ML20246N532
| ML20246N532 | |
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| Site: | 05000199 |
| Issue date: | 02/28/1989 |
| From: | Freese K, Matos J ARGONNE NATIONAL LABORATORY |
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| NUDOCS 8905190467 | |
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Text
_-_ _ - _ - _ _ _
ANALYSES FOR CONVERSION OF THE i MANHA'ITAN COLLEGE ZERO POWER RFACTOR i
FROM HEUTO LEU FUEL i
l J. E. Matos and K. E. Freese RERTR Program Argonne National Laboratory Argonne,IL 60439 ,
I February 1989 f,(($bo 4 P
_ _ ____----__-__-_.-_______________-----___________---___-_-.____-----_J
i
SUMMARY
This report contains the results of design and safety analyses performed by the RERTR Program at Argonne National Laboratory (ANL) for conversion of the Manhattan College Zero Power Reactor (MCZPR) from tta use of HEU fuel to the use of LEU fuel. The objectives of this study were to. (1) maintain the present HEU fuel element design and core size, (2) maintain or improve upon the present margins of safety, and (3) maintain as closely as possible the technical specifications and operating procedures of the present HEU core.
The LEU fuel element has essentially the same design as the present HEU fuel element, but the fuel meat contains LEU U3 Si2-Al fuel instead of HEU U-Al alloy fuel. This LEU silicide fuel has been approved by the Nuclear Regulatory Commission for use in non-power reactors.
Documents that were reviewed by ANL as bases for the design and safety evaluations were the MCZPR Safety Analysis Report, the MCZPR Technical Specifications, and the NRC Safety Evaluation Report related to the renewal of the MCZPR operating license in February 1985.
The methods and codes that were utilized have been qualified using comparisons of calculations and measurements in LEU demonstration cores in the Ford Nuclear Reactor at the University of Michigan and in the Oak Ridge Research Reactor at the Oak Ridge National Laboratory.
Only those reactor parameters and safety analyses which could change as a result of replacing the HEU fuelin the core with LEU fuel are addressed in this report. The attached summary table provides a comparison of the key design features of the HEU and LEU fuel elements and a comparison of the key reactor parameters that were calculated for each core.
The results show that all of the objectives of this study are fully realized and that the MCZPR reactor facility can be operated as safely with the new LEU fuel as with the present HEU fuel.
l I. - -. _ .- _--__-____
l ii l
l
SUMMARY
TABLE HEU and LEU Design Data, Com Physics, and Safety Parameters for Conversion of the Manhattan College Zem Power Reactor i
DESIGN DATA HEU Com IEU Com 15 15 Number of Standard Fuel Elements 1 1 Ring 2 Fueled Number of Partial Fuel Elements i
Fuel Type UAl Alloy U3Si2-Al l Enrichment,% 92.0 19.75 Uranium Density, g/cc 0.7 4.8 Number of Fuel Rings per Element 6 6 Number of Fuel Plates per Ring 3 3 U-235 per Standard Fuel Element, g 200 235 U-235 per Partial Fuel Element, g 24.0 28.2 Fuel Meat Thickness, mm 0.51 0.51 Cladding Thickness, mm 0.38 0.38 I
Cladding Material 1100 Al 6061 Al Natural Boron Impurity Equivalent 10 20 in Cladding and Structural Aluminum, ppm HEU HEU LEU REACTOR PARAMETERS Measured Calculated Calculated Cold Clean Excess Reactivity, % Ak/k 0.32-0.40 1.2 0.4 1.1 0.4 Monte Carlo
- 1.4 1.4 Diffusion Theory Reactivity Bias (Diffusion Theory), % Ak/k -
-1.0 -1.0 0.4 0.4 Adjusted Excess Reactivity, %Ak/k -
-0.9 -1.2 -1.3 Diffusion Theory !
Worth of Regulating Rod, % ak/k Reactivity Bias on Reg. Rod Worth, % Ak/k - 0.3 0.3 l Adjusted Worth of Regulating Rod, % Ak/k - 0.9 -1.0 l Shutdown Margin, % ok/k -0.5 -0.5 -0.6 (with Shim Rod Stuck Out) !
Worth of Shim Rod + Reg. Rod, % Ak/k < -3.4 - 4.3 0.6 -3.9 i 0.5 Monte Carlo
- -4.7 -4.7 Diffusion Theory
< - 2.5 -3.5 -3.4 Diffusion Theory ;
Worth of Shim Rod, % 6 k/k Worth of Emergency Shutdown Rod, %Ak/k < - 3.0 - 3.1 0.5 - 3.6 0.5 Monte Carlo [
t 65 59 Prompt Neutron Generation Time, ps -
t !
EtTective Delayed Neutron Fraction - 0.0078 0.0078
-2.3 -2.0 Fuel Elements l Temperature Coefficient, Ak/k x 10(-4) per *C -
i 0.0 -1.3 Fuel Elements Doppler Coemeient, Ak/k x 10( 5) per *C -
Void Coemeient, Ak/k x 10(-3) per % Void - -1.5 -1.6 Fuel Elements l l
L.
El TABLE OF COhTENI'S Pace
' 1. INTRODUCTION . . . . . . . . . . . . . . . . ..... ... . ...... . .. ... . .... .... ....... ..... ... ......1
- 2. REACTOR DESCRIPTION .... ..... ....... .......... ..... .. ..... .... . . ... ........ .. ... 2
- 3. FUE L ELEMENT DE SCRIPTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
- 4. C A LC U LATI ONAL M O D E LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 4.1 Nuclear Cross Sections for Diffusion Theory Models .. ............ . . . ..... . .... . 6 4.2 Reactor M odels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... ..................................9 i
- 5. DYNAMIC DESIGN EVALUATIONS ...... . ......... . . ....... . . . . . . . . . 10 !
5.1 Cold Clean Excess Reactivities.. ... .. . ....... . .. .. .... . ... . .. . . . . . . . . . . . . . 10 5.2 Sensitivity' Calculations for LEU Reference Core . . . . .. . . . . . . .11 5.2.1 235U Loadings ..... . . .. ... .......... .. .. . .. .... ................11 5.2.2 p om B oro n Equiv alen ts . . . . . . . . . . . . . . . . . . . . . . .. ... . . . . . . . . . . . . . . . . . 11 5.3 Power Distributions and Power Peaking Factors . . . .... .. . ......... .... . 12 5.4 Control Rod Worths and Shutdown Margin ... .. . . . . . . . . . . . . . . . . . . . . . . . . 14 5.4.1 Control Rod Descriptions and Calculational Models. .... .. . .... . . . . 14 5.4.2 Methods for Calculating Control Rod Worths ... ... . . .. ... .. ......... . 14 5.4.3 Reactivity Worths of Control Rods . ....... .. . ....... .. . ... . . . . . 15 l
