ML20140F118

Rev 1 to SAR Chapter IX

ML20140F118
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Site:	Pennsylvania State University
Issue date:	04/24/1997
From:	Hughes D, Witzig W PENNSYLVANIA STATE UNIV., UNIVERSITY PARK, PA
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_ _ _ _ _. _ _. _. _ _ _ _ _ _... _ _ _ _ _ _ _ _. _ _ _ _... _ _ _.

i A

l V i

r SAFETY ANALYSIS REPORT Chapter IX l

For the l

Penn State Breazeale Nuclear Reactor l

By:

Daniel Hughes Warren Witzig Contributors:

T.

Patrick Boyle Terry Flinchbaugh l

Rodger Granlund Samuel Levine Pamela Stauffer Marcus Voth License Number R-2 Docket Number 50-05 The Pennsylvania State University i

University Park, PA 16802 l

i l

January 14,1997 Rev.1 (4/24/97) i i

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11 SAFETY ANALYSIS REPORT TABLE OF CONTENTS 1

TitlePage...........................................................................................

i Table of Con te nts.................................................................................

ii Lis t o f Fi g u re s.....................................................................................

vil

• Lis t of Effective Pages..........................................................................

ix I.

INTRO D UCI'IO N.........................................................................

I-1 II.

S ITE C H ARACTERISTIS..............................................................

II-1 A.

Geography and Dem ography II-l.................................................

II-l

1. Reactor Site Access Con trol...............................................

II-1 B. Nearby Industrial, Transponation, and Military Facilities.....................

II-6 C. M ete o rolo g y..........................................................................

II-6 D. Geology and Hydrology...........................................................

II-7 E.

S ei s m o l o g y.........................................................................

II-10 r

F.

Re fe re nc e s.......................................................................

II-10 V]

III.

RE ACTO R D ES IG N....................................................................

III-l A. In trod u c ti o n......................................................................

III-l B. M ec hanical Des i gn..............................................................

III-l t

1. Reac to r B rid ge...................................................................

III-1

2. Reactor S us pe nsion Tower...............................................

III-3 3. Reac tor G rid Plates.........................................................

III-3

4. Fuel-M oderator Elements......................................................

III-6 5. C o n tro l R od s.................................................................

III-6 6. Co ntrol R od D rives..........................................................

III-9

7. Graphite Reflector Elements..................................................

III-16 C. N uclear Des i g n.....................................................................

III-16

1. S tandard TRIG A Co re........................................................

III-16

2. Extern al Neu tron Source......................................................

III-23 D. Therm al Desi gn.................................................................

III-23 IV.

REACTOR POOL AND WATER S YSTEM...........................................

IV-1 i

A. R e ac tor Poo l.........................................................................

IV-1 B.

PS B R Water Handlin g Sys tem..............................................

IV-2

)

1. General........................................................................

IV-2

2. Pool Recirculation Loo p......................................................

IV-2 3. (D ELETED)..................................................................

IV-2

- -.h_ ~ _.._ Yk I

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1 l

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l l

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' OV TABLE OF CONTENTS (Continued)

4. Transfer of Pool Water..........................................................

IV-2 5. H eat Exc han g er..................................................................

IV-5

6. Liquid Waste Evaporato r.......................................................

IV-5 C. Water Quality Monitorin g and Maintenance.......................................

IV-7 V.

FACILITY CONSTRUCTION..........................................................

V-1 A. Building...............................................................................

V-1 B. Heating and Ventilation.............................................................

V-1 C. U tili tie s...............................................................................

V-6 D. Fi re Pro tec ti on....................................................................

V-6 VI.

FACILITIES AN D EXPERIMENTERS.............................................

VI-1 A. BeamPorts............................................................................

VI-1 B. D O Therm al Column..............................................................

VI-4 2

C. Central Thunble..........................

VI-4 i

p

1. Central Thimble Oscillator.....................................................

VI-6 I

\\

D.

Ve rtic le Tu bes........................................................................

VI-6 1. Jib C rane..........................................................................

VI-8 E.

Pne umatic Transfer System......................................................

VI-8

1. Pneum atic Transfer Sys tem I..................................................

VI-8

2. Pneum atic Trans fer System II...............................................

VI-12 F.

Instrum e nt B rid g e...................................................................

VI-14 G.HotCells..............................................................................

VI-14 H. Co-60 Irradiation Facility.....................................................

VI-14 1

VII. REACTOR S AFETY, PROTECTION, CONTROL AND MONITORING S YSTEM........................................................

VII-1 A. S ystem S u mm ary................................................................

VII-1 B. Sys tem Design Philosophy.......................................................

VII-6 C. Cons ole Functio n Summary......................................................

VII-6 D. RS S Func tio n S um m ary............................................................

VII-7

1. SC RAM Functions...........................................................

VII-7 l

2. In terlock Func tio ns............................................................

VII-7 t

t 4

t l

1 l

1 i

i l

t IV l

i TABLE OF CONTENTS i

(Continued) l E.

RS S Desc rip ti on.....................................................................

VII-8

+

l

1. Wide Range Monitor Descriptio n............................................

VII-10

2. Power Range Monitor Description............................................

VII-ll

3. Control and Alarm S ubsystem Descrip tion.................................

VH-12

a. SCRAM and Control Logic Assembly Description.....................

VH-12 I

b. SCRAM and Rod Control Switch Assembly Description.............

VII-12

c. SCRAM and Alarm Panel Assembly Description......................

VII-13

4. The Po wer Distributio n System................................................

VII-13 F.

PCM S Function S ummary........................................................

VII-13

1. DCC-X Func tions.................................'.............................

VH-14

a. Reactor Co ntrol and Regulation......................................

VU-14 b. Reactor Protec tion..........................................................

VH-16 i

(a)

Reac to r S te pb ac k...............................................

VII-16 (b) Rextor SC RAMS...................................................

VII-16 i

(c)

Re actor Interlocks...............................................

VII-17 i

(d) Facilities Systems S upport.........................................

VH-17 (i)

Emergency Evacuatio n......................................

VH-17 (ii) Reactor Operation Inhibit.....................................

VII-18 l

(iii) Manual Co ntrols..............................................

VH-18 i

(iv) Operatin g History Records..................................

VH-18 (v) Police Services Notification................................

VII-18 (e)

Alarms................................................................

VII-19 (f)

Operator Interface.................................................

VH-19 (g) Se lf Testin g...........................................................

VH-20 2. DCC-Z Fu nc tio ns................................................................

VH-20 G. System s Operation Description..................................................

VII-20

1. Reactor Safety System Descrip tion..........................................

VII-22

2. RSS Relay Logic Design.....................................................

VH-22 a. SCRAM Log ic...............................................................

VII-24 b. Transient Rod Air Interlock Logic........................................

VII-25 c. Rod Drive Interlocks.......................................................

VH-26 H. PCMS Hardware Description...................................................

VII-27 1. C o m p u te rs.......................................................................

VII-27

2. Input /Outp ut Hardware........................................................

VII-29 i

a. Chassis Arrangement and Watchdog / Test Cards.......................

VH-29 b. Analog Signal I/O Cards...............................................

VH-30

c. Digital Signal I/O Cards..................................................

VH-32 d. Watchdog and I/O Self Test Circuits...................................

VII-32

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1 i

l v

TABLE OF CONTENTS (Continued)

3. Mo tors and Associated Controllers...........................................

VII-33 a. M o tor Co ntrol............................................................

VII-34 l

4. Powe r S upplies.................................................................

VII-37 5. I/O Assi gnme nt..................................................................

VII-37 I.

PROTOL Genede S oftware Description.........................................

VII-39 l

1. Control Lan g uage.............................................................

VII-39 VII-40 i

2. The Operating System......................

3. Generic Tasks Running in the PROTROL System..........................

VII-41

4. System Self Checks and Defenses.........................................

VII-42 i

a. Defenses Against Loss of Field Sensor.....'............................

VII-43 b. Defenses A gainst Loss of Power........................................

VII-43 c. Defenses Against I/O Failum..................

VII-43 d. Defenses Against Computational Faults...............................

VII-44

e. Defenses Against Program Corruption Faults...........................

VII-44

5. DCC-X/DCC-Z Self Tests and Robustness Functions.....................

VII-45 a. S e lf Tes ts o n S tart U p.....................................................

VII-45

. O b. Self Tests While On Line.................................................

VII-46 O

J.

A pplic atio n S o ftward.............................................................

VII-48

1. B lock Langua ge Tasks.......................................................

VII-48

2. Non-B lock Language Tasks.................................................

VII-48 K. C o n tro l R o o m.....................................................................

VII-50 l

1. Gene ral Description.......................................................

VII-50

2. Monitor Indications in the Control Room..................................

VII-50 L.

Minimum Safety SCRAM S and Interlocks......................................

VII-53 M. Refere n c es.........................................................................

VII-55 N. Glossary............................................................................

VII-56 i

VIII. CONDUCT OF OPERATION......................................................

Vm-1 l

A.

Organization and Res ponsibility...................................................

Vm-1 B.

Reactor Operating Safety Philosophy.......................................

VM-1 C. Training.............................................................................

VM-3 l

D. Wd tte n Procedu res...................................................................

V m -3 E.

Records.............................................................................

VM-4 F.

Review and Audit of Records...................................................

VIII-4 (m

L M

9

i i

vi TABLE OF CONTENTS l

(Continued)

IX.

S AFETY EVALUATION.................................................................

IX-1 A. In trod uc ti on........................................................

IX-1 i

B. TRIGA FuelTemperature Analysis of the Penn State Breazeale Reactor....

IX-3

1. Steady S tate Analyses.......................................................

IX-4

2. Pulsing Characteristics of the PSB R.........................................

IX-10

3. TRIG A Experimen_t to Measure Fuel Tem peratures........................

IX-13 l

4. Evaluation of the A t for Fuel Element I-14.............................

IX-15 g

5. Evaluation of the Pulse Data for Fuel Element I-14......................

IX-19

6. Evaluation of the Fuel Element I-13 Temperature Data (Pulse and S t eady S ta te)...................................................................

IX-21

7. Conclusion (Temperatun: Analysis).......

IX-23 C. Evaluation of the Limiting Safety System Setting (LSSS)....................

IX-25 D. Loss of Coolant Accident........................

IX-27 E.

M aximum Hypothetical Accident (MHA).........................................

IX-36 F.

Reactivity Accident...................................

IX-42

,s

(

G. Conclusion IX-43 l

H. R e fe re n c es......................................................................

IX-45 l

j i

l l

l

l vii 7

O sirerv initysis aeroar LIST OF FIGURES i

i Figure #

Iille l

2-1 The Location of Centre County in Pennsylvania 1

2-2 Map of Centre County, Pennsylvania 2-3 The PSBR Site Boundary 2-4 Population Within a Five Mile Radius of the PSBR 2-5 The Physiography of Centre County 2-6 The Spring Creek Drainage Basin 3-1 The Location of the PSBR Core, Bridge, and Control Console 3-2 The Layout of the PSBR Grid Plates 3-3 The Arrangement of the PSBR Grid Plates and Safety Plate 3-4 A Standard TRIGA Fuel-Moderator Element 3 An Instrumented TRIGA Fuel Element 3-6 A Fueled Follower Control Rod with Respect to the PSBR Core 3-7 A Transient Control Rod 3-8 A Rack-and-Pinion Control Rod Drive 3-9 The Transient Rod Ddve 3-10 Core Loading #1 Layout 3-11 Core Loading #4 Layout O

3-12 A Graph of Peak Power Versus Prompt Reactivity for the First Nineteen Pulses with Core Loading #4 3-13 A Graph of Peak Fuel Temperature Versus Prompt Reactivity for the First Nineteen Pulses with Core Loading #4 3-14 A Layout of Core Loading #36 J

4-1 The PSBR Water Handling System 4-2 (DELETED) j 4-3 The PSBR Heat Exchanger 4-4 The PSBR Liquid Waste Evaporating System 5-1 The Location of the PSBR on The Pennsylvania State University Campus 5-2a The First Floor Plan of the Original Reactor Building 5-2b The Ground Floor Plan of the Original Reactor Building 5-3 Location of the PSBR Electdcal Supply Transformer 5-4a The First Floor Location of Fire Extinguishers and Fire Alarm Boxes 5-4b The Ground Floor Location of Fire Extinguishers and Fire Alarm Boxes 6-1 The Location of a Number of PSBR Facilities for Experimenters 6-2 The Location of the Beam Hole Laboratory, Hot Cells and Co-60 Irradiation Facility The D O Thermal Column 6-3 2

6-4 The Central'Ihimble Oscillator 6-5 Pneumatic Transfer System I 6-6 Pneumatic Transfer System I Laboratory Terminus

)

6-7 Pneumatic Transfer System 11 j

i 7-1 PSBR Console Layout i

7-2 New PSBR Safety, Protection and Control System j

viii i

II (%

d LIST OF FIGURES l

(Continued) t Figure #

Igle l

7-3 Old PSBR Safety, Protection and Control System 7-4 Functional Block Diagram of RSS 7-5 PCMS and Interfaces to Other Systems 7-6 CMS Equipment Layout (Console Rear View) 7-7 Watchdog and I/O Self Test Circuits 7-8 Instrumentation Pedestal i

