ML20065G704
| ML20065G704 | |
| Person / Time | |
|---|---|
| Site: | University of Missouri-Rolla |
| Issue date: | 12/31/1989 |
| From: | Covington L MISSOURI, UNIV. OF, ROLLA, MO |
| To: | |
| References | |
| NUDOCS 9010230022 | |
| Download: ML20065G704 (79) | |
Text
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hD4 Ab NEUTRONICS STUDY OF THE CONVERSION OF THE UNIVERSITY OF MISSOURI ROLIA REACTOR TO LOW ENRICHED URANIUM RJEL BY LORNE J . COVINGTON, 1963-A THESIS I
/
Presented to the Faculty of the Graduate School.of the i
UNIVERSITY OF MISSOURI ROI.lA I
in Partial Fulfillment of the Requirements'for.the Degrec 1; l
MASTER OF SCIENCE IN . NUCLEAR ENGINEERINC- I 1989 Approved by-( LL 't Dr. Milan Straka. Advisor ~ lDr.;D;Rayhlwards.
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Dr. Albert E. Bolon Dr. Dwight C. l.ook, Ir,
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This thesis .hasiheen f prepared [itif the" style; utilit.ed- byj the1J6urnal o'f 'Nuclea r Technolor.v e Pages):1. 31]will be presentedl for- publicationl N>
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111 ACKNOWLEDGEMENTS My thanks to my advisor Dr. Milan Straka for his guidance, support and encouragement. Special thanks are extended to my committee members !
( Dr. Milan Straka, Dr. D. Ray Edwards, Dr. Albert E. Bolon, and Dr. Dwight !
) C. Look, Jr. for their help in conducting my research.
The author would like to thank Dr. James E. Matos, Dr. Armando Travelli, Dr. William L. Woodruff, and Mr. Roger Remport of the Reduced i
Enrichment for Research and Test Reactors Program of the Argonne National Laboratory for their help and guidance. I also want to thank the staff of the University of Missouri-Rolla Reactor Facility for their help.
The author is grateful for the financial support received from = the Department of Energy (Grant Number DE-FCO2 86ER75272). I am also grateful for the fellowship from the Power Group and also the support from the U.S. Department of Interior.
Many thanks to my wife, Julie, who has given me sup;vre and
' encouragement throughout this research. ,
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Table of Contents Page !
PU B LI CAT I ON TH ES I S 0 PT I ON . . . . . . . . . . . . . . . . . . . . . . . . .11. . . . . . . . . . .!
A C KN0V LE DG EM ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 111- ............
LI S T O F I LLU S TRAT I ONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .vi- ...........
1 LI S T OF TA B LES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
NEUTRONICS STUDY OF THE CONVERSION -
OF THE UNIVERSITY OF MISSOURI ROLIA REACTOR TO lhV ENRICHED URANIUM FUEL ABSTRACT....................................................... 2; I.
INTRODUCTION.............................................. 3 l II.
M ETHOD01DCY AND COD ES . '. . . . . . . . . . . . . . . . . . . . . . . . . . . . S. . . . . . . . :l J
I I I . R EACTOR CA LCU1ATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . 8. ..........
A. Effective Multiplication Factor And Thermal Flux Distribution............................., 8 B. Reactivity Coefficients................................ 11 C.
Irradiation Facilities And Control. Rods................ 14 .;
IV.
CONCLUSIONS............................................... 18 ACKNOWLEDGEMENTS............................................... 19 l l
NOMENCLATURE................................................... 20' i
i REFERENCES..................................................... 21 ADD I TIONA L R E FER ENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 ............ I VITA................................................................ 33
).
APPENDICES.......................................................... 34 A. CODE DESCRIPTIONS AND INPUT' EXAMPLES........................ .34 Al. LE0 PARD........................ ...................... 34 i
A2. 2DB UM................................................ 36 B. CROSS SECTION CENERATION AND CELL DISCRETIZATION............ 41 C . S E LECT I ON O F A LEU COR E CONF I CURAT I ON . . . . . . . .49 ..............
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P Page D. cal.CULATION OF FOWER PEAKING FACTORS AND POWER DISTRIBUTIONS.,..... ......
.......................... 52 E. CALCULATIONS OF REACTIVITY COEFFICl ENTS. . 59. . . . . . . . . "
F. METHODS TO CALCULATE REACTIVITY VORTHS OF 66 CONTRO C . ADD I T I ONAL R EACTI VI TY CA LCULATI ONS . . . . . . . . . . . . . .yo.
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1 vi LIST OF ILLUSTRATIONS I Page 1.
(a) Current HEU and (b) proposed LEU core configurations....... 26
- 2. Computational models of (a) HEU standard element. (b) LEU control element (not to scale)................................. 27 3.
Calculated and measured lateral thermal neutron flux profiles through row D of the HEU core.......................... 28 4
Total moderator coefficient for the HEU and LEU cores. . . . . .... 29
- 5. Reactivity of the UMRR void tube at the HEU core periphery (C-3)........................................... 30
- 6. Lateral thermal flux profile along centerline of irradiation fuel element in grid' position E 5............... 31 Bl. The HEU standard element from UMRP. blueprints (shown with top handle)........................................ 44 B2. The HEU control element from ttRR blueprints. . . . . . . . . . . . . . . . . . . 45 B3, The computational model for the HEU control element (not to scale)................................................. 46
( B4. The computational model for the LEU standard element (not to scale)....................................... ......... '47 B5. The computational model for the LEU irradiation fuel element (not to scale)................................................. 48
- C1.
The LEU cores investigated using the 16 fuel plate element. . . . . 50 C2.
1 The LEU cores investigated using the.18 fuel plate element. . . . . 51 DI. HEU core power peaking 1 factors................................. 54 D2. Proposed LEU core power peaking factors........................ 55 D3.
H EU c o re powe r d is t ri bu t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 D4 Proposed LEU core power distribution...........................
-57 D5 Measured relative thermal flux of the HEU UMRR................. 58 El. Moderator density coefficient versus temperature for the HEU and proposed I.EU cores..................................... 60 E2. Moderator temperature coefficient versus temperature for the HEU- and proposed LEU cores . . .
............................... . 61 E3.
Moderator coefficients versus temperature for the HEU core. . . . . 62
I vi LIST OF ILLUSTRATIONS Page 1
1.
(a) Current HEU and (b) proposed LEU core configurations....... 26-
- 2. Computational models of (a) HEU standard element. (b) LEU control element (not to scale)................................. 27 3.
Calculated and measured lateral thermal neutron flux profiles through row D of the HEU core......................... 28 4
Total moderator coefficient for the HEU and LEU cores. . . . . . . . . . 29
- 5. Reactivity of the UKRR void tube at the HEU core periphery (C 3)........................................... 30
- 6. Lateral thermal flux profile along centerline of irradiation fuel element in grid position E.5............... 31 Bl. The HEU standard element from UMRR blueprints (shown with top handle)........................................ 44 k
B2. The HEU control element from UNRR blueprints........... ....... 45 B3. The computational model for the HEU control element (not to scale)................................................. 46 B4 The computational model for the LEU standard element (not to scale)................................................, 47 B5, The computational model for the LEU irradiation fuel. element (not to scale)................................................. 48
, C1. The LEU cores investigated using the 16. fuel plate element. . . . .
1 50 C2. The LEU cores investigated using the 18' fuel plate element..... 51 Dl. H EU co re powe r peaking f ac to rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 54 D2, Proposed LEU core power peaking factors. . . . . . . . . . . . . . . . . . . . . . . . 55 l D3. HEU core power distribution.................................... 56 [
D4 Proposed LEU core power distribution........................... 57 D5. Measured relative thermal'fitx of the HEU UMRR................. 58 l
El. Moderator density coefficient versus temperature for the H EU a nd p r o p o s e d I.EU c o r es . . . . . . . . . . . . . . . . . . . . . ........... ... 60
- E2. Moderator temperature coefficient versus temperature for the j H EU a nd p ropos ed LEU co res . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 l E3. Moderator coefficients versus temperature for the ifEU core. . .. 62 l
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. vii Page E4 Moderator coefficients versus temperature for the LEU core..... 63 E5. Doppler coefficient versus temperature for the p r o p o s e d LE U c o r e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i . . . . . . . . 64 E6. Reactivity worth of a mid core void versus t' of wa ter for the HEU core . . . . . . . . . . . . . . . . . . pe rcent voiding
.................... 65 i
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l LIST OP TABLES vlii l Page I.
Measured and calculated excess reactivities (tak/k) in the U and T modes for the HEU and proposed LEU cores........ 22 II.
Reactivity coefficients calculated for the HEU a nd p ropos e d LEU _ c o re s ( t ak/k/* C) . . . . . . . . . . . . . . . . . . . . . . . . 23 III.
Measured and calculated void coefficient (tak/k/cm 3 )........ 24 IV. Reactivity worths of the Core Access element for the proposed LEU core (tak/k)........................... 25 a B-I.
Lattice spacing for the HEU and proposed LEU cores (cm) . . . . . 41 B II.
Humber densities and volume fractions for the homogenized z one s o f the H EU and LEU core s . . . . . . . . . . . . . . . . . . . . . . .42. . . . . . .
D I.
Powe r peaking fac tor de fini tions . . . . . . . . . . . . . . . . . . . . . . . . . . .
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Calculated and measured control rod worths for t UKRR HEU core (taak/k)..........................he ............ 68-F II.
Calculated and measured worth of the regulating rod in the HEU core (tak/k)..................................... 69 C-I.
Various irradiation fuel element-designs, power peaking factors and thermal fluxes in grid position E.5............. 70 C II. Raw data for CA experiment.................................. 71 C III.
- Reactivity of CA element for normal' (air filled) and flooded conditions. Cases A.and B are experimental data and 263-U are from computer modeling (tak/k)........................M . 71 i
1 1
e 9
NEUTRONICS STUDY OF Tile CONVERSION
- OF THE UNIVERSITY OF MISSOURI-ROLlA REACTOR TO LOW ENRICHED URANIUM FilEL -
by Lorne J. Covington University of Missouri Rolla Nuclear Engineering Department i Rolla, Missouri 65401 (314) 341 4236 1
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2 ABSTRACT The neutronics calculations for the conversion of.the University 1
1 of Missouri Rolla Reactor (UMRR) from hi 6hly enriched uranium ~ fuel l (IIEU) to low enriched uranium fuel (LEU) are studied. Several computational models of both the present HEU and proposed LEU cores are developed for two dimensional neutron diffusion calculations using the 2DB.UM code. The core multiplication factors, neutron flux pt0 files,.
power peaking factors, moderator and void coefficients are calculated for both cores.
The current llEU irradiation facilitics are modele.1 and an irradiation fuel element for the LEU core is developed. Available experimental data for the HEU core are compared to computer results for ,
a verification of the computational'models. i 1
Results show that the reactor conversion will<not have any major adverse effect on the operation of the UMRR. The criticality should be reached wit.h approximately the same number ' of LEU fuel and ~ control elements as in the present HEU core.
Power peaking' factors for; the LEU core are of the same magnitude as for the HEU core and are well' below i the limits established in the UMRR Safety Analysis Report. All reactivity coefficients remain negative.
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- 1. INTRODUCTION l
In the near future the high enriched uranium (llEU) fuel in the University of Misr.ouri Rolla Reactor (UMRR) will he replaced with low enriched uranium (LEU) fuel. The U. S. Nuclear Regulatory Commission I
requires a submittal of all necessary changes in the licensing _ -1 documents before an order to convert can be issued. This paper presents the results of the neutronics study of such a fuel replacement. A concurrent independent study was performed by the Reduced Enrichment for Research and Test Reactor (RERTR) Program-at the Argonne National 1
Laboratory (ANL). The RERTR Program used a di f fe rent set of computer :
codes to corroborate the analyses performed in this study.