5.4.4 Shutdown Margin . . . .. .. . .. . . . .. . . .... . . . .. . ... . .... . . . . . . 16 !
5.4.5 Emergency Shutdown Rod .. ... ...... .. . ....................... .... . . . . . . 16 J i
- 5. 5 'Re a c to r Kin e ti c s P ai a m e te rs . . . .. . .. . . . .. . ... . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 1 5.6 Temperature, Void,'and Doppler Coefficients . .. ........... .... ...... . . . . . . . . . . . . . 17 4 l
5.6.1 Non-Isothermal Reactivity Changes with Temperature . .... . ... . . . 17 5.6.2 Isothermal Reactivity Changes with Temperature ..... .. . ........ ... . . ... 21 1
l I
- 6. A C CIDENT ANA LYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ... . . .. .. ... . . 22 6.1 Inadvertent Excess Reactivity Insertion ... .. .. ... . . ... ... . . ... . .. ..... .22 6.1.1 Comparison of Calculations with SPERT-I Experiments ., . .. . .. . . . . . . 22 6.1.2 Analyses for the MCZPR ... .. ... ...... . ...... . . . . . . . . . . . . . . 23 REFERENC ES . . . . . . . . . . ... .. . .. . ..... ... . ... ...... .. ... .... . . . 24
- _ _ _ . _ _ - - _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ - _ _ - _ D
- ANALYSES FOR CONVERSION OF THE '
MANHA'ITAN COLLEGE ZERO POWER REACTOR FROMHEUTO LEU FUEL J. E, Matos and K. E. Freese
~
RERTR Program -
Argonne National Laboratory.
Argonne,IL 60439 February 1989
. INTRODUCTION
'This report contains the results of design and safety analyses performed by the
'ERTR' Program at the Argonne National Laboratory (ANL) for conversion of the lanhattan College Zero Power Reactor (MCZPR) from the use of HEU fuel to the use of
.EU fuel.' The objectives of this study were to:(1) maintain the present HEU fuel element -
esign and core size, (2) maintain or improve upon the present margins of safety, and (3) taintain as closely as possible the technical specifications and operating procedures of i
g ae present HEU core, The LEU' fuel element has essentially the same design as the present HEU fuel t.
.lement, but the fuel meat contains LEU U3Si2 -Al fuel instead of HEU U-Al alloy fuel. :
specifications for the LEU fuel element and a fuel element with removeable plates were l l
etermined by EG&G Idaho using measurements obtained by disassembling an MCZPR
- ummy fuel element that was shipped to EG&G Idaho by Manhattan College. A detailed afety evaluation of LEU US Si2-Al fuel can be found in Reference L ,
Documents that were reviewed by ANL as bases for the design and safety evaluations L
iere the MCZPR Safety Analysis Report,2 the MCZPR Technical Specifications,3 and the
. IRC Safety Evaluation Report4 related to the renewal of the MCZPR operating license in
'ebruary 1985.
I.
The methods and codes that were utilized by ANL have been qualified using I 5 i omparisons of calculations and measurements of LEU demonstration cores -9 in the-
'ord Nuclear Reactor at the University of Michigan and in the Oak Ridge Research '
'eactor (ORR) at the Oak Ridge National Laboratory. Additional qualification has been btained via international benchmark comparisons 10-11 sponsored by the IAEA.
l The ' design and safety analyses in this report provide comparisons of reactor i arameters and safety margins for the MCZPR HEU and LEU cores. Only those arameters which could change as a result of replacing the HEU fuel in the core with
.EU fuel are addressed.
I L___________-___-_-_-__-_--- _ _ _ - . _ - - - _
- 2. REACTORDESCRIPTION The MCZPR is a heterogeneous, light-water-moderated, pool-type reactor fueled using 92% enriched U-Al alloy fuel. The reactor was manufactured by AMF Atomics of Greenwhich, CT, and is licensed to operate at a maximum power of 0.1 W. Before being installed at Manhattan College in 1964, the reactor core had been used since 1961 by AMF Atomics to perform critical experiments in the PTR (pressurized tube reactor) located at the Industrial Reactor Laboratory (IRL)in Plainsboro, NJ.
The core consists of 15 full fuel elements and 1 partial fuel element in a hexagonal array (Fig.1) located on a grid plate that is immersed in an open tank of demineralized water that serves as both moderator and reflector. The total fissile loading is 3024 grams of235U. Since the water in the reactor pool has a large heat capacity relative to the 0.1 W power level, no recirculating cooling system is provided.
The reactor is controlled by the vertical movement of two Y-shaped control rods that operate between core fuel elements. One of these rods is a cadmium-stainless steel shim rod and the other is a stainless-steel regulating rod. An aluminum-clad emergency shutdown rod containing B4C is mounted on a wall near the reactor core. If both the shim rod and the regulating rod are disabled in the out position, the reactor can be shutdown by manually inserting the emergency shutdown rod into the core. There has never been a need to use this rod.
The LEU core (Fig.1) also consists of 15 full fuel elements and 1 partial fuel element in the same arrangement as the HEU core. The one exception is that the fuel element in position 46 of the HEU core was moved to position 14 of the LEU core in order to increase the reactivity worth of the regulating rod. The LEU core will use the same shim rod, regulating rod, and emergency shutdown rod that are currently used in the HEU core.
l l
l.
1 .
Figure 1. HEU and LEU Cores HEU CORE l
/ Shl m Hod sourc. (}
=.
- =
@@ g g
==-
= 8g@g8 e ee _ . ,,. .....
< 8 0@ ,
LEU CORE
/ Shim Rod O
._. @g(@ge=- .
- = @@
= Sg@g,9 eSegS
==-
S O S,
4
- 3. FUELELEMENTDESCRIPTIONS The geometries, materials and fissile loadings of the current HEU fuel elements an '
the replacement LEU fuel elements are described in Table 1. A schematic cross sectio
.is shown in Fig. 2. The external dimensions and structural materials of both element are identical, except that the LEU elements utilize 6061 Alinstead of1100 (or 2S) Al.
The HEU full fuel elements consist of 6 concentric cylinders formed by mechanically joining and positioning 18 curved fuel plates within the grooves of 3 aluminum space webs that are located within an aluminum support cylinder. The LEU full fuel element have essentially the same design as the HEU fuel elements, but the fuel meat contain LEU USSi2-Al fuel instead of HEU U-Al alloy fuel and the method for securing the fue plates in the spacer webs is different. Specifications for the LEU fuel element'geometr were determined by EG&G Idaho using measurements obtained by disassembling a MCZPR dummy fuel element that was shipped to EG&G Idaho by Manhattan College.
The HEU partial fuel element is identical to the HEU full fuel element except that contains 3 fueled plates in cylinder number 2 only. All other plates have been removc<
To provide flexibility for fine adjustment of the actual excess reactivity when the LEU co:
is started up, DOE intends to supply Manhattan College with one LEU partial fui element having removable plates in cylinder numbers 2,4, and 6. Cylinder numbers 1, f and 5 will contain secured ahuninum plates.
Each HEU fuel element has a hold-down rod constructed from a lucite rod an aluminum end fittings. Each rod passes axially through the center of a fuel element an is threaded into the grid plate. The portion of the hold-down rod that passes through th active length of the fuel element is a solid lucite rod one inch in diameter. The same hoh down rods will be used to secure the LEU fuel elements to the grid plate.
Figureo. FuelElement Cross Section v .