7-9 Radiation Monitoring System 8-1 Organization Chart t

9-1 PSBR Core Configuration Loading #36 9-2 Comparing Highest Measured Fuel Temperathres During a Pulse with EQ(34) for Fuel ElementI-14 9-3 PSBR Core Configuration Loading #47 i'

9-4 The Time Dependence of Air-Cooled Fuel Body for Center Element with 267 W Input t

9-5 Summary of Equilibrium Data for LOCA Simulation Showing the Fuel-Element j

Cladding Temperature Versus Power In aut to the Element for All Seven Dummy i

Elements Heated with the Same Power 'nput 9-6 Fuel / Cladding Temperature as a Function of Time After LOCA Initiation l

' N 9-7 Maximum Fuel Temperature Versus Power Density After LOCA for Various Cooling Times Between Reactor Shutdown and LOCA Initiation 9-8 Strength and Applied Stress as a Function of Temperature, U-ZrH.65 Fuel with i

Fuel and Cladding the Same Temperature I

I lOG

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l

ix O

)

SAFETY ANALYSIS REPORT LIST OF EFFECTIVE PAGES i

SAR - Title Page Page i January 14,1997 SAR - Table of Contents Pages ii-vi January 14,1997 SAR - List of Figures' Page vii-viii January 14,1997 SAR List of Effectlye Pages Pages ix - x January 14,1997 Rev.1 (4/24/97)

SAR-I Introduction Pages 1-2 March 1,1985 l

SAR-II Site Characteristics Pages 1-10 March 1,1985 l

l SAR-III Reactor Design I

Pages 1-23 April 19,1991 i

l O

OO LIST OF EFFECTIVE PAGES

-(Continued)

SAR IV Reactor Pool and Water System Pages 1-2 April 19,1991 Page 3 September 21 1991 Pages 4-8 April 19,1991 Page 9 March 1,1985 SAR-V Facility Construction Pages 1-4 March 1,1985 Pages 5-6 April 19,1991 Pages 7-10 March 1,1985 S A R-VI Facilities and Experimenters Pages 1-3 March 1,1985 Page 4 September 21,1992 O

Page 5-8 March 1,1985 Pages 9-12 April 19,1991 Pages 13-14 March 1,1985 SAR-VII Reactor Safety, Protection, Control and Monitoring System Pages 1-17 April 19,1991 Page 18 September 21,1992 Pages 19-49 April 19,1991 Pages 50-54 February 28,1992 Page 55 April 19,1991 Page 56-57 August 23,1991 SAR VIII Conduct of Operation Pages 1-2 April 19,1991 Pages 3-4 June 7,1993 SAR-IX Safety Evaluation l

l Pages 1-46 January 14,1997 Rev.1 (4/24/97) d'?

IX-1 IX. SAFETY EVALUATION A. Introduction The Penn State Breazeale TRIGA Reactir (PSBR) was initially loaded with 8.5 wt%

U-ZrH sTRIGA fuel

• in December 19d5.(20 The reactor core was operated with good u

performance with this fuel from 1965 thr mgh the early 1970's. It was then decided to strive to reduce fuel costs for the supplier, the Department of Energy (DOE), by achieving higher fuel bumup through an increased uranium concentration in the fuel. Fuel management studies at the PSBR performed in 1972a.2) showed, by analysis and experiment, that replacing some of the 8.5 wt% fuel with 12 wt% U-ZrH TRIGA fuel would achieve a better fuel utilization and a substantially lower fuel cost. k65he basic "in-out" fuel management method was selected as it would provide the necessary excess core reactivity to achieve a longer fuel burnup. This "in-out" method would start with 12 wt% U fuel being ? aced in some fuellocatiins in the center most ring, the B l

ring. The remainder of the core wouk. be 8.5 wt% fuel. As the fael was consumed, the partially bumed 12 wt% fuel would be moved further out sequentially to '.he C and D rings while removing the 8.5 wt% fuel in those locations. New 12 wt% fuel wouWoe fed into the B ring to replace fuel moved to outer fuel rings. In a given fuellocation unburned 12 wt% fuel will produce a greater power density than the 8.5 wt% fuel by approximately 357o. Thus, increased power density results m higher fuel temperatures which were studied" analyti.: ally and experimentally to avoid exceeding safe operating conditions. The calculations agreed closely with the experimental data.

Since 1972 the PSBR has been refueled with 12 wtv fuel.

o On July 13,1972 six 12 wt% fuel elements were placed in the B-ring replacing 8.5 wt% fuel which was moved to outer rings. This increased the core kf to the level required for a larger fuel burnup I

and also increased the maximum measured fuel temperature to slightly over 400 *C, well below the safety limit of 1150 'C. The maximum radial power peaking factor was about 2.0 and the core reactivity increased by 1.624% Ak ($2.32). This reloading schedule of new fuel going into the B k

ring and after some fuel burnup being moved to the outer rings has been successful over the past 25 years requiring only 26 new 12 wt% fuel elements for the core.

In 1985 (Core Loading 38) a higher steady state maxiumum fuel temperature was observed (in an i

unburned instrumented fuel assembly of 12 wt%U, I-15) compared to previous similar instrumented (ie. I-13) fuel elements located in the same core position. The reason for this temperature increase is due to an increased fuel to fuel cladding gap. This was verified by I

comparing the peak fuel temperature of two similar instrumented fuel elements in the same core position subjected to die same sized pulse. The peak fuel temperature during a pulse of the two instrumented fuel elements were nearly the same whereas the steady state maximum fuel temperature was higher in the newer instrumented instrumented fuel element (I-15). Since the reactor pulse can be considered as an adiabatic process (the reactor pulse is than the thermal time constant of the fuel), there is no instant heat transfer and the conductance of the gap between fuel and clad is immaterial. However under steady state conditions, the maximum fuel element temperature is a function of the gap conductance. Therefore with a larger gap, the fuel temperature willincrease. In the case of I-15, there existed a larger fuel to fuel cladding gap prior to use than with I-13 after use.

• From this point on in Chapter IX, fuel refers to U-ZrH TRIGA fuel with 20% nominal u3 enrichment and zirconium to hydrogen atom ratio of 1 to 1.65 nominal.

January 14,1997 Rev.1 (4/24m)

1 IX-2 I

It also has been found that the measured steady state maximum fuel temperature increases when the fuel element experiences increasing sizes of reactor pulses." Further use of the I-15 instrumented fuel element in larger pulses showed an increased steady state maximum fuel temperature. It.s believed that the larger pulses produced a permanent strain in the fuel cladding due to the fuel '.nermal expansion. At the lower average steady state fuel temperature the fuel expacion is less than that occurring during the pulse, thus creating a fuel to fuel cladding gap.

A new fuel management strategy has been developed to manage the sustained steady state fuel temperature."

The principal computer programs used to perform the calculations are PSU-LEOPARD (8),

EXTERIMINATOR-2(9), MCRAC(10), and SCRAM (II). PSU-LEOPARD incorporates the standard LEOPARD <t2) computer program as originally received and adds additional subroutines. LEOPARD and PSU-LEOPARD calculate the group constants of the core as a function of bumup.

MCRAC is an automatic, multi-cycle, two-dimensional depletion code that gives the power distribution, keff, and isotopic inventory of the core at each burnup step. It is based on the flux and kerr calculation performed by EXTERMINATOR-2, a multi-group two-dimensional diffusion theory code.

SCRAM is a muld-cycle depletion code created specifically for TRIG A reactors and adapted to the PSBR lattice design. It uses analytical equations to compute the power distribution, kerf, e

and isotopic inventory for each cycle. The analytical equations are based on diffusion theory I

and the empirically fitted constants are derived using the PSU-LEOPARD, EXTERMINATOR-l 2, and MCRAC codes. In general, the calculations give good agreement with the measured power distributions and neutron fluxes.(1,2,7) Any equivalent codes can be used as long as they are properly benchmarked.

The fact that one can calculate the power distribution with good accuracy is important to calculating the safety margin in PSBR operation. The calculations identify the fuel element having the maximum elemental power density, MEPD,in the core, and thus the one which will produce the highest fuel temperatures. This is true for both steady state and pulse operations.

The experimental and analytical studies which have been performed to show the safety margin in the operation of the PSBR are described in this section. In particular, the maximum measured fuel temperatures during steady state operation and pulse operation am mathematically related in a unique way to allow predictions of their values during the PSBR operation. All predictions indicate that the design and construction of the PSBR is such that the safety limit of the fuel will not be exceeded during steady state operation. Further, any pulse temperature of too large a magnitude can be prevented by reviewing the steady state temperature measurements prior to pulsing a large excess reactivity into the core as part of administrative control. This is also tme for abnormal operating conditions. The following accidents are analyzed:

1. The loss of coolant accident.

2. The design basis accident which includes cladding rupture.

3. A mactivity accident.

The results of these analyses demonstrate that the reactor can continue to be operated l

safely within bounds of this safety analysis and the regulatory limits.

January 14,1997 Rev.1 (4/24m)

IX-3 B. TRIGA FuelTemperature Analysis of the Penn State Breazeale Reactor 1

There are two limiting conditions for establishing maximum allowed fuel temperatures.

l First, when the reactor is operating in the pool of water, the fuel temperature safety limit is l

1150 *C. Under these conditions, the fuel cladding temperature is less than 500 *C and the cladding will not be ruptured by the internal hydrogen pressure."') During a loss of coolant accident (LOCA), the fuel is not covered with water and must be air cooled. Secondly, when the fuel is air cooled, the cladding temperature will go above 500 *C, where the strength of the cladding decreases. Below a fuel temperature of 950 *C the hydrogen pressure will not mpture the cladding when the fuel and the cladding are the same temperature."') Under these conditions, the fuel temperature safety limit is 950 *C.

Flux gradients across the fuel produce uneven temperature distributions. Pulsing a TRIGA fuel element to high power densities produces sudden expansion and contraction. During the rapid expansion phase, a large temperature gradient in the radial direction can cause uneven axial expansion producing a transverse bend. Experience with the TRIGA fuel elements has shown that they can receive thousands of pulses without being damaged provided their temperature limits are not exceeded. TRIGA fuel elements are considered damaged and no longer useable if their cladding has been ruptured or their dimensions change to where the transverse bend exceeds 0.125 inches over the length of the cladding or the length increases 0.125 inches.

The temperature profile in a single TRIGA fuel element is a function ofits fuel and A

fission product distribution and is different for pulse operation compared to steady state i

operation. During steady state operation, the maximum fuel temperature is at the central fuel.

d zirconium rod interface. Since the thermocouple is placed near this interface, the measured fuel temperature is close to the maximum fuel temperature (4)(the maximum fuel temperature has been analytically determined to be no more than 5% more than the measured fuel temperature). During a pulse, the maximtua fuel temperature is near the fuel-cladding interface and tLe measured fuel temperature is no less than 60-65% of the maximum fuel j

temperature.(4) To know what limits to place on operation, it is important to understand the TRIGA fuel temperature distribution in a fuel element during steady state and pulse operation and to relate the measured fuel temperature to the maximum fuel temperature.

An instrumented TRIGA fuel element is built with three thermocouples placed 0.0226 ft radially from the center, but spaced vertically 1 inch apart. The middle thermocouple is in j

the midplane of the fuel region of the TRIG A fuel element. The thermocouple measures the fuel temperature at a specific point within the fuel element which is not the maximum fuel temperature for pulse operation. During a pulse, the temperature distdbution is the same as that of the volumetric thermal source strength, q"'(d, so that the peak fuel temperature is near the fuel cladding interface. This is due to self-shielding within the fuel. As a result, the measured fuel temperature can be significantly lower than the maximum fuel temperature. On the other hand, the peak fuel temperature during steady state operation is at the inner boundary of the fuel; thus, the measured fuel temperature is slightly less than the maximum fuel temperature. The measured fuel l

temperature in an 8.5 wt% U fuel element is closer to the maximum temperature than it is i

in a 12 wt% U fuel element because the self-shielding of the 12 wt% U fuel U is greater j

than the 8.5 wt% U fuel thereby producing a q"'(D with a steeper gradient.

I January 14,1997 Rev.1 (4/24/97)

~

i I

l IX-4 g)

(v The temperature distribution within the fuel can be calculated from a knowledge of fuel geometry, heat transport parameters, and q'"(I) as shown by Haag and Levine.(3) The volumetric thermal source strength in a fuel element is a function of the core power, the core configuration, and the element's position within the core. For any core configuration, q'"(r) can be determined by neutronic analysis and then used to determine the peak temperature during a pulse or during steady state operation. The steady state fuel temperature is l

determined by the boundary conditions at the cladding water interface and the value of q'"(r).

Studies performed by Haag and Levine have shown that subcooled boiling takes place in the PSBR when the TRIGA core exceeds 200 kW. This helps limit the temperature rise of the cladding surface temperature, tc, because when boiling occurs te increases proportional to approximately (q")"33 '") Hence, the heat flux, q", must increase by a factor of 8 to increase the difference between te and the water saturation temperature by a factor of 2.

It is important to recognize that the q'"(I) produced in a fuel element for a particular core configuration is the heat source that establishes the fuel temperature for both steady state operation and pulse operation. Hence, there is a direct rek: tion between the measured fuel temperature at steady state and during pulse operation for tiie same core configuration and fuel element. The fuel temperature measured in the fuel element having the highest power density in the core during steady state operation can be used to determine the maximum l

fuel temperature in the core during a pulse. This relationship is described in this section.

