimum Test and Calibration Frequency for 3.2-29 Radiation Monitoring. Systems

'4.2-E Minimum Test Calibration Frequency for 3.2-30 Drydell Leak Detection 4.2-F Minimum Test Calibration Frequency for 3.2-31 Surveillance Instrumentation 4.2-G Minimum Test and Calibration Frequency 3.2-34 for. Recirculation Pump Trip 3.6-1 Number of Specimens by Source 3.6-33 4.6-3 Safety Related-Snubbers Accessible Ouring 3.0-42

,' Normal. Operation i

4.6-4 ' Safety Related Snubbers Inaccessible During 3.6-44 i Normal Operation L 4.6-5 Safety Related Snubbers in High Radiation Area 3.6-48 1

During ' Shutdown and/or Especially Difficult to i Remove i

3.7-1 Containment Penetrations Subject to Type "B" 3.7-20' Test Requirements 3.7-2 Containment Isolation Valves Subject to Type "C" 3.7-22

Test Requirements

, 3.7-3 Primary Containment Power Operated Isolation 3.7-25 l Valves 4.7-1 Summary Table of New Activated Carbon Physical 3.7-50 Properties

!. 4.10-1 Summary Table of New Activated Carbon Physical 3.10-7 Propertles.

3.12-1 Deleted 3.12-2 MCPR Limits 3.12-10 .l 3.13-1 Fire Detection Instruments 3.13-11 l

3.13-2 Required Fire Hose Stations 3.13-12 I

! 6.2-1 , ,inimum Sh!ft Crew Personnel and License 6.2-3 Requirements g 1 6.9-1l Protection Factors for Respirators 6.9-8 1 l- 6.11-1 -Reporting Summary.- Routine-Reports 6.11-12 26 .11-2 Reporting Su.anary - Non-routine Reports 6.11-14

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. s 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 .

6.2-1 DAEC Nuclear Plant Staffing 3.12-1 Kf as a Function of Core Flow 3.12-2 Deleted 3.12-3 Deleted 3.12-4 Deleted 3.12-5 Limiting Average Planar Linear Heat Generation Rate (Fuel Type 80274L) l 3.12-6 Limiting Average Planar Linear Heat Generation Rate (Fuel Type 80274H) l 3.12-7 Limiting Average Planar Linear Heat Generation Rate (Fuel Type PdDPB289) l 3.12-8 Limiting Average Planar Linear Heat Generation Rate j

(Fuel Type P80RB299) 3.12-9 Limiting Average Planar Linear Heat Generation Rate (Fuel Type P80RB284H) vii

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SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLAD 0 LNG INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability: Applicability:

i 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 claddir:g. automatic protective action is-initiated to prevent the fuel cladding integrity safety limits from being exceeded.

Soecifications: 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 For operation with the 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 'uel cladding limiting power density integrity safety limit. (MFLPD), the APRM scram trip setpoint shall be as B. Core Thermal Power Limit shown on Figure 2.1-1 and (iteactor Pressure < /85 psig shall be:

or Core Flow < 10% of Rated S < (0.66W + 54)

Whtn the reactor pressure is < ~

785 psig or core flow is less with a maximum setpoint of than 10% of rated, the core 120% rated power at 100?.

thermal power shall not exceed rated recirculation flow or 25 percent of rated thermal greater,

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',' . CAEC-1 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING C. Power Transient Where: S = Setting in percent of rated power (1,593 MWt)

To ensure that the Safety Limits established in Specification W = Recirculation loop flow 1.1.A and 1.1.B are not exceeded, in percent of rated flow.

each required scram shall be Rated recirculation loop initiated by its primary source flow is that signal. A Safety Limit shall be recirculation 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.

For a MFLPD greater than FRP, the D. With irradiated fuel in the APRM scram setpoint shall be:

reactor vessel, the water level shall not be less than 12 in. pgp above the top of the normal 5 < (0.66 W + 54) for two active fuel zone. Top of the MFLPD active fuel zor.e is defined to be 344.5 inches above vessel zero recirculation loop operation, and (see Bases 3.2).

S < (0.66 W + 50.5)

MFLPD for one recirculation loop operation NOTE: These settings assume operation within the basic thermal design criteria. These crit -ia are LHGR< 13.4 KW/ft (8x8 array) and MCPR > values as indicated in Table 3.T2-2 times Kf , where Kf is defined by Figure 3.12-1. Therefore, at full power, operation is nat allowed with MFLPD greater than unity aven t' the scram setting is reduced. If it is determined.

that either of triese design criteria is being violated during operation, action must be taken f tmediately to return to

.peration within these criteria.

2. APRM High Flux Scram When in the REFUEL or STARTUP and HOT STANDBY MODE, the APRM scram shall be set at less tnan or equal to 15 percant of rated power.

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c TTY 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:

FRP 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. For one recirculation loop operation APRM flux noise will be measured once per shift and the recirculation pump speed will be reduced if the flux noise averaged over 1/2 hour exceeds 8% peak to peak, as measured on the APRM chart recorder.

5. 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 Tbove vessel reactor low zero (+170" on water level level instruments)

C. Scram - turoine < 10 percent stop valve 7alve closure closure D. Turbine control valve f ast closure shall occur within 30 milliseconds of the start of turoine control valve fast closure.

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1.1 '8ASESi 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 f' uel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are act directly _ observable during reactor operation the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it.is recognized that a departure from nucleate boiling i 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 a

ccavenient limit. However, the uncertainties in monitoring.the core operating state a,nd in the procedure used to calculate the critical l power result in an uncertainity in the value of_the critical power. Therefore-the fuel cladding

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-integrity safety limit is defined as the critical power ratio in the limiting-fuel assembly for which more than 99.9% of the fuel rods in the core are 4

j expected to avoid boiling transition considering the power distribution within i

the core and all uncertainties.

f 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.

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DAEC-1

8. Core Thermal Power Limit (Reactor Pressure 1785 psig or Core Flow 1107.

of Rated)

At pre 3sures 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 naintained 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 1031bs/hr 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 103 lbs/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern. Full scale ATLAS test data taken at

. pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 Lt. With the design peaking f actors this corresponds to a core thermal power of more than 50%. Thus, a core therma.1 power limit of 25% for reactor 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 1.1.3 will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram frcm neutron flux folowing close of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a Safety Limit violaticn will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.