" - - - - - - - - - = - _ _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _
g Table 1.- Descriptions of the HEU and LEU Fuel Elements lieu lEll Number of Cylinders / Element 6 6 Number of Plates / Cylinder 3 3 Number of Plates / Element 18 18 Fissile Loading / Cylinder, g 235U Cylinder 1 16.8 19.7 Cylinder 2 24.0 27.4 Cylinder 3 29.4 35.2 Cylinder 4 36.9 43.7 Cylinder 5 43.2 50.6 Cylinder 6 49,7 58.4 Fissile Loading / Element, g 235U 200 235 Fuel Meat Composition U-Al Alloy U3Si2-Al Cladding Material 1100 All 6061 A12 Fuel Meat Dimensions Thickness, mm 0.51 0.51 Width ,mm Variable Variable Length ,mm 610 572-110 Cladding Thickness, mm 0.38 0.38 1 10 ppm natural boron was added to the composition of the cladding and all fuel element structural materials to represent the alloying materials, boron impurity, and other impurities in the 1100 Al of the HEU clements.
2 20 ppm natural boron was added to the compostion of the cladding and structural materials of the LEU elements to represent the alloying materials, boron impurity, and other impurities in 6061 Al.
Aluminum with no boron or other impurities was used in the fuel meat of both the HEU and LEU elements.
O
- 4. CALCULATIONALMODELS 4.1 Nuclear Cross Sections for Diffusion Theory Models Microscopic cross sections in ten energy groups (Table 2) were prepared at 23 C using the EPRI CELL code 12 for the HEU and . LEU fuel element geometries and fissile loadings. The integral transport calculations in EPRI-CELL were performed for 69 fast groups and 35 thermal groups (<1.855 eV), which were then collapsed to ten broad energy groups for use in diffusion theory calculations.
Table 2. Ten Group Energy Group Boundaries Group Upper Lower Group Upper Lower
.H1. Enercy Enercy .NA. Enerev Energy 1 10.0 MeV 0.639 MeV 6 1.166 eV 0.625 eV 2 0.639 MeV 9.119 kev 7 0.625 eV 0.417 eV 3 9.119 kev 5.531 kev 8 0.417 eV 0.146 eV 4 5.531 kev 1.855 eV 9 0.146 eV 0.057 eV 5 1.855 eV 1.166 eV 10 0.057 eV 2.53 x 10-4 eV Figures 3 and 4 and Table 3 show the R-Z unit cell geometry and dunensions that were used in EPRI-CELL to generate microscopic cross sections for the HEU and LEU fuel elements. Each material (lucite, cladding, fuel meat, etc.) was modeled as a separate cylinder in R-Z geometry. Only the inner and outer dimensions of each fuel plate are shown in Table 3. The outer water boundary was chosen to preserve the water volume fraction in the physical hexagonal unit cell of each fuel element. Separate cross sections were prepared for the HEU and LEU partial elements, which have three fuel plates forming cylinder number 2. All cell calculations were done using a fixed buckling of 0.00151 cm-2, which corresponds with the anticipated axial extrapole. tion length of about 10 cm in each fuel element in the reactor diffusion theory calculations.
Each EPRI-CELL case was run three times using the local fine-group spectra over the lucite region, the fuel element region, and the outer water region to collapse the fine group cross sections into 10 broad groups. This procedure was performed because the lucite, fuel element, and water outside each fuel element were modeled as separate regions in the diffusion theory model of the reactor. Cross sections for the water reflector and the fuel element end fittings were calculated using a unit cell model consisting of a pure U-235 fission spectrum on a 10 cm thick slab of water.
7 Figure 3. Monte Carlo and Diffusion Theory Models of the HEU Core.
Fuel Shim Rod
@ Element Emergency [h gx Shutdown Rod 12 23
\ N
' Reg. Rod' 25 Partial Element Monte Carlo Model of HEU Core l
Fuel l 1 Support ~
Cylinder 1 32 I l' l'31 LiiJ Element l
I'
\ Shim Rod rf ~'
i Lucite j Rod
! I22 1 1331 144 I I 55 l
./ Fuel
~
Emergency - :
'......' . Shutdown 1 12 I @
~
l 34I I 45 l l Rod
-CELL Model for i i 1 ration of EEU Reg. Rod :
s Sections I 13 l 1 24 I I 35 l- 1 461 I
1 251 ._ Partial Element Diffusion Theory Model of HEU Core
Figure 4. Monte Carlo and Diffusion Theory Models of the LEU Core.
Fuel 54 ' Element Shim Rod
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,p,~s. r -.
f . 1 ( - h.
Shutdown O~ ~._ '
Rod s
P"#'i"1
/ / 2s ---
\ Element
\
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Monte Carlo Model of LEU Core Al Support -
Fuel
~
/ ' Cylinder I 32 l l 43 l l 541 Elemt.
/ Shim Rod i . I Lucite
- ,/ 1221 l 33 l l 44 ] l 55 l
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/ Fuel l
',', ' ..' Plates Emergency
~
Shutdown I 12 I I 23 I l 34 l' l 43 i Rod EPRI-CELL Model for i %. -1, Generation of LEU Reg. Rod -= U Cross Sections 113 1 B lp l 33 ;
,g, 1 _ Partial Element
_ l I
Diffusion Theory Model of LEU Core I
. _ _ _ __-_-______ - _ ____ - _ -____ _ __ _ _ _ b
9 l
Table 3. Unit Cell Geometry for Fuel Element Cross Sections Inner Outer Diameter. cm Diameter. cm Lucite 0.0 2.54 3 Cylinder 1 3.200 3.454 Cylinder 2 4.064 4.318 Cylinder 3 4.928 5.182 Cylinder 4 5.791 6.045 Cylinder 5 6.655 6.909 Cylinder 6 7.518 7.772 Outer Al Cylinder 8.280 8.915 Outer Water Boundary 8.915 10.309 ,
i 4.2 ReactorModels Reactor calculations were performed in three dimensions using the VIM continuous tergy Monte Carlo codel3-14 and using the DIF3D diffusion theory code 15. The Monte arlo and diffusion theory models are shown in Fig. 3 for the HEU core and in Fig. 4 for
.e LEU core.
The Monte Carlo model contains a detailed geometric description of each fuel ement, the shim rod, the regulating rod, and the emergency shutdown rod. In the VIM de, nuclear cross sections are computed using an ultra-fine group cross secti.on library epared from the ENDF/B IV data tapes. i In diffusion theory, the reactor was modeled in rectangular geometry in order to present the control rods and to provide a heterogeneous representation of the lucite hold
.wn rods, the fuel elements, and the water between fuel elements. A heterogeneous satment of these regions is not possible using the hexagonal geometry options availah!e the DIF3D code. Several choices are possible in modeling a hexagonal fuel rangement in rectangular geometry. The diffusion theory representations shown in gs. 3 and 4 were selected because the computed excess reactivities for the HEU and LEU res were consistent with the corresponding Monte Carlo results.
The cadmium shim rod was modeled in diffusion theory by applying a black undary condition to the four energy groups below 0.625 eV. No boundary condition was i
plied to the upper six energy groups. This methodology 6,8,16 provided excellent reement between calculated and measured reactivity worths of the cadmium shim rods l
the HEU and LEU demonstration, cores of the Oak Ridge Research Reactor.