1. Steadv State Analyses n

Standard heat transport calculations are used to analyze the steady state fuel temperatures Q

for the PSBR.

Ixt volumetric thermal source strength at position r within the f" q,"'(r)

=

region.

power generated by the [" fuel element.

P,

=

Then P, = {q,'"hV, (1) where the integral is over the fuel volume, V, of the f" fuel element. The average power, P, produced by a fuel element in the core, and the normalized power for the f" fuel element, NP,, are related to P by the expression j

P, = PNP,

(2) 3 A core of Ne fuel elements producing a total power of Q can be expressed as l

Q = N P = N q'" V,

(3a) c c

p January 14,1997 Rev.1 (4/24/97)

IX-5 g

and a

P=N = q" ' V,

(3b) e where q'" is the volumetric thermal source strength averaged over all fuel in the core, and Ne is the number of fuel elements in the core.

It can be immediately observed that operating at 1 MW, implies that for a core with Ne = 90, then P = 0.0111 MW, and for a core with Ne = 100, then P = 0.010 MW.

Using Goodwin' sos) measured g "'(r,z) as j

q,'"(r,z) = (A, + B r )q "'(z)

(4) 2 o

j where for the j'" fuel element v,"'(r,z)dV = q "'V,

(5) jq j

z is the axial position along the j'" fuel element, and A, and B, are constants. A thermocouple is located at the fuel midplane where z = 7.5". The length of the fuelis

(

l 15". Thus, at the fuel midplane, q,"'(r,7.5) = q "'(r) = dq'"(A, + B r ),

(6) q o

and q '" = f, q,"' = f,N P q'" = f,N P, (7) q j

where f, is the axial peaking factor. The following definitions are used for the j'" fuel elementin these equations:

q '"(r,z) =

point volumetric thermal source strength, j

volumetric thermal source strength averaged

{

q "'

=

j over the radial direction,

q '"

volumetric thermal source strength averaged

=

j over the fuel volume.

When the j'" subscript is missing, q "'and q "' refer to the fuel element producing an average power in the core.

January 14,1997 Rev.1 (4/24/97)

%J l-

i I

I IX-6 O

f For the j"' fuel element P, = q '" V,

(8a) j and P

q"'= V.

(8b)

Equation (Sa) can be written, using Equation (2), in the following form:

P, = N P q'" V.

(9) j The temperature rise between the fuel and the cladding at the fuel element midplane during steady state operation is directly dependent on q,j'"(r). Dropping the j subscript for convenience but remaining in the fuel midplane k,r dt' = -q,"'(A, + B r ), r, s r s R, (10) 11' a

o r dr dr.

where r, =

radius of Zr rod in the center of the fuel rod, I

R=

radius of the fuel rod.

Integradng Equation (10) gives In 5 + B (R' - r') B r,' In "-

(11) t(r)- t, = 2k, _ 2 (R -r )- A,r,2 z

A o

r 8

2 r,

All parameters in Equation (11) are described and given in Table 9-1. Substituting the values for the parameters into Equadon (11) gives, for the thermocouple temperature, t,

ie t, - t, = 6.039x10~5q," ',

(12a) i Subsdtuting Equadon (7) into Equadon (12a) and returning to the j"' fuel element gives (t, - t,), = 6.039x10-5 NP, (12b) i 3

Using Q = 1 MW in Equadon (3b),

1 l

P=N 3.412x10' BTU (13) hr,

e January 14,1997 Rev.1 (4/24,97)

i l

\\

l i

IX-7 r%,

d Table 9-1 Parameters for the 12 wt% U-ZrH TRIGA Fuel Elements (Enriched to less than 20% 235U) l 1

Thermocouple radius, R c 0.0226 ft t t

Fuel mean radius, R 0.0596 ft t 0.0079 ft t Zr rod radius, rz l.

Cladding thickness, C 0.00165 ft Fuel element radius, R + C 0.06125 ft t Conductivity cladding, ke 9.5 Btu /hr ft2p Conductivity fuel, kt 10.5 Blu/hr ft2p Core average volumetric thermal 2.49 x 108 Btu source strength, k" hr ft3 Ao (12 wt% U fuel new) (Reference 4) 0.6534 Bo (12 wt% U fuel new) (Reference 4) 202 ft.2 Number of elements, Ne (Loading 36) 94.6 Number of elements, Nc (Loading 45T)l 88 Axial peaking factor, fa 1.25*

OV Prompt temperature coefficient., a (Ref.19)

-1.4 x 104 Sk/k/*C C, (Loading 36) Correction Factor 0.98**

This value is used for analyses subsequent to the original SAR since it is a more typical value used for TRIGA reactors.

• • See the description of this factor in the text immediately after Equation (15).

i These are nominal values and may vary with the manufacturers specification. It is also assumed that the fuel - cladding gap is zero (0) which in general is not true.

i (rh}

January 14,1997 Rev.1 (4r24/97) v

. _ _ _ _ -. -.=_ -

IX-8 The volume of fuel in a TRIGA fuel element is V = x(R - r,2)H, where H is the fuel height. Using H = 1.25 ft and the value frcm Table 9-1 for the other parameters V = 1.37x10-2 ft*,

(14)

Substituting Equations (13) and (14) into Equation (8b) gives 0'

q"' = C, 2.4 (15) where C, is a correction factor for setting the power instmment to read 1 MW when the actual power is reduced by the factor (1-C,). This reduction'is to provide an extra margin of safety to compensate for an uncertainty in calibration. What value is used for C, depends on how the equations are being used. If they are being used to predict fuel temperature from a given NP, a more conservative C, =1 should be used. If the equations are being used to determine NP i

from measured fuel temperature, a C, $1 may be used depending on the confidence the j

experimenter has in the accuracy of the power calibration.

O Equation (12b) can now be written as:

(t, - t,), = 1.5x10* C, f,NP'

'F, (16) i Ne (t,-t,)j = 8.33x10 C, f,NPC, i

Nc The measured fuel temperature, t, depends on the temperature at the cladding surface in the ie fuel midplane, t,. Because of subcooled boiling above 200 kW, this temperature rises very i

slowly. The At is proportional to (q'")", where At is the difference between t, and the coolant saturation temperature.04) As a result,it is assumed that the surface cladding is superheated by a fixed At degrees and thus at 1 MW, t, = 140 *C. This should be correct withini 10 *C at 1 MW for all NPj's greater than 1 and less than 3. We may write t, = (t - t,) + (t, - t ), + t,

(17) ie ic e

o where the first term on the right hand side of Equation (17) is evaluated using Equation (16). The second term, the temperature change between the fuel cladding and the surface of the fuel rod, is derived assuming a gap, g, between the cladding and the fuel.

Solution of the standard heat equation gives for l

(t, - t,), = (t, - t,); + (t, - t )3, (18a) o

(

January 14,1997 Rev.1 (4/24m)

IX-9 v) the following equation:

(t, - t )' = q ' In R + g + 2nkq ' InR+g+C q

q (18b) c

2nk, R

R+g where temperature of the fuel at the fuel cladding interface t,

=

temperature of the cladding facing the gap t

=

g k

conductivity of the gap

=

g k,

conductivity of the cladding

=

C thickness of the cladding

=

thickness of the gap g

=

q'=

linear heat generation rate q

and the other parameters are as previously defined.

Thus (t, - t,), = q ' In R+g q

(19a)

2nk, R

and q ' InR+g+C qu' i R+C q

s (19b) j (t, - t,)i = 2nk, R+g 2nk, n R

j OC Equation (19a) will be evaluated experimentally as described later, whereas Equation (19b), the temperature drop across the cladding, can be evaluated from the physical values of the parameters.

By definition 2

q ' = n(R r )q '" = n(R* - r )f,NP,q"',

(20) q q

Assuming an average core temperature drop across the gap, 5, Equation (19a) becomes 9

(t, - t,), = C,f,NP,5,,

(21a) where 2

-Ata = R -r* q"'InR+g (21b) 2k, R.

Also, Equation (19b) becomes, using Equation (20)

(t, - t ), = R' - r' f,NP,q'"In 2

R+C c

R gs)

January 14,1997 Rev.1 (4/24/97)

Q)

(

IX-10 l

OV Using the values of Table 9-1, Equation (21c) reduces to i

(t,-t ), = 1.248x10 C, f,NP' l

'F, (21d) o (t,-t,), = 6.936x10 C, f,NP' 2

'C.

Substituting Equations (21a) and (21d) into Equation (18a), and using the result in Equation (17),

it follows that

' 9.024x10' t,q =

+ t,,C,f,NP, + 140 *C.

(22)

(

C j

Equation (22) is used to calibrate an instrumented 12 wt% U fuel element to provide a measured fuel temperature, t, during steady state operation.

io

2. Pulsing Characteristics of the PSBR The temperature distribution in a TRIGA fuel element during a pulse has the same distribution as expressed in Equation (4) up to 89% of the fuel radius.(is) It has been found that adiabatic conditions hold up to 0.07 sec. during which time the maximum fuel temperature is reached.(is)

O Using the values of Table 9-1, it is found(4) that the maximum fuel temperature during a pulse is V

1.6 times that measured by the thermocouple. Thus, during the pulse, the shape of the temperature distribution in a fuel element remains constant, but the magnitude quickly rises.

What we are concemed with here is the maximum fuel temperatures reached during the pulse.

1 To prevent confusion, we use the tenn highest maximum fuel temperature to refer to the highest temperatures reached at any point within the fuel element during the pulse. The highest maximum fuel temperature is thus the maximum fuel temperature reached during a pulse and must remain below 1150 *C. However, the highest measured fuel temperature is 1/1.6 or 0.625 times the highest maximum fuel temperature which corresponds to a measured fuel temperature of 720 *C. Thus, setting the Limiting Safety System Setting (LSSS) at 650 *C, corresponds to a maximum fuel temperature of 1040 *C. The LSS scram will have no effect on the maximum fuel temperature reached during a pulse because the instrumentation time lag allows the peak to be reached before a scram can occur. The maximum fuel temperature reached during a pulse must be limited by the magnitude of the prompt excess reactivity insertion (Sk )

p and/or the q'"(r) produced in a fuel element for a particular core configuration.

A semi-empirical equation, Equation (29), is developed using the definition of the prompt temperature coefficient. The large negative prompt temperature coefficient, a., provides the TRIGA core with its pulsing capability. When excess k.n, Sk x = k n - 1,is inserted into the reactor, the reactor will go on a prompt period, provided Sk = Sk, - p, (23) y is positive, i.e., Sk, > 0.

is the effective delayed neutron fraction (0.007).

a l

[

}

January 14,1997 Rev.1 (4/24m) l V

IX-11 n

5p =

maximum fuel temperature rise averaged over the total core fuel volume for a pulse.

Epq =

maximum fuel temperature rise averaged over the radius of the j*

fuel element at its midplane.

a=

prompt temperature coefficient of reactivity of the TRIGA core.

The prompt temperature coefficient is defined as:

Sk a=_",

(24a)

Stpp or Sk" Stpp = -

(24b) a where Sk prompt excess reactivity insertion.

=

p 79 5pp average maximum rise in com fuel temperature due to the prompt

=

C/

excess reactivity insertion.

Equation (24b) does not include the average core temperature rise due to a pulse insertion of $1 excess reactivity, Stpi. Thus, the total average fuel temperature rise during a pulse, Stp,is:

(25a) l Stp = Stpp y Stpi.

For the j* fuel element, its corresponding temperature increase in the midplane is

_Stpq = f,NP 6tp, (25b) 3 or

-Stpq = f,N P (- Sk' ) + f,N P 5tpi.

(25c) 3 3

G Initially the core fuel temperature is that of the pool water, To, and during the pulse an adiabatic increase in temperature, Stg(r),is assumed. Hence, Stpq(r) = 5,q(A, + B r*).

(26a) o C'g January 14,1997 Rev.1 (4/24/97)

O l

L l

I

a b

l IX-12 l

l Equadon (26a) expresses the maximum temperature rise above room temperature for the -

l j'" fuel element as a funcdon of fuel radius. For convenience l

Stpq(r) = [stp,f(r),

(26b) where l

l f(r) = A, + B r'.

(26c) o he temperature in the midplane of the fuel element at any r posidon, tpq(r)can be l

expressed as:

l tpq(r) = Epqf(r) + To.

(27a)

Let l

tpq = maximum pulse temperature measured by the thermocouple in the j'" fuel element.

i l

Then tpq = Mpqf(r,)+ T.

(27b) i o

Using the values of Table 9-1 l

tpq = 0.7566 Epoi + T.

(27c) o Subsdtunng Equation (25c) into Equadon (27c)

' Sk"'

tpq = 0.7566 f,NP,

+ f,N P Stpi +T (28) i o

\\

a>

l l

Equation (28) is used as the basis for developing the semi-empirical equation, Equation (29), to fit the actual pulse data as a function of NPj, i.e.,

l tpq = K,f,NP, - Sk

+ f,NP,Epo + T,

(29) i o

3 a

l s

s i

where Kt4 and Stpo are empirical constants to be determined experimentally. The i

experimental data may also be represented by:

f tpq - T = M Sk + b,

(30) o 3 p 3

i t

January 14,1997 Rev.1 (4/24/97) l

IX-13

,m

!v) where M = K f,NP' (31a) u j

-a and b = f,NP 5po.