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DAEC-1 The computer provided with Duane Arnold nas a sequence annunciation program which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc., occur. This program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient. Thus, computer information normally will be available for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specification 1.1.C will be relied on to determine if a Safety Limit has been violated.

D. Reactor Water Level (Shutdown Condition)

During periods when the reactor is shut down, consideration must also be gi';en to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation.

The core can be cooled sufficiently should the water level be reduced to two-thirds the core height. Establishment of the safety limit at 12 inches above 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 (See Bases 3.2).

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. OAEC-1

1.1 REFERENCES

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

2. "QuaneArnoldEnergyCenterSingle-LoopOperaticn,"NE00-24272 July 1980.l

Approved Revision at time reload analyses are performed.

E a DAEC-1

2.1 BASES

LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL CL' 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 1653 MWt. The analyses were based upon plant operation in accordance with the operating map given in Figure 3.7-1 of the FSAR. In addition,165S 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.

I Transient analyses performed each reload are given in Reference 1.

Models and model conservatisms are also described in this reference. 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 equations are adjusted for one-pump operation.

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.

Trio Settincs The bases for individual trip settings are discussed in the folicwing paragraphs.

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DAEC-1 A. Neutron Flux Trips

1. APRM Hich Flux Scram (Run Mode)

The average power range

monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads f

in percent of rated thermal power (1593 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 cperational 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 l abnormal operational transients analyzed violate the fuel Safety Limit and there is a substantial margin from fuel damage. Therefore, the use of flow referenced scram trip provides even additional margin. An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering 4

during operation. Reducing this operating margin aculd 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 scr3ms.

.' DAEC-1 i i

The scram trip setting must be adjusted to ensure that the LHGR transient paak is not increased for any combination of MFLPD and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1.A.1, when the maximum fraction of limiting power density is greater than the fraction of rated power.

Analyses of the limiting transients show that no scram adjustment is required to assure MCPR greater than or equal to safety limit when the transient is initiated frcm MCPR > values as indicated in Table 3.12.2.

2. APRM Hic _h Flux Scram (Refuel or Startuo & Hot Standby Mode)

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 accommodate anticipated maneuvers associated with power plant 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 control rod patterns are constrained to be uniform by operating procedures backed up by Worths the rod worth minimizer and the Rod Sequence Control System.

_ _ - _ _ _ - _ _ -_-_-_-_____--_--_-_-__-__--____-________-_____________-____-_x__-_

,' OAEC-1 of individual rods are very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise.

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the beat flux is near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scrara before the power could exceed the safety limit. The 15 percent APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressure is greater than 880 psig.

3. APRM Rod Block (Run Mode)

Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent rod withdrawal beyond a given power level at constant recirculation flow rate, and thus prevents a MCPR less than the safety limit. This rod block trip setting, which is automatically varied witn recirculation loop flow rate, prevents excessive reactor power level increase resulting from control rca 1.1-12 i

. .' DAEC-1 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 specified trip setting versus flow relationship; therefore the worst case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is establishd by specified control rod sequences and is monitored continuously by the in-core LPRM system.

As with the APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum fraction of limiting power density exceeds the fraction of rated power, thus preserving the APRM rod block safety margin. As with the scram setting, this may be accomplished by adjusting the APRM gain.

4. APRM FLUX NOISE APRM flux noise oscillations in excess of these specified in Section 2.1.A.4 could be an indication that a condition of thermal hydraulic instability exists and that appropriate remedial action should be t aken.

5. 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 tne 5 decades are broken down into 10 ranges, each being one-half of a decade in size. The IRM scram trip setting of 120 divisions is active in each range of the IRM. For example, if the instrument were on range 5, the scram would be 120 divisions on that range. Thus, as the IRM is ranged up t'o accommodate the increase in power level, tne scram trip setting is also ranged up. The most 1.1-13 1

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OAEC-1 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 1s 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 the IRM channel closest to the withdrawn rod is by-passed. The results of this analysis show that the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above the 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. Scram and Isolation on Reactor low Water Level The setpoint for the low level scram 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

DAEC-1

. I 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 safety limit in all cases, and system pressure does not reach the safety valve settings. The scram setting is approximately 21 inches below the normal operating range and is thus adequate to avoid spurious scrams.

C. Scram - Turbine Stoo 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 scram setting at 10 percent of valve closure, the resultant increase in surface heat flux is such that MCPR remains above safety limit even during the worst

! case transient that assumes the turbine bypass is closed. This scram l is by-passed when turbine steam flow is below 30 percent of rated, as

! measured by the turbine first stage pressure.

D. Turbine Control Valve Fast Closure (Loss of Control Oil Pressure

! Scram)

The control valve f ast closure scram is provided to limit the rapid increase in pressure and neutron flux resulting from fast closure of

' the turbine control valves due to 3 load rejection. It prevents CR frcm becoming less than safety limit for this transient.

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{ 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 8S0 psig has !

been provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage is tak'en of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing-protection for the fuel cladding integrity. Operation of the reactor at pressures lower than 880 psig requires that the reactor mode switch be in the STARTUP position, where protection of the fuel cladding integrity safety limit is provided by the IRM and APRM high s

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 vacuum initiates autcmatic I closure of the main steam isolation valves.

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DAEC-1 G. H. and I. Reactor Low Water Level Setooint for Initiation of HPCI and RCIC, Closing Main Steam Isolation Valves, and Starting LPCI and Core Soray Pumos These systems maintain adequate coalant 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

, 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

2. "Duane Arnold Energy Center Single-Loop Operation," NED0-24272, July f

i 1980, t I

Approved revision number at time analyses are performed.