10
- 5. DYNAMIC DESIGN EVALUATIONS 5.1 Cold Clean Excess Reactivities The 'MCZPR Technical Specifications limit the excess reactivity to a maximum of
' O.44% Ak/k when the temperature of the pool water is 43.7 C (110.6 F). In the reference core condition, the temperature of the pool water is between 16*C and 27 C (60 F - 80 F) and the measured excess reactivity (Ref. 2, Appendix I)is 0.32 - 0.40% Ak/k.
Table 4 shows the excess reactivities that were calculated for the HEU core and the LEU reference core with the shim and regulating rods fully-withdrawn. In the fully-withdrawn condition, the tips of both rods are 0.25 inches above the top of the fuel meat.
The temperature of the pool water was 27'C in the Monte Carlo calculations and 23 C in the diffusion theory calculations.
Tabic 4. Calculated Excess Reactivities, % Ak/k HEU Com LEU Com Monte Carlo 1.2 1 0.4 1.1 i 0.4 Diffusion Theory 1.4 1.4 These Monte Carlo and diffusion theory results are consistent witn each other, but are about 1% Aldk higher than the measured excess reactivity (0.32 - 0.40% Ak/k) of the HEU core in the reference core condition.
To provide flexibility for fine adjustment of the exces.s reactivity at startup, DOE intends to supply Manhattan College with a partial LEU fuel element having removable fuel plates in cylinders 2, 4, and 6. Table 5 shows the calculated changes in excess.
reactivity of the LEU reference core (Case 1, Cylinder 2 fueled only) as a function of the number of fueled cylinders in the partial element.
Table 5. Sensitivity of LEU Cote to the Number of Fueled Cylinders in the Partial Element Fuelin LEU Reactivity Change, Eartial Element % Ak/k fan 1 Cylinder 2 Only 0.0 (Reference Core) 2 No Partial Element -0.33 3 Cylinder 4 Only 0.14 4 Cylinder 6 Only 0.26 5 Cylinders 2 and 4 0.37 6 Cylinders 2 and 6 0.47 7 Cylinders 4 and 6 0.57 8 Cylinders 2,4, and 6 0.72
11 1
The data in Table 5 show that fine adjustments (0.10 - 0.15% Ak/k) of the core excess reactivity can be made by fueling selected cylinders in the partial element if additional excess reactivity is required at startup. Relative to the reference core, addition of fuel plates to cylinders 4 and 6 increases the excess reactivity by about 0.7% Ak/k. Complete removal of the partial element with fuel in cylinder 2 decreases the excess reactivity by about 0.3% Ak/k.
5.2 Sensitivity Calculations for the HEU Core and LEU Reference Core The as-built LEU fuel elements can have 235U loadings in the fuel meat and ppm boron equivalents in the 6061 Al cladding and structural materials that are different from the nominally specified values. The following calculations were performed to determine the sensitivity of the HEU and LEU cores to variations in these parameters. i 5.2.1 235U Loadings Fuel plate manufacturing specifications usually allow a fissile loading variation of 2%. For the MCZPR LEU elements, this means that 235U loadings between 230.3 and 239.7 g would be acceptable. However, the loading variation in as-built fuel elements is normally less than 1 g. The data in Table 6 show that the sensitivity of the LEU reference core to a loading variation ofil gram of 235U per full fuel element is about 0.09% Ak/k.
For comparison purposes, the corresponding value for the HEU core is 0.14% Ak/k.
5.2.2 ppmBoronEquivalents s
6061 Al will be used to manufacture the fuel plate cladding and structural materials of the LEU e!cments. Spectrographic analyses of the alloying materials in recent 6061 Al samples fi eld about 10 ppn natural boron equivalent for the Fe, Cr, Ni, Cu, Si, Mn, etc.
However, the physical natural bcron imputity content is usually specified in spectrographic analyses as <10 ppm because an additional (and expensive) chemical analysis is needed to measure bgcon impurity contents of <10 ppm.
As stated in Table 1, all calculations for the HEU core included 10 ppm natural boron in the compositions of the fuel plate cladding and fuel element structural materials to represent the alloying elements and impurities in 1100 Al. Similarly, all calculations for the LEU core included 20 ppm natural boron equivalent to represent the alloying materials and impurities in 6061 Al. The calculations shown in Table 6 were performed to determine the sensitivity to several assumed total boron equivalents in both the HEU and LEU cores. The results show sensitivities of about 0.18% Ak/k per 5 ppm natural boron equivalent in the HEU core and about 0.32% Ak/k per 10 ppm natural boron equivalent in the LEU core.
l Table G. Sensitivity Calculations for the HEU Core and LEU Reference Core HEU Com LEU Core Grams 235U ppm Boron React. Change, Grams 235 U pm Boron React. Change ner Elem. EquiL %AM per Elem. h %AM Sensitivity to g 235U per Element 200 10 0.00 (Ref.) 235 20 0.00 (Ref.)
l 201 10 0.14 236 20 0,09 Sensitivity to total ppm Nat. Boron Equivalent in Al Structure of Fuel Elements 200 5- + 018 235 10 + 0.32 200 10 0.00 (Ref.) 235 20 0.00 (Ref.)
200 15 -0.18 235 30 0.32 i
5.3 Power Distributions and Power PeakingFactors The power distributions and nuclear power peaking factors that were calculated for the HEU core and the LEU reference core with the shim and regulating rods fully-withdrawn are shown in Fig. 5. The power distributions show the power per fuel element (in milliWatts) and the power peaking factors show the absolute peak power density in each fuel element (computed at the edge of the mesh interval with highest power) divided by the average power density in the core fuel.
The data in Fig. 5 show that the power distributions and total power peaking factors are nearly the same in the HEU and LEU cores. However, the lirniting fuel element in the HEU core is located in grid position 33 and the limiting fuel element in the LEU core is located in grid position 34 because the location of one fuel element was changed in the LEU core to increase the reactivity worth of the regulating rod.
Because of the very low power level at which the MCZPR is operated, these power distributions and power peaking factors have no effect on the thermal-hydraulic safety -
margins. However, the power peaking factors are important input parameters for analysis of the Maximum Hypothetical Accident discussed in Section 6.
13 Figure 5. Power Distributions and Power PeakinD Factors i
HEU CORE l l
Power / Element, 5.4 6.2 5.2 l 32 l l 43 l l 54 l Peak Power in Elemenu 1.91 1.97 1.84 Average Power b Core Fuel , ,
l I I _
6.5 8.8 8.3 5.5 l 22 l l 33 l l 44 l l 55 l 2.13 2.55 2.52 1.85 5.0 8.1 9.2 7.5 l 12 l l 23 l l 34 l I 45 l 1,68 2.46 2.53 2.28
' i I I !