(31b) j j

NPj and Epo are determined by fitting the experimental data to Equation (30) where NPj, f., and a are known. In Equation (30), Mj Sk represents the temperature rise during a p

pulse due to the prompt excess reactivity (Sk ) insertion and bj represents the p

corresponding temperature rise due to the $1 excess reactivity insertion.

3. TRIGA Exoeriment to Measure Fuel Temocratures Using the analyses of the previous sections, a calibration was made to determine fuel temperatures for steady state and pulse modes of operation. This secdon describes the calibradon techniques.

A series of fuel temperature measurements were made using the 12 wt% U fuel instrumented fuel elements in core configuration loading 36 as shown in Figure 9-1. One

'T instrumented fuel element, I-13, had been in the core since September 1977 (in the B (V

ring, the position of MEPD) and the other, I-14, had never been used. The first series of measurements was taken with I-13 in the G-8 core position and I-14 in the G-10 core position. The core position of a fuel element is identified in Figure 9-1 by a letter for the vertical axis position and a number for the horizontal axis position. The instrumented i

fuel elements have been numbered sequentially with an I prefix. After rotating both fuel elements w.0.5 MW steady state operation to obtain maximum temperature readings, the reactor power was increased in steps to 0.7 MW,0.9 MW, and 1 MW. The actual values of the power were 0.98 (for loading 36) of that read on the recorder because the readout on the linear recorder was adjusted to read 0.4 MW when the actual power, as determined by a thermal power calibration, was approximately 390 kW (see the description of C, immediately after Equation (15)).

After completing the series of steady state mas, the reactor was pulsed sequentially with 2,2.25,2.50, and 2.75 dollar pulses. Before each pulse, the reactor was made subcritical to allow the temperature to reach equilibrium.

'Ihe above experiment was then repeated to determine reproducibility and measure the effect of the gap in I-14 created by the 4 pulses. The above measurements were again repeated with I-13 positioned in G-10 and I-14 positioned in G-8 to again study the reproducibility of the data and obtain another measurement on the temperature drop across the fuel cladding gap, Atg.

l The steady state and pulse measurements were again repeated, first with I-14 in F-10 and j

then with I-14 in H-11; I-13 was in position in G-8 for both sets of these measurements.

/

January 14,1997 Rev.1 (4f24/97) i l 0

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+$

IX-15 The data for all measurements (done prior to 1986) are summarized in Table 9-2. Both

\\

the chart recorder and the meter were used to measure the fuel temperature ofI 13 as shown in Table 9-2. The chart recorder was connected to the thermocouple at the midplane of the fuel element, whereas, the meter was connected to the thermocouple located 1" below. The banked control rods during a pulse causes the position of the highest power density in the fuel element to be displaced slightly downward from the midplane of the fuel. This causes the meter readings to be approximately 24 *C higher than those read on the chart recorder.

4. Evaluation of 5g for Fuel Element I-14 The unused TRIGA 12 wt% U fuel instrumented fuel element,I-14, has been placed in the core configuration of Figure 9-1, Core Loading 36, and experiments performed to evaluate Equation (22). It is assumed that before pulsing the instrumented fuel element, I-14, it had a Atg equal to 0 as the fuel would be in contact with the cladding. When the fuel element is first pulsed, the cladding is stretched introducing a gap which increases the Atg. After a number of pulses of the same size (i.e. $2.50), Atg reaches a maximum value and does not increase with further pulsing.(4)

'Ihe increase in 5g after pulsing I-14 several times is now determined by comparing the steady state temperatures for the same condition after each set of pulses. It has been found that measured fuel temperatures will increase further due to an increase in Ato O

when at some later time pulses of a larger size are performed (i.e. $3.00).m.26) i v

At 1 MW,I-14 measured tic = 372 *C in Core Loading 36 and position G-10 before any pulsing began. (Note that for work that was done for the original SAR C, = 0.98 and f,

= 1.35 and for subsequent analyses 1 and 1.25 are used respectively. The former because of the reasons stated immediately after Equation (15) and the latter because the 1.25 is a more typical value used as a design specification for axial peaking for TRIGA fuel (31))

Using Equation (22),

l 372 = 0.98 '024x10 - x 1.35 NP, + 140, 9

l 94.6 l

l It follows that NP =

~

= 1.84.

I 128.8 x 0.98 l

After 4 pulses tic = 418 *C,

and thus using Equation 22 again while holding NP, constant at 1.84, 418 = 0.98(95.39 + 5g)(1.35)(1.84) + 140, or Ma = 19 *C.

O January 14,1997 Rev.1 (4n4/97)

l IX-16

' O Table 9 2 s

Fuel Temperature Measurement Data for Loading 36 To = 21 *C l

l too(*C) Recorder / Meter j

Fuel Core te(*C) l Element Posidon SS 1 MW Pulse $2.00 Pulse $2.25 Pulse $2.50 Pulse $2.75 l

I-13 G-8 412 353/379 392/421 436/467 478/509 I-13 G-8 411 l

I-13 G-8 411 343/381 387/421 431/461 478/511 I-13 G-8 411 350/381 389/419 435/466 478/511 1-14 G-8 455 389 427 468 517 i

I-14 G-8 466 395 434 482 518 I-13 G-10 381 323/333 359/371 399/412 430/453 I-13 G-10 382 311/332 357/373 400/416 439/453 I-14 G-10 372 339 375 415 456 I-14 G-10 418 l

I-14 G-10 450 348 391 425 466 I-14 G.I1 433 342 373 411 449 I-13 is a 12 wt% U fuel element burned to 2.2 Megawatt days I-14 is a fresh 12 wt% U fuel element l

l l

i l

I g

9 e

l IX-17 1

O These data and analysis show the initial (few pulses) increase in the temperature across the gap. Funher data has shown that the increase in Atg diminishes to zero with successive pulses (see Table 9-2, lines 9 and 10). Element I-14 was then moved from posidon G-10 to position G-8 in the B-ring where the NP, is different from that in G-10.

Assuming 5, = 19 'C and using t = 445 *Cas measured in its new position at 1 ic MW, it follows that 445 = 0.98 (114.4) 1.35 NP + 140.

3 Therefore, NP, = 2.02.

After 8 pulses, the t for element I-14 was measured again in position G-8. This time ic t was 466 *C at 1 MW. Therefore, using Equation-(22) while holding NP constant at ic j

2.02, 466 = (96.4 + Ma)(1.35)(2.02)0.98 + 140, l

we find So = 26.6 *C.

lO It can be observed that after 8 pulses, So = 26.6 *C. Past studies have shown that

's J additional pulses do not alter the 5a ignificantly.*) For I-14, it is assumed that after s

many more pulses of the same size (i.e. $2.50), the 5g increase will be l'C, hence we use

_Ato = 27.6 *C,

and Equation (22) becomes for Loading 36 at Q = 1 MW (t )j = 163C,NP, + 140.

(32) ic Equation (32) can now be used to determine the NP, for I-14 anywhere in Core Loading 36 at 1 MW power as long as larger sized pulses are not performed resulting in an increase in 5.*) To generalize Equation (32) for any core configuration and similar 9

fuel element design specifications,it is only necessary to account for Ne. If this is done, Equation (32) becomes (t )j = 1.57x10' C,NP + 140.

(33) ic j

The steady state data of Table 9-2 has been evaluated using Equation (33) and the results are tabulated in Table 9-3. The t for the G-10 position was not measured after all pulsing ic had ceased and, therefore, is not listed in Table 9-3.

January 14,1997 Rev.1 (4n4/97)

IX-18 l.

Table 9-3 Evaluation of NPis from I-14 Data NP NP (Ave) 3 3

t,,

Steady State Pulse Core Position

(*C)

Eq.(33)

Eq. (34)

G-8 467 2.01 2.07 F-10 450 1.90 1.85 H-11 433 1.80' l.78 G-10 1.84 1.80 Table 9-4 OV Pulse Parameter Characteristics of Fuel Element I-14 M

M /NP, bj i

j X104 X104 b/NP 3

3 Core Position NP

• C/Sk/k

• C/Sk/k

• C

'C i

G-10 1.84 2.23 1.21 162 88 G-8 2.01 2.44 1.21 197 98 G-8 2.01 2.34 1.17 210 104 F-10 1.90 2.25 1.I8 170 89 H-11 1.80 2.04 1.13 178 99

[

i January 14,1997 Rev.1 (4/24s7) l l

v

1 1

IX-19 i

/~N t

t l

\\d Equation (33) is used to evaluate the measured fuel temperature during steady state i

operation. During steady state operation, the measured fuel temperature is close to the maximum fuel temperature (within 5%). In this case, the LSSS of 650 *C is extremely l

conservative because under steady state conditions, the maximum fuel temperature is no greater than 682 *C (650 *C + 5%) and thus, is well below the safety limit fuel temperature of 1150 *C. For loading 36, an upper limit for the measured maximum fuel temperature can be determined by setting NP = 2.2. Extersive calculations have been performed (i.2.5.7.27) to study the maximum power distril.ution produced by different core configuradons with fresh 12 wt% U fuelin the B-ring and the other core configurations containing a mixture of both 12 wt% U fuel and 8.5 wt% U fuel Future maximum steady state measured fuel temperature will be below the 650 *C LSSS.

5. Evaluation of the Pulse Data for Fuel Element I-14 Each series of pulse data using I-14 is fitted to a straight line to determine M and b, of j

Equation (30). Table 9-4 summarizes the results, wherein the data in the last column show that b /NP, is constant within 8% for the different core positions. The constants 3

K and Stpo are determined to be 1.22 and 71 respectively using Equation (31).

u Substituting these values and those from Table 9-1 into Equation (29) the result is d

tg = 1.177x10 NP Sk, + 95.8NP, + T.

(34) j o

Equation (34)is now used to evaluate NP, for the various core positions. The results are shown in Table 9-5 give consistent values for NP using different pulse magnitudes.

3 This validates Equation (34). The value of NP as obtained from steady state Equation 3

(33)is in good agreement with the corresponding value of NP, obtained using the pulse Equation (34). The highest measured fuel temperatures in fuel element I-14 are compared to that using Equation (34)in Figure 9-2. It can be observed that the measured temperatures are in good agreement with Equation (34).

The Penn State in-core fuel management codes were employed to determine the power distribution and NP,'s for Core Loading 36. In a recent Ph.D. thesis,(16) the group constants of the individual fuel elements were evaluated as a function of their burnup using the SCRAM code. The SCRAM code provides a simple but reasonably accurate method of depleting the PSBR core as has occurred since December 1965. These constants were input into the EXTERMINATOR-2 code to obtain the NP's for Core j

Loading 36. The NP's for G-9 and H-10 with new 12 wt% U fuel varied between 2.03 i

and 2.10. This is to be compared with the measured values of 2.01 steady state and 2.07 by pulse. In general, the steady state and pulse NP's agree with 3%.

It is now possible to eliminate the NP from Equation (32) and Equation (34) to give for 3

I-14

-T = (72.2Sk, + 0.588) t* - 140'C.

(35) tg o

[

January 14,1997 Rev.1 (4/24/97)

V i

1

1 IX-20 i

(x

(>

550 1

f N P-2.07 500 6

- NP-1.85 2.- ;;;-

Es NP-1.80

$ j 450 --

N P-1.78

8. 2 E2 h

hA C

O I?

$.E 400 --

• E x

g-

+

4 Core Position G-8 (2)

M Core Position G-8 (1)

A Core Position G-10 X Core Position F-10 350 l

X Core Position H-11 l

300 1.50 1.75 2.00 2.25 2.50 2.75 3.00 3.25 3.50 Step Reactivity Insertion (Dollars)

Figure 9-2. Comparing Highest Measured Fuel Temperatures During a Pulse With r

EO(34) For Fuel Element 1-14

(

January 14,1997 Rev.1 (4/24/97)

i IX-21 l O l V Equation (35) is an equation developed for fuel element I-14. It can be used anywhere in l

the core to predetermine the highest measured fuel temperature as a function of pulse prompt excess reactivity insertion (Sk,). It will require using the I-14 measured temperature when operating at 1 MW with I-14 in the same core position. For Core Loading 36 and using a maximum value of NP = 2.2 and corresponding t, = 499 *C, i

see Equation (32), the maximum value for t oj is 684 *C for a pulse reactivity insertion of p

$3.50 assuming T, = 21 *C and C, = 1.0. This corresponds to the maximum fuel temperature of 1095 *C which is below the safety limit of 1150 *C for fuel damage.

Thus, in the future, Equation (28) can be used to evaluate and predict t oj. This requires p

placing I-14 in the hottest spot in the core and running at 1 MW to evaluate t. Then ic starting with a $2 pulse, verify Equation (28) and predict t oj for the high values of Sk.

p p

The t oj related to a $2 pulse will be more than 100 *C below t oj for a $2.75 pulse and p

p even much lower than that for a $3.50 pulse. Hence, these initial pulses will produce maximum measured fuel temperatures well below 700 *C and allow determining the maximum fuel temperature attainable at the maximum allowable reactivity insertion pulses. A 700 *C fuel temperature measured by the thermocouple in a 12 wt% U fuel element corresponds to a maximum fuel temperature of 1120 *C. This is below the maximum allowed 1150 *C.