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TABLE 3.1-1 ,.-

REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMEtiTAfl0N REQUIREMENT ,

Hinimum No. of Modes in Which Number of Operable Function Must be Instrument

Instrument Operable Channels Chcnnels Provided by for Trip Refuel Startup Run Design Action (1)

System (1) Trip function Trip Level Setting (6) 1 Mode Switch in X X X 1 Mode Switch A Shutdown (4 Sections) 1 l

1 Manual Scram X X X 2 Instrument A Channels 2 IRM Iligh Flux < 120/125 of Fuel Scale

~

X X (5) 6 Instrument A Channels 2 IRM Inoperat ive X X (S) 6 Instrument A Channels -

2 APRM liigh Flux for two recirc loop operation X 6 Instrument A or ti

<(.66W+54)(FRP/MFLPD) (11) (12) Channels Tor one recirc loop operation l ~~<(.66W+50.5)(FRP/MFLPD) (11) (12) 1 l

2 APRM Inoperative (10) X X X 6 Instrument A or ts Channels 2

APRM Downscale > S Indicated on Scale (9) 6 Instrument A or B Channels 2 APRM liigh Flux i 15% Power X X 6 Instrument A in Startup Channels 2 liigh Reactor i 1035 psig X(8) X X 4 Instrument A Pressure Channels ;

l l - - . .

~

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Minimum No. tiumber of of Operable Instrument - -

Instrunent Channels -

Channels Per Provided by Trip System Instrument Trip Level Setting Design Action for 2 recirc loop operation I

2 APRM Upscale (Flow Biased) 1(0.66 W + 42) ( D for 1 recirc loup operation l

1(0.66 W + 38.5) (MFLPD )(2) 2 APRM Upscale (Not in Run Mode) i 12 indicated on scale 6 Inst. Channels (1) 2 APRM Downscale 1 5 indicated on scale 6 Inst. Channels (1) 1 (7) Rod Block Monitor for 2 recirc loop operation (Flow Biased) FRP 2 Inst. Channels (1) 66 W + 39M MFLPD)(2) for 1 recirc loop operation 1(0.66W+35.5)(MfN D )( }

l (/) Rod Block Monitor Dawnscale 1 5 indicated on scale 2 Inst. Channels (1) 2 IRM Downscale (3) 1 5/125 full scale 6 Inst. Channels (1) l 2 IRM Detector not in (8) 6 Inst. Channels (1)

Startup Position 2 IRM Upscale i 108/125 6 Inst. Channels (1) 2 (5) SRM Detector not in (4) 4 Inst. Channels (1)

Startup Position l 2 (5)(6) SRM upscale i 105 counts /sec. 4 Inst. Channels (1) l 1 Scram Discharge Volume -< 24 gallons 1 Inst. Channel (9)

Water Level-lingh

. o i

( , DAEC-1 LIMITING CONDITIONS FOR OPERATION LIMITING SA_FETY SYSTEM SETTING 3.3.0 Reactivity Anomalies 4.3.0 Reactivity Anomalies The reactivity equivalent of Ouring the startup test the diference between the program and startup following actual critical rod refueling outages, the configuration and the critical rod configurations expected configuration will be compared to the during power operation shall expected configurations at not exceed l", a k. If this selected operating limit is exceeded, the conditions. These reactor will be shut down comparisons will be used as until the cause has been base data for reactivity determined and corrective monitoring during subsequent actions have been taken as power operation throughout appropriate. the fuel cycle. At specific -

power operating conditions, the critical rod configuration will be compared to the configuration expected based upon appropriately corrected past i

data. This comparison will be made at least every full a power month.

E. Recirculation Pumps When the reactor mode switch is in startup or run position, the reactor shall not be operated in the natural circulation flow mode.

See Specifications 3.6.F.2 for operation with one recirculation loop out of service.

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 0 above cannot be met, an orderly shutdown shall be initiated and the reactor shal' be in the Cold Shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

-__-_-_-__ ___________-___ -_-_ _ _ J W _ _ -___-__-___ _ _ _____ _ _ ____ _ _ _

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

2. 2. At least one of the relief valves shall be disassembled and

a. From and after the date that the inspected each refueling outage.

safety valve function of one relief valve is made or found to be 2 inoperable, continued reactor operation is permissible only during the succeeding thirty days unless such valve function is sooner made operable.

b. From and after tne date that the -

safety valve function of two relief valves is made or found to be inoperable, continued reactor operation is permissible only during the succeeding seven days unless such valve function is sooner made operable.

3. If Specification 3.6.0.1 is not 3. With the reactor pressure > 100 met, an orderly shutdown shall be psig and turbine bypass flow to initiated and the reactor coolant the main condenser, each relief pressure shall be reduced to valve shall be manually opened and

)

atmospheric within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . verified open by turbine bypass valve position decrease and d pressure switches and thermocouple readings downstream of the relief valve to indicate steam flow from i the valve once per operating cycle.

E. Jet Puros E. Jet pumos

1. Whenever the reactor is in the 1. Whenever there is recirculation l

startup or run modes, all jet pumps flow with the reactor in the j shall be operable. If it is startup or run modes, jet pump j determined that a jet pump is operability shall be checked daily inoperable, an orderly shutdown by verifying that the following i shall be initiated and the reactor conditions do not occur shall be in a Cold Shutdown simultaneously:

Condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The two recirculation loops have a a.

flow imbalance of 15% or more when l!

1 the pumps are operated at the same speed, f

i

b. The indicated value of core ficw rate varies from the value I

derived from loop flow measure-i ments by more than 10%.

3.6-6

Y

( '

'DAEC-1 ,

l t LIMITING CONDITIONS FOR OPERATION LIMITING SAFETY SYSTEM SETTING

' c. The diffuser to lower plenum differential pressure reading on an individual jet ,

pump varies from the mean of all jet oump differential pressures by more than 10%.