5.1 6.7 6.5 4.2 l 13 l --
l 24 l l 35 ,1 l 46 l 1.80 -
2.13 2.13 1.56 0.8 l 25 l 0.23 LEU CORE Power / Element, _
5.4 6.1 4.9 milllWatts Peak Power in Element / l 32 l @j [ 54 l Average Power in Core Fuel 1.94 1.99 1.80 I l 6.7 0.7 8.0 5.1 M l 33 l l 44 l l 55 l 2.24 2.58 2.52 1.73 5.4 8.5 9.1 6.9 l12l l 23 l l 34 l l 45 l 1.80 2.56 2.59 7.21
' i l l I 5.8 7.1 6.3 l 13 l .-
l 24 l l 35 l 2.02 -
2.20 2.16 4.2 1.0 l 46 l l 25 l 1.61 0.26
l 1 i
l 5.4 Control Rod Worths and Shutdown Margins p The MCZPR Technical Specifications state that the minimum shutdown margin l provided by the control rods shall not be less that 0.46% Ak/k at 43.7 C (110.6 F) with the most reactive rod (the shim rod) stuck out of the core.
5.4.1 Control Rod Descriptions and Calculational Models The reactor is controlled by two Y. shaped control rods which operate in aluminum guide assemblics located between adjacent fuel elements. The shim rod has blades formed by sandwiching a 1/16" layer of cadmium between 1/16" layers of stainless steel.
The blades of the regulating rod are composed of stainless steel which is 3/16" thick. The cadmium poison sections of the shim rod are 24"long.
When fully withdrawn, the tips of the shim rod and the regulating rod are 0.25 in.
above the top of the fuel meat. When fully inserted, the tips of both rods are located just below the bottom of the fuel meat in the adjacent fuel elements.
In the rea'ctor ditTusion theory mohl, the cadmium poison section of each shim rod blade was modeled as one region. The stainless steel cladding was homogenized with the water surrounding each blade. The aluminum volume of the guide assembly was modeled as three separate aluminum blocks located between fuel elements (See Figs. 3 and 4). The stainless steel regulating rod was modeled in a similar manner, except that each blade was modeled as a single region.
5.4.2 Methods for Calculating Control Rod Worths Normal diffusion theory was used to calculate the reactivity worth of the regulating rod because stainless steelis not a highly absorbing material. Howevet , normal diffusion j theory is not valid for the highly absorbing cadmium poison conta:ned in the shim rod.
The cadmium blades of the shim rod were modeled in diffusion theory by applying a black boundary condition to the four energy groups below 0.625 eV. No boundary condition was applied to the upper six energy groups. This methodology 6,8,16 provided i excellen agreement between calculated and measured reactivity worths of the cadmium shim rods in the HEU and LEU demonstration cores of the Oak Ridge Research Reacto:-
In the Monte Carlo model, the shim and regulating rods were represented in explicit detail. The aluminum volume .of the guide assemblies was modeled as blocks of aluminum located between the fuel elements.
I l
15 5.4.3 Reactivity Worths of Contol Rods The reactivity worths of the regulating rod and the shim rod of the HEU and LEU reference cores that were calculated with diffusion theory and Monte Carlo methods are shown in Table 7.
Table 7. Calculated Reactivity Worths of Regulating Rod and Shim Rod HEU Core LEU Core Diffusion Monte Diffusion Monte
.C_antrol Rod Configuration Theory Carlo Theory Carb Regulating Rod Fully-Inserted Shim Rod Fully-Inserted (% AM) - 4.7 -4.310.6 - 4.7 - 3.9 i 0.5 Regulating Rod Fully-Withdrawn Shim Rod Fully-Inserted (% AM) - 3.5 -3.4 Regulating Rod Fully-Inserted Shim Rod Fully-Withdrawn (% AM) - 1.2 - 1.3 The Technical Specifications for the HEU core state that the regulating rod has a negative worth of 0.9% AM and that the shim rod has a negative worth of 2.5% AM.
However, the Supplement to the MCZPR Hazards Summary Report l7 states in Table III that:(1) the worth of the regulating rod is - 0.9% AM, (2) the maximum worth of the cadmium control rod is - 3.4% AM, and (3) the minimum worth of the cadmium control l
rod is - 2.5% AM. These values may have been measured by AMF Atomics for a critical
- asu mbly of the PTR reactor at IRL in Plainsboro, NJ. We also note that AMF Atomics estimated 18t hat the vorf.h of the s dm rod in the then-proposed MCZPR core may be in the range - 2.5% Ak/k to - 4.5% AM and that the worth of the regulating rod would be
, approximately the same (- 0.9% AM) as that measured in a PTR critical assembly.
l We conclude that:(1) the magnitude of the calculated worth of the regulating rod in the HEU core is probably high by about 0.3% AM (1.2% - 0.9% AM ) , (2) a conservative negative reactivity worth of - 2.5% AM for the MCZPR HEU shim rod was probably adopted from measurements made in the PTR reactor, and (3) the actual worth of the l MCZPR shim rod in the HEU core is probably more negative than - 2.5% AM.
r
{
l w_--------------- - - -
5.4.4 ShutdownMargin The MCZPR Technical Specifications limit the excess reactivity of the core to a maximum of 0.44% Ak/k when the temperature of the pool water is 43.7'O (110.6 F). The minimum shutdown margin with the shim rod stuck out of the core is - 0.46% Ak/k, based on a negative reactivity worth of 0.9% Ak/k for the regulating rod.
For the LEU core, we estimate that the minimum shutdown margin with the shim rod stuck out of the core will be about - 0.56% Ak/k (1.3 - 0.3 - 0.44), assuming that the calculated worth of the regulating rod is high by 0.3% Ak/k and that maximum excess ,
reactivity is limited to 0.44% Ak/k. '
5.4.5 Emergency Shutdown Rod As mentioned in Section 2, an aluminum-clad emergency shutdown rod containing .
B4C is mounted on a wall near the reactor core. If both the shim rod and the regulating rod are disabled in the out position, the reactor can be shutdown by manually inserting the emergency' shutdown rod into the core. There has never been a need to use this rod.
Utilization of this emergency shutdown rod was approved by the NRC in the late-1960's in lieu of adding boric acid to the pool water in the event that it is needed.
l The emergency shutdown rod 19 has an cuter diameter of 15/16", an aluminum wall thickness of1/16", and a length between plugs of about 11 feet (about 334 cm). It contains I
1.22 lb (553 g) of B4C powder. According to MCZPR procedures, the most efTective position for insertion of the rod (ifit were needed)is in the gap between core positions 23,33, and 34 (See Figs. 3 and 4). The negative reactivity worth of the fully-inserted rod was measured 20 to be not less than 3% Ak/h.
The reactivity worth of the emergency shutdown rod located in the gap between core positious 23,33, and 34 was calculated using the Monte Carlo models of the HEU and LEU cores. In the model, full density B C 4 (2.52 g/cm3) was assumed to fill the rod over the 24" active length of the adjacent fuel elements. Tha shim rod and the regulating rnd were located in their fully-withdrawn positions with the ends of the rods 0.25" above the top of !
the fuel meat. The calculated negative reactivity worths wara (3.1 i 0,5)% sk/k in the HEU core and (3.6 i 0.5)% Ak/k in the LEU reference core.
17 5.5 ReactorKineticsParameters The prompt neutron generation times and effective delayed neutron fractions of the HEU and LEU reference cores were calculated using standard perturbation theory techniques in the PERT 2D code.21 Axial extrapolation lengths were first determined using fluxes from the 3D reactor calculations. A two-dimensional reactor model was then used to compute the real and adjoint flux distributions needed for the perturbation calculations. The results are shown in Table 8.