Q

6. Evaluation of the Fuel Element I-13 Temnerature Datn (Pulse and Steady State)

O Table 9-2 shows the temperature data taken for I-13 and I-14. As expected, a review of these data shows that the measured temperature ofI-13 during 1 MW steady state operation, t, is significantly lower than the t e f I-14 for the same core positions. On ic t o the other hand, the I-13 measured pulse temperature data is not significantly lower than that ofI-14 for the corresponding conditions. This is because the Ao and Bo constants of Equation (14) are not the same for I-13 and I-14. The depletion of the outer rim of fuelin I-13 during burnup,in addition to the burnup of 235U in all of the fuel element, lowers the self-shielding of thermal neutrons. As a result, the q"'(r) distribution for I-13 is much flatter than that of I-14. Using Equation (4) for I-13, and setting Ao = 0.9267 Bo = 40 yields the results of the I-13 data as shown in Table 9-6. The equivalent of Equation (33) and Equation (34) for I-13 are Equation (36) and Equation (37) respectively.

(t ), = 1*68x10' C,NP, + 140 (36) ic tpoi = 1.475x10* NP,5k, + 80NP, + T.

(37) o n

January 14,1997 Rev.1 (4/24/97) f

IX-22 Table 9-5 Table of NPj Determined for I-14 Using Pulse Data in Eq. (34) l Core Positions Skex Sk G-10 G-8 G-8 F-10 H-11 Dollars Dollars 2.00 1.00 1.79 2,06 2.10 1.84

.1.80 t

2.25 1.25 1.79 2.04 2.08 1.86 1.78

.l 2.50 1.50 1.79 2.04 2:10 1.85 1.78.

2.75 1.75 1.81 2.07 2.07 1.86 1.78 l

l Ave NPj 1.80-2.05 2.09 1.85 1.78 I

I i

l j

l l

January 14,1997 Rev.1 (4/24,97)

LX-23 0)

V It can be observed that the agreement between the pulse data and steady state data for determining NPj is not as good as that for I-14. This is due to the approximations made in deriving Equations (33) and (34), namely,

a. The Ao + Bor2 shape of q"'(r) approximates the excess burnup of U-235 at the perimeter of the U-ZrH fuelin I-13.

b. The is for I-13 is probably different from that of I-14.

However, temperatures measured by I-13 are consistent for the purposes of monitoring the core fuel temperatures.

7. Conclusion (Temocrature Analysis)

A major conclusion of this section (based upon the present fuel specifications) is that an unused instrumented 12 wt% U fuel element can be calibrated and used to monitor the maximum fuel temperatures in the core. Once calibrated, the fuel element will only be used to measure maximum fuel temperatures in new core configurations. For steady state cperations a measured fuel temperature of 650 *C results in a maximum f uel temperature well below 1150 *C. Under these conditions, the measured fuel temperature is close to the maximum fuel temperature. For pulse operation, a measured 700 *C fuel temperature corresponds to a maximum fuel temperature of 1120 *C which is below 1150 *C. The safety limit shall not be exceeded during pulse or steady state operation.

O

(-~)

Once a fuel element has been depleted, its maximum steady state temperature decreases as long as the gap between the fuel and the cladding remains the same. As experience has shown, pulsing at higher levels will increase the fuel / cladding gap and the maximum steady state temperature may increase before it starts to decrease. The fuel temperatures measured during steady state operation with a depleted fuel element are related to the fuel temperatum of a new fuel element by a simple ratio. Hence, this ratio can be used to assess the maximtun fuel temperature during steady state in a new fuel element. The maximum fuel temperatures measured with a depleted fuel element during a pulse are close to that in a new fuel elemcat. The preferential depletion of the periphery of the fuel element causes the power distribution and hence, the temperature distribution during a pulse, to be flatter than that of a new fuel element. Thus, the measured fuel emperature in a depleted fuel element corresponds to a lower maximum fuel temperature. It is also closer to the average fuel temperature. The core average fuel temperature rise for a given Sk insertion is the same for all cores. The lower NP for a depleted fuel element p

l accounts for its being closer to the average core fuel temperature.

l l

f ]i January 14,1997 Rev.1 (4/24/97)

/

V l

~- -

l l

IX-24 i

Table 9-6 Evaluation of NPj's from I-13 Data I-13 (1.2 MWD Depleted) l 1

1 NPJ NPj(Ave) tto Steady State Pulse Core Position

(*C)

Equation (36)

Equation (37)

I i

G-8 411 1.56 1.62 G-8 382 1.39 1.54

((

l For comparison, see I-14 data in Table 9-3.

(

J

~

n January 14,1997 Rev.1 (4/24s7) k i

-- i

1 IX-25

/%kj C. Evaluation of the Limiting Safety System Setting (LSSS)

The limiting safety system setting is a measured fuel temperature of 650 *C as defined in the Technical Specifications.

If the core power were at 1.15 MW (15% over power) steady state, the measured fuel temperature in the B-ring using extrapolated exp*erimental data for, Figure 9-3, Core Loading 47' with 95.5 elements would be 650 C, (using the 12 wt% U fuel element, I-15, which had been pulsed at the $3 level 20 times). The maximum temperature will be slightly higher, but the fuel temperature near the cladding will be approximately half this temperature. The extrapolated 650 *C fuel temperature is close to the maximum fuel tempeinture (within approximately 5%) due to the radial temperature distribution. A sudden insertion of reactivity with power at 1.15 MW, close to but less than $1, into the core will initially increase the reactor power exponentially at a period faster than one second. Using a negative temperature coefficient of 1 x 10-4 5k/*C,' the increase in average core fuel temperature is less than, 0.0076k /k

= 70 *C 1x10"Sk /k'C and for an NP = 2.2 and f, = 1.25, the maximum fuel temperature increase is 193 *C (2.2 x 1.25 x 70 *C = 193 *C). Adding this increased fuel temprature in the hottest fuel element to the 650 *C steady state temperature results in 843 *C. much less than the safety limit of 1150 *C. For this to occur at power levels above the power level scram 9(V setpoint will require that both power level scrams fail. The temperature scram will be initiated when the measured temperature exceeds its set point. The equilibrium temperature of 843 *C wiil be achieved at least within two to three periods (seconds) after reactivity insertion. A control rod drop time less than one second assures an early decrease in reacdvity and fuel temperature. At this point, the control rods moving into the core will begin to decrease the reactor power in less than a second after the scram.

Control rods are checked semiannually to assure their rod drop time is less than one second. The kinetics of the reactor cause the reactor power to decrease as soon as the control rods move a few inches into the core. Thus, the maximum fuel temperature will remain well below 1150 *C since the measured fuel temperature is close to the maximum fuel temperature for these quasi-static conditions.

The maximum allowed pulse reactivity of $3.50 is established to prevent the fuel temperature from exceeding the safety limit of 1150 *C. A $3.50 pulse, the maximum measured temperature staning from pool ambient temperature, using Equation (34) and

• Core Loading 47 is considered an extrema loading relative to steady state measured peak fuel temperatures.

• The temperature coefficient during a fast period is slightly less than the prompt temperature coefficient.

n

(

)

January 14,1997 Rev.1 (4/24/97)

%./

O O

O t

I Nonh Bottom Grid Plaie i

f Top Grid Plate A

B zL C

4 SA = Safety Careral Rod D

i N

Sil = Shirn Carent Rod y

RR = Regulating Conrol Rod E

TR = Transient ControlRod g

CT = Ccesralihenble Waner Filled F

i D

115 = 12 =% Insinunerned NI Bernew G

d Rt = Puneumatic Transfer I

System 5 = Neueron siernsp Source 3g PR = Power Range Gemena Ion p

Ournber

/

WR2 = Wide Range Rssion Oiamber I4 I

i Onendiers are locased above 3

Il 15 K

(f :(Th -h q9p.- (ior ri3 g o 14 i.

weersnea 4^

] (( [h gh h b 12. s m iBen==

2 M

2 4

5 v

i s.5 ws Ni nemers

~"

^

g ~, _ R Air Fonower Transas Control Rod Figure 9-3 PSBR Core Configuration Loading 47 t

IX-27 O

NP = 2.2,is 684 *C. This corresponds to a maximum fuel temperature of 1095 *C.

The temperature scram will not lower the maximum fuel temperature attained during a pulse once the pulse is initiated; however, it does protect the core from high temperatures during steady state operation. The core average fuel temperature is independent of core i

size for a given Sk insertion, therefore the maximum fuel temperature attained during a p

j pulse for an NP = 2.2 is also independent of core size for a given Sk insertion.

p D. Loss of Coolant Accident l

l The PSBR pool contains 71,000 gallons of water. For a loss of coolant accident to occur, l

a break in the pool wall or break in a connecting pipe must occur below the bottom of the l

core. A series of alarms will occur as the water level drops more than 26 cm below l

reference pool full level. Just below 26 cm, alarms will notify the reactor operator in the l

control room and the University Police Services. If the reactor is operating at 1.0 MW, a low pool level alarm will alert the operator who is required by admmistratively approved procedure to shut down the reactor. There exists a mdveable gate that can be used to isolate either side of the pool after the leak is noticed.

Emergency procedures call for moving the reactor to the non-leaking side of the pool and isolating that side of the pool with the gate to prevent the water level from dropping below the reactor core.

t i

If the reactor is operating when the leak occurs the reactor operator will shut down the reactor upon receipt of the low poollevel alarm. Within two minutes after the shut down the maximum fuel temperature will drop more than 350 *C.07) Three minutes after the shut down the maximum fuel temperature is within 20 *C of the water temperature."

The largest conceivable LOCA is a break in the 6" pipe connected to the bottom of the l

pool. For this LOCA, it will take more than 1360 seconds (22.6 min.) before the water i

falls below the bottom of the reactor core. Therefore, the minimum time before air l

l convection cooling occurs is about 23 minutes after a LOCA. The fuel has been within l

l 20 *C of the water temperature for approximately 20 minutes before air convection i

j cooling begins.

As soon as the water falls below the reactor core, the fuel temperature will begin to rise, because the natural air convection cooling is less effective than water. The rate of rise of the fuel temperature will depend on the previous operating history of the reactor, and the effectiveness of natural air convection to cool the fuel elements. The time it takes for the water to fall below the bottom of the core once LOCA occurs is Os. The time it takes once air cooling begins until the fuel temperature reaches its maximum temperature is Oe.

Thus, the total time, t, starting when the LOCA occurs until the fuel reaches its maximum temperature,is the sum of two times Os and Oe.

i I

January 14,1997 Rev.1 (4/24/97) i

.=

IX-28 1 Q General Atomic conducted a set of LOCA experiments for TRIGA reactors.UU In these l

experiments dummy TRIGA fuel elements were electrically heated in a grid to determine the rate of temperature increase of TRIGA fuel elements when cooled by natural air convection. The dummy fuel elements were wound with n:sistance wire to simulate a cosine distribution similar to that produced in the core. The standard TRIGA grid-plate assembly pitch for a circular (non-hexagonal) core was used with a seven element assembly to mock-up the central ponion of a standard core. The LOCA experiments were more conservative for two reasons.* The PSBR does not have a central fuel element in the core to block the air flow in the hottest pan of the core.* In addidon, the hexagonal pitch of the PSBR is less likely to produce hot spots on the cladding. When the core is uncovered, the central part of the PSBR core will allow more efficient cooling of the fuel elements in the B-ring increasing the safety factor associated with these calculations.

The time Os may be computed assuming the 6" drain pipe at the bcttom of the pool mptures. In this case, Os = 1360 sec.(23 min.)

To calculate the time Oe,it is necessary to review GA's results as summarized in Figure 9-4 and 9-5. Figure 9-4 shows that with a constant cosine shape power input of 267 watts, it takes approximately 0e = 300 minutes before the maximum fuel temperature reaches the equilibrium temperature, T.qw, of 600 "C. The maximum fuel temperature attainable, i.e., T.qw as a l

function of the source power in watts, is given in Figure 9-5. The thermal time constant of the

(]

fuel after a LOCA is approximately the same for all values of decay power. Thus, it will take v

300 minutes once air cooling begins before the fuel temperature reaches T.qw. The fitted equation for T.qw s a function of s and Tmu (which in the case of Figure 9-4 is 600 'C) is a

e as follows:

Teqruei = Tmu (-8.063 x 10 -3 + 2.193 x 10 20e -2.263 x 1040e2 (37a)

+ 1.193 x 10 40e - 3.031 x 10-90e + 2.929 x 10a29,5h 0<0e<300 minutes.

3 4

Before the maximum fuel temperature reached during a LOCA is determined, the core operation history and maximum fuel element powers must be established. As shown in l

Figure 9-5, the maximum fuel temperature reached during a LOCA is directly related to the decay power and the decay power is a function of the pre-LOCA reactor operating history.

The following assumptions are conservative and are used to determine the decay power.

The PSBR is licensed for a maximum steady state power of 1 MW and normally operates on a 40 hr/wk schedule. Even when the reactor operates on a two shift schedule for reactor operator instruction or for laboratory experiments, the average power per 40 hr week is much less than 1 MW. For the 15 years from 1967-81, the PSBR was operated an average of 15 MW hr/wk. For the 13 years from 1981-94, the PSBR was operated an average ofless than 5.5 MW hr/wk. Assuming the PSBR operates for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at 1 MW l

during the week establishes an upper limit for its operation. However, to cover any future increase in operational activities, it will be assumed that the reactor is operated continuously for one week at 1 MW. With this assumption, the fission product decay power can be determined in the following manner.

O January 14,1997 Rev.1 (4/24s7)

. o i

O O

O L

700.00 i

I t

+

?

i i

500,00 -

l' h 400.00.

f + CL+6.5-l l

5e&

300.00 -

200.00 i

f W

l E

100.00 g

rL O

'S 0.00 l

0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 L

Dapse Time (Minutes)

L M

i s

Figme 9-4 The Time Dependence of Air Cooled Fuel Body for Center Element With 267 W Input (Data from s

l

'c d

Reference 18)

. ~.

~

.m.

s s

N L

800 ir R

M 700 -

5 c

O j

3

\\

e 600 -

,~

E

.~

ng p

E p-d b

.B g" 500 -

R f 400 -

x E

.4...-

m 2 300

~,-

1 u

8

$ 00 -

2 100-O O

50 100 150 200 250 300 350 400 Power Input per Bement (PD)(Watts)

Figure 9-5 Summary of Equilibrium Data for LOCA Simulation Showing the Fuel Element Cladding Temperature versus Power Input to the Element for all Seven Dummy Elements Heated with the Same Power Input (Data from Reference 18).

t m.

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

IX-31 El-Waki100 gives the following equation for decay power:

P

'O.1(0, + 10)*2 - 0.087(0, + 2x10')**

!- =

P, l

- 0.1(0, + 0, + 10)*2 - 0.087(0, + 0, + 2x10 )*2' (38) 7

where, Ps =

the power after shutdown produced by the fission product

decay, Po =

the steady state power before LOCA, i.e.1 MW,

(

Os =

the time after LOCA initiation, i.e.,1360 sec, 00=

the time of operation at power before LOCA,i.e.,

5 168 hr. x 3600 sec/hr = 6.05 x 10 sec.

Usin,g these values, at time Os fter LOCA initiation or at the beginning of air convection a

coohng:

P*

-- = 1.65x10,

t l

l or if we assume a maximum elemental power density, MEPD, of 24.7 kW, the maximum power due to fission product decay, P at time Os, is O

o l

l Po = P,N,1000 kW (38a)

P,

= 1.7 x 10-* x 24.7 kW/ fuel element

= 410 watts Assuming the core is uncovered and reaches equilibrium fuel temperature 1360 sec (~ 23 l

min.) after a LOCA, the hottest fuel element in the core will have less than 410 watts of power which will continue to decay.

Figure 9-5 shows the equilibrium fuel cladding temperature as a function of fuel element power input. The data of Figure 9-5 have been fit with a straight line equation,i.e.,

T"" = P + 62.89 (39) o 0.5498 where P is the decay power (watts) producing temperature in 'C.

o When P = 410 watts then tmax = 860 *C.

o O

'~"~~"~

p IX-32 This shows, that once air cooling begins and if the average decay power remams constant at l

410 watts for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the maximum fuel temperature is 860 *C. However, the decay power does not remain constant but decreases exponentially. By combining l

Equations (37a), (38), (38a), and (39) an equation of T w.i as a function of time after the l

LOCA inidadon is obtained. This equation indicates that the fuel reaches a maximum l

temperature of 468 *C approximately 7160 seconds after LOCA inidation, with a Os (dram 1

l down time) of 1360 seconds, and then continues to decrease (see Figure 9-6). If Os is l

reduced by half to 680 seconds the fuel reaches a maximum temperature of 480 *C l

approximately 6250 seconds after LOCA and then continues to decrease. For the most i

severely conceived LOCA for the PSBR, the maximum fuel temperature remains well below the 950 *C limit throughout the LOCA. It should also be stated that the experimental data of Figure 9-4 shows that because of the long time it takes to heat up a TRIGA fuel element, i.e.

Ge, once air cooling begins, the time Os is less critical.

l The LOCA for a TRIGA core may also be analyzed analytically instead of by the results of experiments used above. General Atomic used one of their own two dimensional transient-heat transport computer codes to calculate the TRIG A fuel element temperature after the loss of pool water.(19) It was assumed that the reactor was operating for an infinite period of time.

l Their results are plotted in Figure 9-7 for several cooling or delay times showing maximum l

fuel temperatures in the TRIGA fuel element as a function of its operating power. It can be observed that a fuel element having approximately 24.7 kW before the LOCA, will attain <

950 "C maximum temperature,1360 seconds after the LOCA.

l O l Q The 950 *C maximum fuel temperature is important when analyzing the TRIGA core for a l

LOCA. During a LOCA, the fuel element is uncovered producing cladding temperatures greater than 500 *C. Under these conditions, if 950 'C fuel temperature and 950 'C cladding temperature is reached or exceeded, the TRIGA cladding could be ruptured.03 30) l Below a fuel temperature of 950 "C the cladding remains intact. The strength of the fuel l

element cladding is a function ofits temperature. The yield strength of the stainless steel l

cladding (assuming that the cladding design specifications of the TRIGA fuel elements do not change between procurements from the as-tested cladding) under LOCA conditions (heated in air),is shown in Figure 9-8. Also shown in Figure 9-8 is the cladding stress produced by any l

gas in the gap. This gas pressure consists of the hydrogen gas pressure plus the pressure of the volatile fission products plus the pressure of the trapped air. It can be obst :d that the cladding stress equals the cladding yield strength at approximately 950 'C. This is different l

from the safety limit (1150 *C) which is the case when the cladding is in water and cladding temperatures remain below 500 *C.

l The following results are applicable to the PSBR experiencing a LOCA:

1. The maximum temperature that the PSBR TRIG A fuel element can have during a LOCA without damage to the cladding is 950 *C. As long as the 950 *C temperature limit is not exceeded, there will be no stress sufficient to rupture the cladding thereby allowing the escape of fission products.

January 14,1997 Rev.1 (4/24/97) l l

O O

O i

500 L

%-qMm d

j 400-i b

G t

Assumptions:

g 300'

l. Time from IDCA initiation to completely uncovered core-1360 seconds.

l 1

2. Maximum fuel element decay power at the beginning of air cooling of the Q

uncovered cote-410 watts.

8

3. It is assumed that the fuel and the cladding temperature are equal with an D

air cooled core.

i O

b vo G

t--

100 i

E C

M

'r 8

0 0

2500 5000 7500 10000 12500 15000 17500 20000 Tune After LOCA Inidation (sec) 2 a

Figum 9-6 Fuel / Cladding Temperatum as a Function of Time After LOCA Initiation 6

u d

O O

O v

2000 y

> 0 Cooling Time c:

t.a

.10e3 Cooling Time "

1800-

-e- 0 Cmling Time J

i

_E

- -A- - 10c3 Cmling Time

- - - - 10e4 Cmling Time e

/

/

1600 -

/

- - - 10e5 Cmling Tune w

f

- - O- - 10e6 Cmling Time

• 10e4 Cooling Time L

_ 1400-f 2

P

/

?

X

. +. -

1:

2d f,i 1200-8.

'g.

5

'4'

" 1000-

.. ~

.s g-4

.010e5 Cooling Time O

800 ef S

.A'

,.o-

/

,/

,g ~~

-W

• y..-

.V

,/,,.-

.O-10e6 Cooling Time

' - o

_.. o

~ ~. _o.

- e

~. - 0

/'.-

O'-

400

,k'

/.4..-

C'. -

s.-

es<

A.-.O'

,,.. o

-O 200-x -I' M ', -

, *. 4 ".. O ' -

r 0$

0.00 5.00 10.00 15.00 20.00 25.00 30.00 35.00 40.00 45.00 l

Operating Posw Density (kW/ Element)

Figum 9-7 Maximum Fuel Temperature Versus Power Density After LOCA For Various Cooling Times Between Reactor Shutdown and LOCA Initiation (Data from Reference 19)

- ~

IX-35 O

1.00E+05 i

l A

in-

/

'W.

/

N

/

s2--+ ___.

N f

^

g 's f*

1 x.

/

\\

\\

1.00E404 y

x

/\\

t ?

N

'/

N 1

/

N

=

/

=

?

I M

lO j

m t

1.00E+03 f

/

4 l

/

c

?

I

- -+- - 2% Yield llll Ultimate Tensile e

J, Stress Duc To Hydrogen i

1.00E+02 400 500 600 700 800 900 1000 1100 1200 Temperature (*C)

Figure 9-8 Strength and Applied Stress as a Function of Temperature, U-ZrH1.65 Fuel With Fuel and Cladding the Same Temperature (Data from Reference 13)

)

January 14,1997 Rev.1 (4/24/97)

l IX-36 V'

2. Reviewing the complete history of the TRIGA cores at Penn State, the PSBR has never been operated for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at 1 MW during a week with the 12 wt% U fuel. In addition, no core configuration is permitted to operate at a power level such that the MEPD is greater than 24.7 kW. Assuming continuous operation at 1 MW with the 12 wt% U fuel l

producing 24.7 kW, a LOCA will not cause the fuel element to heat up to 950 *C under any condition.

i In conclusion, a LOCA with the PSBR will not result in damaged ~ fuel and, thus, fission l

product containment within the fuel is assured.

l E. Maximum Hypothetical Accident (MHA)

The maximum hypothetical accident (MHA) is an assumption that a fuel element cladding ruptures in an air cooled core releasing volatile fission products to the reactor bay. The MHA is defined as a postulated accident with potential consequences greater than those from any event that can be mechanistically postulated. The assumptions create conditions far more l

severe than is actually possible. Nevertheless, the accident is bounded by the regulatory limits.

I The potential hazards associated with the MHA are related to the escape of fission products from the ruptured cladding of a fuel element into the reactor bay and then from the bay to the environment. The fission product buildup in a fuel element is a function of the fuel element power history and sustained measured steady state fuel temperature. It is assumed in these l

calculations that the fuel element is a 12 wt% U-ZrH fuel element operated in the core position l

fm of highest power density. The core is assumed to operate continuously at 1 MW throughout its Q

life. A review of the operating history of the 12 wt% U-ZrH fuelelement I-15 shows that when operating at 1 MW it reached a maximum measured fuel temperature of - 600 *C with Core Loading 47, however 650 *C is used as the bounding sustained measured steady state fuel temperature used to calculate the release fraction. Since all of the volatile fission products reach their maximum after a few weeks of operation, a value of MEPD = 24.7 kW is used to compute the fission product activity, Arp, in a fuel element. It is assumed that the rupture occurs when the reactor is just completing continuous 1 MW operation. The saturated activity of the core, R, for one fission product (f ) nuclide is p

l 3.1x10

i R = 1 MW x Yo 3.7x10' curies (40a)

~

l

where, fission product cumulative yield Y

=

o l

1MW 3.1 x 1016 fission /sec

=

l 1 curie =

3.7 x 1010 disintegration /sec The fission product activity in the core is an accumulation of f activity from the previous p

weeks operation provided the halflife is greater than approximately a day. The contribution to the f activity at the time of the rupture is as follows:

p AV)

January 14,1997 Rev.1 (4/24/97)

/

l l

l t

IX-37 D

(V R(1 - e~"')

Activity produced during the week the rupture occurs, l

R(1 - e-"' )e-*' +)

Activity produced during the week before the rupture occurs, and R(1 - e~"' )e* +)

Activity produced during the nth week before the rupture occurs.

l The total acdvity of this f in the core at the time of the rupture is the sum of the above p

activities. It can easily be shown that the core activity is

1. EAT' Co = R (40b) 1 - e4T +T2) i

and, 1 - e-"',g (40c)

A, = NP R where in the case of continuous operation at 1 MW, T

number of hours operated in one week =168 hours

=

3 number of hours not operated in one week = 0 rc urs T2

=

T +T2=

number of hours in one week = 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> i

O The PSBR presently operates on a 40 hr/wk schedule. The actual operation produces (d

much less than 40 MW hours (MWhr) per week. For the 15 years from 1967-81, the PSBR operated an average of 15 MWhr/wk. For the 13 years from 1981-94, the PSBR was operated an average ofless than 5.5 MWhr/wk. Continuous operation at 1 MW, therefore, establishes a large safety factor for this calculation, but allows for future increase of operating time.

Tables 9-7 and 9-8 lists the activity of each of the important gaseous fission products.

The fission product yields were taken from the Katcoff, et.al., report.(2 These yields include direct production plus precursor decay from fission products. Only a fraction of these fission products escape from the fuel element into the reactor bay. The fraction that escapes from the fuel element is called the release fraction, fr.

l The fission product release fraction, determined experimentally at General Atomic,is a i

function of the sustained fuel temperature.(20 The only fission products that escape are those that have diffused into the gap between the fuel and cladding during the operation of l

the reactor. Therefore, the release fraction in accident conditions is characteristic of the sustained normal operating temperature and not the temperature during an accident transient.

l A review of the operating history of the instrumented fuel element I-13 in the PSBR B-ring shows that its maximum measured temperature during steady state operation at 1 MW was less than 460 'C. After the first year of operation,its measured temperature at 1 MW dropped to approximately 400 *C. During this period, its burnup was approximately 0.65 MWD. This temperature drop occurs because as burnup increases, O

January 14,1997 Rev.1 (4/24/97)

U l

l IX-38 Ov the 35U core inventory decreases with a corresponding drop in NP and temperature.

2 Operation after one year lowers the maximum measured fuel temperature in the D-ring to 400 t

l

• C or less. However, the experience mendoned above with I-15 in Core Loading 47, which l

is considered as an extreme (having a large peak to average temperature) core loading, l

indicates that measured fuel temperatures as high as 600 *C are possible. Thus, it is l

conservative to use a maximum measured fuel temperature of 650 *C, to compute the release fraction. Operadon using the mixture of 12 and 8.5 wt% U fuel eternents will only be allowed l

l by Technical Specifications if the MEPD is s 24.7 kW and the maximum measured fuel l

temperature, of an instrumented fuel element in the position of MEPD, is 650 *C. Thus the initial assumptions and the limits on operation resulting from this accident analysis are:

1. A maximum power operating history of continuous 1 MW operation.

)

2. A maximum sustained measured steady state fuel element temperature of 650 *C.

3. A MEPD of 24.7 kW.

Experiments demonstrated at General Atomic (20 generated release fraction data for the U-ZrH fuel under various conditions. Below 400 *C the release fraction, fr, is a constant, 1.5 x 10-5, and above 400 *C the followmg equation is used l

f, = 1.5x10~5 + 3.6x10 exp( 1.34x10" f-l (41)

To

[V-)

where, i

1 T, is the fuel temperature in Kelvin.

i l

For low temperature results, i.e., below 400 *C, the release fraction for a typical U-ZrH, i

I fuel element is constant, independent of operating history or details of operating temperatures. Averaging Equation (41) over the volume and temperature profile of the fuel element gives a release fraction of 3.1 x 104 using maximum measured fuel temperature of 650 *C. Applying the release fraction of 3.1 x 104to a single element operating at a MEPD l

of 24.7 kW yields, for each fission product activity in Table 9-7 and 9-8, a release to the reactor bay of Co curies. The bay concentration, Cb(Ci/ml) is based on a minimum free air bay volume of 1900 m3. An immediate and complete mixing is assumed to occur.

'Ihe concentration in the unrestricted area (outside the reactor building) is obtained by dividing the activity release rate through the emergency exhaust system by the dilution rate.

The release rate for the emergency exhaust system is equal to the flow rate (3100 cfm or 1.46 x106 ml/sec) times the bay concentration Cb. The dilution rate is the wind velocity (1 m/s)

I times the cross-section area of the building (200 m2) or 2 x 108 ml/sec. Thus, at the instant the fuel element cladding ruptures, the maximum concentration in the unrestricted area, l

Cu(Ci/ml)is:

Cu = C x 7.30x10-3 Ci/ml.

b January 14,1997 Rev.1 (4/24/97)

aA a

.s IX-39 i

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January 14,1997 Rev.1 (4/24/97) l l

l l

7

\\'

ACTMTY. Cl UNHESTf9CTED PERCENT OF ANNUAL L!MIT YlELD HalfOle CORE MAxlMUM Decayed RE ACTOR BAY AREA BAY (Mastnaed Ares W

Sewula TOTAL (Col ElEtENT Aalvtry (Cb1 pCVmt DAC FRACTION t!MIT Fraction 1 MIN 1 HOUR 24 HOtRS Br-83 0 5100 8 64E+03 4 27E+03 106E+02 106E+02 1.72E45 3 00E-05 5 74E41 9 00E48 1.40E+00 4 66E44 4 97E43 5 22E43 Br-84 0 0190 3 60E+02 1.59E +02 3 93E+00 3 93E+00 6 42E-07 2 00E45 3 21E42 8 00E4R 5 86E-02 2 47E45 6 90E45 6 90E45 Br44m*

0 9200 1.91 E +03 7 71E,03 1.90E +02 190E+02 311E-05 1.00E 47 311E+02 1.00E49 2 27E+02 2 50E41 6 25E41 6.36E41 Br 85*

1.1000 1.72E+02 9 22E+03 2 28E+02 228E+02 3 71E45 1.00E47 3.71 E +02 1.00E49 2.72E+02 2 69E41 180E-01 180E-01 A

Br 87*

2 0000 5 61E+01 1.68E +04 4.14E+02 4.14E+02 6.75E45 1.00E 47 675E+02 100E49 4 94E+02 3 89E41 1.19E-01 11CE41

• g Brty 3 8400 164E +01 3 22E+04 7 95E +02 7 95E+02 1.30E 44 100E-07 1.30E+03 1.00E -09 9 48E+02 3 67E41 6 99E42 6 99E42 Br 89*

5 4000 4 40E+00 4 52E+04 1.12E +03 1.12E +03 18?E44 100E47 182E +03 100E49 133E+03 160E41 2 67E 42 2 67E42 Ar-90*

58000 180F +00 4 93F+04 122F +03 1??F+03 199F-04 100F47 199F+03 100E 49 145E +03 7.15E 42 119F 42 119F42 RROMtNF TOTAt 165F+05 4 07F+03 4 07F+01 6 46F+03 4 73F+03 15 10 1J y

L129 0 0000 5 02E+14 6 70E+03 166E+02 166E +02 2 70E45 4 00E49 6 75E+03 4 00E-11 4 94E+03 5 50E+0C 1.91E+01 2 03E+01 Q

L131 3.1000 6 95E+05 2 60E +04 6 42E,02 6 42E+02 105E44 2 00E46 5 23E+03 2 00E-10 3 83E+03 4.26E+00 1.48E+01 1.57E+01 L132 4 3800 8 24E+03 3 67E+04 9 06E +02 9 06E +02 1.48E44 3 00E 46 4 93E +01 200E48 5 41E +01 4 00E42 1.92E41 201E41

[

L133*

6 9000 7.49E+04 5 78E+04 1.43E+03 1.43E+03 2.33E44 100E-07 2 33E+03 1.00E49 1.70E +03 1.90E +00 6 51E+00 693E+00 0134 7 8000 3.16E+03 6 54E+04 1.61 E +03 161E +03 2 63E44 2 00E-05 1.32E+01 6 00E48 3 21E+01 1.07E42 1.00E 41 1.03E-01 g

1-135 6.1000 2 37E+04 5.11 E +04 126E+03 1.26E+03 2 06E-04 7 00E47 2 94E+02 6 00E 09 2 51E+02 2 39E41 9 40E41 9 96E41 g

t136*

3.1000 e.50E+01 2 60E +04 6 42E+02 6 42E+02 105E44 100E47 105E +03 1.00E-09 7 65E+02 6 75E41 2 72E41 2 72E41 0137*

62500 2 45E+01 5 24E +04 129E +03 1.29E +03 211E 04 100E 47 211E +03 1.00E -09 154E +03 8 32E 41 1.68E 41 1.68E41 L138' 5 6600 6 50E+00 4 74E+04 1.17E +03 1.17E+03 191 E-04 100E47 191E+03 1.00E49 140E +03 2 47E41 413E42 4.13E-02 L139' 5Eino 2 40F+00 4 72F+04 1 17F +03 1 17E +03 19nF 44 100F 47 190F+01 100F 49 139F +03 912042 1 52F -02 152E42 IODINF TOTAL 417F+05 103F+04 103F+04 216F +04 1 59F +04 14 42 45 HALOGEN TOT A1 5 B' F + 05 144E +04 1 4 4E + 04 2 81E +04 2 0f E +04 15 43 40 NOBt F G AS Pi tJS hat OGFN 1 11 F +06 2 74F.04 2 74F+04 3 9FF+04 2 92F+04 21 47 50 Aporovmate TEDE.

1038 24 20 Apprommata Thyroid Dose.

6415 698 744 Table 9-8 F4ssion Produa Data. Fisson Produa Release Ior the Halogens and the Restr6cted and Unrestrided TEDE for the MHA TABLElk7.8 NOTES COOL DOWN TIME o 00 hr RELE ASE FRACTION 310E44 SAY VOLUME 190E+09 mi EXHAUST VE NTLATION 146E+06 musec l

BUlt.DNG OLUTION 2 00E+0e ersec NUMBER OF FUEL ELEMENTS 100 HRADITION TIME TO MAxmai2E ACTIVITY Core actwey-R yield 11+"(4mmtusa*T1)M1+*(lamtxsa(T1+T2)]

P.1 MW powerlevet it.16s hrima aperateg tsee T2= 0 hrAn4 shuesown R-31E+1e tasonstnec or 1 MW NP 2.47 rate of mammum to average bei elemers actway

• A DAO veke of 1E-7 m used tar those r=*a=atopes weh hall Me.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and not hated a 10CFR20 Appendia B

• A--'

inn Ema et 1E e a used for unreatncied areas for those en -

. weh has we.2 houg wNch are nos Ested m 10CFR20 Appendu 8 Percere of annuma ime a based en 2000 DAC4eurs for tw seacsor bay and 8780 *hme-hours

• 9er the _

^amt Thysoid Dose incbdes percers at annual ama lor r

_ and edemal does Dom 4129 through 4135.

Annual 1hyveaf dose 4me a osannamed hem a 50.000 mrom hms ter resmceed areas and 1870 mrera for unresenceed areas (50 mroen CEDEJD 03 tesus weigherg tacsork

IX-41 In the reactor bay (restricted area), the concentration in air of airbome activity decreases with time because of radioactive decay and the removal of air by the emergency exhaust system.

Consider a model in which activity, released fonn the single fuel element, instantly mixes completely with air in the bay. The resulting uniform concentration is C. We have, then, b

l i

id+*')'

(42) i C =C e 7

o

where, bay air concentration of activity at elapsed time t after release from j

Cr

=

the single fuelelement (Ci/ml)

A decay constant (sec-')

=

o ventilation rate constant (sec-')

A,

=

ventilation rate 1.46 m sec-'. = 7.7x10" sec-'.

bay volume 1900 m l

l The exposure to an airborne concentration is equal to the integral of Cr over the period l

T, labeled ICr.

Yd * *')*

(43) l IC = A + A, s1-e T

o l

)

O The reactor bay exposure in DAC-hours (derived air concentration hours) is equal to ICT divided by the DAC value from 10 CFR Part 20 Appendix B. The same technique is used to determine the exposure in effluent limit concentration hours for the unrestricted area. The i

exposures determined for the reactor bay and the unrestricted area are then compared to the limits of 2000 DAC-hours for the reactor bay and 8760 effluent limit concentration hours for the unrestricted area, respectively, to arrive at the percent of the annuallimit.

The activity is removed rapidly from the reactor bay and about 92% of the TEDE in the I

unrestricted area is received in the first hour. Essentially all activity has been released to the unrestricted area within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and doses in both the reactor bay and the unrestricted area l

have reached their maximum values. Release of the activity from a fuel element over an l

extended period of time would reduce the dose because of the decay of short half-life l

l radioisotopes before release.

l i

The total exposure in 1 minute for the reactor bay (1 minute is considered a maximum evacuation time)is a TEDE of about 1038 mrem, corresponding to about 21% of the annual 10 CFR Part 20 limit. An exposure time of 7.5 minutes in the reactor bay would be required to reach a TEDE of about 4988 mrem, corresponding to about 99% of the annual 10 CFR Part 20 limit.

The total exposure in the unrestricted area is about 26% of the annuallimit, which corresponds to a TEDE in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of about 26 mrem (the annuallimit in the unrestricted area is 50 to 100 mrem depending on the fractions of the dose from radioisotopes with an internal dose l

component).

i f

January 14,1997 Rev.1 (4n4/97) l

_~

l IX-42 O

\\

/

Since the emergency exhaust system removes the bay air through a charcoal filter, the iodine exposure in the unrestricted area should be less than 10% of that given in Table 9.7. Since credit is not taken for the emergency exhaust system filtration in determination of the consequences of this accident, a Technical Specification requiring filtration as a limiting condition for operation is not required.

j The above calculations assumed an MEPD of 24.7 kW and a maximum measured fuel temperature of 650 *C. Using the higher temperature increases the release fraction and the higher power density increases the fission product inventory. In addition, it was assumed that there are no mitigating circumstances such as fission products plating out on surfaces or dissolving in the pool water. It was further assumed that the rupture occurs to the element with the highest fuel temperature, the highest power density, and occurs immediately after 1

continuous 1 MW operation. Therefore using these parameters provide very conservative (hypothetical worst case) results.

The MHA creates conditions far more severe than are actually possible. Therefore, the fact that the MHA consequences are within the 10 CFR Part 20 limits outside the reactor bay in the unrestricted area shows that the PSBR's operation is safe to the public. The emergency evacuation alarm which is initiated by the reactor bay radiation monitors and/or the bay air monitors will minimize the time the operation personnel spend in the reactor bay during the l

MHA to much less than the 7.5 minute time interval used in determining the consequences in the reactor bay. Therefore the operation personnel can also avoid any significant hazard from the MHA (within the 10 CFR Part 20 limits for radiation workers).

(7 F. Reactivity Accident

(~)

In this accident, it is assumed that the reactor is taken to a 1.15 MW power level with the transient rod inserted in the core and then the reactor is pulsed with a $3 reactivity insertion. This accident requires a breakdown in the PSBR Standard 'gerating Procedures, the overpower scrams, and a failure of the interlocks.

l When the core is operating at 1.15 MW, its total reactivity has been reduced by more than $4 depending on the average core fuel temperature. The measured fuel temperature in the B-ring using extrapolated experimental data for Core Loading 47' with 95.5 elements would be 650 'C in the 12 wt% U fuel element, I-15. The maximum temperature will be slightly l

l higher, but the fuel temperature near the cladding will be approximately half this temperature.

~

The maximum allowed core reactivity of $7 leaves less than $3 available for pulsing l

(actually the reactivity loss for Core Loading 47 would be = $4.75 at 1.15 MW leaving only $2.25 for a pulse).' Should a $2.25 pulse occur while the reactor is at 1.15 MW, the measured fuel temperature will rise from 650 *C to 1030 *C as calculated using the maximum measured $2.25 pulse temperature' for I-15 and Core Loading 47. In this case, when the core is pulsed from an initial power of 1.15 MW, the maximum fuel temperature is the measured fuel temperature. This is because the temperature rise during l

d Core Loading 47 is considered an extrema loading relative to steady state measured peak fuel temperatures.

• These data are based on that obtained in 1994 & 1995.

(

January 14,1997 Rev. I pn4n7)

l IX-43 r'

)

(

a pulse has a different radial shape than that attained during steady state operation. During a pulse, the increase in fuel temperature is a maximum near the edge of the fuel. Superimposing this shape of the fuel temperature on that attained at a steady state power of 1.15 MW produces, at the end of the pulse, a relatively flat radial temperature distribution at approximately 1030 *C.

j However, since the negative temperature coefficient acts immediately as the transient rod moves i

upward, the final maximum fuel temperature will be less than 1030 *C. If the steady state fuel temperature is higher for a particular loading the reactivity loss due to temperature feedback is greater and the reactivity available for a pulse is subsequently less. Effectively the core excess reactivity limit also limits the maximum temperature of the fuelin this accident independent of the initial steady state fuel temperature. Therefore with the limit on the core excess reactivity the initial steady state core temperature and the maximum reactivity available for the accidental pulse work against each other (raising one lowers the other and conversely) to limit the final maximum fuel temperature to less than 1030 *C.

j 1

Administratively, if the reactor is operating above 900 kW all four control rods must be balanced.

This implies that the transient rod would not be available for the $2.25 pulse. In addition, the Technical Specification required pulse interlock (section 3.2.4), which prevents initiation of a pulse when reactor power is greater than 1 kW, will prevent the postulated accident. Also the Technical Specification required high power scrams will prevent operation at 1.15 MW. The result of the reactivity accident are peak fuel temperatures less than the safety limit of 1150 *C.

The $5 ramp analysis was performed for AFRRI by General Atomics.(24) It indicated that even i

with a reactivity addition rate of $2.50/second (averaged over the full rod travel) the safety limit j

q (1150 *C) was not reached. The rod withdrawal was terminated with a high power scram less Q

than 1 second into the event. A reactivity of $1.86 was added after criticality was achieved and before the SCRAM occurred. The maximum power in the transient was 330 MW with a maximum fuel temperature of 330 *C. Compared to the above analyzed reactivity accident this excursion is inconsequential. It should be noted that the amount of reactivity available in the ramping rods does not impact on the final result as long as the reactivity addition rate does not exceed the $2.50/second rate and the SCRAM time is not significantly longer than that analyzed.

G. Conclusion There are two limits which,if not exceeded, will prevent mpture of the cladding of a TRIGA fuelelement. They are:

1. Limit the fuel temperature to a maximum 1150 *C when the cladding temperature remains below 500 *C, i.e., when the fuel is covered with water.

2. Limit the fuel temperature to a maximum 950 *C when the cladding temperature is the same as the fuel temperature i.e., as with an air cooled core after a LOCA.

The Technical Specifications for the PSBR are established to prevent reaching these two limits.

l The 1150 *C temperature limit is not reached as the fuel temperatures are limited during pulse l

mode operations. Equation (34) provides a direct method for determining the maximum fuel temperature based on the measured fuel temperature during a pulse. Using this equation the following limits are established:

January 14,1997 Rev.1 (4/24/97)

IX-44 O

v

1. The maximum allowed reactivity insertion for the pulse mode and the maximum allowed worth of the pulse rod is $3.50. A sudden insertion of $3.50 excess reactivity results in a maximum peak fuel temperature of 1095 *C and a measured peak fuel temperature of 684 *C if the NP s 2.2.

2. With any core loading the maximum radial peaking factor, called the normalized power, NP, in the SAR is 2.2 if the transient rod wonh is $3.50. This ensures that a pulse with the full travel of the maximum allowed transient rod wonh will not cause the fuel temperature in any fuel element to exceed the safety limit of 1150 *C. If the maximum allowed pulse is less than

$3.50 for any given core loading (i. e, the pulse can be limited by the worth of the transient rod, by the core excess, or administratively) the maximum NP can be increased as long as a calculation by an accepted method (documented in an administratively approved procedure) is done to show that the safety limit is not exceeded with the allowed pulse and NP. The limits shall be either physical or administrative or both.

3. The maximum allowed excess reactivity of the core is $7. Thus, when the core is operating at 1.15 MW steady state, a maximum of $2.25 of excess reactivity is available for pulsing, a minimum of $4.75 of excess reactivity is needed to reach 1.15 MW. Based on core loading 47

)

and I-15 ( an extrapolated measured fuel temperature of 650 *C) at 1.15 MW prior with a pulse insertion of $2.25 ( measured fuel temperature of 380 *C) the temperature would equal 1

1030 *C. However the Technical Specification required interlock prevents pulse initiation when the power is above 1 kW and the Technical Specification required high pwer SCRAMS prevent operation above 1.1 MW.

4. Core configuration limitations are also established to prevent a fuel element from producing toc much power relative to the other fuel elements. The maximum elemental power density, MEPD, allowed is 24.7 KW. If core size and or NP leads to a MEPD greater than 24.7 kW when the reactor power is 1 MW the maximum allowed reactor power must be administratively reduced to reduce the MEPD to 24.7 kW or less. The maximum allowed reactor power to maintain the MEPD less than 24.7 kW for a given core configuration shall be determined by calculation by an accepted method (documented in an administratively approved procedure).

Limits set for steady state operation prevent the maximum fuel temperature reaching 1150 *C.

Limits imposed here prevent the fuel temperature during a LOCA from reaching 950 *C. If operated at 1 MW continuously, a single fuel element could operate at its maximum power level of 24.7 kW and still not have its fuel temperature reach 950 *C during any conceived LOCA. In addition, the 24.7 kW MEPD and the maximum measured fuel temperature of 650 *C during steady state operation limit the release of fission products such that the consequenses are within the 10 CFR Pan 20 limits if the cladding ruptures.

The MHA analyzes the effect of a fuel element cladding rupture in air after continuous operation at 1 MW In addition, the reactor was assumed to have operated continuously during the previous year. Under these extreme conditions, the maximum TEDE to a person in the unrestricted area is l

29 mrem after 1 hr and 30 mrem after 24 hr.

1 O

January 14,1997 Rev.1 (4/24/97)

V l

l

IX-45 V(3 In conclusion, the analyses described in this section shows that under no possible accident conditions will the regulations in either 10 CFR Part 20 or 10 CFR Part 100 be violated. Thus, the PSBR can be operated safely within the regulatory limits.

H. References

1. Naughton, W.F., Cenko, M.J., Levine, S.H., and Wi..ig, W.F., "TRIGA Core Management Model," Nucl. Technology. vol. 23, p. '56 (Sept.1974).

2. Naughton, W.F., Cenko, M.J., Levine, S.H., and Witzig, W.F., Increasing TRIGA Fuel Lifetime with 12 wt% U TRIGA Fuel, TOC-5, TRIGA Owner's Conference III (February 1974).

3. Haag, J. A., and Levine, S.H, " Thermal Analysis of The Pennsylvania State University Breazeale Nuclear Reactor," Nucl. Technology. vol.19, p. 6 (July 1973).

4. Levine, S.H., Geisler, G.C., and Totenbier, R.E., Temperature Behavior of 12 wt% U TRIGA Fuel, TOC-5, TRIG A Owners' Conference III (February 1974).

5. Levine, S.H., Totenbier, R.E. (Penn State Univ.), and Ahmad T. Ali (PPAT Ismail -

Malysia), Fourteen Years of Fuel Management of the Penn State TRIGA Breazeale Reactor (PSBR), ANS Trans. vol. 33 (November 1979).

6. Levine, S.H. and H. Ocampo, "The k -Meter Concept Verified via Subcritical/Cridcal TRIGA Experiments," Proceedings of the International Symposium on the Use and (qj Development of Low and Medium Flux Research Reactors, MIT, Cambridge, MA (October 1983).

7. Kim, S.S. and S.H. Levine, " Verifying the Asymmetric Muldple Posidon Neutron Source (AMPNS) Method Using the TRIGA Reactor," Ninth TRIGA User's Conference, Anaheim, CA (March 1984).

8. Levine, S.H., " Module 5 - In-core Fuel Management," Nuclear Fuel Cycle Educadonal Module Series, N.D. Eckhoff, gen.ed., Kansas State University (July 1980).

9. Fowler, T.B., et.al., " EXTERMINATOR-II: A FORTRAN IV Code for Solving Multigroup Neutron Diffusion Equations in Two Dimensions," ORNL-4078, Oak Ridge

. National Laboratory (April 1967).

10. Huang, H.Y. and S.H. Levine, "An Automated Multiple-Cycle PWR Fuel Management Code," ANS Trans. (November 1978).

i 11.Cenko, M.J., " Comparison of PSBR Operation's History with the TRIGA Core Management Model," M.S. Thesis, The Pennsylvania State University (1972).

12. Barry, R.F., " LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for IBM-l 7094," WCAP-3269-26, Westinghouse Electric Corporation (September 1963).

O January 14,1997 Rev.1 (4/24/97)

<V

IX-46 O

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Lj

13. Simnad, M. T., F. C. Foushie, and G. B. West, " Fuel Elements for Pulsed Reactors," GA Report E-117-393 (January 1975).

14. El-Wakil, M.M., " Nuclear Heat Transport," ANS (May 1978).

15. Goodwin, W.A., "The Measurement of Radial Power Distribution in a TRIGA Fuel Element During Reactor Excursion," Ph.D. Thesis, University of Illinois (1967).

16. Kim, S. S., " Development of an Asymmetric Multiple Position Neutron Source (AMPNS)

Method for Monitoring the Criticality of the Degraded Reactor Core," Ph.D. Thesis, The Pennsylvania State University (1984).

17.PSBR Log Book 37, page 265 (November 21,1984).

18. Shoptaugh, J. R., Jr., " Simulated Loss-of-Coolant Accident for TRIGA Reactors," GA-6596 (August 1965).

19. West, G. B., " Safety Analysis Report for the Torrey Pines TRIGA Mark III Reactor," GA-9064 (January 5,1970).

20. Katcoff and Seymour, Nucleonics. vol.18, p. 201 (November 1960).

21. Foushee and R. H. Peters " Summary of TRIGA Fuel Fission Product Release Experiments," GULF-EES-A10801 (September 1971).

tk

22. Regulatory Guide 1.109, "Calculadon of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance With 10 CFR Part 50, Appendix I."

23. Intemational Commission on Radiation Protection Report #.,.

24. General Atomics, Analysis of5 Dollar Ramp insertion over 2 SecondIntervalin AFRR1 TRIGA Reactor, General Atomics Publication of work performed for Anned Forces Radiobiological Research Institute, Bethesda, Maryland, April,1988.

25. Coffer, C. O., Dee, J. B., Shoptaugh, Jr., J. R., West, G. B., and Whittemore, W. L.,

" Characteristics of Large Reactivity Insertions in a High Performance TRIGA U-ZrH Core," GA-6216, General Atomics (April 12,1965).

26. W. F. Naughton, " Core Management Program to Optimize Fuel Utilization in TRIGA Research Reactors," Ph.D. Thesis, Nuclear Engineering Department, The Pennsylvania State University, University Park, Pennsylvania (September.1972).

27. D. Hughes, P. Boyle, and S. H. Levine, "A New Management Plan for the Penn State l

TRIGA Reactor with Supporting Experiments and Calculation.s," Unpublished paper l

Nuclear Engineering Department, The Pennsylvania State University, University Park, Pennsylvania (Mar.1996).

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28. NUREG-1282, " Safety Evaluation Report on High-Uranium Content, Low -Enriched Uranium-Zirconium Hydride Fuels for TRIGA Reactors," Docket No. 50-163, U. S.

Nuclear Regulatory Commission (August 1987),

29. M. T. Simnad, G. B. West, J. D. Randall, W. J. Richards, and D. Stahl, " Interpretation of l

Damage to the Flip Fuel During Operation of the Nuclear Science Center Reactor at Texas l

A&M University," GA-A16613, General Atomics (December,1981).

30.M.T. Simnad, "The U-ZrHx Alloy: Its Properties and use in TRIGA Fuel," GA E-117-l 833, General Atomics (February,1980).

31. Stationary Neutron Radiography System Final Safety Analysis Report prepared by Argoone National Laboratory, McClellan Air Force Base (January,1992).

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