F. Jet Pumo Flow Mismatch 2. whenever there is recirculation flow from the

1. Wnen botn recirculation reactor in the Startup or pumps are in steady state Run mode, ano one.

operation, the speed of the recirculation pump is faster pump may not exceec operating, the diffuser to 122% of the speed of tne lower plenum differential >

slower pump when core power pressure shall ce enecked is 80% or more of rated cally and the differential power or 135% of the speed pressure of an individual of the slower pump when jet pump in a loop snall not "

core power is below 80% of vary from the mean of all rated power. jet pump differential

, pressures in that loop by ,

2. If specification 3.6.F.1 more than 10%.

cannot be met, one recirculation pump shall be ,F . Jet Pumo Flow Mismatch tripped. The reactor may -

be started and operated, or 1. Recirculation pump speeds operated with one shall be checked and logged recirculation loop out of at least once per day, service provided that:

2. For one recirculation loop

a. MAPLHGR multipliers as out of service the core indicated in section plate delta p noise will ce ,

3.12.A are applied. measured once per shift and >

tne recirculation pump speed

b. The power level is will ce reduced if tne noise limited to maximum of exceeds 1 psi peak to peak.

50% of rated power.

c. The idle loop is isolated by electrically disconnecting tne reCirC. pump prior to startup, or if disabled during reactor operation, witnin 24 ,

hours. Refer to specification 3.6.A for i startup of the idle i.

recirculation loop. l

d. The recirculation system controls will be placed in the manual flow I control mode, t

c-

-

DAEC-1 l

l

c. The jet pump 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 f ail 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 summation 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 f ailure 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 gener3te the coincident failure of a jet pump body; however, the converse is not true.

I

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

I

. - - _ - - - _ _ - _ _ _ m

i l

k DAEC-1 80% power cases, respectively. If the reactor is operating on one pump, the loop select logic trips that pu.,o before making the loop selection.

An evaluation has been provided for ECCS performance during reactor operation with one recirculation loop out of service ($ec. 3.12, Ref. 11). Therefore, continuous operation under sucn conditions is appropriate. The reactor may in any case be operated up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with one recirculation loop out of service without isolating the idle loop. This short perico of time permits corrective action to be taken to re-activate the idle loop or to implement tne cnanges for continuous operation with one recirculation loop out of service.

During periods of Single Loop Operation (SLO), the out-of-service recirculation icop is isolated by electrically oisarming the recirC. pump.

This is done to prevent a cold water injection transient caused by an inadvertant pump start-up. It is permissible to leave the suction and discharge valves open during SLO to allow flow tnru 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 tne 1 cop, prior to opening the LPCI injection valve.

Core Plate AP oscillations in excess of tnese specified in Section 4.6.F.2 could be an indication that a condition of thermal hydraulic instaoility exists and that appropriate remedial action snould be taken.

Requiring the discharge valve of the lower speed loop to remain closed until the speed of f aster pump is below 50% of its rateo speed provides assurance wnen going from one to two pump cperation that excessive vioration of tne jet pump risers will not occur.

- - _ - - - -____ _ _ ________ _ __M.m _ ___ __ _ _ __ ___ __ _

, , OAEC-1 LIMITING CONDITIONS FOR OPERATIO,N SURVEILLANCE REQUIREMENTS 3.12 CORE THERMAL LIMITS 4.12 CORE THERMAL LIMITS -

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

rod operating conditions.

Objective Objective The Objective of the Limiting 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, i Specifications Specifications A. Maximum Averace Planar Linear A. Maximum Averace Planar Linear heat uenerat15n Kate (McLnun) heat beneration kate ( w Lnun)

During reactor power cperation, 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 determineo planar exposure shall not exceed during reactor operation at the limiting value shown in Figs. daily

> 25 % rated thermal power and any 3.12-5, -6, -7, -8 and -9. For -c hange in power level or t

single-loop operation, the values distribution that would cause i in these curves are reduced by operation with a limiting control multiplying by 0.7. If at any rod pattern as described in the time during reactor power bases for specification 3.3.2.

operation (one or two loop) it is During operation with a limiting determined by normal surveillance control rod pattern, the MAPLhGd that the limiting value for (LAPLHGR) shall be determinec at i

MAPLHGR (LAPLHGR) is being least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

exceeded, action shall then be initiated within 15 ininutes to restore operation to within tne

! prescribed limits. If the i MAPLHGR (LAPLHGR) is not returned

< to within the prescribed limits l

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce reactor power to < 25% of rated thermal power, or to sucn a power level that the met, limits within theare nextagain being

hours.

If the reactor is being operated l witn one recirculation loop out of service and cannot be returned to within prescribed limits within this 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period, tne reactor shall be brougnt to the cold snutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

For either the one or tao loop operating condition surveillance and corresponding action shall continue until tne prescribed limits are again being met.

- - -- _ _ _ _ _ _ _ _ _ _ -- U ua l - -- --- - _ _ _ _ _ - - - _ _ - . - - - _ _ - _ - - - - - _ - - _ _ _ _ . - _ - - _ - - - _ - _ - _

. e

- o UALC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS C. Minimum Critical Power Ratio C. Minimum Critical Power Ratio (MCFR) (MCPR)

During reactor power operation MCPR shall be determined daily MCPR for one or two during reactor power operation recirculation loop operation at > 25% rated thermal power.

shall be > values as indicated and Tollowing any change in in Table 3.12-2. These values power level or distribution are multiplied by Kf which is that would cause operation with shown in figure 3.12-1. Note a limiting control rod pattern that for one recirculation loop as described in the bases for operation the MCPR limits at Specification 3.3.2. During -

rated flow are 0.03 higher than operation with a limiting the comparable two-loop values, control rod pattern, the MCPR If at any time during reactor shall be determined at least power operation (one or two once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , loop) 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If the reactor is being operated with one recirculation loop out of service, and cannot be returned to within prescribed limits within this 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period the reactor shall be brought to cold shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

For either the one or two loop operating condition surveillance and corresponding action shall centinue until tne prescribed limits are again being met.

3.12-3 J

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

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

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 and 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 oper3 tion were derived in Reference 11.

3.12-4

a O DAEC-1 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 resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument 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.

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 bounos the sum of a safety limit MCPR and the largest aCPR.

DAEC-1

2. MCPR Limits for Core Flows Other than Rated Flow The purpose of the Kf 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 f actor provides the required thermal margin to protect against a flow increase tnansient. The most limiting transient initiated from less than f

rated flow conditions is the recirculation pump speed up caused by a motor-generatcr speed control failure.