Table 8. Prompt Neutron Generation Times and Effective Delayed Neutron Fractions Parameter HEU Core LEU Com Prompt Neutron Generation Time, s 65 59 Effective Delayed Neutron Fraction 0.0078 0.0078 5.6 Temperature, Void, and Doppler Coefficients Non-isothermal and isothermal reactivity feedback coefficients as functions of temperature and void fraction were computed for the HEU and LEU cores using difrusion theory and 3D reactor models for each of three physical effects: (1) the hardening of the neutron spectrum due to increasing the water temperature only, (2) the increase in .
neutron leakage due to decreasing the water density only, and (3) the increase in absorption of the 238U epithermal resonances dee to increasing the temperature of the fuel meat only (Doppler Effect). Calculations were performed for non-isothermal conditions because different regions of the reactor would heat at different rates during an unplanned power increase.
5.6.1 Non-Isothermal Reactivity Changes with Temperature For these calculations, the reactor was div4 ded into three regions: ('1) the 15 fuel l clements and the partial fuel element (including all water inside the aluminun support i cylinders),(2) the water between fuel elements, and (3) the reflector. On the outer edges of the core, a water channel thickness equal to one-half of the water channel thickness between fuel elements was included as part of the inter-element water. The remaining water in the pool is referred to as the reflector. During an unplanned ' power increase, the fuel element and water inside the aluminum support cylinder would be heated first, followed by heating of the water between fuel elements, and lastly, heating of the reflector i water. Specific heating rates in the three regions would depend on the time constants of l
)
I
- ---- -- - - - - -- _ u
18 the power increase.
The calculations were performed by separately changing the water temperature (four values), the water density (three values), and the fuel temperature (three values) in each region while holding the materials in the other two regions at 23 C. Least-squares fits were then done to obtain reactivity values at intermediate temperatures. Slopes of the reactivity feedback components between 20 C and 30 C are shown in Table 9 along with the void coefficient for a 1% change in only the fuel element water density. The detailed reactivity changes relative to 20 C are tabulated in Table 10 and are plotted in Fig. 6.
Table 9. Non-Isothermal Temperature, Void, and Doppler Coefficients With Shim and Regulnting Rods Fully-Withdrawn Aldk x 10-4 per C (20-30 C)
Reactor Recion/Effect HEU Core LEU Core FuelElements Water Temperature Only - 1.9 - 1.6 Water Density Only, - 0.4 - 0.4 Fuel Temperature Only .QQ -0.1
- 2.3 - 2.1 Inter-Element Water Water Temperature Only + 1.1 + 1.0 Water Density Only, _Q1 - 0.1
+ 1.0 + 0.9 ReflectorWater Water Temperature Only + 0.9 + 0.9 Water Density Only, - 0.1 _Ql
+ 0.8 + 0.8 Void Coefficient (0-1% Void), - 1.5 - 1.6 Aldk x 10-3 Fuel Element Water Only per % void 1 !
The reactivity feedback coefficient is negative for the fuel elements in both the HEU and LEU cores, but is slightly larger for the HEU core. The fuel f emperature component has, been summed with the water temperature and density ccaponents in Table 9.
However, the Doppler coefficient of the LEU core actually has a larger weight becar se the fuel temperature normally increases more rapidly than the water temperature. The net reactivity feedback coefficients for the inter-element water and for the reflector are positive. The main reason is that the hydrogen absorption cross sections become smaller as the neutron spectrum hardens with increasing temperature. The fuel-element void coefficient is slightly more negative in the LEU core than in the HEU core because the LEU core has a harder neutron spectrum and slightly more leakage. ,
9 19 Table 10. Calculated Non-Isothermal and Isothermal Reactivity Changes with Temperature.
(Relative to 20'C)
HEU CORE - NON ISOTHERMAL REACTIVITY CHANGES (% Ak/k)
Sum:
Fuel Element Inter-Element Water Reflector Water F. E.
Fuel F. E. Water Water, l. E. Water Water, Refl. + 1. E.
Temp. Water Water,
'C Temp. Density Temp. Sum Temp. Density Sum Temp. Density Sum + Refl.
0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 20 0.0 30 -0.186 -0.039 0.0 -0.226 0.107 -0.008 0.099 0.092 0.006 0.087 -0.040
-0.459 0.210 -0.017 0.193 0.182 -0.012 0.170 -0.097 40 -0.372 -0.088 0.0 50 -0.557 -0.145 0.0 -0.701 0.310 -0.028 0.282 0.270 -0.020 0.250 0.170 60 -0.741 -0.211 0.0 -0.952 0.406 -0.041 0.365 0.357 0.030 0.327 -0.260 70 -0.925 -0.287 0.0 -1.212 0.498 -0.055 0.443 0.442 -0.040 0.402 -0.367 80 -1.110 -0.371 0.0 -1.481 0.586 -0.071 0.515 0.526 -0.052 0.474 -0.493 90 -1.295 -0.465 0.0 -1.760 0.670 -0.089 0.581 0.608 -0.065 0.543 -0.636 100 1.482 -0.567 0.0 -2.049 0.750 -0.108 0.642 0.689 -0.080 0.610 -0.798 l
LEU CORE - NON ISOTHERMAL REACTIVITY CHANGES (% Ak/k)
Sum:
Fuel Element Inter-Element Water Reflector Water F. E.
Fuel F.E. Water Water, l. E. Water Water, Refl. + 1. E.
Temp. Water Water, O Temp. Density Temp. Sum Temp. Density Sum Temp. Density Sum + Refl.
0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 20 0.0 0.0 0.013 -0.213 0.098 -0.011 0.087 0.089 -0.006 0.083 -0.043 30 -0.155 -0.044 40 -0.310 -0.098 -0.026 -0.434 0.193 -0.024 0.169 0.175 -0.013 0.162 -0.104 50 -0.464 -0.162 -0.039 -0.665 0.284 -0.039 0.245 0.260 0.022 0.238 -0.182 60 -0.617 -0.236 -0.052 -0.905 0.372 -0.057 0.315 0.344 '0.032 0.312
- -0.279 70 -0.771 -0.319 -0.065 -1.155 0.456 -0.077 0.379 0.426 -0.044 0.382 -0295 80 -0.925 -0.413 -0.078 1.416 0.536 0.100 0.437 0.507 -0.056 0.450 -0.529 90 -1.080 -0.516 -0.090 -1.686 0.614 -0.12.5 0.489 0.586 -0.071 0.516 -0.681 100 -1.235 -0.630 -0.103 -1.967 0.681 -0.152 0.529 0.664 -0.086 0.578 -0.860 ISOTHERMAL REACTIVITY CHANGES
(% ok/k)
Measured Calculated Temp. HEU HEU LEU
'C Core Core Core 20 0.0 0.0 0.0 25 0.031 -0.035 -0.038 30 0.055 -0.070 -0.076 35 0.071 -0.107 -0.116 40 0.081 -0.144 -0.156 43.7 0.083 -0.171 -0.186 45 0.082 -0.181 0.197 50 0.077 -0.220 -0.239 55 0.064 -0.259 -0.281 60 -0.298 -0.324
. Figure 6. Non-Isothermal Reactivity Changes with Temperature.