The present HEU core consists of 14 fuel elements, 4 control elements and I half element. A standard fuel element consists of_10 i
curved fuel plates connected by two aluminum side plates. A ~ control.
element consists of 6 curved fuel. plates connected by two aluminum side
-plates with i.n aluminum guide tube for the control rod located in the center of the element. The half' element is similar to the standard fuel 1
element, except that five fueled plates in one half of the element are replaced with solid aluminum plates. The HEU - fuel plate- consists of aluminum cladding surrounding 3g U O -Al fuci'enric M to % b .
The LEU fuel material will be U3Si2 in an aluminum matrix with a 235 U enrichment of 19,75%. The standard fuel element contains 18 ci';ved fueled plates connected by two aluminum sideplates. The control element-is similar to the LEU standard element,: except 'that 10 of the center plates are removed and an-aluminum guide tube is inserted. The LEU half element is similar to the LEU standard element with 9 fueled plates .
1 replaced with aluminum plates in either half of the element. l
1 e f, The UMRR is a pool type reactor presently licensed for-200 kW and
.t is cooled by natural convection. The UMRR core is suspended from a movable bridge which allows operation in. two different modes: a completely water reflected mode and a thermal column reflected mode in which one side of the reactor faces a graphite thermal column. The UMRR ,
has one bare and one cadmium lined pneumatic tube facility and one beam port facility. The reactor control is accomplished by three oval.
borated steel shim / safety rods and one oval stainless steel regulating' ,
rod. The current HEU core and the proposed LEU core configuration. as developed within the course of this study.'are shown in Figure 1.
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6 D 'V 4 g -E r,g4g; +S g -O g-1 to N '(1) g where the source term is defined as y N g-1 S - vE),d,+
g (E,) ,,, d, (2)
The third dimension, which is in che axial direction, is handled by the input of an axial buckling term uhich'is defined as:
(w/( active fuel height + 2* extrapolation length)) (3)
The axial extrapolation length was provided by ANL- using a 3D diffusion model of the UMRR (1). Two energy group LEOPARD cross sections were input for the 2DB-UM~ calculations (hence N-2 in Eqs.(1) :
and (2)).
The standard element is divided into five homogenized zones; two i side plate zones, two end plate zones and one zone for the interior . .
fuel plates. The control element is divided into seven homogenized zones: two side plate, two fueled end- plate, two interior fuel plate zones, and one zone containing the - control rod guide tube and water channel. The pneumatic tubes were modeled by homogenizing all materials l over the cell. The number densities were calculated for each material in each zone and then -smeared over the volume of the ' cell. The fuel- ;
t i plates were assumed to be straight to simplify mesh generation in the diffusion code. Figure 2 shows the computational model of the HEU <
standard fuel element (a) and a proposed'-LEU control fuel element (b ) '.
The dashed lines show the division of the elements into the different 1
homogenized zones.
A detailed computational mesh in the X Y plane was then developed. l The number of mesh lines per element was '10 in the X direction and ' 12 l
in the Y direction with an approximate spacing'of I cm. The X Y mesh 1
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7-included very fine mesh lines at - the edges of all fueled regions to allow for a determination of thermal flux peaking in the element.
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8 111, REACTOR CALCULATIONS '
The first phase of the neutronic calculations was performed for '
the prer.ent UMRR HEU core. The - reactor geometry and materials were
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carefully modeled so that a comparison could be made vith measured data.
Ill.A Effective Multiplication Factor and Thermal Flux Distribution Using the procedure described above the effective multiplication '
factor, k,7f, was- calculated for- the HEU core in both the water reflected and the graphite reflected ' mode. The values are in good-agreement with measured data and are shown in Table I.
The LEU core was then modeled following the same guidelines.used in modeling of the HEU core. A' problem in generating cross sections for g the LEU fuel material U 3Si2 was .that silicon cross sections were unavailable in LEOPARD. A discussion with the RERTR Program staff determined that aluminum cross sections could be used in place of those
-of silicon [5) . Two different LEU element types and several different core configurations were developed. Both element types contained 18
-plates with the difference .being that one element type had all the plates fueled and the other had the two outer plates replaced with aluminum plates. The cores developed using the 18 fueled plate element I
contained approximately the same number.of elements as the current HEU core and one core configuration was found which resembled the present core. It was, therefore, decided pursue to only this core configuration. Results of k df cal ulati ns for this core are shown in Table I.
Besides the computational mesh described in Section II, a second l
9 mesh, called the fine mesh, was ~ created for_ modeling the LEU l irradiation fuel element. The difference between the two meshes is that row E had 20 mesh lines in the Y direction in order to model individual-fuel plates.
The 2DB-VM k,gg values for the LEU core are higher than the ANL results. The difference is attributed to the use of aluminua cross sections instead of those for silicon and to an uncertainty associated with the axial buckling term. This term was. found to--bei a very -
sensitive parameter and could have been adjusted to produce ai vide range of values for k,gg. The . difference between the fine and' coarse-mesh values of k,fg is attributed to the fact that the_ row containing the fine mesh enables a more detailed description of thermal neutron-flux in the core and reflector. . The spacing of the fine mesh (0.10;4 cm) is on the order of the thermal ~ diffusion coefficient for the 1
reflector and fuel regions 0.17 and 0.24 cm , respectively., This is l especially important when the neutron leakage - (and return) term is'
- calculated. The fast neutrons, which have leaked out._ of' the core, are thermalized in the reflector and subsequently these neutrons. then a '
diffuse back into the core. It:is this portionL of - the flux- which is -
described better with the fine mesh model. A detail ~ comparison of the two different meshes show = a 64 increase in the number of thermal neutrons returning into the core. There is a similar effect seen in = the other rows, but to a lesser extent. Overall,- this effect manifests itself with.an increase in the value of k,gg for the' fine. mesh.
The power . peaking factor and power distribution were determined for each element position in the HEU and LEU core. The - power peaking .
factor is defined to be the peak 'pever in the element divided by.the !
_____r---_ w-- v- . - - - _ _ _ _ _ -- w -- e-e = w IT *F ' * * '"
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average power in the core. Average power densities were obtained directly from 2DB.UM by selecting edits that perform averaging over '
requested tones. The peak thermal- flux in the element- was found by
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j scanning 2DB UM thermal flux outputs in the particular e1ement for the maximum value and then determining the X Y coordinates where that value occurred. The peak power density was then calculated by summing the l
product of flux and fission cross section for all energy. groups at the determined X Y coordinates.
The maximum power peaking factor for the HEU core as - determined from the 2DB.UM and ANL calculations was 2.00 and 2.19, respectively.
For the LEU core, the power peaking factors were. 2.22 and 2.34, respectively. The element in which the maximum power peaking factor '
occurred was, in all cases, the control element.C3.~1he increased value
(
of the power peaking factor = in the 1.EU case is still well within the power peaking limits L of - 3.0 to 4.0 prescribed in the UMRR Safety Analysis Report (6).
A comparison was made of the measured and calculated the'r mal flux distribution at the midplane of the HEU core. Figure 3 shows the lateral flux profile through row D compared with measured values of the thermal flux. The accuracy of _ the measured data within the core is about i 304. Outside the core the flux values are known with much less accuracy. The comparison. shows -that the calculated thermal flux distribution is slightly flatter across the core than the measured values, except the peak occurring in the control element. This i
flattening of = the thermal flux was investigated 'and was determined to I
come from the use of the two energy group cross sections. A sensitivity' l
study was performed which consisted of varying the magnitude of fast i
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11 diffusion coefficient, D in Eq. (1). Varying D g changes' the fast i.
3 leakage of the core, which causes a change. in t he fast and thermal flux shapes. The study showed that a slight decrease in the fast transport cross section and consequently an increase in Dg caused the calculated i thermal flux to better fit the measured data.
III.B Reactivity Coefficients The moderator temperature, moderator density and total moderator reactivity coefficients were calculated.for both the LEU and HEU cores.
The Doppler coefficient was calculated only for: the LEU core.
The procedure to calculate the reactivity coefficients consisted -
of generating sets of macroscopic cross' sections. for the fueled regions of the reactor with only one parameter affecting the reactivity changed at a time. These cross sections sets were ' then used as input for the global 2DB-VM problem to obtain values for k,77 from which the-pertinent reactivity coefficient was determined. Cross sections for the
. Doppler coefficient were generated by- changing only- the resonance fuel a temperature input data in LEOPARD in the range from 20 to 600 *C, The i
resonance fuel temperature -is used by LEOPARD in 'the calculation of U
resonance absorption in the epithermal energy. range. Cross sections for the moderator density coefficient ' vere generated in a !
similar manner, but only the moderator volume fraction was changed for each LEOPARD run. The volume fractions were calculated for water densities at temperatures from 20 to 100 *C, When calculating the moderator temperature coefficient cross sections, the only value that should be changed is the moderator temperature. In LEOPARD, hovever, a change of the moderator temperature causes the code to adj o:s t - the
12 moderator number density corresponding to the thermal expansion of the moderator. To negate this adjustment, the volume fraction (vf) of the water must be adjusted such that the number density used in calculating the macroscopic cross sections is the same-as that for the base case.
These adjusted volume fractions were determined by running LEOPARD at each selected moderator temperature using the base volume fraction and ,
then calculating the adjusted volume fraction by taking the ratios of the hydrogen number density in the moderator gN as follows:
(NH }T - 20 'C (vf) adjusted ' M }-
H}T selected The temperatures for the moderator temperature coefficient ranged from 20 to 100 *C, For the base case, the temperature of all materials was i chosen to be 20 *C and the moderator volume fraction to be 1.0.
i The calculated reactivity coefficients are . tabulated in Table 11. f The results show that the total moderator. coefficient, while still i negative, is of smaller magnitude for the proposed LEU core than the
- HEU core. This was to be expected because the greater number of plates -
in the proposed LEU core reduces the amount of moderatorLin the core.
The Doppler coefficient. for the ' LEU core is also negative, but is an i
order of magnitude smaller than the total moderator coefficient for'the operating range of the reactor.
Figure 4 shows the calculated total. moderator coefficient for the HEU and LEU cores along with the measured moderator coefficient for the-HEU core. The measured total moderator coefficcint is shown only for the normal operating range of-the reactor. It shows a good agreement between the measured and the calculated HEU value for low temperature. l
!' However, one should l
note, that the calculated value slightly
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'1 overpredicts the measured value.
The void coefficient was calculated for a midcore position (E-5) and a peripheral position (C 7 in the HEU and C 3 in the LEU core). Two
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different approaches were used to model a void. The first approach was to slowly decrease the water numoer . densi ty in the void region and calculate k,fg in corresponcing 2DB UM runs. For both the> midcore and peripheral calculations. this lead to increasing instabilities as the percent voiding of the region approached 100%. A strong non linear relationship between the reactivity and voided fraction was also observed, The second approach was developed in that air was - replaced by aluminum whose cross sections were modified (2). A base set of macroscopic aluminum cross sections was produced from LE0 PARD. In each
( energy group the absorption cross section was reduced by a factor of 10' .
The difference between the new absorption cross section and the
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base absorption cross section was then subtracted from the transport
- cross sec tio .. No changes were made to the scattering cross sections.
The modified cross sections werd then used as cross sections for air in the 2DB-UM global calculations.
Results of both methods along with a . measured void are shown in 1
Figure 5 for the HEU peripher al calculations. It shows that using-the
' method of adjusted aluminum c. oss. section yields the result which is ,
close to a measured value of the peripheral void coefficient. This i method was also- stable during the calculation. This method- was therefore used to calculate the void coefficient of the LEU core. The results of this calculation are shown in Table III together with the results for the HEU core.
16 I
Ill.C Irradiation Facilities and Control Rods '
With the current 10 plate llEU fuel element, Irradiations can be made using stringers suspended in between individual fuel plates. It is desirable to have a variation of this feature in the LEU core. The proposed 18 plate LEU fuel element - will have limited access to the interior of the element. Therefore, an irradiation fuel (IF) element was designed. The areas of concern in the design of the IF element are.
maximization of the space used for irradiation, physical protection of-the IF and minimization of the thermal load in the IP caused by thermal neutron flux peaking, 1
The geometry of the IF element is the same as the LEU standard '
fuel element except that a number of fuel plates were removed from one half of the IP element. In addition to simply removing fuel plates, two
( aluminum plates were inserted in the outermost positions from which the L
fuel plates were removed. Thesn' aluminum plates are inserteo tu protect the fueled plates on either side of the water channel from any damage
-that could potentially occur during: insertion or removal 'of samples from the IF during operation.
Several different combinations of removed fuel plates and inserted t i
aluminum plates were investigated for the IP element placed in the grid plate position E 5 of the LEU core. The combinations included runs with-the removal of 2,4,6 and 8 fueled ' plates and runs with the removal of '
4,6,8 and 10 fueled plates and the insertion of two aluminum plates.
The design of the _ IF element selected by the reactor staff was element with 8 fueled plates removed and 2 aluminum plates inserted.
This design gives tbc maximum working area while not imposing any-extreme thermal loads within the element. Figure 6 shows'the centerline I
i
. . -i 15 geomet ry of the. proposed IF element and the shape of the thermal neutron flux across the element. There is, by a factor of 1.5, an increased value of the thermal flux in the IF clement as compared to l ~
the current HEU bare pneumatic tube facility, 1
The calculations have shown that the maximum and average thermal neutron flux in the water gap increased with the removal of each. pair l of fueled plates. The insertion of the aluminum plates did not appreciably lower the magnitude of the thermal neutron flux. The maximum power peaking factor for. the least favorable position of the IF element (E 5) is very comparable to the maximum power peaking factor.
determined for the base LEU core design. The respective values are 2.'31 in the IF element and 2.22 in the control rod element C3. Both of these >
power peaking factors are well within . the thermal limits of the = UMRR.
i The value of k gf for the LEU core with this IF element derign is 1
O 9873, therefore additional.elemeens will:have to be added to achieve criticality.
The core access -(CA) element is - a .non fuuled element, made of '
aluminum and graphite. It is clad with an aluminum jacket which has two i
opposite sides curved to the same curvature as :s fuel element, The interior of the element contains a piece' of graphite with a . circular center channel. The computational model of the CA used straight sides 1^
for the curved exterior sides and a square center chanr.el.'The CA~ !
element can be used either in.a core ' periphery . position or in the interior of the reactor core.
The reactivity worth of the CA element in the periphery position C-7 of the HEU core was determined experimentally and by a 2DB UM calculation. Two different cases were investigated. The first case was '
q
l 16 with the center channel containing air, which is'the normal mode of 1 1
operation. The second case was with the center channel flooded with I water. To simulate air in 'the center channel of the CA clement, the l method developed for the calculation of the void coefficient was used.
There is a good agreement between the measured reactivity, 0.425t, and calculated reactivity 0.479%, for the flooded CA element in the !
HEU periphery. The measured reactivity worth of replacing the water q
with air is 0.161t, while the calculated worth'is -0:058t.
The CA element was - modeled : for two reactor positions in the LEU core: the peripheral position C 3 and the midcore position E-5, The i reactivity worth of the CA element in~ the periphery position was calculated to be 0.1424. Flooding of the CA element had a worth of 0.1124. The midcore calculation showed the worth of the air to be 0.5174. Results of the calculations for the CA element in the flooded condition are summerized in Table IV.
The important consideration is 'that negative - reactivity is added -
, to the reactor for all cases in which the . CA . element goes from the-normal air-filled operation to abnormal flooded operation.
i Control rods ~ for the UMRR are solid stainless steel 304 (SS 304) with boron added to increase neutron absorption.
Such strongly absorbing control rods can not be modeled directly using LEOPARD and -
2DB VM. Indirect modeling can be performed by using a reaction rate matching method.
The reaction rate matching method consists of using a Monte Carlo code to determine the ratio of . the absorption rate in the. control rod / guide tube region to the fission rate in a unit cell. This ratio is then matched in a 2DB UM calculation by adj us ting the control rod
17 1
absorption cross sections. The unit cell, for which the calculation is performed, consists of a control element surrounded by 8 standard fuel elements with reflected boundary conditions on all four sides (g).
Monte Carlo calculations were performed by.the RERTR group and the data provided to us (1), When the reaction. rate method was applied to the llEU core, the calculated reactivity worths were = nearly identical for all shim / safety control rods. Their measured worths, however, vary greatly' depending on their core position. This lead to the conclusion that the unit cell model cannot be used for the small core of the UMRR.
Rather the whole core with all shim safety rods present would have to be modeled and the reaction rates for each rod adjusted simultaneously.
This, however, was judged as being beyond the scope of the work and was not attempted for either the llEU or LEU core, !
The regulating rod for the UMRR is a SS .%4 tube , ,It is considered to be a weak absorber and therefore was mcJoled using 1.EOpARD and 2DB-F UM directly. The SS 304 is- homogenized across the ; entire guide tube l i
. reg on and the resultin6 cross sections used in a = global 2DB UM-calculation. The calculated worth of 0.31% agrees well with the measured worth of 0.35% for the HEU core, l
I 9
18 IV. CONCLUSIONS The results of this study show that a conversion to the LEU core l
will not have adverse e f fec ts on the operation of the UMRR. The criticality of the new core will be reached with approximately the same number of fuel elements and in a similar configuration as the HEU core.
Power peaking factors are only slightly larger for the LEU core..but are still well within the requirements ' of the thermal liydraulics analysis. The . moderator temperature coefficient is about 404 smaller i
than for the HEU eore. The moderator temperature coefficient and Doppler coefficient provide the desirable negative reactivity feedback with increasing temperature. The irradiation fuel element in the. LEU core will improve the UMRR irradiation capabilities by providing a thermal flux twice the magnitude previously obtained in'the HEU core.
The codes used in the study,. LEOPARD and 2DB UM. provided adequate results for most cases. They were - easy to set up, run L and likewise their input was simple. Their convergence and run times were short. A disadvantage of LEOPARD is that it ,
l only provided two energy or four-energy group cross section sets. It is believed that.a4 1arger number of energy groups would have provided better results during the
- calculations of the moderator and void coefficients. The main limitation of the two dimensional model was in determining the-value of the axial buckling term. The best assumption that could be made was >
choosing a chopped cosine representation of the- axial flux-distribution. Experimental data show that this definitely was not the case. The large alun.inum grid plate at the bottom of the core and' strongly absorbing control rods present at the top of the' core l contribute to the deformation of the flux shape.
)
19 ,
l Future work could be better performed using a 3 D diffusion code .. .
and a cross section collapsing code that is reore flexible with respect to energy group structure.
ACKNOWLEDGEMENT This study was supported ~ by the U.S. Department of Energy under Crant No. DE-FCO2 86ER75272.
- I.
i l
~\
20 NOMENCLATURE j
D -
diffusion coefficient i g or g' -
subscript denoting energy group
, i k,ff '
effective multiplication factor Ng -
hydrogen number density I S -
neutron source T -
temperature
\
Greek v -
average number of neutrons released per fission -
E - !
r mact SC Pic removal cross section.
I -
f macroscopic fission cross section.
E -
E,~E macroscopic scattering cross section from energy group g' to energy group g -
x -
, E probability that a fission neutron'will be born in energy group g 4 -
neutron flux t
i e
1 4
?!
REFERENCES
- 1. R. F.
Barry.. " LEOPARD . A Spect rum Deper. dent Hon.spatlal Depletion Code for the IBM 7094", Vestinghouse Elec t ric Corporation, VCAP.
3269 26, 1963.
- 2. " Leopard Manual". Argonne National Laboratory, RERTR group, May 1981.
- 3. V. V. Little, Jr. and R. V.
Hardie., "2DB Users Manual Revision 1",
Battelle Memorial Institute, BfNL.831 REVI, 1969.
4 '2DB VM Users Manual". Argonne National 1.aboratory, RERTR group, May 1980.
- 5. J.
Matos, Personal Communication, Argonne National Laboratory.
RERTR group, Summer 1986.
- 6. Safety Analysis Report, University of Missouri Rolla Reactor Facility, License number R 79 Sept. 1984
- 7. J . Matos, Personal Communication. Argonne Nationn) Laboratory, RERTR group, February 1987.
- 8. L. J . Covington.
"Neutrontes Study of the Conversion of the University of Missouri Rolla Reactor to Low Enriched Uranium fuel *,
MS Thesis University of Missouri Rolla, 1989
l j . .
22 TAbli 1 Measured and calculated excess reactivities in the W and T modes for the HEU and proposed LEU cores (tak/k).
HEU Core W mode T mode Measured 0.27 0.35 0.73 0.80 s
Calculated (this work) 0.21 0.57 Calculated (ANL) 0.23 0.48 j
120 Core W. mode T mode Calculated (this work) 0.81 1.19 1.22* 1.52*
Calculated (ANL) 0.17 n/a
- very fine mesh in row E '
'l
23 TABLE 11 Reactivity coefficients calculated for the HEU and proposed LEU cores (46k/k/*C).
Coefficient UMR ANL moderator (20 50 'C) HEU 0.0210 0.0141 total LEU -0.0135 0.0121 (50 100 'C) HEU 0.0262 0.0186 LEU .O.0183 0.0167 Doppler LEU 0.00108 0.00176 i
l l
} )
d
. 24 I
t l
i TABLE III Measured and calculated void coefficient (tok/k/cm3 ),
Core Periphery Midcore HEU measured HEU calculated '
"/*-
1.EU calculated 2.36E 04 9.03E.05 3.26E 04 k
9 e .m 4.
25 ;
1
. t TABLE IV Reactivity worths of Core Access elemene for the proposed LEU core (tak/k).
Periphery Mid core CA with air 0.030 4.501 CA with water 0.142 -S.018
at 1 2 3 4 6 6 7 8 9 5y Ts
(( 5%%.j%gfN/jhyNy $R'.Q"
~
.$bkNb5?dNIkh'hh
, i -
iA s-g C .
r i Ca p"h; D r Cl. r r r r r
E C2 r C3 lr F
,ggg,,, gg (a)
F - standard fuel element BRT - Bare rabbit tube
(
S - source tube CR7 - Cadmium rabbit tube C1 I C4 - control fuel element
- 2 3 1 4 5 6 7 8 9 l
^
'm;ggg. gy "
MrdHErTENIaw3End%g'-L.4 F 8
h.3$$6$$,!dh)h C T .
r r C4 -
x D r Cl r r i r E,Ri L I F UN E r C2 lr C3 l r- r $@fg.1 1
I bmi l F %)'jg BRT r r r SMMi ,
CRT. ): @?j b!NMN A _ _ . k.%%d l
(b) f i ;*, 1.
(a) Curresit Ill:ll neul (la) pits luit.cel 1.l:11 cnte r utifir,ut ot iissi+.
j
i l
\
27 <
i g ... . .
.; .p.
$t.
3 --
i@
-q --
f, .
- +-
g3;
. A .
i P: - e i L. :
n --: ... ,l
- g/m
- 's
i dQ '
- i; .
- m -
j J l !
HEU interior fuel, clad and lJ,p. g 1
moderator zone pi l ;
!Mjl~l 2 Standard element side plate
! I' i .
3 Standard element end plate 4
97,4 j............................................; s
(
I;""""k$Ni
.. 1 i". . . . . . . . . . . . . . .'.'.""'; . . . . . . . . . . . . . . . . . . . . . . . . . . ~. .T (a) i
. ... v . . ... . ... .
. . ~ + . - ~ ~ . . . . .
......i
!j I '
i _
fT- : ., -
j 1 q j..........................
4 1.EU interior fuel, clad and ,
- e '
..........z c moderator zone
[',4 s
R 5 - control element end plate t
j "J y
, 6 Control element side plate
'n .
6
-7
% <X i 4 cuide tube and water region jl+Ql qg t
}..........................................:
'f'j, l
y
- 4 a
_= ..
a.
i
........7_..................................... -.;.
. m: .
.~....
~
l- (b) l Fig. 2.
Computational models of (a) HEU standard element, (b) LEU control element (not to scale).
l n
1 1
1 i
.i 1
- 4. 0E+ 12 -
s W(A$URCO (CRROR: +/- 30s)
) CALCULAff0 n
u a t w
m3.0E+12 s
N m =
u N
1
=
e t
x ?. 0E+ 12 -
2 -
a a A
a '
2 E
w
.z- 1 . 0 E + 1. . .
- coat -
e e
' t i
O.0E+00- . .
i 0 20 40 60 80 100 120 DISTANCE ACROSS CORE (CM) D Fig. 3. .
Calculated and measured lateral thermal flux profiles through row D of the HEU core.
l 4
,..-.e... . . . . _ . . . , _ , . . . , , , _ . , - . . -,_,,,,,..,y.,_,.,__...y.,-,,,-,_y, , _ _ , , , , _ . . , , . ..#, . , ~ . , .,,-.,..._...r...~_,.,,. . , _ . _ , . , , _ , ,
1 as k
i i
1 l
i i
0*000
... __ nre stasunto i
'., Mtu calcutatte
-0.001 - '
s ., . . . . . . . . Ltu cateutaito
! s s ..
- i. -0.002 -
\s s
>- s s.
> -0.003 6- '
m
$-0.004- '.,
. ( a: .
- i W -0.005-
', 6- -
o .
a .
o .
m m -0.006-
-0.007-l l
i
-0.008-
-0.009 .
- 20 30 40
, 50 60 DECREES CELSIUS
l
! i l \
1 i
l 30 J
.1 1
f e
4 i
- 1 l
J' I
0.0 . .... . . . . . . . . .
. . . . l
. uslag adjusted j -
, , , aluminum cross sections M . .
N M .0.05 . .- . .
4
% .0,10
- . . MT.MURf.D g .
./
a m . . .
p .
,4 .
O(A 0.16-l -
-~~~~~--:---~~~
4 .
~ ~ . .
020,- . . -. . . . .
,f .
oQj$ .
e a f
i t i e g g g 10 20 30 40 50 60 70 80 90 100 i
PERCENT VOIDED l 1 1
Fig. 5.
Reactiv!ty of the UNRR void tube at the HEU core periphery (C 3).
u
1 31 i
4-Thermal 3-Flux /
d (x10')2-
' Dis-ca" '
j.
1 i
I I
Fig. 6.
lateral thermal flux profile along centerline of irradiation fuel element in grid position E 5.
32 ADDITIONAL. REFERENCES 9.
RERTR Pror, ram Report , " Properties of Fuel Heat Materials", Argonne National Laboratories RERTR.
- 10. C. L. Copeland and others, " Performance of Low Enriched U Si 3 2 Aluminum Dispersion Fuel Elements in the Oak Ridge Research Reactor" Argonne National Laboratory, July 1987.
- 11. M. M. Bretscher, " Blackness coefficients, Effective Diffusion Parameters and Control Rod Worths for Thermal Reactors", Argonne National Laboratory, ANL/RERTR/TH 5,1984
's
- 12. J. L. Snelgrove, " Method for Computing Axial Extrapolation Lengths" Argonne National Laboratory, 1984
- 13. "Research Reactor Core Conversion from the Use of Highly Enriched Uranium to the Use of Low Enriched Uranium Fuels Guidebook",
t International Atomic Energy Agency, Vienna, Austria, 1980.
I l
l i
l l
f 1
33 ,
VITA ,
1.ntne Joseph Covington was born January 20 1963 in Normandy, Missouri. He received his elementary education in Denver, Colorado and St. Louis, Missouri. He received his high school education at Riverview Cardens High School in St. l.ouis, Missouri. He is married with no children.
He entered the University of Missouri Rolla, Rolla, Missouri in August 1981. He received his B.S. in Nuclear Engineering from the University of Missouri Rolla in May 1986.
He entered the Graduate School of the University of Missouri.
Rolla in August 1986 as a candidate for the degree of Master of Science of Nuclear Engineering.
I 1
l l
l
i 34 APPENDIX A CODE DESCRIPTIONS AND INPUT EXAMPLES A1. LEOPARD LEOPARD is a zero dimensional cross section collapsing code using a MUIT.50FOCATE model of the neutron spectrum. The code was developed by R . F.
Barry of the Westinghouse Electric Corporation in 1963 The code assumes a regular lattice of fuel, clad and moderator. An " extra" region is included to develop cross sections for non fueled regions such as guide tubes, sideplates, graphite reflector, etc. The thermal spectrum is modeled with a Wigner Wilkins spectrum at 172 energy points from 0 to 0.625 eV. The fast spectrum uses a consistent B 1 MUFT IV ,
i spectrum. The cross section library consists of isotopes commonly used in light water reactor analysis.
Input for LEOPARD is broken into 3 groups: input flags, lattice geometry and n;aterial compositions. The lattice geometry is divided Anto a fuel material region, a clad and void region, a moderator region and the " extra" region. For the plate type fuel used in this study, the region thicknesses were measured from the center of the plate to the '
center of the water channel. Material compositions are inputted in two different ways depending on the isotope described. Trace elements such as 2M U, B U, and O B are inputted with number densities in w/ barn /cm.
Other materials such as Hy o, aluminum, graphite, SS 304, are inputted by their volume fraction in the homogenized zones. '
On the following page, the JCL and input data for running a sample LEOPARD problem are given.
ey- - . , , ,
3 's
//L11 JOB
//
(0368\ S1b,een=), 'COVINGTON,IARNE' ,MSGl.EVEl.-(l ,1) .T1ME-1, MSCC1. ASS-H
//51 EXEC PCM-LEOPARD
//STEPLI B DD DSN-USER.XO368.1.EOPARD. I. MOD. DISP-SitR
//CO. FT0l F001 DD DSN-US ER . X2903. LEOPAR D . I.! BMI N2. DI S P-SilR
//* LEOPARD ENDF/B 1V DATA LlBRARY Fi1.E ,
//CO. IT03F001 DD DSN-6k'RESTRT, UNIT SYSDA,SI' ACE-(TRK ,( 3,1)),
//
D I S P-( , D ELETE ) , DCB- ( R ECFM-VBS , lJt ECl.-X , Bl.KS l 7.E-6136 )
//* RESTART FILE k'RITTEN BY LEOPARD OUTPUT
/f t:0.1704F001 DD DSN &RRESTRT, UNIT-SYSDA, SPACE-(TRK,(3,1)),
//
DI S P-( , DELETE ) , DCB-(R ECFM-VBS . LR ECL-X , B LKS 12E-6136 )
//* RESTART FILE READ BY LEOPARD - INPUT FOR RESTART
//00.FT06F001 DD SYSOUT-V
//CO.1707F001 DD DSN-6LINX, UNIT-SYSDA,
// SPACE-(TRK,(5,1)). DISP-(, DELETE),
DCB-( R EC FM-VBS , LR ECL-X , B LKS I Z E-6136 )
//* BlNARY Fil.E OF CROSS SECTION DATA FOR LINX CODE BURNUP DEPENDENT
//CO.lT08F001 DD SYSOUT-V
//* FORMATED FILE OF CROSS. SECTIONS FOR 2DBUM CODE
//CO. IT10F001 DD DSN-6ENER, UNIT-SYSDA. SPACE-(TRK ,(2,1)),
//
DI S P-( , DELETE) , DCB-( R ECFM-VB S . LR ECL-X , B LKS 12 E- 6136 )
//* ENERCY AND TEMPERATURE DATA
//CO. TT16F001 DD DSN-6SPECTRM, UNIT-SYSDA. SPACE-(TRK,(2,1)) ,
//
DI S P-( , DELETE ) , DC B- (R ECih-VBS . LR ECL-X , B1):S 12 E-6136 )
//* SPECTRUM DATA
//CO. FT20F001 DD DSN-&MICROXS, UNIT-SYSDA. SPACE-(TRK ,(5,1)),
//
D1 S P-( , DELETE) , DCB-(R ECFM-VBS , LRECL-X , BLKS 1 Z E-6136 )
//* MICRSCOPIC CROSS SECTION DATA BY BURN STEP
//GO.FT0$F001 DD
- 11E0 CORE 1 0 0 29 1 0 0 0 0 0 -0.9037 1.0077 0.0 1.0077 2 6.6680E 3
, 18 2.2540E 3 20 2.4720E 4 100 0.0 0.0 1.000 0.0000 777 777 70.0 70.0 70.0 70,0 0.0254 0.0762 0.010 0.8103 1.0 25.6 0.0 0.5
/*
l //
l
1 36 A2. 2DB.UM i 2DB VM is a two dimensional, multigroup diffusion code. 2DB UM is a modification of the 2DB code developed in 1969 by Little and liardie of Battelle Laboratory in Richland, k'A , The modifications were made at the University of Michi Ean and included the addition of FIDO. the free .
format input processor, edit capabilities, improved calculational methods and additional input options. Version w6, 1980 of 2DB.UM was used in this study.
2DB VM solves the multigroup diffusion in the following form; Dg v4 g -E 4 +S -O g-1,N r.S 6 6 (A 1) where x N g1 i
S - - E-U k 77
[(vI)6,(6,+g['-l g'-1 f
(I )6,~E
- 4, E
(A 2)
The difference equations are applied with placing the mesh point placed at the center of a homogeneous mesh interval, Boundary conditions uced in this study were zero flux and zero flux gradient at the boundary.
Input consists of a set of input flags, description of the XY
- mesh, a l igning to each mesh interval a zone number, associating a material for each zone and input macroscopic cross sections, Listed is JCL and input data for an example run of 2DB UM, I
l 9
37
//llAP JOB (0368Vsi n, mena), 'COVINGTON LORNE' , TIME-30,
// MSCLEVEl.-(1,1 ) , MSCCI ASS-V
//S) EXEC PCM-UM2DH
//STEPl.lB DD DSN-USER.XO368.UM2DB.lMOD. DISP-SilR
//GO. FT06F001 DD SYSOUT-V, DCB-(RECFM-FMA ,l.RECL-13 3 hl KSI 7.E-l $96) .
// OUTLIM-20000 *
//CO. FT04F001 DD DSN-6SCRATCil, UNIT-SYSDA, SPACE-(CYl.,(l ,1)), '
//
DI S P-( , DELETE) , DC B- ( R ECFM-VBS , lREC!rX , BLKS ! ZE-6136 )
//* SCRATCil FILE FOR UM2DB DATA
//CO.FT08F001 DD DSN-6FLUXOUT, UNIT-SYSDA SPACE-(CYL (1,1)),
//
DI S P-( , DELETE ) , DCB- (R EC}N-VBS . LR EC L-X , B LKS 12 E-6136 )
//* FLUX OUTPUT FILE READ AND VRITTEN BY UM2DB
//CO.FT09F001 DD DSN-6 SOURCE UNIT-SYSDA. SPACE-(CYL,(1,1)),
//
D I S P- ( , D ELETE ) . DCB- ( R EC FM-VB S , lA ECtrX , B LKS I Z E-6136 )
//* SOURCE VRITTEN BY UM2DB
//G0.FT10F001 DD DSN-hXSECDMP, UNIT-SYSDA, SPACE-(TRK,(3,1)),
//
D I S P-( , DELETE) , DCB-(REC}H-VBS , lRECir X , BIES I ZE. 6136 )
//* SINARY FILE OF CROSS SECTION DATA FOR PERTV
//CO. FT14F001 DD DSN-6FLUXIN, UNIT-SYSDA, SPACE-(CYL,(1,1)) , '
//
DI S P-( , DELETE) , DCB-(REC FM-VBS , LRECL-X , BLXS I ZE-6136 )
//* INPUT FLUX CUESS FOR UM2DB CODE
//CO.FT15F001 DD DSN-&XSECIN,0 NIT-SYSDA. SPACE-(TRK,(2,1)),
//
DI S P-( , DELET E) , DCB-(RECFM-VBS , LRECL-X , B LKS I ZE-6136 )
//* INPUT CROSS SECTION LIBRARY FOR UM2DB
//CO,FT17F001 DD DSN-&XSBURND. UNIT-SYSDA SPACE-(TRK,(2,1)),
//
DI S P-( , DELETE) , DCB-(REC FM-VBS , lR ECL-X , BLKS I ZE-6136 )
( //* KATERIAL BURNUP DATA
//C0.FT18F001 DD DUMMY
//* FLUX PLOTTING DATA OUTPUT FILE
//GO. FT19F001 DD DSN-6FLXSCRT, UNIT-SYSDA SPACE-(CYI,, (3,1)),
//
DISP-(, DELETE) DCB-(RECFM-VBS.LRECL*X,BLKS12E-6136)
//* FLUX COEFFICIENT SCRATCil DATA FILE FOR ARRAYS CXS, CXR, AND CXT
//Co.FT0$F001 DD
- 100 C UATER DENSITY
- HEU CORE MODER TEMP C0fff .
'-------------l$$ ARRAY INTEGER CONTROL PARAMET).RS---*-----------
IS$
1RECAD IEVTYP ISERCll 0 -1 0 NGPS NXD0VN NXCARD NXTAPE IGUESS 2 1 15 00 0
MOUTER MINNER 1 Al.DIR 10VIAY ICEOM IM I
100 5 0 JH NZONES 0 0 98
) '
NMAT NMIXCD IBCL IBCR 71 58
- iPCT IBCB N120El. NJZDEL 15 0 0 0 0 0 0 0
NPRT NOTUSED NOTUSED IXSTRT IXENIC IEDIT ICPBUM IDUI'l ? E O O O 00 0 3
2 0 0E p** *
- AR RAY R EAl. CONTRO L PARAM ETER S------------------
EV- EVM PAREV 1.0 BUCK ALAL AIAll EPS EPSPAR 0.0 0.0 1.71000 3 0.0 EPS Fl.X POD 0.0 5.0-4 0.0 ORF POWER FISMEV XMUFA 1.0 3 0.0 1,55 ORFFS AHLYVO E T
- 3.281 3 193.1 0.0 1.0 5 -590.16 E T 23** 2 3 *
3B j 1.628 1.27 4.127 2.54 1.264 2.54 2.051 10RI .925 4RO.9638 0.6765 0.1 6RI.0262 0.1 0.6765 5Q10
)
4RO.9638 6RI.925 2.051 2.54 1.264 2.54 4.127 1.27 1.628 '
)-------------------
' ------------ 24 *
- ARRAY Y M ESil . I NCR EM ENTS- (.1M 24**
3R4.05 3.589 2RI.27 9R2.54 2R).27 5RO.9843 0.1 0.5726 SRI.3~51 0.5726 2RO.1 0.5726 2R2.025 4R4.05 1.66 0.1 2RI.617 0.1 1.66 0.5726 0.1 2Q10 4Rl.025 t
'- ------5$$ ARRAY ZONE !? UMBER FOR EACil MESH-(IM X
JM)-----------
1 32 94R34 32 1 3Q98 1 3R32 90R34 3R32 1 3R1 32 90R34 32 3R1 5Q98 3R1 3R32 S6R33 3R32 3R1
$R1 32 86R33 32 SRI 2Q98
$R1 88R32 $R1 IQ98 31R1 10R29 10R30 10R31 37R1 4Q98 21R1 10R35 2 BR38 2 2 8R39 2 2 8R40 2 10R36 27R1 21R1 10R35 2 38 6R26 38 2 2 39 6R27 39 2 2 40 6R28 40 210R36 27R16Q 21R1 10R35 2 8R38 2 2 SR39 2 2 8R40 2 10R36 27R1 21R1 2 BR41 2 37 8R42 37 2 8R43 2 37 8R44 37 2 8R45 2 2 8R46 2 17R1 21R1 2 41 6R18 41 2 37 42 6R19 42 37 2 43 6R21 43 2 37 44 6R22 44 37 2 45 6R24 45 2 2 46 6R25 46 2 17R1 IQ98 21R1 2 41 6R18 41 2 37 BR42 37 2 43 6R21 43 2 37 8R44 37 2 45 6R24 45 2 2 46 6R25 46 2 17R1
- 16R1 1 3R1 1 2 41 6R18 41 2 37 8R20 37 2 43 6R21 43 2 37 8R23 37 2 45 6R24 45 2 2 46 6R25 46 217R1 IQ98 21R1 2 41 6R18 41 2 37 8R42 37 2 43 6R21 43 2 37 BR44 37 2 45 6R24 45 2 2 46 6R25 46 2 17R1 '
21R1 2 41 6R18 41 2 37 42 6R19 42 37 2 43 6R58 43 2 37 44 6R22 44 37 2 45 6R24 45 2 2 46 6R25 46 2 17R1 IQ98 21R1 2 8R41 2 37 8R42 37 2 8R43 2 37 8R44 37 2 8R45 2 2 8R46 2 17R1 .
21R1 2 8R47 2 37 8R48 37 2 8R49 2 2 8R50 2 2 BR51 2 2 8R$2 2 17R1 21R1 2 47 6R10 47 2 37 48 6R13 48 37 2 49 6R14 49 2 2 50 6R15 50 2 2 51 6R16 51 2 2 52 6R17 52 2 17R1 lQ98 21R1 2 47 6R10 47 2 37 8R48 37 2 49 6R14 49 2 t 2 50 6R15 50 2 2 51 6R16 51 2 2 52 6R17 52 2 17R1 1
21R1 2 47 6R10 47 2 37 8R12 37 2 49 6R14 49 2 2 50 6R15 50 2 2 51 6R16 51 2 2 52 6R17 52 2 17R1 IQ98 21R1 2 47 6R10 47.2 37 SR48 37 2 49 6R14 49 2 2 50 6R15 50 2 2 51 6R16 51 2 2 52 6R17 52 2 17R1 21R12 47 6R10 47 2 37 48 6R13 48 37 2 49 6R14 49 2 l
2 50 6R15 50 2 2 51 6R16 51 2'2 52 6R17 52 2 17R1 IQ98 i 21R1 2 8R47 2 37 8R48 37 2 8R49 2 2 8R50 2 2 8R$1 2 2 8R$2 2 17R1 21R1 2 8R$3 2 2 8R54 2 2 8R55 2 37 8R56 37 2 8R55 2 27R1 21R1 2 53 6R$ $3 2 2 54 6R6 54 2 2-55 6R7 55 2 37 56 6R8 56 37 1 1 6R1 1 1 27R1 IQ98 21R1 2 53 6R$ $3 2 2 54 6R6 54 2 2 55 6R7 55 2 37 8R$6 37 1 8R1 1 27R1 l 21R1 2 53 6R$ 53 2 2 54 6R6 54 2 2 55 6R7 55 2 37 8R09 37 i BRI 1 27RI l l 21R1 2 57 6R4 57 2 2 54 6R6 54 2 2 55 6R7 55 2 37 8R09 37.1 8P.1 1 27R1 l 21R1 2 57 6R4 57 2 2 54 6R6 54 2 2 55 6R7 55 2 37 BRS6 37 1 8R1 1 27R1 21R1 2 57 6R4 57 2 2 54 6R6 54 2 2 $$ 6R7 $5 2 37 56.6R8 56 37 1 8R1 1 27R1 IQ98 21R1 2 BR57 2 2 8R54 2 2 8R55 2 37 8R$6 37 1 8R1 1 27Rl-
39 98R1 9Q98
--6$$ ARRAY IZMAT 6$$ MATERIAL NUMBER FOR EACH ZONE-(NZONES)-
l i
12434444064406444444406440644444111 1 1 1 10 11 12 19R4 3 4 )
- -----7$$ ARRAY MATID BU LIBRARY XSEC SET ASSIGNMENTS-(NXT 7$$
'-- --8$$
8$$ ARRAY REGION EDITS BY ZONE-(IEDIT X NZONES)--------
- EDI T s 1---- ,
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 a7 18 19 20 21 22 23 24 25 -
26 27 28 29 30 31 32 33 34 35 36 37 38
'~ 39 40 41 42 EDIT 43 44 2 45 46 47 48 49 50 51 52 53 54 55 56 57 58 AVERACE POWER IN EACH ELEMENT-- --
1 2263 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 27 14 15 2816 2917 29529 6 730 8 573158 32 33 34 35 26 27 28 18 19 21 22 24 25 10 13
-EDIT 3 AVERACE POWER ACROSS CORE --------
1234555565175555555855955555111 10 11 12 13 14 15 19R$ 4 56
-9**
9** ARRAY FISSION SPECTRUM-(ICM)------ -
i '
' FISSION SPECTRUM FROM ENDF/B IV CENERATED IN EPRI CELL ENERGY BOUNDARIES -
1.0 0.0 10 MEV, 0.821 MEV, 5.53 EV, 0.625 EV, O EV t
, 13** 13** ARRAY CROSS SECTIONS R OM CARDS-(ITL X NGPS X NXC
- 0.0 2 CP REG NONLAT **** 934 HEU WATER REFLECTOR 4.58252E 04 0.0 0.0 1.87576E 02 0.0 2.65154E 01 2.15893E 01 0.0 2.09724E+00 2.07848E+00 4.88029E 02
' 0.0 2 CP REG 3.86035E NONIAT04 **** 0.0 934 HEU STANDARD ELEMENT SIDEPIATE 0.0 1.28098E 02 0.0 1.80141E 01 1.67355E 01 0.0 5.97867E 01 5.85057E 01<1.23995E 02
' 0.0 2 CP REG 4.22391E NONIAT 04 **** 934 HEU DUMMY ELEMENT NON.WELED 0.0 0.0 1.69640E 02 0.0 2.60253E 01 2.21962E 01 0.0
' 2 CP REC TOTAL 1.65508E+00 1.63811E400-3.78687E 02
- 934 HEU STANDARD FUEL, CIAD, MODERATOR RECION 1.2(A88E 03 2.42961E 03 3.14696E 03 2.68234E 010.00000E 010.0000
'2 5.60711E CP REC TOTAL 02 8.19225E 02 1.35636E 01 1.63013E+00 0.00000E+00 3.5
- 934 HEU CONTROL WEL, CIAD, MODERATOR REGION s 1.34364E 03 2.52650E 03 3.27171E 03 2.64068E 01 2.24750E 01 0.0 5.82553E 02 8.50610E 021.40920E 01 1.71259E+001.62753E+00 3.679
' 0.0 2 CP REC NONIAT **** 934 HEU CUIDE TUBE RECION 4.12352E 04 0.0 0.0 1.64707E 02 0.0 2.56560E 01 2.20295E 01 0.0 1.57236E+00 1.55589E400 3.58520E 02
, ' 0.0 2 GP REC ' NONIAT **** 934 HEU CRAPHITE REFLECTOR
! 3.31330E 05 0.0 0.0 2.16348E 04 0.0 2.32656E 01 2.30476E 01 0.0 3.58813E 01 3.58596E 01 2.14693E 03
- 0.02 CP REG NONIAT **** 93t HEU ALUMINUM 6061 3.92668E 04 0.0 1 0.0 1.25551E 02 0.0 1.09891E 01 1.09382E 01 0.0
' 2 CP REC NONLAT_ 9.06178E 02 7.80627E-02 1.16865E 04 j 0.0 **** 934 HEU LEAD FROM EPRI-CELL 1.49720E 04 0.0 3.07090E 01 0.33645 I 0.0 l
40 0.0 1.95330E 03 0.0 3.700600 01 0.36936 1.81970E 04
' 2 CP REC NONI.AT **** 43t HEU BARE RABBIT TUBE REC 10N 8.39352E 09 5.0151?E 04 0.0 8. 39352E 09 2.4 7598E.01 2.10037E 01 0.0
?.03414E 07 0.0
'2 CRP CADMlUM RABBIT TUBE CROSS SECTION FROM El'R ,
- 8. 39352E 0.0 09 5.07182E 04 8.39352E 09 2.47598E 0) 2.10037E 01 0.0 2.87380E 02 0.0 1.62419E400 1.60385E*00 3.70592E 02
' 0.0 2 CP REC NONLAT **** 93% HEU CONTROL El.EMENT SIDE PLATE 3.78051E 04 0.0 1.63955E 01 1.54716E 01 0.0 0.0 1.21723E 02 0.0
'4.45142E 01 4.32969E 01 8.86137E 03
- 2 CP REC NONLAT **** 93% HEU REGULATING ROD 1.34674F. 07 0.0 2.09360E 03 1.34674E 07 2.80300E 01 2.60343E 01 0.0 7.85950E 02 0.0 1.15060E*00 1 07185E+00 1.78559E 02
' 2 CP REC NONLAT **** 934 HEU SHIM ROD WITH BORCN 1.87221E 08 0.0 1.73950E 02 1.87221E 08 3.59780E Ol 3..'1760E 01 0.0 0.31346E+00 0.0 0.95059E400 6.37223E 01 1.06183E 02
'0.0 004 VOIDED VATER CROSS 4.58252E 04 0.0 SECTIONS FOR THE PERIPHERY OF Tile CORE
'0.0 2.65154E 01 2.11893E 01 0.0 1.87576E 02 0.0 2.09724E*00 2.07848E*00 4.88029E 02
'0.095% VOIDED VATER CROSS SECTIONS FOR THE PERIPHERY OF THE COR 2.29110E 03 0.0 1.32577E 02 1.0'947E 02 0.0
'0.0 9.37880E 04 0.0
' 2 CP 1.04862E 01 1.03924E 01 2.44015E 03 0.0 93%
3.92668E 07 0.0 HEU AL TO APPROXIMATE AIR X SECS X 10 3 0.0 1.25551E 05 0.0 1.09491E Ol'l.09382E 01 0.0
'-----15** 7.80630E 02 7.80627E 02 1.16865E 04 i 35** ARRAY BUCELING MODIFIERS-(NZONES OR N7.ONES X NCRS)-----
F 1.0
' ------------17 17** *
- ARRAY BURNUP I N t OF FI S S I LE-( NXTA PE)------T ! ---
BURNUP T IN t OF FISSILE EDIT PROVIDED IN LEOPARD
'-----------91 $ $ AND 9 2 *
- ARRAYS TIM E STEPCONTROLe---------------
IBUCON 91$$ 0 NPRT NTUSD IXSTRT IDUPl?'IRECAD E T 1 0 00 0 0 E 92** 1.0E 10 1.0
/* E T
//
L
o .
41 APPENDIX B CROSS SECTION CENERATION AND CELL DISCRET12AT10N The two main inputs to LEOPARD are lattice geometry and material compositions.
There are three regions that describe the LEOPARD lattice: c fuel region, clad and void region and rooderator channel region.
For slab geometry, lattice spacing is measured from the center of the fuel plate to the outside edge of the region. Information for U S1 3 2 was obtained from references 9 and 1. S e B1 shows the lattice descriptions for the HEU and LEU cores , e B .I shows the number density or volume fraction for the dif ferent hor.ogenized zones of the HEU and LEU cores. A trace amount of boron.10 is added to all j i regions containing aluminum for the LEU core to account for the I impurities
( in AL 6061. Volume fractions for graphite and water !
reflectors, aluminum and lead in the thermal column were all l.000.'The cladding on the present llEU fuel plate is 20 mils (0.0508 cm) thick, l i
and the cladding on the proposed LEU fuel plate will be only 15 mils (0.0381 cm) thick.
TABLE B 1 i
Lattice spacing for the llEU and proposed LEU cores (cm),
region llEU LEU fuel material 0.0254 0.0254 clad 0.0762 0.0635 moderat.or 0.8103 0.4426
.=
O O 42 TABLE B ll Number densitics and volume fractions for the homogenized zones of the HEU and I.EU coren.
Zone description Isotope or Number Volume material density fraction l*/ barn /cm) [-)
HEU fuel material 28 U 2.254E 3 0 2.472E 4 16 0 6.668E 3 A1 0.9037 HEU clad Al 1.0077 HEU moderator HO 1.0000 2
HEU standard Al 0.7262 element side plate HO 0,2738 2
HEU control Al 0.7561 element side plate HO2 0.2439 i
HEU control Al guide tube 0.1589 HO 0.8411 2
HEU standard and U 1.829E 4 control element 0 2.006E 5 end plate 16 0 5.411E 4 Al 0.2200 HO 0.7566 '
2 LEU fuel material U l.761E 3 38 0 7.064E 3 10 B 2.539E 7 Al 0.8504 LEU clad B 2.986E 7-Al 1.0077 1.EU moderator HO 2 1.0000 LEU standard B 1.987E 7 element side plate Al 0.6654 HO 0.3346 2
LEU control B 2.414E 7 element side plate Al 0.8084 11 0 0,1916 2
! i l
l
TABLE B 11 cent.
Zone description Isotope or Number Volume material density fraction (w/ barn /cm) *[ - )
LEU control B 6.262E 8 element guide tube A1 0.2097 HO 2 0.7804 LEU standard and U 1.757E 4 control element U 7.048E 5 end plate B 6.687E 8 A1 0.3088 HO 2 0.6763 The computational model, in the Y direction, for the thermal column consists of 1 inch of aluminum, 4 inches of lead, and 17 inches of graphite. The Y mesh was 2 mesh lines (1.27 cm) in the t
aluminum, 4 mesh lines in the lead (2.54 cm), and 11 mesh lines in the graphite (some 2.54 and some 4.05 cm) .
Figures B1 and B2 show the HEU standard and control elements, respectively,
. (reproduced from the UMRR blueprints). Figures B3 through B5 show the computational models of various elements used in this study.
I i
1
44 l j 1
.i l
l
,% 3.00" q .
- p. -
Mr w,, ' ;
~m '
i F
1
- 4. . . ..
p -- ; , ,
4 3 2.6" f , ;- m'
( . 7 ? -N . .
- f. -~~-
i
~
l i
ir 7 ..I
+
AN ;
,*. ~' y Fig. 81. The HEU standard element from UMRR blueprints (shown with top handle).
- - - . ., v,-, - , , . . ,.. ee-.. ,, , .. , , , , , - . ,
4$
l i
i i
i l -
1 l
i if
, ~3.00
. . . .. ... 1 l , s:cs-c.rmn7'l.Q T e- ,
.x. .
/
/
,7 W
K. hY
-.m _ .& ,
}. l
,- , q x a s uu_m - -
gj
:, . . )..
e ,
,' .% ) !
- s. . . ,
M.,.
. u '
4f*yq; -
r-"rfe**.y r.-. r ,
y%
f.37f ..
.;..> .:: . 4
. , \
l 9 ..
g '
y '
s% ' )8 Aumm.muu>
l . , .
h . . . + .
/ - ] *;, ,
,i .- _
' f,
, . (,w.v . .. s. . * $' '
. .. 1s ., .
< l u .
e '.- . , c. r t
. ' .fb . .CG. l- .., ;'. r,,
{k ,.0 4.
f J
l.
Fig. 82. The HEU control element from UMRR blueprints.
)
l l
1 l
i
- - -~.....-.- . _ - . _ - - _ . _ - . . - _ . _ - _ _ _ _ _ - _ _ _ - _ _ _ . _ _ _
l----.---.-..
40
- 5. U b
E i
Fig. B3. The computational model for the HEU control element (not to scale).
l
o l
l ll l
(
l l
< _E j
t _
i l
l l
l l
l l
Fig. B4. The computational model for the IEU standard element (not to scale).
I
. , , _ , . , - - - . . ~ - . - - . - ,
48 H -
i i
l, M -
t 4
I Fig. AS. The computational model for the IEU 1rradiation fuel element (not to scale).
49 APPENDIX C SELECTION OF A' LEU CORE CONFICURATION Uhen the investigation to determine a new LEU clement and LEU-core began, the only constraint was that the element must be of the
{
same outer dimensions as an llEU ement so that it would fit into the grid plate. Two different elements were investigated: an element with 16 fuel plates and 2 outer aluminum plates, and an element with 18 fueled plates. The aluminum plates on the 16 plate element were i added to better protect the thinly clad fuel plates of the element during handling. The 16 fuel plate: element cores were typically = .
determined to be 3 elements -larger tha,a cores constructed with the 18 fuel plate element. The 1
different core configurations investigated are shown in Fi6 ures C1 and C2 for the 16 fuel plate !
i element and the 18 fuci plate element, respectively.
It was decided by the reactor staff that the 16- fuel plate element would be the most suitable element and that- core configuration (e) of Figure C2 would represent the proposed 4 LEU-core, i
i 4
4
i i
\
i ANL: -3.75 7.ok/k 2DB: 3.517.ok/k ANL:-1.45 7.ak/k 2DB: 1.827tk/k A!Mb!$$$i$$@lM$$h3 A($N0@$l$$$$$$di:Ii[ I i
Bhi . :.::. a u,u, a ..... BN.wh.!ME8hNN,ua!I{!!.!.!!.!$@ea:"3!
e :.:.
C $..$,:!8I8NA' '
+N ' O (d Ne
- M, Y N,.
- O C Nz
- l!! '
[ $
,uuku.,uu 8
0,!a.
' l Cd
@y@!!$
m .:. $.,,,
1 1
..ESi ?!
D T{
r ci r r r r yLl$ D .t' ci r i s:.::,
i ljj;.ij E : ' '
c2 .
c3 E c? ' c) '
. '.l$.i:8.i h
Fj l;{iffygi,' ,' ' {,{h,,
Fj .... h. i_j,,51,' .glyy;[,ij,
!,gj( i 1 2 34 56789 1 2 3.4 5 6 789 '
(a) (b)
ANL: -0.83 7.ok/k 2DB:-1.527.ok/k ANL: 0.10 7.ak/k 2DB:-0.28 7.ok/k AEMM$$$NMk A
$$$$$$@@!!Md BRMMMMMMMM 8 EEBMMMMMM t
c gg,
,- n ,g@-.
r c g;g,,,o gq
, , p 1 D[ ci '
M M' '.@.' .
D - c2 il' i l ' j!3,.
E c2 c3 '
'$ E $%( c2 5 c) I ' ijiik' Fi b38h 1 23456789 Fkk 1 2345
' ' r 6 789
@ i h!!f; (c) -(d)
ANL: 0.37 7.ok/k 2DB:-0.077.ak/k !
A
!MMMMM$!MN!!!!!!!
- T-MODE B j@$$$$$N!! CALCULATION C %j' it r. cd r i hh! ;
D EF l
Cl l i V l^
f
[ l C? I C3 f f , ?
1-2-3-4 5 6 789 (c)
Fir,. C). The 1.EU cotes i nves t l y,a t est us l ey', t he 1 (, l ue l . l>la t e elemen:
.i 1
i ANL: 0.37 7.ak/k 2DB: 0,30 7.ak/k ANL:-0.03 7.ak/k 2DB: 0.70 Y./ k/k l A ,QQgl[,f'!
g ,,
- t+3:v:v+f:f,fl;f;;f>
v 1+ ,,,
yv.v,.sv';,,,;,S+'w.
,:" 03, EMv+.3ss 3 A [((!j[lj!;$${$[jg%(3:q:.:<a:.
- :2..f.f.gp.,.qt.s:
v.ve.v. B .tf0P(fpi.NN
.e. - u/N, NNd, '.I'm+,d.yi;. 0;,
, :,,v,. .s.. . .v - e. vie ,
i -3 3,8,4,1 " " . ' " ' " 8,8,8,2,&, nna
'r i C $3-(f i i cd g,(3,j,t'44:g: C:.3: r cd si:ijij:i:'!:!!!:.:i:! ,
D r ci r i e s hh 9 ,'ff'(, ci r r i Iff,5fff,!
i E :t2.! r er ygp ;---
i c) j!!!!l3:I 03 E,IINjkil'd i
.,ye;7 cr 3
i c) i N[f.!
i F[';,u.s,eiv,g)gi ir r ,cR g:.,3,:, 5,'+: . y 294 nci.; _ l # : ;cci[3.....:++:
n ?2?s :.:-
v .n n. 2 _-
s,na.n, , ,1ux,3 _
_ w cccc, cc,y, 1 2 34 56 789 1 2 3 4--5 6 789 i (a) ( t> > . !
ANL 1.40 7.ok/k 2DB: 1.917.ok/k ANL: 1.15 7.ok/k 2DB: -1,62 7ok/k
{ A EMMEN A $$$$$$!@f$$$ i B
5,[. hhhh '
B MMMMM%$$$!
( C I r c4 d:jjjj:]l[ C @]l@ r i r .. c4 %g@j DME$$i '
w-C2 r r r .' N Dims % $8icii i. l i
.!$5:!!
- 3. :,;
E MM' c2 ' c3 ' t-
$! Eijiiiil!!$'c2 i- c3 e i j:i:.'.i:!: ;
rM28.< 82m r r rjg?;nei, , , gi.fi!!!!![ "
123456 789 1 2 3456 785 (c) (a) l ANL: 0.17 7.ok/k 2DB: 0.87 7.ok/k- )
A$$$$$$$$$
B.MWSMENEME C RMSr- r o
?Wfs D -r cl r ' r! h
- r. f r 'f EM,@ L r- c2 8 ca t' s %j f TRMB'r 1234 56
'BRM '
789
<c)
L F i r, C?. The 1. Ell cores ' invest 1;;a t ed usin;, t he 18 fucl plate e l enie n t j i
.,- 5. -
52-APPENDIX D cal.CULATION OF POWER PEAKING FACTORS AND POWER DISTRIBUTIONS The power peaking factor was calculated by dividing it into l'st radial, elemental, and axial components. The definitions of these components are given in Table D-I.
TABLE D I-Power peaking factor definitions.
Power generated in element
- dI"I Average power per element in the core
~
Maximum power generated in the axial direction
' axial Average power-generated in axial direction. I Local maximum power generated at z
( * ' clement
~
Average power generated at z-Peak power in element t
~
' total Average power in the core P.P. total - .P.
radial * '
' clement
- P.P.,xg,y l The radial power peaking factor was determined by selecting' edits l
in 2DB VM that perform the averaging of the power in l'ndividual elements and in the entire core. The elemental' power
_ peaking factor calculation required scanning each' mesh interval in I the element for the maximum value for the thermal flux. The area o1~.
i this mesh interval was calculated and the peak power density determined by summing the product. of the flux and fission cross L
section for each energy group and dividing by the area. The average.
- F l
1
- o 53 !
power density was determined by 2DB VM edits described above'for the radial power peaking factor. The axial power peaking factor was determined by using a chopped cosine representation: of- the power distribution. It was assumed the same for all elements in the core.
Power distributions were determined by selecting a different set of edits for 2DB UM, On the following pages, Figures B1. through B4[ show the power peaking factors and. power distributions for the HEU and proposed 1.EU cores. Figure B5 shows the measured relative thermal flux for the current HEU core. The values for the thermal flux were normalized to 2
1.0+12 n/cm /s. It was from comparison of this figure to- the calculated HEU power distribution that lead to the investiEation of the flattening of the thermal flux (thermal flux .is direc*1y. l i; proportional to power), .
i.
l I
l I
l
a ;
1 l
1 1 2 3 4 5 6 7 8 9 A ,
i B
1 0.85 0.95 0.99' C 1.17 120 1.13 Ben. '
1.33 1.52 1,49 O.S2 1.15 1.11 1.03 0.84 p 1.25- C1 1.13 1.12 120 1.34 1.53 1.73 1.65 1.65 1.50 .
0.s0 1.15 1.01 '0.33 E 1.25 c2 1.13 c3 1.28 125 1.49- 1.73 1.72 1.38 p 0.s1 123 0.s9 128 0.s3
-123 p
F BRT CRT V! 1.61 1.68- 1.65 V Ax!&1.33 -
Relial 1.18 1.18 .1.21 1.02 Demet 122 123- 124 3.30 Total 1.91 1.94 2.00 1.76 C1 C2 03 REG. i 3
. i i
Fig. D1. HEU core power peaking factors. ;
i I . , _ . - _ _ . _ . . . . . _
53 I
1 2 3 4 5 6 7 8 9 A
B 0.89 ' O.95 0.96 C 12s las geg, 1.46 1.63 1.76 1.86 O.87 1.07 1.05 0.96- 0.85 y 1.30 Cl- 120 12s 127 1.45 t
1.51 1.84 1.72 1.62 1.64 0.86 1,09 0.99 0.83 E 1.41 c2 1.32 c3 1.37 1.47 1.61- 1.91 1.80 1.62-p I
. 0.8s 8#
0.s4 128 0.s p F BRT 128 CRT U 1,66 1.72 1.67 V -
Axial =1.33 Radial 1.17 1.16 1.24. .1.11 Dement 1.32 1.33 1.35 1.26 Total 2.06 2.06 2.22 1.86 i C1 C2 C3 REG. :
l Fig. D2. Proposed LEU core power peaking factors, i
i
, - , - . - - . - . - , . . .. -. . . , , - . , . - - - - . . ~ . - . - - - - - - - - - .
l l
i 1 2 3 4 5 6 -7 8' 9 A
B 5.01 1120 11.70 C "#'
2.51 6.60 6.85 f
-10.90 13.66 13.10 12.17 3.89 D c1 5.45 6.83 6.55 6.09 4.94 .
(
10.65 13.78 11.98 9.86 E
6.89
" 5.99 4.93 i 5.32 l
l OBRT 10.79 11.17 11.03 O CRT F
V 6.39 5.86 6.51 V Kilowatts 8.41 8.27 8.61 7.27 L % Power 4.21 .4.13 4.31 3.64 C1 C2 C3 . REG.
L Fig..D3. HEU core power distribution. >
n 1 l 1 1 2 3 4 5 6 7 8 9 A i B
10.98 11.68 5.88 C
5.49 5.84 D- 2.94 10.74 13.18 12.94 11.80 10.54 0 5.37 C1:
6.59 6.47 5.90 5.27 I "
10.64 13.44 12.18 10.24 E
5.32 c 6.72 Q
6.09- 5.12 O 10.94 11.58 11.26 p F BRT CRT V
5.47 - 5.79 5.63.
V Kilowatts 7.96 7.94 8.46 - 7.58 4
- 7. Power 3.98 3.97 4.23 3.79 C1 -C2 C3 REG.
Fig. D4. Proposed LEU core power distribution.
i,-,e w , , , - - , , . , ..s,.na
, - , .,,.-,m .--,.e..., ,----,e e -~ -~em,,e -w-> a w - - ,
. 58 ,
i 1 2 3 4 5 6 7 8 9-
.I J A l l
1 B 0.55 S 0.55 j i
C 0.35 0.67 D 0.53 1.70 c1 3.30 2.80 2.00 1.30 0.36 E 2.20 c2 5.00 c3 3.30 2.00 F 2.40 1.40 3.00 0.63 0.10 V i l
'l C1 C2 C3 REG.
Fig. D5. Measured relative thermal flux of the HEU UNRR. j
1 59-APPENDIX E i l
CALCU!ATIONS OF Rt. ACTIVITY COEFFICIENTS The temperature coefficient of the moderator is the= sum of two 1
component coefficients, the moderator density coefficient 'and moderator temperature coefficient. The moderator density coefficient is due to changing only the density of the moderator. The moderator '
density coefficient is shown in Figure El for the llEU and 'EU' L cores.
The moderator temperature coefficient is due to' changing only the temperature of the moderator. The moderator temperature coefficient !
is shown in Figure E2 for the llEU and LEU. cores. The summing of the. ,
individual coefficients to get the total moderator are shown in Figures E3 and E4 for the HEU and LEU cores, respectively.
The Dopple r - coe f ficient is due to the- resonance - absorption in i U. Increased fuel- temperature causes a -broadening of. the resonance peaks and hence increased absorption during the - slowing down of the neutron. There is no measurable or calculable Doppler coefficient for the llEU core because of the relatively small amount of U. Figure E5 shows the calculated Doppler coefficient for the l
proposed LEU core.
The reacivity worth of a 2~31 cm void at the midcore position E-5 for the llEU is shown in' Figure E6. This figure shows the two ,
methods used in modeling -a void: the slow reduction in the number density in water and the use of adjusted aluminum cross: sections.
-Both methods appear to show a positive reactivity for the void..
Since the void coefficient has never been measured for an interior core position, no comparison could be made to measured data.
t -
4
u
. . 60; )
.t
. 2
-I 0.000 <
e
~l
.)
l 0.00l <
I l
-0.002-i l
0.003-l o- '
O
- -0.004- I
-0.005;
(
1 i
')
-0.004-I eLEU- )
-0.0072 .;
.av 1 l
1
-0.00e-5' 'W' 'W' E t' 'l' W '5' g to 30 40 50 60 70 so s0 tog Occatts ca stus .l
.l
)
l Fig. El. Moderator density coefficient versus temperature l for HEU and proposed LEU' cores.
l J
1
. _ . ~ . -
64 l
l 1
l i
I t 0.0000 - e ,
L
-0.0001 -
0 0002 -
1
-0.0003 <
, l
- 0.0004 -
\
-0.0005 -
0.0004 <
+0 000F -
, -0 0006 <
-0.0009 -
i 0.0010 -
-0.0011 -
- 0 0012 - "
-0.0013<
- 0 0014-
-0.0015-U +0.0016<
g -0 00lF<
0 -0.0010<
0 -0 0019-a
-0.0020-
-0.0011-
\
=0 0022-
-0.0023-
-0.0024-
-0.0025-
-0.0026-
-0.002F<
l . -0 0020-
-0 0029<
! 0.0030-
=0.0031< *
-0.0032- 3
-0.0033-0 0034-0.0035<
- 0 900 200 300 400 500 ,600-3 0(CREES MLSIUS Fig. E5. Doppler coefficient versus temperature for proposed LEU core. ,
i:
I 4
5 J
, , _, , . , . . , , . , . . n ,
6) l l
l
,. o.oco ' I o.ool -
t
-o.002 -
-0.003 -
-o.004 -
=o.005-E-0004-u ,
- LfU a
3 -0.cor-
=o.000-(
-o.oos-
-o.clo<
-o.011-i l
l
-o.oit <
. estu
-o.oss , .,. . , . ... ... .,.
to ao 40 so se to so so s oo .
- McMES CELSIUS Fig. E2. Moderator temperature coefficient versus temperature for HEU and proposed LEU' cores.
+
63 1
1 4
0.000-
-0.001 -
- 0.002 -
+0.003 -
~0 004 -
I 0.005 -
-0.004 - l
-0.001-eOCasiTT
- 0 000-g -0.000-
- -0 010-Y E -0 011- i
-0 012-y eTEPP
-0 013-l
-0 014 I
-0 015-
-0 016-
- 0 0lf-
-0.010-
-0.019-e TOTAL l -0.020 ,. .,- ... , .,. ... .,. ..
20 30 40 50 s0 70 e0 oQ 00 O(CRCES M LSIUS .
Fig. E3. Moderator coefficients versus ternperature for HEU Core.
l l-I l
l l
a l l
l 1
-i 0.000- l
- 0.001 -
I 1
0.002 -
0.003 -
0.004 - .
-0.005 -
l 0.006 - i y itW
-i
- -0 001- i
u D(NSITY 4
E
-0 008-
.1
( -0 009-
-0.010-
-0.011-
-0.012<
I l
-0.013- i'
- TOTAL
-0.0 e, to 30 40 50 60 .r0 eo e0 800 M0RCES CEttlus l '(
t-Fig. E4 Moderator coef ficients versus temperature. for LEU Core.
t f
f a
k
, ,,e -
,,,-..w - , . , , , . , , . - - , , - . -, e--a n-v-, r w ,
l . ..
o l 1
l I
1 0.3
. usug .. .. a. usted . . . .. .. d j. /e l
l . aluminum cross sections
- 1 o,4 . .
Extrapolated
- s. 4-o,s - ... .. .
g . .
s 0.2 - '- -
i-M -
- Q . . : :
go o,i .................
M o ,o g 10 20 30 40 60 60 7.0 80 90- 100
, e H
% Voided g .0,3 .. .
0,2- . . . . . . . . . . . . . . . . . . . . . ...,..............;..... . . . . . . . . . . .
l 0,3 . . . : : : :. .. : .. ..
-0.4 - . .
l . .
d
-0.5 1
. Fig. E6. Reactivity worth of a mid-core void versus percent . voiding of water for the HEU core. '
i i
- v. . . _ . . _ _ , . . _ . . . . - . . _ _ . . . . . ..
66 t t
TAPPENDIX F METHODS TO CALCULATE REACTIVITY WORTHS OF CONTROL RODS
- Direct reactivity calculations of highly absorbing tontrol rods are impossible for neutron diffusion codes. An approximation for control rod behavior can be made by performing _ a reac tion ' ' rate i matching between a Monte Carlo code calculation and a diffusion code q
as described in _ reference (11). '
The first step is to create.a unit cell consisting of a control rod element surrounded by 8 standard fuel elements using reflective +
boundary conditions on the four ' sides. This unit cell is- the geometry used in both codes. The Monte Carlo code is run once and the ratio of the absorption. rate in the control : rod ( control rod + -
-j i water in guide tube + guide tube) and the. fission rate in the rest of the unit cell (fuel' + moderator + clad + sideplates) .is . then !
determined for each' energy group that will be used -in the diffusion -j code.
a
.R E
-a .l rod ,
a
- p
- V-l rod FW f I cell f
- 4'* v I cell i These reaction rates as determined- by the Monte Carlo code :are the base rates that will be matched by the diffusion code.
Initial diffusion. cross sections _ are . generated by a cross-section collapse code (LEOPARD) . for all regions in the unit cell including the homogenized control rod -
guide . tube region. .The ,,
i diffusion code Is then run with these initial cross sections and the reaction rate ratios are calculated in each energy group, The a
}
-67 l
following is the stepwise method used to match all group reaction rate ratios:
1
- 1. Run diffusion code and determine reaction rate ratios for each energy group.. j
- 2. Start with the highest energy group and proceed down i through-each energy group, l
- 3. Calculate the new I and I t for the control rod region for the current ene8sy group'as follows; a- ' f I cell y , ,
- "*" F 2)
Rg MC.
4
- V l rod AE, E, l new '
a I old '}
E I
~ # 0 tr Iold
~
tt new a i 4. Rerun the diffusion code replacing the old ~ absorption
-and transport cross sections with the new values.
- 5. Repeat steps 3 and 4 until the reaction rate ratios for the current energy group match for both the diffusion and Monte Carlo _ codes. Then proceed ( to - the next.Iower
, energy ' group and repeat steps' . 3 and' 4 Repeat process for all ener6y groups.
- 6. The process of sweeping down' through' the energy groups has the effect of slightly changing the reaction rates in all the other energy groups. Because the reaction rates ' in the higher energy-groups _have changed steps 2 through 5 must be repeated until there-is convergence of ratios in all-energy groups.
Results of the reaction rate matching method calculations-for the HEU core are shown in Table F 1.
68 1 TABLE F-1 l
Calculated and measured control rod worths for the UMRR HEU. core (uk/k).
rod I calculated I measured I l
' l l C1 l -2.799 l 2.64 C2 l- 2.712 .l 2.65 C-3 l, 2.859 l 3,35 reg. l' 0.744 l 0.35 Rod worths for rods 1 and . 2 compare favorably with measured data, while rod worths' for rod 3 and the regulating rod are l significantly different. The cause of this difference arises from the-assumptions of using the unit cell geometry to calculate : adjusted cross section as described in the previous section. ' The ..relatively small HEU core has a large spatial dependence of the flux and hence the reaction rates vary considerably locally. In the case of the shim rods only one unit cell calculation was performed and the Esame initially-generated cross sections were used in all three shim' rods, 1
l causing the calculated reactive worths of the rods to be similar. The l
large difference in the case of - the regulating -rod ' ik because the geometry of the unit cell is much 'too different form the global calculation. In the unitcell, the regulating rod is surrounded by.8 standard fuel elements with reflective boundary conditions on all sides (no. XY flux ' gradient). In the global calculation', the regulating rod is at the edge ' of the core and is surrounded by' 4 standard fuel elements on -one side and 4 water cells on the other side, a significant flux gradient .is present:in this geometry.
To eliminate the : effects of < geometry in the reaction rate matching method the complete global . geometry must be used for both ; l the Monte Carlo and diffusion codes. Changing.co the global geometry
69 wil1~ significantly increasc'the time to determine the adjusted cross sections because of the increased computer , time and the neutronic coupling of the control rods. This increased time is undesirable but seems the only way to achieve accurate results. -
Since the regulating rod is a stainless steel 304 tube filled with water and the SS 304 is homogenized across the entire rod region, the cross ' sections . generated by the cross section collapse . .
code (LEOPARD) should give adequate results. The unadj us ted cross sections were used in a global calculation and the results are in good comparison with the known worth of the regulating rod. The results are in Table F-II.
TABLE F II
( Calculated and measured worth of the regulating.
rod in the HEU core (tok/k).
rod l calculated l measured- .{
reg. l 0.31 l -0.35-i 1
l i ?
-p.
4
-,w., -
70-APPENDIX C l ADDITIONAL REACTIVITY CALCULATIONS -
1 l
The reactivity of a standard element on the periphery of the core was determined for both the HEU and LEU cores. The position of the element in the HEU core was C-7 and C 3 for the proposed LEU core. The reactivity worth for the HEU core was 1.484 and 0.95% for l the LEU core.
Table GI shows the various Irradiation Fuel (IF) element designs that were investigated along with the power peaking factor *
(P.P.) and thermal flux values. The IF was inserted in positionn E 5 I of the grid plate '
TABLE C-I I
Various irradiation fuel element designs. power peaking factors and thermal fluxes in grid, position E-5.
- of w of width Total Total Peak Average Fueled Al o f- P.P. -P.P. Thermal Therrnal plates plates trap in in flux in flux in k"*
removed added (cm) IF E6 IF IF 2 0 1.202. 2.02 -2.17 2.20+12 2.07+12 1.0090 4 0 2.088 2.20 2.20 2.85+12 2.57+12 1.0082 4 2 1.202 .2.12 2.18 2.65+12 2.61+12 1.0063 6 0 2.970 2.29 -2.28 3.40+12 3.09+12 0;9994 I 6 ~2 2.088 2.26 _2.25 3.20+12 3.04+12 0.9992 8 0 3.750 2.38 2.36 3.90+12 3.41+12 0.9923 8 2 2.970 2.31 2.31 3.70+12 3.36+12 0.9873 10 2 3.750- 2.43 2.36 3.95+12 3.48+12 0.9845 all --- --- -- -
2.71 5.20+12 3.71+12 0.9689 ;
none -- --
1.91 2.22 -- ---
1.0124 i
An experiment was performed using the Core. Access (CA) element.
The experiment was split into two runs on two differenti days. The l,
first was on 10/23/87 with the CA being' filled with air, and the-second run was on 11/4/87 with the water - filled. For ' reasons of
71 ,
/
f med in two
/ shim heights ' was per orthe same I of the rods to cecparison, measurements set all shim he same first was to different manners. The set shim rods 1 and 3 to t results are i
height. The second method achieve was tocriticality. The a rod 2 to hiight and adjust shim ls
-in the following tab e .
TABLE G-1I-t Raw data for CA experimentho . ,,,,,,
hes) ;e ,
Control Rod Heights (inc . reg.
3' 2
1 0.405 t 24.0 0.400 t 10/23/87temo inlet 68 F 21.12 21.12 24.0 21.12 21.00 I 1.) 21.44 0.670 t No CA 2.) 21.00 24.0 20.12 0.665 4 20.12 24.0 21.00 20.12 18.60 CA with 1.) 2.) 21.00 air 0.390 4 11/4/87temp 70 F 24.0 0.385 4 inlet 21.18 24.0 21.18 21.001 % + N-1.) 21.18 21.63 No CA 21.00 0.825 % l 2.) 24.0 0.800 t 19.63 24.0 19.63 21.00 19.63 17.55 CA with 1.) 2.) 21.00 i
l water <
TABl.E C III air-element Cases for normal (
A and 8.are Core Access modeling computer Reactivity of filled ) and flooded conditions.d 2DB tm are ' from experimental data an l
(SAk/k).
2DB UN .,!
. CASE B-CASE A ~ 0.421
, 0.265
.0.265 0.479 CA with air ' O.415 0.435 i CA with H2O '?
..:~.. '!
, - -. ., , . -