I For operation in the automatic ficw control mode, the Kf f actors assure that the operating limit MCPR of values as indicated in Table 3.12-2 will 4 not be violated should the most limiting transient occur at less than rated i flow. In the manual flow control mode, the Kf factors assure that the +

1 Safety Limit MCPR will not be violated for the same postulated transient i event.

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

l i

The Kf f acters shown in Figure 3.12-1 are consvvative for Ouane Arnold operation because the operating limit MCPR of values as indicated in Table

! 3.12-2 is greater than the original 1.20 coerating limit MCPR used for tne generic derivation of Kf .

3.12-7

. l i * !

DAEC-1 4.12 BASES: CORE THERMAL LIMITS C.- Minimum Critical Power Ratio (MCPR) - Surveillance Recuirement At co're 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 small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin.

With this low void content, any inadvertent core flow increase would only place operation in a more conservative 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 pump speed. The MCPR margir.

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 J

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.

i i

l l

l i

l 2 , x

^

I

l. }

DAEC-1 3.12 REFERENCES

- 1. Duane Arnold Energy Center Loss-of-Coolant Accident Analysis Report, NE00-21082-02-1A, Rev. 2, June 1982.

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.

8. 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 Amendment for Duane Arnold Energy Center, Supplement 5: Revised Operating Limits for Loss of Feedwater Heating, NEDO-24987-5.

11. Duane Arnold Energy Center Single Loop Operation, NE00-24272, July 1980.

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

3.12-9 1

s DAEC-1 TABLE 3.12-2 MCPR LIMITS For two recirculation For one recirculation Fuel Type loop operation loop operation 8x8 1.25 1.28 8 x 8R/P8 x SR 1.27 1.30 e

3.12 . _ _ _ _, . . - . , .

1.4 e i t i i i I K g FACTOR -

1.3 -

AUTOMATIC FLOW CONTROL i

1.2 _

f 1.1 - -

MANUAL FLOW CONTROL -

Scoop-Tube Set-Point Calibration .

positioned such that -

Flowmax = 102.5%

1.0

= 107.0% -

.= 112.0% -

= 117.0%

I f f I I t_

_f ___ ,

30 40 50 60 70 80 90 100 CORE FLOW, %

DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGIIT & POWER COMPANY TECIINICAL SPECIFICATIONS Kg AS A FUNCTION OF-CORE FLOW FIGURE 3.12-1

, 6 1

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10 ,'

Plana: Averar,e Exposure (WD/T)

JJ 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.

i DUANE ARNOID ENERGY CENTER ICh*A ELECTRIC LIGliT AND PCb'ER CCMPANY TECFNICAL SPECITICATICNS LIMITING AVE 3 AGE ?IANAR LINEAR REAT GENERATION RATE AS A FUNCTICN CF PIANAR AVERAGE EX?OSURE FUEI,TY?E: 8D27/+L FIGURE 3.12-5 1

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O 5,000 10,000 1'5,000 20,000 25,000 30,000 Planar Average Exposure (WD/T) l_/ When core ficw 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 f actors for one recirculation loop operation are given in Section 3.12.A.

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0 5,000 10,000 15,000 20,000 25,000 30,000 P1- Average Exposure (WD/T) i 1/ When core flow is equal to or less than 70% of rated, the MAPLHGR shall 4

Wat exceed 95% of the limiting values shown. Values shown are for two recirculation loops. Reduction factors for one recirculation loop operation i.

are given in Section 3.12.A.

a i

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l DUANE A?.NOLD ENERGY CENTER IOWA "Tr'TRIC LIGHT AND POWER COMP.AliY t

i

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...__ Plaaar Avera~se Exposure (G'.Jd / 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.

w,A.ir. A.. rm ,3 _ u.c,a. .. C.,.ER .

IC'w'A ~~ %..CC LIGE"' A'O ?C*.iER COM?AW TECENICAL S?ICIIICATIC53 LDiITING AVERAGE P' ANAR LD'I.G EZA.T GENERATICN .EATE AS A FUNC:!ON OF F.'.AS.G n7. ,AG . . s O ew- .n e J rt ,:. .

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                                                                                          ~-

TC, UNITED STATES

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3 .;%^y,/,_G Ii NUCLEAR REGULATORY COMMISSION WASMNGTON. O. C. 20555 5 Sectember 23, 1931 Cocket No. 50-331 Mr. Duane Arnold, President Iowa Electric Light & Power Company P. O. Box 351 Cedar Rapids, Iowa 521C6 1 RE: DUANE ARNCLD ENERGY CENTER ,

Dear Mr. Arnold:

My letter to you of August 24, 1981 infomed you that we were proposing a meeting with licensees who have recuested aoproval to operate on a single recirculation loop. The purpose of the meeting is to detemine what may have caused variations in jet pump ficw at Browns Ferry Unit No. I wnile operating on a single loop and what impact this should have en approval of other facilities to operate on one loop. Licensees and applicants who have not requested approval for single loop operation were also invited to the meeting. As you were advised by your licensing project manager, the meeting scheduled for September 9,1981 had to be postponed to allow more- time for analysis of relevant operating data. We apologize for this incon-venience. The meeting will be held at 9:00 AM on Thursday, October 22, 1981 in Room ?-118, Phillips Building, 7920 Norfolk Avenue, Bethesda, Maryland. It would be appreciated if you would infom your licensing project manager of the number of people who will be attending this meeting from your organi:ation. Sincerely,

                                                                  = -'

E,Thomas

                                                   -<,!-yf.*     ac $h Chief 4 Ippolito,  -

Operating Reactors Branch 42 Division of Licensing cc: See Next Page g-qf U { )_7-( "

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l Mr. Duane Arncid Iowa Ele::ric-Light & P wer Company - - c::

                 ' Mr. Ecber: Lowenstein, Esquire Harold F. Reis, Esquire Lowenstein, Newman, Reis and Axelrad 1025 Connecticut Avenue, N. W.

Washiagton, D. C. 20035-Cedar Rapids Public Library 42S Third Avenue, S. E.

Cedar Rapids,. Iowa 52401 ,

                 ..L'. S. Nu: lear Regulatory Commission Residen: Inspect:rs 0ffice
                  . Rural Route il Faio, Icwa 5232?