Reactivity Change Components HEU Core LEU Core e "
d
O l.E. Water Temp. .-
l.E. Water Temp. ,,-
. \, . ' , , , . -. . *
- f' - -s,\.3: -
w -
\
"'~',, il. T emp. ~ M ;r$'# Rett. T(emp. _
Re d -
-$.'. llll y," o _
. ;;:n.y.
..e.$
- fff ~ s - Refl. Densliy 69 4
' p .,v$,, Refl. Dennity so q ,,,***,,v' Fuel Temp.-
gd ,,,
MO ~ ' '
of ------------.........,,,,,,,!s,r,....
1.E. Wa Density . c) ,,,;;;,;g,- p[. . .
O) O 1.E. Wat r Density C
j#d -
O *. ~
~
L O ~
oi y
/
F.E. Water Density o
g F.E. Water Density
'E m
- .D 9 y6 oI Do - ~
O o8 O /
Z E /
F.E, Water Temp.
N F.E. Water Temp. ~
y ~
\
f f I f f f i i 20.0 40.0 60.0 80.0 10 0.0 20.0 40.0 60.0 80.0 10 0.0 Temperature, degrees C Temperature, degrees C Region Sums for HEU and LEU Cores e
d F--HEU inter Element Water
. :, HEU Reflector 3 '"E LEU Rollector o -
.ggM, ...j;nti'd". -' LEU Inter. Element Water o .,,***p.r.'
M cr ,e N O) d
^
C S O
.C U9 h7
~
~ "
\
.T s OU .
~
0 l W
fo C! -LEU Fuel Elements y ' ' ' -HEU Fuel Elements 20.0 40.0 60.0 E0.0 10 0.0 Temperature, degrees C
21 5.6.2 IsothermalReactivity Changes withTemperature In the reference condition, the pool water temperature of the MCZPR is 16 - 27 C (60 -
80 F) and the excess reactivity with the control rods fully-withdrawn is 0.32-0.40% ak/k.
In 1966, Manhattan College measured (Ref. 2, Appendix I) a positive reactivity change in the MCZPR when all of the water in the reactor tank was heated isothermally. In this experiment, the excess reactivity increased from 0.32% Ak/k at a pool water temperature of 15.6 C (60 F) to a maximum value of 0.44% Ak/k at a pool water temperature of 43.7 C (110.6 F) and decreased at temperatures above 43.7 C. Based on this experiment, the MCZPR Technical Specifications limit the core excess reactivity to a maximum of 0.44%
Ak/k at 43.7 C.
Table 10 provides a comparison of measured 2 and calculated reactivity changes relative to 20 C as a function of temperature for isothermal heating of all materials inside the reactor tank. The measured reactivity changes for the HEU core are positive, while the calculated reactivity changes for both the HEU and LEU cores are negative.
' Examination of the non-isothermal reactivity change components in Table 10 shows that the contributions to isothermal reactivity changes involve the sum of positive and negative contributions that result in a small net value. Additional investigation of the calculational models (to attempt bringing the measured and calculated data into better agreement) was judged to be inappropriate because possible accidents in which the reactor pool would heat isothermally are considered to have an extremely low probability.
[The sums of the non-isothermal reactivity change components as a function of temperature for the fuel elements, inter-element water, and reficctor water (Sum F.E. +
I.E. + Reft) in Table 10 are slightly different from the computed isothermal reactivity changes because interaction among the three regions was not accounted for in the non-1 isothermal calculational procedure (material temperatures were increased in each region, independent of the other two regions). Better agreement between the results for the sums of the non-isothermal components and the isothermal data could have been obtained by using a calculational procedure in which the temperatures of the materials in all three regions were increased incre nentally.]
l In our judgment, there are no significant safety issues releted to the pesitive j isotherma! temperature coefficient up to 43.7 C in the HEU core of the MCZPR. Based on the similarity of the calculated non isothermal and isothermal reactivity change data shown in Table 10, we eucct that the LEU core wili also exhibit a positive iscthermal temperature coefficient over approximately the same limited range as in the HEU core.
]
In our judgment, there are also no significant safety issues related to the anticipated positive isothermal temperature coefficient over a limited temperature range in the LEU ,
core of the MCZPR.
l i
L_____-__________-______________ _
l
J 22
]
- 6. ACCIDENT ANALYSIS Several accident scenarios were evaluated by Manhattan College in its Safety ,
Analysis Report3and by the Nuclear Regulatory Commission (NRC) staff in February 1985 as part its review 4 of an application by Manhattan College for a renewed operating license. The accidents that were considered included: (1) inadvertent excess reactivity insertion (nuclear excursion), (2) natural phenomena, and (3) mechanical rearrangement of fuel.
Of these, only one scenario (inadvertent excess reactivity insertion) could be affected by changing the core fuel from HEU to LEU, and only this scenario is addressed here. An excess reactivity insertion is designated as ns maximum hypothetical accident (MHA)in the MCZPR.
l 6.1 Inadvertent Execss Reactivity Insertion ANL has ev'aluated the consequences of an inadvertent stepwise reactivity insertion of 0.44% Ak/k in the HEU core and the LEU reference core of the MCZPR using the PARET code.22 This excess reactivity is the maximum value allowed in the MCZPR Technical Specifications.
6.1.1 Comparison of Calculations with SPERT-I Experiments The PARET code was originally developed at the Idaho National Engineering Laboratory for analysis of the SPERT-III experiments, which included both pin-type and plate type cores and pressures and temperatures in the range typical of power reactors.
The code was modified by the RERTR Program at ANL to include a selection of flow instability, depar!ure from nucleate boiling, single- and two-phase heat transfer correlations and a properties library applicable to the low pressures, temperatures, and flow rates encountered in reseerch reactors.
2 To validate the PARET cele, calculated and measured data were compared 3 for three SPERT-I HEU cores 24,25: B-24/32 (32 element core with 24 fuel plates per element),
B-12/64, and D-12/25. These cores were similar in design to many plate-type research reactors in current operation. The tests performed in the D-12/25 core included both '
nondestructive and destructive transients. The results of these analyses are in good agreement with the measured data and validate the PARET code for use in calculating research reactor transients.
23 6.1.2 Armlysis for the MCZPR The same model and methods that were used for analysis of the SPERT-I HEU cores were also used to analyse the HEU and LEU reference cores of the MCZPR.
Inputs to the code included the prompt neutron generation time, effective delayed neutron fraction, temperature coefficients of reactivity, and axial power distributions. To be conservative, the temperature coefficients of reactivity included both the negative contribution from the fuel elements and the positive contribution from the inter-element water. Axial power distributions for the average channel of the HEU and the LEU core were represented by chopped cosine shapes having peak-to-average power densities of 1.32. This value corresponds with calculated core-average extrapolation lengths of about 7.7 cm in both cores. In the hot channel, these axial shapes were scaled to produce peak to core-average power densities of 2.55 and 2.59 (see Fig. 5) at the centerlines of the HEU and the LEU cores, respectively.
