AFRRI TR91-1

AFRR TECHNICAL REPORT

I

Maximum Temperature Calculation

and Operational Characteristics

of Fuel Follower Control Rods

for the AFRRI TRIGA Reactor Facility

M. Forsbacka

M. Moore

DEFENSE NUCLEAR AGENCY

ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE

BETHESDA, MARYLAND 20889 5145

APPROVED FOR PUBLIC RELEASE, DISTRIBUTION UNLIMITED
REVIEWED AND APPROVED

MARK MOORE

Reactor Facility Director

GEORGE W. IRVING, III

Colonel, USAF , BSC

Director
NSN 7540 .01 .280-5500 Standard Form 298 (Rev . 2-89)

Prescribed by ANSI Sta. 239-18

298-102

REPORT DOCUMENTATION PAGE

Form Approved

OMB No . 0704-0188

Public reporting burden for this collection of information is estimated to average 1 hour per response . including the time for reviewing instructions. Searching existing data sources,

gathering and maintaining the data needed, and completing and reviewing the collection of information Send comments regarding this burden estimate or any other aspect of this

collection of information, including suggestions for reducing this burden, to Washington Headquarters Services, Directorate for Information Operations and Reports, 1215 Jefferson

Davis Highway. Suite 1204 . Arlington, VA 22202-4302 . and to the Office of Management and Budget . Paperwork Reduction Project (0704-0188), Washington, DC 20503

1 . AGENCY USE ONLY (Leave blank) 2 . REPORT DATE

May 1991

3 . REPORT TYPE AND DATES COVERED

Final

4 TITLE AND SUBTITLE

(see cover)

5 . FUNDING NUMBERS

NWED QAXM

6 . AUTHOR(S)

Forsbacka, M ., and Moore, M .

7 . PERFORMING ORGANIZATION NAME(S) AND ADDRESS(ES) 8 . PERFORMING ORGANIZATION

REPORT NUMBER

TR91-1

Armed Forces Radiobiology Research Institute

Bethesda, MD 20889-5145

9 . SPONSORING/MONITORING, AGENCY NAME(S) AND ADDRESS(ES) 10 . SPONSORING/MONITORING

AGENCY REPORT NUMBER

Defense Nuclear Agency

6801 Telegraph Road

Alexandria, VA 22310-3398

11 . SUPPLEMENTARY NOTES

12a . DISTRIBUTION/AVAILABILITY STATEMENT 12b . DISTRIBUTION CODE

Approved for public release ; distribution unlimited .

13 . ABSTRACT (Maximum 200 words)

Opeational requirements of the Armed Forces Radiobiology Research Institute

(AFRRI) TRIGA reactor necessitate the implementation of fuel follower control

rods (FFCR's) . This technical report discusses the operational and safety

aspects of FFCR installation . Thermalhydraulic analysis shows that FFCR's

can be operated safely in the AFRRI TRIGA reactor . The maximum calculated

fuel temperature at the maximum steady-state power level is well within the

limit of the maximum safe operating temperature set by the Technical

Specifications .

14 . SUBJECT TERMS 15 . NUMBER OF PAGES

24

16 . PRICE CODE

17. SECURITY CLASSIFICATION

OF REPORT

UNCLASSIFIED

18 . SECURITY CLASSIFICATION

OF THIS PAGE

19 . SECURITY CLASSIFICATION

OF ABSTRACT

UNCLASSIFIED

20 . LIMITATION OF

ABSTRACT

UNCLASSIFIED
SECURITY CLASSIFICATION OF THIS PAGE

CLASSIFIED BY:

DECLASSIFY ON:

SECURITY CLASSIFICATION OF THIS PAGE
Contents

Introduction	 1

General Description of Fuel Follower Control Rods 	1

FFCR Maximum Fuel Temperature Calculation	 2

Power Density in FFCR Fuel Element	 3

Maximum Temperature in FFCR Fuel Element	4

Fuel Temperature in Pulse Mode Operation	6

FFCR Operational Characteristics 	 8

Conclusion	 9

References	 9

Appendix A: Determination of Free Convective Heat Transfer Coefficient	11

Appendix B : Reactor Core Loading and Unloading 	15

iii

INTRODUCTION

Operational requirements of the Armed Forces Radiobiology Research

Institute (AFRRI) TRIGA reactor facility necessitate the implementation of

fuel follower control rods (FFCR's) . Fuel follower control rods are like the

standard TRIGA control rods as described in section 4 .10.1 of the AFRRI

TRIGA Safety Analysis Report (SAR) except that they have a fuel-filled

follower rather than an air or aluminum follower . The primary purpose of

the FFCR's is to offset the long-term effects of fuel burnup .

The Code of Federal Regulations (CFR ; Title 10, Part 50 .59) requires that

modifications of a portion(s) of a licensed facility, as described in the facility

SAR, be documented with a written safety analysis . The SAR ensures that

all safety issues associated with the implementation of FFCR's have been

reviewed . This technical report will show that implementing FFCR's will

allow the standard control rods to function in their intended purpose and

will restore core reactivity economically . FFCR's have been implemented in

approximately a dozen TRIGA reactors and have been used for over 20

years without reported failure .

This report has been submitted to the AFRRI Radiation Facility Safety

Committee to ensure that all safety questions have been reviewed before

submission to the U .S . Nuclear Regulatory Commission (NRC), as required

under 10 CFR 50 .59 .

GENERAL DESCRIPTION OF FUEL FOLLOWER CONTROL

RODS

The current AFRRI TRIGA standard control rods were installed in 1964 .

The standard control rod consists of a sealed aluminum tube (0 .065 inch

thick) approximately 1 .25 inches in diameter and 31 inches long . The upper

15 .25 inches of the tube contain a compacted borated graphite rod (B 4 C

with 25-percent free boron or other boron compounds), which functions as a

neutron absorber or poison . The lower end of the tube contains a 15 .25-

inch long and 1 .125-inch diameter solid aluminum rod called the aluminum

follower. The follower functions as a mechanical guide for the control rod

as it is withdrawn from or inserted into the reactor core .

The proposed FFCR's differ from the current standard control rods in the

following respects :

•

The aluminum cladding is replaced by smooth stainless steel (SS304)

cladding with a wall thickness of 0 .020 inch . The inner and outer

diameters are 1 .085 inches and 1 .125 inches, respectively .

•The length of the control rod is increased to 37 .75 inches ; the absorber

and fuel follower section are both nominally 15 inches long .

•

The outer diameter of the absorber section and the fuel follower are

both 1 .085 inches .

•The fuel follower has a solid zirconium rod as its central core with an

outer diameter of 0 .225 inch .
The absorber or poison material of the proposed FFCR's is, however,

identical to the standard control rods presently installed .

The fuel contained in the FFCR consists of a fuel-moderator element in

which zirconium hydride is homogeneously mixed with partially enriched

uranium . The FFCR fuel element contains 12 percent uranium by weight

and has a nominal enrichment of 20 percent in the

235U isotope. The FFCR

fuel element contains about 30 .0 grams of

235 U--this is 79% of the235U

loading of a standard AFRRI TRIGA fuel element . The nominal hydrogento-zirconium ratio in the FFCR fuel element is 1 .7 with a range between 1 .6

and 1 .7 . The FFCR fuel element contains no burnable poison . The

stainless steel cladding on the FFCR fuel element has a hardness greater

than the aluminum control rod guide tubes, so wearing will occur on the

guide tubes rather than on the FFCR fuel elements .

FFCR MAXIMUM FUEL TEMPERATURE CALCULATION

A thermal-hydraulic analysis of the FFCR fuel element to determine the

maximum fuel temperature uses the following model:

•The neutron mean free path for neutrons of all energies is smaller than

the diameter of the TRIGA fuel rods, so the reactor must be treated as

a heterogeneous reactor . Thus, the active volume of the core is taken

to be the volume of fuel contained within the reactor core .

•The ratio of power in a fuel element with 12 wt-% uranium versus 8 .5

wt-% uranium is 1 .21 . This is determined by General Atomics design

calculations . 1

• The reactor is operating at a steady-state power level of 1 .0 MW, and

the heat flux across the fuel element is described by Fourier's law of

thermal conduction:

q" (r) (r) = -kVT (r) (1)

where

q" (r) = heat flux at position r

k = thermal conductivity

T(r) = temperature at position r

For steady-state heat transport, the heat production rate and the rate of

energy loss due to heat transport are equal. This can be generally expressed

as

q ''' (r) = V• q'' (r)	(2)

where

q"'(r) = volumetric heat rate (heat production rate) at position r .

Substituting equation 1) into equation (2) yields the time-independent

equation of thermal conduction :

q''' (r) = -V •kVT(r)

2

(3)
Equation (3) is, thus, the second-order ordinary differential equation that

must be solved to determine the maximum temperature attained in the fuel

portion of the FFCR .

Using this model to determine the maximum fuel temperature divides the

analysis into two separate tasks : determining the power density in the FFCR

in a D-ring grid position and solving equation (3) for the given power

density.

Power Density in FFCR Fuel Element

The anticipated fuel loading for the AFRRI TRIGA reactor core with

FFCR's installed will consist of 77 standard TRIGA fuel elements and the

three FFCR fuel elements . Presuming that the control rods are fully

withdrawn to achieve a power level of 1 .1 MW, the total active fuel volume

will be 30,597 .9 cm3 . Thus, the average power density at 1 .1 MW will be

36 .0 W/cm3 .

The maximum fuel temperature is the important parameter, so only the

radial variation of the core centerline power density is considered . To

determine the maximum power density in the D-ring location of the FFCR

fuel element, the following calculations are made :

For the AFRRI TRIGA, the radial and axial peak-to-average power ratios

are 1 .55 and 1 .30, respectively . 3 Thus the maximum power density (heat

rate) will be

q) max = (1

.55) (1 . 3 0) q ''' ave

(4)

= 72 .4 W/cm3

To determine q"'D-ring relative to q"' X , it is useful to compute a scaling

factor from the gross variation of Thermal neutron flux in the radial

direction thermal flux and power density are directly proportional) . The

normalize radial flux distribution for the AFRRI TRIGA core is best

represented by a Bessel function of the first kind of order zero :

2 .405r

) 	 O ( R J - therm

e

where

Re = 21 .78 cm, the extrapolated core radius

r = 11 .99 cm, radial position of D-ring element

and the Bessel function scaling factor is J o (1 .3240) = 0 .6074 .

The power density for the D-ring is thus computed to be

q ''' D-ring = (0

.6074) q ''' max = 44.0 W/cm3(6)

Because the FFCR fuel element differs from the standard fuel element in

concentration of uranium, the power density in an FFCR fuel element is

3

(5)
greater than the power density in a standard fuel element by a factor of

1 .21 . 1

Taking the above scaling factor into account, the power density of an FFCR

fuel element is found to be

q ) FFCR = (1

.21)q"'D-ring

( 7)

= 53 .2 W/cm3

Note that the calculation of q"'FFCR takes into account the most limiting

condition for power peaking in a 12 w/o fuel element versus an 8 .5 w/o fuel

element . As developed in equation (7), q"'FFCR is still considerably less

than the theoretical maximum q"' as determined in equation (4) . A less

conservative approach would have also accounted for the reduced volume in

an FFCR fuel element .

Maximum Temperature in FFCR Fuel Element

Equation (3) takes the following form for cylindrical geometry with axial and

azimuthal symmetry (see Figure 1) :

[

drkd +q'''=0

(8)

r

The boundary conditions required that constrain equation (8) are as follows :

dT

=Oat r=Ri andT=T i at r=R i(9)

dr

Figure 1 . Cross-section of FFCR fuel element .

4
Integrating equation (8) twice yields a general solution of the form

2

T(r) = _q'''

4k

+ C1ln(r) + C2

(10)

f

Applying the boundary conditions to solve for the temperature difference

between the outer edge of the zirconium rod and the inner surface of the

cladding,

q '' 'R 2

C _	i

12k f

2

C 2 = Ti	q'2kRi (ln (Ri)

2 ) f

Ti - To =

q''

' Rig [ (RO)

2

21n ( R°)1]

4kf R i R i

To account for the transfer of heat from the fuel through the cladding to

the coolant, we must consider the heat conduction between the inner and

outer surfaces of the cladding, qc lad , and the heat conduction from the outer

surface of the cladding to the coolant, qfluid. We make the assumption that

no heat is produced in the cladding or coolant, so the heat conduction

from the outer surface of the fuel, q f	must be equal to gclad and gfluid.

The heat conduction leaving the fuel is given by

gfuel =

x (Ro 2 - Rig) Lq'' ' = 7r ((Ro/Ri) 2

where

L = length of fuel element

dT

gclad -k CA dr

qfluid = hA(Tc - Tf ) = 2x(R0

27rk cL (T0 - T c)

cladR + c

ln( 0 )

R0

+ c) Lh (T c - Tj)

5

- 1)LR i

2q'''

Note that the area, A, used in computing gclad

is the

of the cladding . Recall that

qfuel - gclad =

qf luid

So we can solve for the temperature differences in the above equations in

terms of q"':

logarithmic mean area
2

R + c

Ro 2 Ri

T o - Tc = In

	0

() [ (-) 1]

q

)))

R R . 2

0i

2

T c Tf =1 [ (RO)

2

- 1] Rl

q `

Adding equations (16) with equation (12) and solving for T 1 gives us the

expression for the maximum temperature in the FFCR .

"'Ri ( R°) 2 -

21n(Ro)

- 1

4k

f

2

+

R1

[ (Ro) 2 - 1] [1In (

R° +

2R i kc Ro

where

T . = maximum fuel temperature

Tf = bulk coolant temperature

q`

R = 1 .38 cm (radius of FFCR fuel element)

Ro = 0 .286 cm radius of Zr rod)

c ' = 0 .051 cm cladding thickness)

4

kf = 0.18 W cm- C (thermal conductivity of UZrH)4

k = 0.138/cm-°C (thermal conductivity of SS304) 2

h` = 1 .339 W/cm2-°C (free convective heat transfer coefficient of water)

q"'= 53 .2 W/cm3 (from equation (7))

Note that the free convective heat transfer coefficient, h, was an

experimentally derived quantity . The method by which h was determined is

presented in Appendix A . Solving equation (17) using a volumetric heat rate

of 53 .2 W/cm3 and a bulk water temperature of 48.6°C (the conditions at

which h was determined) yields a maximum fuel temperature of 210 .2°C .

The maximum temperature achieved in the FFCR is nearly 180°C less than

the normal temperature of 390°C in a standard fuel element in the B-ring

during a 1 .0 MW steady-state power operation .

R .

R

Ti =Tf +

(R

o

+c) hRi 2

2

q

1i

Fuel Temperature in Pulse Mode Operation

The Nordheim-Fuchs model predicts the maximum fuel temperature achieved

in a pulse mode operation .' The fundamental assumptions of this model are

as follows :

• The neutron flux in the reactor is separated into a spatial component

(shape factor) and a time-dependent component (amplitude factor), such

that

0 (r,t) = vn (t)'' (r)(18)

c 1

)		]

h (R° + c)

6
where

v = neutron velocity

n t = neutron density (amplitude factor), proportional to power

41 (r = shape factor

The shape factor is assumed to remain constant during a pulse . This is

called the point-reactor model .

•

The production of delayed neutrons and the effects of source neutrons

are neglected .

• The pulse from a thermodynamic standpoint is adiabatic, so

dT

Kn(t)(19)

dt

where

T = fuel temperature

K = reciprocal of heat capacity

From the first and second assumptions we can write the time-dependent

neutron density as

dn

p - fl

dt Q .

n (20)

where

Q = mean lifetime of neutrons in the reactor

beta = delayed neutron fraction

p = reactivity

To account for a step insertion of reactivity, we can write

p = po - adeltaT	 (21)

where

a = negative of the temperature coefficient of reactivity

p o = step insertion of reactivity

Taking the derivative with respect to time of the above equation and

substituting the result from equation (18),

7

dp

dt

Applying

= -alphaKn

the chain rule to equation (20) yields

(22)

dn

dp

(p - fl)

(23) alphaK1
Integrating equation (23) and solving for the constant of integration gives us

the result

1

n =

F (P o - fl) 2 - (P - fl) 2]

( 24)

The pulse is terminated when n becomes negligibly small . This occurs when

p = 2beta - po	(25)

Equation (25) gives us the condition for the total energy release from the

pulse, which manifests itself as a temperature rise in the fuel element when

it is substituted into equation (21) .

2 (Po - beta)

DeltaTcore,ave -	(26)

a

Calculations by General Atomics show that a complete core of 12 w/o fuel

would have a temperature coefficient of reactivity, a, that is 75% of the

value for an 8 .5 w/o fueled core .' The value for alpha 8.55 w/oistaken to be

-$0.0118 °C . (This value was experimentally verified

with

a series of 23

pulses ranging from $1 .30 to $2 .00 that resulted in an average a of

-$0 .0128°C--within 8 .5% of the published value .) The effect of addingthree

12 w/o FFCR's would, however, have a negigible effect on the overall

temperature coefficient of reactivity for the entire core .'

Applying the Nordheim-Fuchs model for self-limiting power excursions, we

can determine the maximum average increase in temperature for the entire

core using equation (26) . For a maximum allowed reactor pulse with a

$4 .00 step insertion o reactivity, the maximum attained average temperature

rise is calculated to be 333°C . Applying the power-peaking factors from the

previous section, the maximum calculated temperature rise in an FFCR

would be 1.48•DeltaTcore, av e ; the temperature rise for an FFCR for a $4 .00 pulse

is calculated to be 493°C. Assuming an initial temperature of 25°C, the

maximum temperature value would be 518°C . Note that even in the

limiting case, neither the technical specification safety limit of 1000°C nor

the limiting safety systems setting of 600°C is violated .

FFCR OPERATIONAL CHARACTERISTICS

FFCR's are a standard design offered as a stock item by General Atomics

and have been used in several TRIGA reactors for over 20 years . FFCR's

are currently implemented in approximately a dozen TRIGA reactors . There

has been no reported evidence of fuel failure as a result of FFCR use in the

United States . The operational issues to be resolved are the effects of

burnup on the FFCR and the influence of FFCR's on the temperature

coefficients of reactivity, shut-down margin, and rod worth .

FFCR control rod worth curves will be generated the same way that

standard control rod worth curves are . Since the poison section of the

FFCR will be the same as that of the currently installed standard control

8
rods, the worth of the poison section of the FFCR will be that of the

currently installed control rods . Measurements made by AFRRI reactor staff

of control rod worth of the currently installed standard control rods yielded

a nominal rod worth of $1 .90 . The fuel follower is expected to add at least

$0 .70 of reactivity 6 when the control rod is fully withdrawn, so the total rod

worth for an FFCR is estimated to be $2 .60. The transient control rod

and its follower have been measured and have a total nominal worth of

$4 .01 . The shutdown margin, as established by ANS/ANSI 15 .1, is

computed as follows :

Total rod worth$11 .81

kexcess (maximum) $ 5 .00

$ 6 .81

Worth of TRANS rod $ 4 .01

Shutdown margin$ 2 .80

The shutdown margin with the most reactive control rod removed from the

reactor is $2 .80--well in excess of $0 .50 minimum allowed value .

Once operational rod worth curves are established and power monitoring

channels have been calibrated by the thermal power calibration method,

power coefficient of reactivity curves will be generated . The issues regarding

the measurement of shut-down margin and excess reactivity are addressed in

Appendix B, Reactor Core Loading and Unloading .

Structural changes in the FFCR's will be monitored on an annual basis as

part of the annual shutdown and maintenance . Specific effects to be

monitored are the elongation and lateral bending of the fuel . FFCR fuel

elements that have an elongation greater than 0 .100 inch or a lateral bend

greater than 0 .0625 inch will be removed from service .

CONCLUSION

The analysis in this report shows that installing FFCR's in the AFRRI

TRIGA reactor core will not result in an unsafe condition or violation of

technical specifications. The primary parameter of interest, the maximum

fuel temperature, was computed to be 210°C in the limiting case for steadystate operation and 518°C in the limiting case for pulse operation .

Operational issues regarding maximum excess reactivity, shutdown margin,

and burnup have also been addressed, and it has been determined that

sufficient surveillance capabilities exist to prevent any unsafe or illegal

condition .

REFERENCES

1 . General Atomics, letter to M . Moore on fuel follower control rods, 28

October 1988 .

2 . El-Wakil, M . M ., Nuclear Heat Transport, The American Nuclear

Society, Lagrange Park, IL, 1978 .

9
3 . Defense Atomic Support Agency, AFRRI/USAEC Facility License R-84,

Complete with Applications and Amendments, Bethesda, MD, 1962 .

4. Wallace, W . P ., and Simnad, M . T ., Metallurgy of TRIGA Fuel

Elements, GA-1949, General Atomics, San Diego, CA, 1961.

5 . Hetrick, D . L ., Dynamics of Nuclear Reactors, The University of Chicago

Press, 1971 .

6 . DNA Contract DNA001-89-R-0030 to General Atomics for fuel follower

control rod construction .

7 . Jaluria, Y ., Natural Convection Heat and Mass Transfer, Pergamon

Press, 1980 .

1 0
APPENDIX A : DETERMINATION OF FREE CONVECTIVE

HEAT TRANSFER COEFFICIENT

Introduction

We can measure the bulk water temperature within the AFRRI TRIGA core

to determine the average free convective heat transfer coefficient of the

cooling water. This experiment involves inserting a temperature-measuring

probe between the B- and C-ring fuel elements while the reactor is operating

at a steady-state power level of 1 .0 MW and measuring the water

temperature at various axial positions . Once the bulk water temperature has

been determined, Newton's law of cooling can be used to calculate the

average free convective heat transfer coefficient .

Experimental Apparatus and Procedure

The equipment used in this experiment consists of two approximately 18-foot

lengths of chromal-alumel thermocouple wires fused together at one end,

encased in a 16-foot-long, 0 .375-inch-diameter aluminum (Al) tube, and the

thermocouple display readout on the AFRRI computerized reactor control

console (Figure A-1) .

The potential difference generated at the thermocouple junction as the water

is heated by the reactor is amplified and displayed by the thermocouple

circuitry in the AFRRI computerized reactor control console . The

thermocouple is initially inserted into the core to correspond to position I .

The thermocouple resides in each region for several minutes to allow it to

attain thermal equilibrium . Once thermal equilibrium is attained, ten

temperature readings are taken at 10-second intervals . After each

temperature measurement, the thermocouple is withdrawn to the next

position, and the temperature measuring procedure is repeated .

Figure A-2 shows that the temperature is measured in five axial positions :

(I) 3 inches below midpoint (14 inches of thermocouple wire inserted into

the core) ; (II) midpoint in axial dimension ; (III) halfway between midpoint

and bottom of graphite slug ; (IV) at top of fuel region ; (V) 1 .5 inches

above top of fuel region .

Figure A-1 . Experimental apparatus .

1 1
Figure A-2 . Axial measuring points .

Safety Considerations

There are two safety considerations associated with this experiment : radiation

streaming and an unintentional positive change in reactivity if the

thermocouple wires are rapidly withdrawn from the reactor core while it is

at power. Radiation streaming is avoided by flooding the aluminum tube

with water and bending the tube so that it is at an angle not normal to

the top of the core . The thermocouple wire displaces only 0 .043 inch 3 of

water when it is fully inserted in the core, so using the void coefficient of

reactivity, the thermocouple wire represents a negative reactivity insertion of

only 0 .001 cents . If we were to estimate conservatively that the

thermocouple wire had the same neutron-absorbing properties of a control

rod, the maximum negative reactivity would be only 0 .01 cents . Thus, there

is no possibility of a reactivity accident associated with the apparatus used

in this experiment .

Data

Table A-1 summarizes the data gathered during a 1 .0 MW steady-state run

of the AFRRI TRIGA reactor . The variation in the temperature

measurements is most likely due to variance in the radial position of the

temperature probe in the channel .

Table A-1 . Bulk Water Temperature at Each Axial Position

in the AFRRI TRIGA Reactor Core

1 2

Axial

position

Inlet

temp ( ° C)

Measured core

bulk water temp ( ° C)

I 22 72 .9

II 24 65 .0

III 25 48 .6

IV 26 51 .6

V 27 59 .7
Analysis and Conclusion

The purpose of this experiment is to determine the bulk water temperature

within the core shroud ; thus, it is the lowest measured value of the water

temperature that is sought . Figure A-3 illustrates the temperature variation

within a cooling channel .

Table A-1 shows that the measured value that most closely represents the

bulk water temperature within the core shroud is 48 .6°C .

The free convective heat transfer coefficient, h, is found by solving equation

(8) for boundary conditions given by a standard TRIGA fuel element .

Equation (A-1) gives the solution in terms of h.

where

T. = measured fuel temperature at 1 .0 MW

Tf = measured bulk coolant temperature in the core

r° = fuel outer radius, 1 .816 cm

r .= fuel inner radius, 0 .229 cm

c = cladding thickness, 0 .051 cm

kf = thermal conductivity of fuel, 0.18 W/cm-°C

k = thermal conductivity of clad, 0 .138 W/cm-°C

q`" = volumetric heat rate .

The measured fuel temperature in the B-ring at 1 .0 MW steady-state power

level is 390°C, and the calculated volumetric heat rate is 65 .9 W/cm3 .

Using the measured value of the bulk coolant temperature of 48 .6°C yields a

value of 1 .339 W/C°-cm2 for the free convective heat transfer coefficient .

Figure A-3 . Temperature variation within a cooling channel .

1 3

2

q"

'ri r0)

2

((T . - Tf) - 21n (r0) - 1-

1 4kf r ir i 1 r o + c0

h=( )

2

- ln( )

r + c 0

0 q '''r i 2

kc

(ro r0

-)

1

2 r •i
Newton's law of cooling expresses the linear relationship between the heat

transfer rate, Q, and the temperature difference between the clad surface

temperature, T,, and the bulk water temperature, T f, as

Q = hA (T f - TC)

where h is the overall convective heat transfer coefficient and A is the area

of the fuel element . ? The value of h determined for the AFRRI TRIGA is

unique in that it takes into account the flow configuration, fluid properties,

and the dimensions of the fuel elements . Assuming that the dependence of

Q on the temperature difference T f - Tc is roughly linear, then the value of

h, computed using data from B-ring elements with their higher heat transfer

rate and local temperature difference, will be close to the value for h for Dring positions .

1 4
APPENDIX B : REACTOR CORE LOADING AND UNLOADING

General

Loading and unloading of the reactor core shall be performed under the

supervision of the Reactor Facility Director or the Reactor Operations

Supervisor .

Specific

1 . Setup

a. Ensure that at least one nuclear instrumentation channel is

operational .

b . Ensure that the source is in the core .

c . Ensure that an operator monitors the reactor console during all fuel

movements .

d . Check new FFCR's before insertion into the core ; this includes

cleaning, visual inspection, and length and bow measurements .

e.Install all control rods .

f . If irradiated fuel elements are to be removed unshielded from the

pool, obtain a Special Work Permit (SWP) from the Safety and Health

Department (SHD) ; do not remove fuel elements with a power history

(greater than 1 KW) in the previous 2 weeks from the reactor pool .

2 . Core Loading

a. After each step of fuel movement perform the following :

(1) Record detector readings .

(2) Withdraw control rods 50% ; record readings .

(3) Withdraw control rods 100% ; record readings.

(4) Calculate 1/M .

(5) Plot 1/M versus number of elements (and total mass of

(6) Predict critical loading .

(7) Insert ALL rods ; continue to next step .

b . Load elements in the following order :

(1) Load the B-ring and C-ring thermocouple elements .

1 5

235U)
(2) Connect thermocouple outputs to reactor control console display .

(3) Install any other thermocouple elements .

(4) Complete loading of B- and C-ring elements (total of 18

standard elements plus 3 FFCR's) .

(5) Load D-ring (total of 33 standard elements plus 3 FFCR's) .

(6) Load the following E-ring elements in order :

16, 17, 18, 20, 6, 8, 9, 10 (total of 41 standard elements plus 3

FFCR's) .

(7) Complete the E-ring by loading the following elements in order :

15, 21, 11, 5, 14, 22, 4, 12, 13, 1 (total of 51 standard

elements plus 3 FFCR's) .

(8) Load the following F-ring elements in two elements per step

until criticality is achieved, using the following loading order :

22, 23, 24, 21, 20, 25, 26, 27, 28, 29, 30, 1, 2, 3, 4, 5, 19, 18,

17, 16, 15, 14, 13, 6, 12, 7, 11, 8, 10, 9 .

Once criticality has been achieved, perform control rod worth measurements

at core position 500 by rod drop technique . Calculate shutdown margin

(SDM) :

SDM = total control rod worth - K- TRANS rod worth excess

(9) Load core to $2 .00 excess reactivity by loading two elements

per step using the loading order in instruction 8 .

(10) Verify control rod worth using rod drop techniques ; calculate

SDM .

(11) Load the core to achieve a Kthat will allow calibration of excess

the TRANS rod based on the last available worth curve of the TRANS rod

(approximately $4 .00) . Calculate the reactivity value of each element as it

is added.

$5.00) .

(12) Calibrate all control rods .

(13) Calculate SDM .

(14) Estimate Kexcess

with a fully loaded core (must not exceed

(15) Load core to fully operational load using loading order in

instruction 8, and recalibrate all control rods . Calculate SDM .

1 6
(16) Adjust the core loading pattern to meet operations requirements

if necessary . Recalibrate all control rods . Calculate SDM .

3 . Core Unloading

a. Unload the reactor core starting with the F-ring and ending with the

B-ring.

b. Remove the fuel elements individually from the reactor core, identify

them by serial number, and place them in the fuel storage racks or a

shipping cask .

c . If elements are to be loaded into a shipping cask, clean the cask

completely, and check for radiological contamination before placing the cask

in or near the pool . Load cask in accordance with procedures specific to

the cask.

d. Once the cask is loaded, perform an air sample and survey ; check

temperature and pressure inside cask, if necessary .

e. If elements are placed in temporary storage away from core

monitoring, ensure that criticality monitoring in accordance with 10 CFR 70

is in place.

1 7

DEPARTMENT OF DEFENSE

ARMED FORCES INSTITUTE OF PATHOLOGY

ATTN : RADIOLOGIC PATHOLOGY

DEPARTMENT

ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE

ATTN : PUBLICATIONS DIVISION

ARMY/AIR FORCE JOINT MEDICAL LIBRARY

ATTN : DASG-AAFJML

ASSISTANT TO SECRETARY OF DEFENSE

ATTN : AE

ATTN: HA(IA)

DEFENSE NUCLEAR AGENCY

ATTN: TITL

ATTN: DDIR

DEFENSE TECHNICAL INFORMATION CENTER

ATTN: DTIC-DDAC

ATTN: DTIC-FDAC

FIELD COMMAND DEFENSE NUCLEAR AGENCY

ATTN: FCFS

INTERSERVICE NUCLEAR WEAPONS SCHOOL

ATTN: RH

LAWRENCE LIVERMORE NATIONAL LABORATORY

ATTN: LIBRARY

UNDER SECRETARY OF DEFENSE (ACQUISITION)

ATTN : OUSD(A)/R&AT

DEPARTMENT OF THE ARMY

HARRY DIAMOND LABORATORIES

ATTN: SLCHD-NW

ATTN: SLCSM-SE

LETTERMAN ARMY INSTITUTE OF RESEARCH

ATTN: SGRD-UL-B1-R

SURGEON GENERAL OF THE ARMY

ATTN: MEDDH-N

U .S . ARMY AEROMEDICAL RESEARCH LABORATORY

ATTN: SCIENTIFIC INFORMATION CENTER

U .S . ARMY ACADEMY OF HEALTH SCIENCES

ATTN: HSHA-CDF

U .S . ARMY CHEMICAL RESEARCH, DEVELOPMENT, AND

ENGINEERING CENTER

ATTN: DIRECTOR OF RESEARCH

U .S . ARMY INSTITUTE OF SURGICAL RESEARCH

ATTN: DIRECTOR OF RESEARCH

U .S . ARMY MEDICAL RESEARCH INSTITUTE OF CHEMICAL

DEFENSE

ATTN: SGRD-UV-R

U .S . ARMY NUCLEAR AND CHEMICAL AGENCY

ATTN: MONA-NU

DISTRIBUTION LIST

U .S . ARMY RESEARCH INSTITUTE OF ENVIRONMENTAL

MEDICINE

ATTN: DIRECTOR OF RESEARCH

U .S . ARMY RESEARCH OFFICE

ATTN: BIOLOGICAL SCIENCES PROGRAM

WALTER REED ARMY INSTITUTE OF RESEARCH

ATTN: DIVISION OF EXPERIMENTAL

THERAPEUTICS

DEPARTMENT OF THE NAVY

NAVAL AEROSPACE MEDICAL RESEARCH LABORATORY

ATTN: COMMANDING OFFICER

NAVAL MEDICAL COMMAND

ATTN : MEDCOM-21

NAVAL MEDICAL RESEARCH AND DEVELOPMENT COMMAND

ATTN : CODE 40C

OFFICE OF NAVAL RESEARCH

ATTN : BIOLOGICAL SCIENCES DIVISION

DEPARTMENT OF THE AIR FORCE

BOLLING AIR FORCE BASE

ATTN: AFOSR

BROOKS AIR FORCE BASE

ATTN: USAFOEHL/RZ

ATTN: USAFSAM/RZ

ATTN: USAFSAM/RZB

NUCLEAR CRITERIA GROUP . SECRETARIAT

ATTN: WL/NTN

SURGEON GENERAL OF THE AIR FORCE

ATTN: HQ USAF/SGPT

ATTN: HQ USAF/SGES

U .S . AIR FORCE ACADEMY

ATTN: HQ USAFA/DFBL

OTHER FEDERAL GOVERNMENT

BROOKHAVEN NATIONAL LABORATORY

ATTN: RESEARCH LIBRARY, REPORTS

SECTION

CENTER FOR DEVICES AND RADIOLOGICAL HEALTH

ATTN: HFZ-110

DEPARTMENT OF ENERGY

ATTN: ER-72 GTN

GOVERNMENT PRINTING OFFICE

ATTN: DEPOSITORY RECEIVING SECTION

ATTN: CONSIGNED BRANCH

LIBRARY OF CONGRESS

ATTN: UNIT X

LOS ALAMOS NATIONAL LABORATORY

ATTN : REPORT LIBRARY/P364
NATIONAL AERONAUTICS AND SPACE ADMINISTRATION

ATTN: RADLAB

NATIONAL AERONAUTICS AND SPACE ADMINISTRATION,

GODDARD SPACE FLIGHT CENTER

ATTN : LIBRARY

NATIONAL CANCER INSTITUTE

ATTN: RADIATION RESEARCH PROGRAM

NATIONAL LIBRARY OF MEDICINE

ATTN: OPI

U .S . ATOMIC ENERGY COMMISSION

ATTN: BETHESDA TECHNICAL LIBRARY

U .S . FOOD AND DRUG ADMINISTRATION

ATTN: WINCHESTER ENGINEERING AND

ANALYTICAL CENTER

U .S . NUCLEAR REGULATORY COMMISSION

ATTN: LIBRARY

RESEARCH AND OTHER ORGANIZATIONS

BRITISH LIBRARY (SERIAL ACQUISITIONS)

ATTN: DOCUMENT SUPPLY CENTRE

CENTRE DE RECHERCHES DU SERVICE DE SANTE DES

ARMEES

ATTN: DIRECTOR

INHALATION TOXICOLOGY RESEARCH INSTITUTE

ATTN: LIBRARY

INSTITUT FUR RADIOBIOLOGIE

ACADEMIE DES SANITATS UND GESUNHEITSWESESNS DER

BW (WEST GERMANY)

ATTN: DIRECTOR

KAMAN TEMPO

ATTN: DASIAC

NBC DEFENSE RESEARCH AND DEVELOPMENT CENTER OF

THE FEDERAL ARMED FORCES (WEST GERMANY)

ATTN: WWDBW ABC-SCHUTZ

NCTR-ASSOCIATED UNIVERSITIES

ATTN : EXECUTIVE DIRECTOR

RUTGERS UNIVERSITY

ATTN : LIBRARY OF SCIENCE AND MEDICINE

UNIVERSITY OF CALIFORNIA

ATTN : LABORATORY FOR ENERGY-RELATED

HEALTH RESEARCH

ATTN : LAWRENCE BERKELEY LABORATORY

UNIVERSITY OF CINCINNATI

ATTN: UNIVERSITY HOSPITAL, RADIOISOTOPE

LABORATORY