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                  'j                       July 2, 19?1 Docket No. 50-331 Mr. Duane Arnold, President Iowa Electric Lign: & power Company p.0. Box 351 Cedar Rapids, Iowa 52406

Dear Mr. Arnold:

Ey le::er dated October 17,1980 (LDR-50-277) you submitted an application to permit Duane Arnold to be operated with a single recirculation loop in service rather than both loops. Three otner facilities have requested similar authori:ation and we excect other BWRs will recuest approval for single loop operation in the near future. Several SWRs have previously been authorized to operate for a short period of time with one recirculation loop and two BWRs are currently authorized to operate routinely on a single recirculation loop. In all but one case, power level nas been been limitec to 50 percent; the ene exception was Browns Ferry 'Jni t No.1. On September 29, 1979, based on analyses performed for TVA by :ne General Electric Company (GE), we authorized TVA to operate Erowns Ferry i for aceut tw0 months at coaer levels up to $2 percent of full rated power. During porer ascension witn Erowns Fer"y I in single loop operaticn, jet cam? flow variations were notec in :ne active locp acove a puns speed of 65 percent of rated flow (about 59 percent of rated power). ,nenever TVA tried to increase the power level above nis point, they noted variations in jet pump flow, neutron flux, and relatec parameters. Accorcin;1y, TVA acministratively limitec Erowns Ferry 'Jnit 1 cperation to less tnan 60 percent for tne a: proximately twc months tne ur.it 0;erated on a single loco. While ana,1yses indicate tna: it shoulc be safe to cperate EWRs on a single loop in the range of S5 percent of ratec power, the experience at Erowns Ferry Unit I has raised concerns about authori:ing single locp operation for BnRs above 50 percent rated power until tnere is a cetter uncerstancing of wnat may have caused the variations in this facility.

gO Mr. Duane Arnold 2 To try to develop a better understanding of what occurred at Browns Ferry 1 in tne fall of 1979, we are proposing a meeting with you and the other licensees who have requested approval for single loop operations. We also propose to invite other BWR applicants and licensees since they may desire to have approval ftr single loop cperation of their facilities in the future. The questvons we wisn to address in tne proposed meeting are discussed in the enclosure to this letter. Since GE has provided the analysis to you to support your application, it 1; pears highly cesirable that representatives of GE be present in trying tt determine what occurred at Browns Ferry Unit i ena tne implications, if any, to other SWRs. To accommodate the appropriate personnel from your organization and other licensees, we have proposed a range of dates for the above meeting, specif-ically, the weeks of either July 27, 1981, August 10, 1081, August 17, 1981, August 24, 1981, or Septemoer S, 1981. If you will acvise your project manager of the date cr cates most convenient to you, we will try to find a day that is rest suitaole to all pa-ties and so advise you. i Sincerely yours, ggss c- - incmas 4, ppo i to, Chief Operating Reactors Branch d2 Division of Licensing

Enclosure:

Orc?csed Meeting Agenda cc w/ enclosure - See next page l h l I

f Mr. Duane Arnold Iowa Electric Light & Power Company - - cc: Mr. Ecbert Lowenstein, Esquire

     - Marcid F. Reis, Esquire Lowenstein, Newman, Reis and Axelrad 1025 Connecticut Avenue, N. W.

Washington, D. C. 20035 Cedar Rapids Public Library 425 Third Avenue, S. E. Cedar Rapics, Iowa 52401 L'. S. Nu: lear Regulatory Commission Resident Inspectors Office Rural Route #1 csc, icua ocat-Mr. Ron E. Engel, Manaaer Reload Fuel Licensing (MC 582) General Electric Company San Jose, California 95125

s . . J ', o pro eosed . Meeting with EWR Ar. plicants and

                 .              Licensees en Single-' Loop Operation purpcse of !:eeting: 1. To determine what may have caused the jet pump flow variations and other variations experienced by Browns Ferry Unit I curing single lccp operaticn and

2. Evaluate whether the Erewns Ferry experience shculd result in power limits for other EWRs cperating en a singi e ,s co p.

Agenda: 1. Discussicn cf what may have caused the unexpected variations in cperating parame.ers w.,en a s c rowns Ferry Uni: 1 exceedad about 60 percent rated power wnile cperating with on,iy one recircu,sa.1on icop. E. Discussion of parameters affected at Erewns Ferry 1 (i.e. , jet pump flew, neutron flux, core flow, cere pressure drep, etc.) ,

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wculd be exeected at c her EWRs Operating on ene recirculatica loop. If so, are safety lici s likely to be violated er cause cc:plications with respect tc core s ta bil i .v , c o re fl ow s.ven e t r.v , pump cavita:icn er damage to the je: pumps and reacter vessel internals.

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      % y*.... )j Docket No. 50-331                          August 24, 1981' Mr. Duane Arnold, President Iowa Electric Light & Power Company P. O. Box 351 Cedar Rapids, Iowa 52406 RE: DUANE ARNOLD ENERGY CENTER

Dear Mr. Arnold:

By letter dated July 2,1981, we informed you that we were proposing to hold a meeting with you and other licensees who have requested approval to operate at power levels above 50% with only one recirculation loop in service in the event the other loop is inoperative. The announced purpose of this meeting is to obtain a better understanding of what might have caused variations in jet pump flow and related parameters at Browns Ferry Unit No.1 during single loop operation and how this incident affects approval of your application. We had proposed a range of dates for the meeting to accomodate the people expected to attend. Based on expressed preferences, the meeting on single loop operating experience will be held at 9:00 A.M., Wednesday, September 9,1981 in Room P-118, Phillips Building, 7920 Norfolk Avenue, Bethesda, Maryland. The agenda, a copy . of which is enclosed, is the same as that included with our previous letter. You are requested to advise your NRC project manager of the personnel who will be attending from your organization. Sincerely, .., 0tThom JAS Chief

                                                       &as /7Ippolito, Operating Reactors Branch #2 Division of Licensicg

Enclosure:

Meeting Agenda cc w/ Enclosure See next page

       ..       i.JF                                                                                                                                       l l

e, .. . 5 .- Proposed Meeting with BWR Applicants and Licensees Lon Stngle Loop Operation, . ' Purpose of Meeting: 1. To determine shat may have cau. sed the jet pump.f1mv and other variations experienced by Browns ~ Ferry Unit 1

                                            'during single loop operation and ~

2. Evaluate whether the Browns Ferry experience should result in power ~1imits for other BWRs operating on a single loop.