Calculations were performed for a stepwise reactivity insertion of 0.44% Ak/k with the pool water at a temperature of 43.7 C, the reactor at a power of 0.1 W and no reactor scram. The results of these calculations are shown in Table 11.
Table 11. Results of Step Reactivity Insertion Transient in HEU and LEU Cores Parameter HEU Com IEUCom Step Reactivity Insertion, % Ak/k 0.44 0.44 Asymptotic Period, s 3.8 3.0 Peak Power, kW 221 183 Peak Fuel Centerline Temperature, *C 116 115 Peak Surface Cladding Temperature, C H6 115 Time to Peak Power, s 58.5 58.0 Time to PerJc Fuel Temperature, s 66 69 Time to Peak Cladding Temperature, s 66 8 Because the peak temperatures in the cladding are far below the solidus temperature of 660 C in the 1100 Al cladding of the HEU core and far below the solidus temperature of 582 C in the 6061 Al cladding of the LEU core, no damage to the fuel and no release of fission products is expected.
24 REFERENCES
- 1. U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Non-Power Reactors", NUREG-1313, July 1988.
- 2. Safety Analysis Report for the Manhattan College Zero Power Reactor Submitted to the United States Nuclear Regulatory Commission for Renewal of Facility License R-94, August 1983.
- 3. Appendix A to Facility License No R-94, Technical Specifications for the Manhattan College Zero Power Reactor (Rev. 4), March 15,1985.
- 4. U.S. Nuclear Regulatory Commission, " Safety Evaluation Report Related to the Renewal of the Operating License for the Research Reactor at Manhattan College, NUREG-1098, February 1985. '
- 5. M. M. Bretscher and J. L. Snelgrove, " Comparison of Calculated Quantities with Measured Quantities for the LEU-Fueled Ford Nuclear Reactor," Proc. International Meeting on Research and Test Reactor Core Conversions from HEU to LEU Fuel, Argonne National Laboratory, Argonne, IL, November 8-10, 1982, ANI/RERTR/TM-4, CONF-821155, pp. 397-425 (1983).
- 6. M. M. Bretscher, " Analytical Support for the Whole-Core Demonstration at the ORR,"
Proc.1986 International Meeting on Reduced Enrichment for Research and Test Reactors, Gatlinburg, TN, November 3-6,1986, ANI/RERTR/TM-9 , CONF-861185, pp.
287-301(1988).
i
- 7. R. J. Cornella and M. M. Bretscher, " Comparison of Calculated and Experimental Wire Activations," Proc.1986 International Meeting on Reduced Enrichment for i Research and Test Reactors, Gatlinburg, TN, November 3 6,1986, ANIAERTRfrM-9, ,
CONF-861185, pp. 302-309 (1988).
f i
- 8. M. M. Bretscher, " Evaluation of Differential Shim Rod Worth Measurements in the !
Oak Ridge Research Reactor," Proceedings of the 1987 International Meeting on Reduced Enrichment for Research and Test Reactors, Buenos Aires, Argentina, !
September 28 - October 2,1987 (to be published). I
- 9. M. M. Bretscher, J. J. Snelgrove, and R. W. Hobbs, "The ORR Whole-Core LEU Fuel Demonstration", Trans. Am. Nucl. Soc. .Qfi, 579-581 (1988). *
)
4
_ _ _ _ _ _ - _ . - - _ - . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ ~ ~ ~ - - ~ ~ - ~ ~ ~ ~ ~
25
- 10. IAEA Guidebook on Research Reactor Core Conversion from the Use of Highly Enriched Uranium to the Use of Low Enriched Uranium Fuels, IAEA-TECDOC-233, Appendix F, pp. 443-628 (1980).
- 11. J. E. Matos, E. M. Pennington, K. E. Freese, and W. L. Woodruff, " Safety-Related Benchmark Calculations for MTR-Type Reactors with HEU, MEU, and LEU Fuels,"
paper included in IAEA Safety and Licensing Guidebook on Research Reactor Core Conversions from HEU to LEU Fuel, Volume 2, Analytical Verification, Draft #7, June 1985.
- 12. B.A. Zolotar et al., "EPRI-CELL Code Description," Advanced Recycle Methodology Program System Documentation, Part II, Chapter 5 (Oct.1975).
- 13. E. M. Gelbard and R. E. Prael, " Monte Carlo Work at Argonne National Laboratory," ;
in Proc. NEACRP Mtg. Monte Carlo Study Group, July 1 3, 1974, Argonne, Illinois, ANL-75-2 (NEA-CRP-L-118), Argonne National Laboratory , p. 201(1975).
- 14. R. Blomquist, " VIM - A Continuous Energy Neutronics and Photon Transport Code",
Proc. Topl. Mtg. Advances in Reactor Computations, Salt Lake City, Utah, March 28-31,1983, p. 222, American Nuclear Society (1983).
- 15. K.L. Derstine, "DIF3D: A Code to Solve One , Two , and Three-Dimensional Finite-Difference Diffusion Theory Problems," ANL-82-64, April 1984.
- 16. M. M. Bretscher, " Blackness Coefficients, Effective Diffusion Parameters, and Control Rod Worths for Thermal Reactors," ANI/RERTR/TM-5, Sept.1984.
- 17. Supplement to Hazards Summary Report - Manhattan College Low Power Critical Reactor, " Replies to the Atomic Energy Commission on Additional Information Requested for the Manhattan College Low Power Critical Reactor", undated, but probably around 1962.
- 18. Letter from 1.. Crevoiserat, Project Manager, AMF Atomics, Greenwich, CT, to Brother Conrad Gabriel, Manhattan College, November 15,1962.
- 19. MCZPR Maintenance Logbook, October 17,1967.
20." Emergency Shutdown Rod Experiments", Memorandum from Paoshu Ko, Chief Reactor Supervisor, to MCZPR Reactor Hazards Committee, November 3,1967.
- 21. T. A. Daly, et al., "The ARC System Two-Dimensional Adjunct Calculations," ANL-7720 (Oct.1972). 1 I
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ - _ _ _ _ _ _ _-_-__-_-__ _ D
l 26 22 C. F. Obenchain, "PARET - A Program for the Analysis of Reactor Transients," IDO-17282, Idaho National Engineering Laboratory (1969).
- 23. W. L. Woodruff, "A Kinetics and Thermal-Hydraulics Capability for the Analysis of Research Reactors," Nuclear Technology SA, pp. 196-206, (Feb.1984). And W. L.
Woodruff, "The PARET Code and the Analysis of SPERT I Transients," Proc.
International Meeting on Research and Test Reactor Core Conversions from HEU to LEU Fuel, Argonne National Laboratory, Argonne, IL, November 8-10, 1982, ANURERTR/fM-4, CONF-821155, pp. 560-578 (1983).
- 24. A. P. Wing, " Transient Tests of the Fully Enriched, Aluminum Plate-Type, B Cores in the SPERT I Reactor," IDD-16964, Idaho National Engineering Laboratory (1964).
l 25. M. R. Zeissler, "Non-Destuctive and Destructive Transient Tests of the SPERT I-D, Fully Enriched, Aluminum Plate-Type Cores: Data Summary Report," IDO-16886, Idaho National Engineering Laboratory (1963).
/
l
_ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _