ML17256A526
| ML17256A526 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 11/30/1982 |
| From: | Robbins T PLG, INC. (FORMERLY PICKARD, LOWE & GARRICK, INC.) |
| To: | |
| Shared Package | |
| ML17256A520 | List: |
| References | |
| NUDOCS 8303010392 | |
| Download: ML17256A526 (38) | |
Text
for Rochester Gas
& Electric Corporation Ginna Plant CRITICALITY ANALYSIS FOR THE SPENT FUEL STORAGE RACKS by Thomas R. Robbins
- Pickard, Lowe and Garrick, -Inc.
November 1982 83030i0392 830223 PDR ADOCK 05000244 P
- PDR,
T
TABLE OF CONTENTS 1: 0 INTRODUCTION 2.0 ANALYTICALTECHNIQUE 3.0 EVALUATION OF CRITICALITY SAFETY 4.0 TOLERANCES AND UNCERTAINTIES
~
- 5. 0 ACCIDENT ANALYSIS
- 6. 0 ANALYSIS CONSERVATISMS REFERENCES
l
LlST OF TABLES TABLE l Summary of LEOPARD Results for Measured Criticals TABLE 2 Westinghouse UO2 Zr-4 Clad Cylindrical Core Critical Experiments TABLE 3 Battelle Criticals TABLE 4 Fuel Assembly Characteristics TABLE 5 Summary.of Perturbations to the Multiplication Factor of the Basic Cell
LiST OF FIGURES FIGURE l RGE Rack Reference Design FIGURE 2
Ginna Spent Fuel Racks Neutron Multiplication Factor vs Fuel Assembly Initial Enrichment Ginna Spent Fuel Racks Heutron Multiplication Factor vs Water Denstiy in Pool (Constant Temperature of 68'F)
I PDQ-07 Accident Geometry FIGURE 5
FIGURE 3
Ginna Spent Fuel Racks Neutron Multiplication Factor vs Temperature in Pool FIGURE 4
CRITICALITY ANALYSIS FOR THE SPENT FUEL STORAGE RACKS
1.0 INTRODUCTION
The following discussion summarizes the evaluation of the spent fuel racks with respect to criticality safety.
The analytical techniques described here are the same as those used to license the existing spent fuel racks for Ginna.
- 2. 0 ANALYTICALTECHNIQUE The LEOPARD computer program was used to generate macro-(1) scopic cross sections for input to four energy group diffusion theory calculations which, are performed with the PDQ-7.(2) program.
LEOPARD calculates the neutron energy spectrum over the entire energy range from thermal up to 10 Hev and determines averaged cross sections over appropriate energy groups.
The fundamental methods used in the LEOPARD program are those used in the HUFT and SOFOCATE programs which were developed (3)
(4) under the Naval Reactor Program and thus are well founded and extensively tested techniques.
In addition, Westinghouse Electric Corporation, the developers of the original LEOPARD
- program, demonstrated the accuracy of these methods by exten-sive analysis of measured critical assemblies consisting of slightly enriched UO fuel rods.
2 In addition, Pickard, Lowe and Garrick, made a number of improvements to the LEOPARD increase its accuracy for the calculation of
'n systems which contain significant amounts mixed with UO2.
PLG has tested the accuracy inc.
(PLG) has program to reactivities of plutonium of these modifi-
modifications by analyzing a series of UO and PuO
-UO2
'critical experiments.
These benchmarking ana3.yses not only demonstrate the improvements obtained for the analysis of Pu02-U02 systems but also demonstrate that these modifications have not adversely affected the accuracy of the PLG-modified LEOPARD program for calculations of slightly enriched U02 systems.
The UO2 critical experiments chosen for benchmarking include variations in H20/UO2 volume ratios, U-235 enrichments, pe3.3.et diameters and cladding materials.
Although the LEOPARD model also accurately calculates the reactivity effects of soluble boron, these experiments have not been included in the LEOPARD benchmarking criticals since the spent fuel pool cal-culations do not involve soluble boron.
Neutron leakage was represented by using measured buckling input to infinite lattice LEOPARD calculations to represent the critical assembly.
A summary of the results is shown in Table 1
for the 27 measured criticals chosen as being directly applicable for benchmarking the LEOPARD model for generating group average cross sections for spent fuel rack criticality calculations.
The average, calculated k ff is 0.9979 and the standard deviation eff from this average is 0.0080 hk.
Reference 5 raised questions concerning the accuracy of the measured buckling reported for the experiments number 12 through 19.
If these 'data are
- excluded, the average calculated k ff for the remaining 19 eff experiments is 1.0006 with a standard deviation from this value of 0.0063 hk.
In all of these experiments, there are significant uncertainties in the measured bucklings which are necessary inputs to the LEOPARD analysis.
These uncertainties are the same order of magnitude as the indicated errors in
the LEOPARD results, and therefore a more definitive set of experimental data is used to establish the accuracy of the combined LEOPARD/PDQ-7 model used for the criticality analysis of the spent fuel racks.
The PDQ series of programs have been extensively developed and tested over a period of 20 years, and the current version, PDQ-7, is an accurate and reliable model for calculating the subcritical margin of the proposed spent fuel rack arrangement.
This code or a mathematically equivalent method is used by all
- the U.S. suppliers of'ight water reactor cores and reload fuel.
In addition, this code has received extensive utilization'n the U.S. Naval Reactor Program.
As a specific demonstration of the accuracy of the calcu-lational model used for the spent fuel rack calculations, the combined LEOPARD/PDQ-7 model has been used to calculate fourteen measured just critical assemblies.
The criticals are high neutron leakage systems with a large variation in 'H20/U20 volume ratio and include parameters in the same range as those applicable to the spent fuel rack design.
Experiments including soluble boron are included in this demonstration since the ability of PDQ-7 to calculate neutron leakage effects is of primary interest.
The use of soluble boron allows changes in the neutron leakage of the assembly while maintaining a uniform lattice and thus
, allows a better test of the accuracy of the model.
Furthermore, it eliminates the error associated with the measured bucklings, which is inherent in the LEOPARD benchmarks, thus permitting determinations of the actual calculational uncertainty which much be accounted for in 0he spent fuel rack criticality analysis.
These combination LEOPARD/PDQ-7 calculations result in a calculated average k ff of 0.,9928 with a standard deviation eff about this value of 0.0012 hk.
These results, as shown in Table 2 demonstrate that the proposed LEOPARD/PDQ-7 calculational model can calculate the reactivity of the proposed spent fuel rack arrangements with an accuracy of better than 0.010 5k at the 95 percent confidence level.
In addition to the fourteen critical assemblies in Table 2, the LEOPARD/PDQ model was used to calculate-the k ff for seven eff additional critical assemblies, two of which incorporated thin stainless steel plates in an array which is geometrically similar to a section of the spent fuel racks.
These seven criticals were performed by Battklle:Pacific Northwest Laboratories specifically for the purpose of pro-viding benchmark critical experiments in support of spent fuel criticality analysis.
They are described in detail in Reference 17.
The results of these critical experiments are summarized in Table 3.
The overall average k ff calculated for eff these seven just critical assemblies was 0.9933, with a standard deviation around this value of 0.0013 hk.
Combining the results of benchmarking against both the Nestinghouse (Table 2) and the Battelle (Table
- 3) critical experiments results in a mean calculated k ff for 21 experi-ments of 0.9929 with a standard deviation of.0012 as shown in Table 3.
- Thus, the final bias to be applied to the combined LEOPARD/PDQ-7 model is +.0071, and the 2a uncertainty to be applied corresponding to a 95 percent probability at a
95 percent confidence level is.0024.
As a result of this.approach to separately benchmark both the cross sections and the diffusion theory calculations against applicable critical assemblies, the "transport theory correction factor" is implicitly included in the derived calculational uncertainty factor.
3.0 EVALUATION OF CRXTlCALXTY SAFETY The PDQ-7 program is used in the final predictions of the'ultiplication factor of the spent fuel storage pool.
The calculations are performed in four energy groups and take into account all the significant geometric details of the fuel
- bundles, fuel boxes and major structural components.
The geometry used for most of the calculations is a basic cell representing one quarter of the area of a repeating array of two iden"ical stainless steel boxes.
The specific geometry and dimensions of this basic cell are shown in Figure l, and the fuel assembly characteristics are listed in Table 4.
The calculational approach-is to use the basic cell to calculate the reactivity of an infinite array of uniform spent.
fuel racks and to account for.any deviations of the 'actual spent fuel rack array from this assumed infinite array as perturbations on the calculated reactivity of the basic cell.
The effects of
~ mechanical tolerances are also treated as perturbations on the calculated reactivity of the basic cell.
The fuel assemblies were assumed to be unirradiated with a U-235 enrichment of 4.25 w/o in the enriched center section which is higher than any anticipated reload enrichment for the Ginna core.
This corresponds to a U-235 loading of 4l.9 gm U-235 per axial cm of fuel assembly.
Host of the calculations were performed at a uniform pool temperature of 68'F, but the reactivity effects
.of pool temperature are also taken into account as a pertur-bation on the basic cell calculations.
The reference basic cell calculation is performed with the minimum dimension on all the stainless steel boxes which results in a k
= 0.9305.
Other tolerances on.the geometric array representing the racks are treated as perturbations on this reference basic cell calculation.
The calculated variation of the basic cell k as a function of the initial enrichment of the fuel assembly is shown in Figure 2.
Most of the calculations with the basic cell geometry utilized a 40x20 two-dimensional array of mesh points.
To test I
the adequacy of this mesh description, a calculation was run with a 80x40 mesh size and the resulting k was
.9296.
- Thus, the perturbation on the basic cell due to mesh spacing effects is
.0009 hk.
Based on the, results of 'the benchmarking of the combined LEOPARD/PDQ-7 analysis
- model, the bias in the calculated multiplication factor compared to measured just critical arrays is.007l, and this bias must be added to the calculated basic cell reactivity.
The k of the basic cell as a function of temperature is shown in Figure 3.
With a maximum pool temperature of 200 F, under the worst possible conditions the k is.9383, which results'in a perturbation due to temperature effects of
+0.0078 hk.
Although the overall steady state reactivity temperature coefficient of the spent fuel pool is positive, the temperature coefficient of the fuel assemblies is negative.
a
I
~
~
The sensitivity of the spent fuel storage rack multi-plication factor to the simultaneous and uniform variation of water density in both the fuel box and water box is illustrated in Figure 4.
Simultaneous variation of water density in both boxes is conservative and unrealistic since the water density in the fuel box will always be less than or equal to the water density in the water box.
Furthermore, it has been established that due to the large flow area provided by this rack design it is not possible to trap steam or air in the water boxes when the fuel boxes are filled with water.
As shown in Figure 4, there is no increase in reac-tivity until the relative density drops below about 70% of full water density and since there is no mechanism by which the water density within the pool could be reduced to a value anywhere near this,. the effect of water density variation on spent fuel rack reactivity is considered to be zero.
The basic cell was also used to evaluate the reactivity effect of axial neutron leakage.
A one-dimensional axial model, using flux weighted cross sections rom two-dimensional radial planar cell calculations, was used to simulate the center enriched. section and the. short natural uranium sections located at. the top and bottom of the enriched. section.
The" r
resulting reactivity perturbation due to axial neutron leakage is
.0026 5k.
A summary of the perturbations and biases to cell.reactivi"y calculation is shown in Table 5.
calculated reactivity of the spent fuel pool with unirradiated bundles with 4.25 w/o U-235 is
.9419 temperature of 200'F.
the basic
- Thus, the 595 for a pool
- 4. 0 TOLERANCES AND UNCERTAINTIES There are a number of tolerances and uncer ainties which result in perturbations which must be considered in the criticality analysis'he reactivity effect of all such positive perturbations is then combined statistically in accordance with Reference l8 to determine a single reactivity perturbation. which is added to the calculated basic cell multiplication factor (including biases) to determine the final conservative evaluation of the spent fuel rack maximum possible multiplication factor.
1 Based on the results of the calculational model benchmarking described in Section 2, the 2a uncertainty in the model, which corresponds to a 95/95 confidence statement, is.0024.
The stainless steel fuel and water boxes are nominally
.090 inches thick with a tolerance of +.004 inches.
Assuming a worst case in which all boxes were at the minimum thickness of.086 inches, the k of the basic cell is.9333.
Therefore, the maximum perturbation on the reactivity of the basic cell due to variations in the stainless steel box thickness is
+.0028 5k.
With the fuel bundles located in their most reactive positions inside the stainless steel boxes, the k of the basic cell is.9327.
- Thus, the perturbation on the basic cell reactivity due to positioning uncertainties is +.0022 Dk.
The nominal density of the pellets contained in the fuel assemblies is 95% of theoretical density.
Increasing this to the maximum possible theoretical density of U02 pellets of 964 results in a positive reactivity perturbation of.00l2 hk.
As shown in Table 5, the total reactivity perturbation to be added to the biased basic cell reactivity to account for tolerances and uncertainties is
.0044 hk.
This results in a final conservatively calculated spent fuel rack multi-plication factor of.9463.
Realistically, the spent fuel pool coolant contains a
minimum concentration of 2000 ppm boron at, all times.
When the reactivity effect of this minimum boron concentration is
- included, the actual spent fuel pool multiplication factor including all biases, tolerances and uncertainties 'is.6622 as shown in Table 5.
5.0 ACCIDENT ANALySIS The fuel racks are designed to prevent any criticality threat caused by a dropped fuel bundle which penetrates and occupies a position other than a normal fuel storage location.
The only positive effect of. such a bundle on the reactivity of the rack would be by virtue of reduction in axial neutron leakage from the rack.
Since. the calculations reported here.
show the total axial neutron leakage effect to be only
~.0026
- LQc, a dropped fuel bundle lying on top of the rack would not have any significant effect on the reported maximum possible reactivity of the spent fuel storage rack.
The lattice of the fuel assemblies results in an under-moderated configuration so that any crushing or compaction of the fuel assemblies would tend to reduce the multiplication factor of the spent fuel pool.
Therefore, the dropping of
heavy objects into the fuel pool or deformations from the
.effects of earthquakes or tornadoes could not produce a
criticality accident.
Figure 5 depicts a hypothetical accident condition which involves an extra bare fuel assembly inadvertently placed in a side water channel and next to the fuel rack fully loaded with unirradiated Westinghouse 14 x 14 opti-mized fuel assemblies.
When such an assembly is located in a non-fuel storage location, it is considered to be an abnormal condition and appropriate credit is taken for the soluble boron that is present in the pool water.
PDQ-07 calculations.
show that 2000 ppm of" soluble boron is worth 0.268 5k which. is more than enough to compensate for the positive reactivity effect caused by the extra fuel assembly.
The neutron multiplication factor of the rack system in this accident condition is 0.725 which is substantially less than the basic cell k of 0.9305 (see Table
Because of the well founded, conservative technique used for determination of the.infinite multiplication factor, there is more than reasonable assurance that this spent fuel rack.
design will not cause undue risk to the public health and
. safety resulting from criticality considerations.
10
6.0 ANALYSIS CONSERVATISHS As discussed previously, the basic cell calculations make the conservative assumption that the fuel pool water contains no boric acid when in fact the minimum boron content in the pool water is 2000 ppm.
As shown in Table 5, a
realistic evaluation of the maximum multiplication factor of the spent fuel racks, even when completely filled with un-irradiated fuel, is
.66'22.
11
REFERENCES 1.
R. F. Barry, "LEOPARD A Spectrum Dependent Non-. Spatial Depletion Code for the IBM-7094," SCAP-3269, September 1963.
2.
W.
R. Caldwell, "PDQ-7 Reference Manual," WAPD-TM-678, January 1967.
3.
H. Bohl, E. Gelbard and G.
- Ryan, "MUFT-4 -- Fast Neutron Spectrum. Code for the IBM-740, "WADP-TM-72, July 1957.
4.
H. Amster and R. Suarez, "The Calculation of Thermal Constants Averaged Over a Wigner-Wilkins Flux Spectrum:
Description of the SOFOCATE Code,"
WAPD-TM-39, January 1957.
5.
L. E. Strawbridge and R. F. Barry, "Criticality Calculations for Uniform Water-Moderated Lattices," Nuclear Science and
.Engineering, 23, 59, 1965.
6.
"Large Closed-Cycle Water Reactor Research and Development Program Progress Report for the Period January 1
March 31, 1965," WCAP-3269-12.
7.
"List of Equipment, and Apparatus at, WREC," Westinghouse Reactor Evaluation Center, February 1967.
8.
N. L. Orr, H. I. Sternbert, P.
- Deramaix, R.
H. Chastain, L. Binder and A. J.
Impink, "Saxton Plutonium Program, Nucle'ar Design of the Saxton Partial Plutonium Core,"
WCAP-3385-51, December 1965.
(Also EURAEC-1490.)
9.
R.
D. Le'amer, W. L. Orr, R. L. Stover, E.
G. Taylor, J.
P.
Tobin and A. Bukmir, "Pu02-UO2 Fueled Critical
'xperiments,"
WCAP.-3726-1, July 1967.
10.
A. F. Henry, "A Theoretical Method for Determining the Worth of Control Rods,"
WAPD-218, August 1959.
ll.
P.
W. Davison, et al.,
"Yankee Critical Expe'riments Measurements on Lattices of Stainless Steel Clad Slightly Enriched Uranium Dioxide Fuel Rods in Light Nater,"
YAEC-94, Westinghouse Atomic Power Division (1959).
12.
V. E. Grob and P.
N. Davison, et al.,
"Multi-Region Reactor Lattice Studies Results of Critic'al Experiments in Loose Lattices of UO2 Rods in H20," WCAP-1412, Westinghouse Atomic Power Division (1960).
s
~
~
)
~
REFERENCES (cont:)
13.
W. J. Eich and W. P. Kovacik, "Reactivity and Neutron Flux Studies in Multiregion Loaded Core,"
WCAP-1433, Westinghouse Atomic Power Division (1961).
14.
W. J. Eich, Personal Communication (1963).
15.
T. C. Engelder, et al., Measurement and Analysis of Uniform Lattices of Slightly Enriched HO2 Moderated by D20-H2 Mixtures," BAW-1273, the Babcock
& Wilcox Company (1963)-
16.
A. L. MacKinney and R.
M. Ball, "Reactivitv Measurements on Unperturbed; Slightly Enriched Uranium Dioxide Lattices, " BAW-1199, the Babcock
& Wilcox Company (1960).
17.
Battelle Pacific Northwest Laboratories, "Critical Separa-tion Between Subcritical Clusters of 2.35 Wtb 235-U Enriched UO2 Rods in Water with Fixed Neutron Poisons,"
PNL-2438.
18.
"OT Position for Review and Acceptance of Spent Fuel Storage and Handling App3.ications," U.S.
- NRC, April 14, 1978.
TABLE 1 SU)5)ARX OF LEOPARD RESULTS FOR HEASURED CRITICALS Case>>>>
mu bcc 1
2 3
4 5
6.7 8
9 10 11 12 13 14 15 16 17l0 19 20 21 22 23 24 25 26 27
'Reference Number ll 11 11 12 12 12 13 13 13 13 14 15 15 16 16 16 16 16 16 8
8 0
6 9
9 9
9 Enrichment
~cue 2.734 2.734 2 '34 2.734 2.734 2.734 2.734 2.734 3.745
- 3. 745
- 3. 745 4.099 4.099 4.099 4.069 4.069 4.069 3.037 3.037 0.714" 0.714*
0.714" 0.714*
0.729*
0.729*
0.729>>
0.729>>
H20/U Volume 2.10 2.93 3.00 7.02 8.49 10.13 2.50 4.51 2.50 4.51 4.51 2.55 2.14 2.59 3.53 0.02 9.90 2.64 0.10 1.60 2.17 4.70 10.76 1.11 3.49 3.49 1.54 Fuel Densify
~(lcm')
10.18 10.10 10.10 10.10 10.10 10.10 10.10 10.10 10.37 10.37 10.37 9.46 9.46 9.45 9.45 9.45 9.45
- 9. 20 9.20 9.52 9.52 9.52 9.52 9.35 9.35 9.35 9.35 Pellet Diameter (cm) 0.7620 0.7620 0.7620 0.7620 0.7620 0.7620 0.7620 0.7620 0.7544 0.7544 0.7544 1.1278 1.1270 1.1260 1.1260 1.1268 1.1260 1.1260 1.1268 0.0570 0.8570 0.0570 0.8570 1.2827 1.2827 1.2027 1.2827 Clad Diameter
~cm 1
0.0594 0.0594 0.0594 0.0594 0.0594 0.0594 0.0594 0.0594 0.0600 0.8600 0.0600 1.2090 1.2090 1.2701 1.2701 1.2701 1.2701 1.2701 1.2701 0.9931 0.9931 0.9931 0.9931 1.4427 1.4427 1.4427 1.4427 Clad Thickness (cm) 0.04085 0.04085 0.04005 0.04005 0.04005 0.04085 0.04085 0.04005 0.0406 0.0406 0.0406 0.0406 0.0406 0.07163 0.07163 0.07163 0.07163 0.07163 0.07163 0.0592 0.0592 0.0592 0.0592 0.0800 0.0000 0.0000 0.0000 Lattice Pitch
~cm )
- 1. 0287 1.1049 1.1930 1.4554 1.5621 1.6091 1.0617 1.2522 1.0617 1.2522 1.2522 1.5113 1.450 1.555 1.604 2.198 2.381 1.555 2.190 1.3200 1.4224 1.0669 2.6416 1.7526 2.4705 2.4705 1.9050 Critical Buckling m
40.75 53.23 63.20 65.64 60.07 52.92 47.5 60.0 60.3 95.1 95.60 88.0 79.0 69.25 05.52 92.84 91.79 50.75 60.81 100.0 121.5 159.6 120.4 09.1 104.72 79.5 90.0 Calculated eff 1.0015 1.0052 1.0043 1.0090 1.0110 1;0072 1.0008 0.9907 1.0010 1.0025 1.0009-0.9809 0.9030 0.9999 0.9958 1.0040 0.9072 0.9946 0.9009 0.9912 1.0029 0.9944 1.0000 0.9902 1.0055 0.9948 0.9070
'These axe PuO2 in Hatural UO2.
- 'ases 1 through 19 are with stainless steel clad, Cases 20 thxough 27 are zircaloy.
~
~
TABLE 2 WESTINGBOUSE UO Zr-4 CLAD CYLINDRICAL CORE CRITICAL.EXPERIMENTS 2
BORON PITCB CONCENTRATION I
MATERIAL DUCKING FOR LEOPARD CM-2 CRITICAL NO.
OF PINS RADIUS OF FUEL REGION (cm) eff LEOPARD/PDQ-1 2
3 4
5 6
7 8
9 10 11 12 13 14 0.600 0.690 0.848 0.976 0.600 0.600 0.600 0.600 0.600 0.600 0.600 0.600 0.848 0.848 306.
536.4 727.7 104.
218.
330.
446.
657.1 104.
218.
.008793
.009725
.008637
.006458
.007177
.006244
.005572
.008165
.007599
-007106
.006661
.005809
.007320
.006073 489.4 317.0 251.6 293.0 659.9 807.2 950.2 546.3 607.1 669.5 735.3 895.3 321.0 420.5 19.021 17.605 19.276 23.935 22.088 24.429 26.504 20.097 21.186 22.248 23.315 25.727 21.772 24.919 0.9912 0.9941 0.9927 0.9935 0.9927 0.9937 0.9940 0.9919 0.9917 0.9916 0.9909 0.9944 0.9938 0.9925 Fuel Re ion Data Enrichmdnt Fuel Density Pellet Radius Clad IR Clad OR
- 2. 719 w/o Q-235 10.41 g/cm~
c 0.20 in
~ 0.2027 in
= 0.23415 in (b) Thickness of water reflector is that required to attain total radius of 50 cm for model.
(c)
BZ
~.000527 cm 0.9928 Mean 0.0012 Std.
~
g
TABLE 3 BATTELLE CRITICALS Case Length Times
~l4 dth No. of Assemblies
~rn nrra Absorber
~T e
Thickness Distance 'ritical To Fuel SeParation Cluster of Clusters keff LEOPARD/PDQ 028 027 20x16 20x16 m
3 S.S.
.485 cm S.S.
.302
.645 cm
.645 6.88 cm 7.43 0.9922 0.9919 002D 015 013 022 021 20x18.075 20x17 20x16 20xl5 20x16 3
3 None 11.94'm 8.42 6.39 4.46 0.9956 0.9932 0.9921 0.9933 0.9946 Battelle
.9933 Statistical Summar
~EE eriments member mean kaie a
.0013
.Ba tte1 le and Westinghouse (Table 2) 21
.9929
.0012 Fuel region data:
Enrichment 2.35 w/o, Pellet radius mm 0.5588 cm, Clad OR
.635 cn, Wall thickness am.0762 cm, Pitch
~ 2.032 cm
~
~
TABLE 4
FUEL ASSEMELY CHARACTERISTICS Numbe" of rods conta'ning UO2 Rod pitch (in)
Overall envelope dimensions (in)
Weight of U
(KgU)
Active fuel length (in)
Enriched uranium region Length (in)
Enr ichment (w/o)
Natural uranium blanket region Lenght (in)
Enrichment (w/o)
Instrument tube Material O.D. (in)
I.D. (in)
Guide tubes Material O.D. (in).,
,O.D. (in),
I.D. (in),
I.D. (in),
Fuel pellet Material above
- dashpot, in dashpot above dashpot
'n dashpot Density
(% theoret'cal,)
O.D. (in)
Cladding O.D. (in)
I ~ D. (in)
Spacer Grids Number Location Weigh" s of materials" Inconel grids (2), lbs total Zircaloy grids (7), lbs total 179 0.5560 7.763 350.5 141.4 128.98 4 '5 12.42
- 0. 711 Zr-4 0.4015 0.3499 Zr-4 0.5280 0.4825 0.4900 0.4425 UO2 95 +1.0 "1.5 0.3444
+ 0.0019
.400
.3514 9
(see outline drawing) 3.00 19.46
- Does not include weight of the stainless steel sleeves or inserts.
17
TABLE 5
SUMMARY
OF PERTURBATIOiNS TO THE MULTIPLICATION FACTOR OF THE BASIC CELL
~eeecri tice Basic cell at 68'F, 4.25 w/o U-235 in the enriched center section of the assembly rk effect k
.9305 Calculat'onal Biases Inc ease in temperature from 68oF to 200 F
LEOPARD/PDQ Model Bias Mesh spacing effect Net axial leakage ef feet Basic Cell including biases
+.0078
+.0071
-.0009
".0026
.9419 Tolerances and Uncertainties Tolerance on SS box thickness Fuel position uncertainty Maximum fuel pellet density Calculational uncertainty Total Uncertainty (statistica).)
+.0028
+.0022
+.0012
+.0024
+.0044 Maximum k including Biases
- and, Uncertainties
. 9463 Basic Cell 8 68'F,
- 4. 25 w/o with 2000 ppm Boron
.6622 18
FIGURE l NOTE:
Boundary Condition at the Top ol'his Figure is 180 Rotational Syn.retry RGE RACIC REFERENCE DESIGN 02 02 Region No.
Materials Qi Q2 03 Qq Q5 Fuel Pin Cells Guide Tube Cells Instrument Tube Cell Explicit Stainle'ss Steel Explicit Water
- 8. 430
- 0. 233
0
~
0 0 ~
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~-
~ ~
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g e
~-
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Figure S
PDQ-07 Accident.Geometry
~
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-F <<
=F~
I
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Bare fuel Bundl I~ ~
I I..
~ ~
~
~ ~ ~ ~
~
W II ~
I ~
=F" F
I
~
infinite water..
reflector F
~
~
~ ~
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-F-0 I
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~ l
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I t,.
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~"'-~--"':::l:-"-::-:~--:::l:'::::::-':.=-:-:.=:=-':"-:::-.~.-::l:-~:-::F Normai fuel storage avoca tion ii I:.,
' "'::"-i'"';,"""'". "
~ ~;:::f": ":"]::::::::)...l.."..'""!"':."I:
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23
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