Agenda: 1. Discussion of what may have caused the unexpected - variations in operating parameters when Browns Ferry Unit 1 exceeded about 60 percent rated power while operating with only one recirculation loop. ,

2. Discussion of parameters affected at Browns Ferry 1 (i .e. . jet pump flow, neutron flux, core flow, core pressuredrop,etc.)

                        ,             3. Discussion of whether the Browns Ferry 1 experience would be expected at other BWRs operating on one
                                     . . . recirculation loop. If so,_are safety limits likely.to be violated or cause complications with respect to core stability, core flow symmetry, pump cavitation' or damage to the jet pumps and reactor vessel internals.

4. Discussion of the benefits vs. potential problems and cost of testing single loop operation in another SWR that is instrumented to detect what parameters are 4

a f fected. .

5. ' Evaluation of whether single toop operation at p5wer levels about 50 to 55 percent is a safe and prudent means of reactor operation.

J e 9 1 9

                                                               ,     ,-        , - , - , , - - -                . - - + ,           -,,-a--.     -. .., ,

    .        1 Mr. Duane Arnold Iowa. Electric Light & Power Company  -   -

cc:

              'Mr. Robert Lowenstein, Esquire Harold F. Reis. Esquire Lowenstein, Newman, Reis and Axelrad 1025 Connecticut Avenue, N. W.

Washington, D. C. 20036 Cedar Rapids Public Library 428. Third Avenue, S. E. Cedar Rapids,-Icwa 52401

           .  .U. S. Nuclear Regulatory Commission Resident Inspectors Office                                            -

Rural Route #1 Falo, Iowa 52324 , Ron E. Engle, Panager Reload Fuel Licensing (MC 682) General Electric Company San Jose, California 95125 e e e O O e e 9 9 0 0 -n,. ,, , -

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     %, ' v * #                             April 7,1981 Docket No. 50-331 Mr. Duane Arnold, President Iowa Electric Light & Power Ccmpany Post Office Sox 351 Cedar Rapids, Iowa 52406

Dear Mr. Arnold:

Reference is made to your application of October 17, 1980 (LDR-80-277) requesting aucharization for single recirculaticn 1000 operation cf the Duane Arnold Energy Center. To complete our review, we need responses to the enclosec request for additional information within 60 days of receipt of this letter. Sincerely,

                                                                    . /
                                                  -) -             /C
                                              % -( .   ,         :.fv&

ppolito, Chief . Thomas / - Operating Reactors Branch #2 Division of Licensing Encicsure: As stated cc w/ encl: See attached page L

141!P 1 i Mr. Duane Arnold Iowa Electric Light & Power Company - - cc: Mr. Robert Lowenstein, Esquire

Harold F. Reis, Esquire Lowenstein, Newman, Reis and Axelrad 1025 Connecticut Avenue, N. W.

Washington, D. C. 20036 Cedar Rapids Public Library 423 Third Avenue, S. E. Cedar Rapids, Iowa 52401 U. S. Nuclear Regulatory Commission Resident Inspectors Office Rural Route #1 Palo, Iowa 52324 G 9 1 l bL

4 REQUEST FOR ADDITIONAL INFORMATION DUANE ARNOLD ENERGY CENTER SINGLE-LOOP OPERATION i 1. Specify expected minimum and maximum operating core power / as percentage of Rated Core Power / Flow for Single-loop i Operation. 2. At the specified minimum and maximum operating Core Power / Flow C for Single-Loop Operation, provide the following: (1) Safety Limit MCFR values, Fuel Loading Error MCFR analysis results, (2) (3) Local Rod Withdrawal Error (with limiting instrument failure) Transient Summary, and Core Wide Transients Analysis and Operating Limit MCPR results (4) [ I for all the fuel types in the core for the following transientsFlow per NECE-24011-P-A-i: Pressurization. d In Section 2.0, a 65 Core Flow Measurement Uncertainty i has been

3. for single-lcop operatica (. compared to 2.5% for two-loop operation).

Explain how the contributien to the total core flow i l measurement unc value of 65 was calculated and justify that this value asure-l conservat ve y reflects the one standard deviation accuracy of the core f ow me ment system.

                  '        Describe how the change from norral two recirculation                   l h icooling       loop operatien to one loop oceration would be accomplished,         i   l        with  what p y and administrative controls, and while complyingd with their branch control, techn    ca position EICSB 12 (attached) regarding multiple setpoints an and with IEEE STD. 279 c.15.                          for decreased F

5. What provisions would be made in the technical specification flow stability in single loop operation? 5. Describe changes made to the flow ccmputer topumps loop jet automatically accou r.agnitude and sense change for reverse flow in the idle during single loop cperation. 7. Is there a requirement for the recirculation flow equalizer loop valves to be closed and tagged prior to commencing single recirculation i t operation as stated in NED0-24272 Page 1-1 and how is this requ rem ensured in the technical specification change? um i

. rx c BRANCH TECHNICAL POSITICN ICSB 12 l PROTECTION SYSTEM TRIP POINT CHANGES FOR OPERATION WITH REACTOR COOLANT PUMPS OUT OF SE'tVICE A. BACKGROUND For the past several years, including a time prior to the development of IEEE Std 279, the staff has required automatic adjustment to more restrictive settings of trips affect-ir3 reactor safety by means of circuits satisfying the single failure criterion. The basis for this requirement is that the function can be accomplished more reliably by automatic circuitry than by a human operator. This design practice, which has also been adopted independently by the national laboratories and by much of industry, served as the basis for paragraph 4.15. " Multiple Set Points," of IEEE Std 279. More recently, all applicants have stated that their protection systems were designed to ceet IEEE Std 279. Paragraph 4.15 of IEEE Std 279 specified that where a mode of reactor operation requires a more restrictive set point, the means for ensuring use of the more l restrictive set point shall be positive and must meet the other requirements of IEEE Std 279. A number of designs have been proposed and accepted which reliably and simply satisfy this requirement. During the review of some applications, however, certain design deficiencies have been found. The purpose of this position is to provide addi-tional guidance on the application of Section 4.15 of IEEE Std 279. B. BRANCH TECHNICAL POSITION

1. If more restrictive safety trip points are required for operation with a reactor coolant pur.p out of service, and if cperation with a reactor coolant pump out of service is of sufficient likelihod to be a planned mode of operation,- the change to themorerestrictivetrippointsshouldbekccomplishedautomatically.

2. Plants with designs not in accordance with the above should have included in the plant technical specifications a requirement that the reactor be shut down prior to changing the set points manually.

C. REFERENCES

1. M111 store-3 Safety Evaluation Peport, September 24, 1973.

2. Seaver Valley-2 Safety Evaluation Report, October 10, 1973.

3. IEEE Std 279, " Criteria for Protection Systers for Nuclear Power Generating Stations."

7A-9

Iowa Electric Light and Ibwer Company March 28, 1983 NG-83-1039 LARRY D. IMXTr t GL E RAff0 Mr. Harold Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Duane Arnold Energy Center Docket No: 50-331 Op. License No: DPR-49 Single Loop Operation for the Duane Arnold Energy Center

Dear Mr. Denton:

The attachment to this letter provides the final responses to the request for additional information in Mr. Ippolito's letter to Mr. Arnold, dated April 7,1981 on our submittal for a technical specification amendment on single recirculation loop operation. Should you have further questions, please contact this office. Very truly yours, l'

                                                                                 )f, W. Yh\ N@o lk'/

L/

                                                                          !; [..LAssistant parry D. Root Vice President I       Nuclear Generation LDR/RAB/dmh*

Attachment cc: R. Browning o0I D. Arnold L. Liu S. Tuthill F. Apicella NRC Resident Office

   - dug 4040124 h g PDR ADOCK 05000331 P                       PDR General Qtfice . fu fkg 33;

redar /f.ymb. lou.11?406 + 319/3%443;

    .            .                                                                          ATTACHMENT
        *                                                                                 NG-83-1039
                   -                                                                        March 28,1983 i                                 . IOWA ELECTRIC RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION ON SINGLE LOOP OPERATION The following is the Iowa Electric response to the Commission's request for additional information on single loop operation (SLO) at the Duane Arnold Energy Center (re: letter, Thomas Ippolito to Duane Arnold, " Request for Additional Information, Duane Arnold Energy Center, Single Loop Operation,"

Docket No. 50-331, April 7, 1981.) NRC Question #4 Describe how the change from normal,two recirculation cooling loop operation to one loop operation would be accomplished, with what physical and administrative controls, and while complying with branch technical position EICSB 12 (attached) regarding multiple setpoints and their control, and with IEEE STD. 279-4.15. IE Response In order to ensure conformance with the assumptions used in the GE analysis of SLO at DAEC, Iowa Electric proposes to operate under the following restrictions:

1) The suction valve will be closed and electrically isolated in the inoperable recirculation loop per proposed Technical Specification 3.6.F.2.C. This is to prevent degradation of LPCI flow during LOCA events.

2) DAEC does not have equalizer lines between the A and B loop jet pump risers, thus the requirement that the valves be verified to be closed is not applicable.

3) The recirculation system controls will be placed in the manual mode per proposed Technical Specification 3.6.F.2.d., thereby eliminating the need for a control systems evaluation.

4) The steady state thermal power level will not exceed 50% of the rated value, per proposed Technical Specification 3.6.F.2.b.

5) The settir gs for the Rod Block Monitor (RBM), APRM Rod Block and Scram flow-biased setpoints will be modified, per the proposed Technical Specifications 2.1.A.2, 2.1.A.3 and 3.2.C.1 and will be implemented by the appropriate adjustment in APRM gain f actor, per Surveillance Test Procedure (STP) 42A001, Item 4.2.K.6. Setdown of Rod Block and Scram setpoints by amplifier gain adjustment is an accepted procedure per Amendment No. 30 to the OAEC Technical Specifications.

6) The fuel operating limits (MCPR and MAPLHGR) will be adjusted for SLO in the Technical Specifications, Sections 1.1. A, 3.12. A, 3.12.C. The Minimum Critical Power Ratio (MCPR) Safety Limit will be increased by 0.03 and th. Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits will be reduced by 30%.

V

  , f.. ~_ .
        ,                                                                               ATTACHMENT NG-83-1039
                         -                                                          March 28, 1983

7) Increased surveillance requirements on core plate AP and APRM flux noise will be implemented per proposed Technical Specifications 4.6.F.2 and 2.1.A.4, respectively.

NRC Question #6 Describe changes made to the flow computer to automatically account for magnitude and sense change for reverse flow in the idle loop jet pumps during single loop operation. s IE Response No changes a e necessary to account for reverse flow in the idle jet pumps as the existing circuitry will sc5 tract this flow when the recirculation > pump is deactivated. The coefficient of 0.95 applied to the reverse jet pump flow, to account for the difference in flow coefficient between forward and reverse flow, is accounted for in the calibration of the flow l summer amplifier gains. g l t l L __ _ _ _ _ _ _ _ _ _}}

ML20099K559

Contents

Text

2.1 REFERENCES

Subject:

Dear Mr. Denton:

1.1 REFERENCES

2.1 BASES

2.1 REFERENCES

Dear Mr. Arnold:

Dear Mr. Arnold:

Enclosure:

Dear Mr. Arnold:

Enclosure:

Dear Mr. Arnold:

Subject:

Dear Mr. Denton:

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ML20099K559

Text

2.1 REFERENCES

Subject:

Dear Mr. Denton:

1.1 REFERENCES

2.1 BASES

2.1 REFERENCES

Dear Mr. Arnold:

Dear Mr. Arnold:

Enclosure:

Dear Mr. Arnold:

Enclosure:

Dear Mr. Arnold:

Subject:

Dear Mr. Denton:

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